JAERI-M - International Nuclear Information System (INIS)

J A E R I - M 88-221

REACTOR ENGINEERING DEPARTMENT ANNUAL REPORT

(April I, I 9 8 7 - M a r c h 3 1, 1988)

November 1988

Department of Reactor Engineering

Japan Atomic Energy Research Institute

JAER 1-M

88-221

REACTOR ENGINEERING DEPARTMENT

ANNUAL REPORT

(April 1， I 987-March 3 1， 1988)

?、~ovem !)er 1988

日本原子力研究所Jopan Atomic Energy Research Institute

JAERI M u . t : - l;.t, H*ifi r->JWffiii>r*'^'tltl|l-i>fi|L O •6W£8!i'r, 'fT-1' A. r-^'rliKf -b-^.'i, H *i# r- >i W-^*ttttrtfiSiSBtffftlVt*:f** (T:IU> - 11**WIMffidufcf i

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JAERI M reports are issued irregularly Inquiries about availability of the reports should lx- addressed to Information Division, Department

of Technical Information. Japan Atomic Energy Research Institute. Tokai mura. .Maka gun. Ibaraki ken 319 11, Japan

'© Japan Atomic Energy Research Institute. 1988

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Ihafilkl ken 319 11. Japan

'[) ]apan Aloml仁 Enげ日vResearch In日ll1ule. 1988

制限維11':li 11 ~ I原(}J側先叫

I/J 刷 u也尚述1:11刷株式会付

JAERI-M 88-221

Reactor Engineering Department Annual Report

(April 1, 1987 - March 31, 1988)

Department of Reactor Engineering Tokai Research Establishment

Japan Atomic Energy Research Institute Tokai-mura, Naka-gun, Ibaraki-ken

(Received October 6, 1988)

This report summarizes the research and development activities in the Department of Reactor Engineering during the fiscal year of 1987 (April 1, 1987 - March 31, 1988).

The major activities in the Department concerns the programs of the high temperature gas-cooled reactor, the high conversion light water reactor, the advanced fission reactor system and the fusion reactor at JAERI and the fast breeder reactor at PNC.

The report contains the latest progress in nuclear data and group constants, theoretical methods and code development, reactor physics experiments and analyses, fusion neutronics, shielding, reactor and nuclear instrumentation, reactor control/diagnosis and robotics, as well as the new topics from this fiscal year on advanced reactors system design studies and technique developments related the facilities in the Department. Also described are the activities of the Research Committee on Reactor Physics.

Keywords: Reactor Engineering Annual Report, JENDL-3T Tests, Advanced Reactor System, Reactor Physics, Shielding, Reactor Instrumentation, Reactor Control, HTTR, HCLWR, FBR, Fusion Neutronics

Board of Editors for Annual Report T. Nakamura (Chief Editor) K. Tsuchihashi (Associate Chief Editor) M. Obu, I. Kanno, K. Sakasai, K. Suzuki, J. Kusano, Y. Ikeda, H. Kotegawa, H. Takano, T. Yamashita

JAERI-M 88-221

、‘，，

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p・今

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，，‘、

Tokai Research Estab1ishment

Japan Atomic Energy Research Institute

Tokai--mura， Naka-gun， Ibaraki-ken

(Received October 6， 1988)

This report summarizes the research and deve10pment activities in the

Department of Reactor Engineering during the fisca1 year of 1987 (Apri1 1，

1987 -March 31， 1988).

The major activities in the Department concerns the programs of the

high temperature gas-coo1ed reactor， the high conversion 1ight water

reactor， the advanced fission reactor system and the fusion reactor at

JAERI and the fast breeder reactor at PNC.

The report contains the 1atest progress in nuc1ear data and group

constants， theor巴tica1methods and code development， reactor physics

experiments and ana1yses， fusion neutronics， shie1ding， reactor and

nucle.ar instrumentation， reactor contro1/diagnosis and robotics， as well

as the new topics from this fiscal year on advanced reactors system design

studies and technique developments related the facilities in the

Department. Also described are the activities of the Research Committee

on Reactor Physics.

Keywords: Reactor Engineering Annual Report， JENDL-3T Tests， Advanced

Reactor System， Reactor Physics， Shielding， Reactor Instru-

mentation， Reactor Control， HTTR， HCLWR， FBR， Fusion

Neutronics

Board of Editors for Annual Report T. Nakamura (Chief Editor) K. TSllchihashi (Associate Chief Editor) M. Obu， 1. Kanno， K. Sakasai， K. Suzuki， J. Kusano， Y. Ikeda. H. Kotegawa， H. Takano. T. Yamashita

J A E R I - M 8 8 - 2 2 1

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]AERI-M 88-221

昭和62年度防(-{ーが L学古f5年報

LJ本JJj([-ノJ研究Jifi*泌研究所

JJ;(-(カ L・'{:8ii

( 1 988 "f' 10) J 6 LJ ~ fl]1)

昭和62年度における除、子炉 L'1:部の研究活動状況をとりまとめた。 Jf，i-[-カj 1乍剖5U)研究

は. JJ;( liffの諸プロジェク卜 lfb JMカスが向転換経水炉.次世代新型自カiおよび十五融

合;炉ーならびに動燃事業団における高速増殖炉の開発iこ密接に関係するものである。本報

告には.絞データと群定数. 'II-; f望論とコード開発.?:;i物庖!積分実験と解析.骸融fTニュー

トロニクス.放射線遮蔽.原子炉.HiJlリ・計装. 1.息子炉制御・診断・ロボット技術における

研究の進展 i乙加え今期は.新しく新型炉概念設計研究およびか物J~!脳設技術 l泊先がìÆべら

れている。またかi物路!に閲する研究委H会活動も含まれている。

東海研究所:〒319-11淡減県郷珂郡東海村白}j~f-白似 2 -4

原子炉工学部年報編集委員会

中村知1夫(委員長).土橋敬一郎(副委員長).大部誠.神野郎犬.坂{牛JH時，鈴木勝男.革野謙一.

池田裕二郎.小手川洋，高野秀機.山ド哲行

JAERI-M 88-221

Contents

Foreword i

I. Nuclear Data and Group Constants - Benchmark Tests of JENDL-3T - < H. Takano - 4

1.1 Benchmark Test of JENDL-3T for Thermal Reactor and High Conversion Light Water Reactor < H. Takano et al. • ... 5

1.2 Benchmark Test of JENDL-3T on Fast Reactors < H. Takano et al. -- 8

1.3 Problem of Total Cross Section for 0, N, Na, Fe, SUS in JENDL-3T by the Analysis of BROOMSTICK Experiment < A. Hasegawa - \\

1.4 Assessment of Natural Iron Cross Sections of JENDL-3T through ASPIS Deep Penetration Shielding Experiment < A. Hasegawa -' yt

1.5 Benchmark Test of JENDL-3T for Materials Used in Fusion Reactors < T. Mori et al. > 17

1.6 Integral Test of JENDL-3T Through Benchmark Experiments Using FNS < H. Maekawa et al. > 20

1.7 Angular Neutron Flux Measurement and Nuclear Data Test on Slabs of Fusion Blanket Materials < Y. Oyama et al > ....23

1.8 Compilation of MCNP Data Library Based on JENDL-3T < K. Sakurai et al. > 26

1.9 Development of the JSSTDL-295 Group Cross Section Library System for Shielding Calculation < A. Hasegawa > "29

2. Theoretical Method and Code Development < Y. Ishiguro > 32 2.1 Dependence of Spallation Product Distributions on the

Level Density Parameter < T. Nishida et al. > 33 2.2 Evolution Analysis of Spallation Products in Their

Buildup and Decay Processes < Y. Nakahara et al. > 3^ 2.3 Improvements on Intranuclear Cascade Model

< H. Takada et al. > 39 2.A Development of a Vectorized Monte Carlo Code GMVP

< T. Mori et a l . > 42

JAERト M 88-221

Contents

Foreword ............... 1

1. Nucle.ar Data and Group Constdnts

-Benchmark Tests of JENDL同 3T- 〆 H. Takano > ・・・・・・・・・・・・・・・ 4

1.1 Benchmark Tes t of JENDト 3Tfor Thermal Reactor and

High Conversion Light Water Reactor 〆 H. Takano et al. ・・・ 5

1.2 Benchmark Test of JENDL-3T on Fast Reactors

< H. Takano et al. > ••••••••••••••••••••••••••••••••••••••• 8

1.3 Problem oE Total Cro自呂 Section Eor O. N. Na. Fe. SUS

in JENDL-3T by the Analysi呂口fBROONSTICK E.，，<per !ment

< A. Haseg!ma :--・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 11

1.4 Assessment oE Natural Iron Cross Sect!ons of JENDL-3T

thrロughASPIS Deep Penetration Shielding Experiment

< A. Hasegawa > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 1/~

1.5 Benchmark Test of JENDL-3T for Materials Used in Fusion

Reac tors < T. Mori et al. > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 17

1.6 Integral Test of JENDL-3T Through Benchmark Experiments

Using FNS く H. Haekawa et al. > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 20

1.7 Angular Neutron Flux Measurement "md Nuclear Data Test

on Slabs of Fusion Blanket Materials < Y. Oyama et al > ・・・・ 23

1.8 Compilation of HCNP Data Library Based on JENDL-3T

く K. Sakurai et al. > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 26

1.9 Development of the JSSTDL-295 Group Cross Section Library

System for Shielding Calculation < A. Hasegawa > ・・・・・・・・・・・29

2. Theoretical Method and Code Development く Y. Ishiguro > ・・・・・・・ 32

2 .1 Dependenc巳 ofSpallation Product Distributions on the

Level Density Parameter < T. Nishida et al. >・・・・・・・・・・・・・・・ 33

2.2 Evolution Analysis of Spal1ation Products in Their

Buildup and Decay Processes < Y. Nakahara et al. > ・・・・・・・・・ 36

2.3 Improvements on Intranuclear Cascade Model

< H. Takada et a1. 、・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・39

2.4 Deve10pment of a Vectorized Honte Car10 Code GMVP

く T. ~lori et al. > ............・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・42

J A E R 1 - M 8 8 - 2 2 1

2.5 Development of the Standardized Intelligent Shielding Analysis Code Package: INTEL-BERMUDA •• A. Hasegawa et al. • 45

3. Reactor Physics Experiment and Analysis •- M. Nakano > ^g 3.1 Measurement of Reactivity Worth of Experimental Control

Rod in VHTRC-1 Core with Integral Version of PNS Method • I. Kanno et al. > 40

3.2 Measurement of Reactivity Worth of Burnable Poison Rod in VHTRC-1 Core < F. Akino et al . > 5 1

3.3 Measurements of 63Cu(n,y) Reaction Rate Distribution in VHTRC-3 Core < T. Yamane et al. > 5 3

3.4 A Simple Method for Reactivity Determination Based on Integral Version of Pulsed Neutron Area-Ratio Method < T. Yamane et al. > cc

3.5 Measurements and Analyses on K .. and K. . at the J elf m f

F~A-XV-1 Core < T. Osugi et al . 5 g

3.6 Absorber Material Re^tivity Worth in the FCA-XIV Cores < S. Oka j ima et al . > i

3.7 Measurement of Reaction Rate Ratios at FCA-XIV Cores < M. Obu et al. > 6 4

3.8 Measurement of Adjoint Flux Weighted Infinite Multiplication Factor Using Central Cell Reactivity Worth a t the FCA-XIV-1 and -XIV-2 Cores < T. Sakurai et al. > ... 67

4. Advanced Reactor System Design Studies < T. Hiraoka > JQ

4.1 The Concept of High Conversion Light Water Reactor with Flat Core and Its Applications < Y. Ishiguro et al. > 72

4.2 Note on Burnup Calculation of Pu Isotopes in Blanket Used in HCLWR < K. Okumura et al. > ? 4

4.3 Burnup Calculation Method for Spectral Shift Reactor Using Fertile Rod < K. Okumura et al . > -,r

4.4 NEACRP HCLWR Cell Burnup Benchmark Calculation < H. Akie et al. > 7 g

4.5 Treatment of the 1.06eV Resonance of Pu-240 in Tight Lattice Cell < H. Akie et al. > g 2

4.6 Conceptual Design of SPWR (System-Integrated Pressurized Water Reactor) < K. Sako > 8 5

]AERI-M 88-221

2.5 Deve10prnent of the Standardized Inte11igent Shielding

Ana1ysis Code Package: lNTEL-BERMUVA

、 A. Hasegawa et al. 、・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 45

3. Reactor Physics Experirnent and Analysis ヘ M. Nakano :> ・・・・・・・・・ 48

3.1 Measu r巴rnentof Reactivity Worth of Experirncntal Control

Rod in VHTRC-l Core with lntegral Version of PNS Method

1. 1く孔nnoet a1. > ・・・・ 49

3.2 Measurernent of React:ivity Worth of 13urnable Poison Rod

in VHTRC-l Core 、F. Akino et al. > .......................・ 51

3.3 Measurernents of 63Cu(n，y) Reaction Rate Distribution

in VHTRC-3 Core く T. Yarnane et a1. > ・・・・・・・・・・・・・・・・・・・・・・・ 53

3.4 A Sirnp1e Method for Reactivity Deterrnination Based

on Integra1 Version of Pu1sed Neutron Area-Ratio Method

T. Yarnane et a1. >

3.5 ~leasurernents and Ana1yses on Ke[fニndk in f a t the

・・ '55

FぐA-XV-lCore < T. Osugi et al. ・・・・・・・・・・・・・・・・・・・・・・・・・ 58

3.6 Absorber Naterial Re;:>("'tivity Wort:h in the FCA-XIV Cores

て S. Okajirna et 81. :> ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 61

3.7 Measurernent of Reaction Rate Ratios at FCA-XIV Cores

く M. Obu et a1. > •.•..••••.• 64

3.8 Measurernent of Adjoint F1ux Weighted lnfinite

Multiplication Factor Using Central Ce11 Reactivity Worth

at the FCA-XIV-l and -XIV-2 Cores く T. Sakurai et a1. > ・・・ 67

4. Advanced Reactor Systern Design Studies く T. Hiraoka > ・・・・・・・・ 70

4.1 The Concept of High Conversion Light Water Reactor with

Flat Core and lts Applications < Y. Ishiguro et al. ¥ ・・・・・ 71

4.2 Note 011 Burnup Calculation of PU Isotopes in Blanket

Used in HCLWR < K. Okurnura et al. > ・・・・・・・・・・・・・・・・・・・・・・・ 74

4.3 Burnup Ca1cu1ation Method for Spectral Shift Reactor

Using Ferti1e Rod < K. Okurnura et a1. > ・・・・・・・・・・・・・・・・・・・ 76

4.4 NEACRP HCLWR Cell Burnup Benchrnark Ca1cu1ation

く H.Akie et a1. > .... ......•...... 79

4.5 Treatrnent of the 1.06eV Resonance of Pu-240 in Tight

Lattice Ce11 < H. Akie et a1. > .••••• 82

4.6 Conceptua1 Design of SPWR (Systern-Integrated Pressurized

Water Reactor) < K. Sako > .••••••.••••• ・・・・・・・・・・・・・・・・・・ 85

J A E R I - M 8 8 - 2 2 1

4.7 Design Study of the Valves for Reactor Shutdown System of SPWR < K. Sako •• 8 8

4.8 Preliminary Characteristic Analysis of a System-Integrated PWR in Inherently Safe Concept < T. Ise et al. gQ

4.9 Controllability of Loud Reduction of 1SF.R (Intrinsically Safe and Economical Reactor) < Y. Asahi > y-j

4.10 A Very High Burnup Pressurized Water Reactor with Highly Enriched Plutonium Fuel Assemblies Using a Spectral Shift Concept •• 11. Ichikawa ct al. • 95

4.11 Analysis of Fission Product and Minor Actinide Build-up in Very High Burnup Pressurized Water Reactor < H. Takano et al. > no

4.12 Conceptual Study of Higher Actinide Burner Reactors < T. Mukaiyama et al 202

4.13 Concepts of TRU Fuel and Predictions of Phase Diagrams •- T. Ogawa > 1 0 5

4.14 Effect of Nuclear Data Uncertainty on Nuclear Characteristics in TRU-Burner Fast Reactor •- H. Takano 208

4.15 High Breeding / High Burnup Fast Reactor — Proposal of FP Gas Purge / Tube-in-Shell Metallic

Fuel Fast Reactor < T. Hiraoka et al. > -Ql 4.16 Study of Nuclear Characteristics of Nitride- and

Carbide-Fueled LMFBRs < S. Iijima et al. - n 4

5. Fusion Neutronics < T. Nakamura > 217 5.1 Phase-IIB Experiment of JAERl/USDOE Collaborative Program

on Fusion Blanket Neutronics < Y. Oyama et al. > 218 5.2 Analysis of Neutronics Parameters Measured in Phase II

Experiments of the JAERl/USDOE Collaborative Program < M. Nakagawa et al. > 221

5.3 Blanket Benchmark Experiment on a Beryllium-Sandwich Lith ium—Oxide Cylindrical Assembly < H. Maekawa et al. > ... 224

5.4 Measurements and Analyses of Gamma-Ray Heating in Lithium-Oxide, Graphite and Iron Slab Assemblies Irradiated with D-T Neutrons < S. Yamaguchi et al. > 227

jAERI-M 88・221

4.7 Design Study of the Valves for Reactor Shutdown System

of SPWR ピ K. Sako ・・・ 88

4.8 Preliminary Characteristic Analysis of a System-Inte民rated

PWR in Inherently Safe Concept ¥ T. 1 se 日仁、11. 9な

4.9 Contro11abillty of Loud Reduction of JSER (Intrlnsic‘111 v

Safe and Econom1cal Reaじtor) < Y. '¥1'はh1、.• • • • • • • • • • • • • • • •• 93

4.10 ^ Very H1gh surnup I'ressllrized Water I~eactor with Highly Enriched 1'111toDillm Fllel Assemb1ies Using a

Spectral Shift Concept . 11. Ichikawa l't a1. ・・・・・・・・・・・・・ 96

4.11 Analysis of Fission Product and Minor Actinide suild-llP

in Very High Bllrnllp Pressurized Water Reactor

H. Takano et al. > 99

4.12 Conceptual Study of Higher Actinide Burner Reactors

く T. ~lukaiyama et al. ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 102

4.13 Concepts of TRU Fuel and Predictions of Phase Diagrams

¥ T. Ogawa > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 105

4.14 Effect o[ Nllclear Data Uncercainty on Nllclear

Characteristics in TRU-Burncr Fast Reactor

H. Takano ・・・・・・・・ 108

4.15 High Breeding / High Burnup Fast Reactor

ー- Proposal of FP Gas Purge / Tube-in-Shel1 Metallic

Fuel Fast Reactor < T. Hiraoka 巴tal. > ・・・・・・・・・・・・・・・・ 111

4.16 Study of Nuclear Character~~tics of Nitride-and

Carbide-Fue1.ed U1FBRs < S. Iij ima et a1.. 、、・・・・・・・・・・・・・・・・ 114

5. Fusion Neutronics < T. Nakamura > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 117

5.1 Phase-IIB Experiment of JAER1/USDOE Collaborative Program

on Fusion Blanket Neutr0nics く Y. Oyama et al. > ....・・・・・・・ 118

5.2 Analysis of Nelltronics Parameters Measured in Phase 11

ExpeLiments of the JAERI/USDOE Collaborative Program

< N. Nakagawa et a1.. > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 121

5.3 Blanket Benchmark Experiment on a Beryl1ium-Sandwich

Lithium-Oxide Cylindrical Assembly く H. 1、laelくawaet al. > ・・・ 124

5.4 Neasurements and Analyses of Ganrrna-Ray Heating in

Lithium-Oxide， Graphite and lron Slab Assemblies

Irradiated with D-T Neutrons < S. Yamaguchi et al. > ・・・・・・・ 127

JAERI-M 88-221

5.5 Absolute Cross Section Measurements for '7A1(n.cO'^Na, 9 0Zr(n,2n) 9 sZr and '';iNb(n,2n) 9->mNb Activation cross Section at Neutron Energy Range from 13.3 to 15.0 MeV by Means of Associated a-Particle Counting Method • Y. Ikeda et al . • j JQ

5.6 The Formulation of Systematics for (n,p), (n,np), (n,'<) and (n,2n) Reaction Cross Sections Based on the Data Measured at FNS • C. Konno et al. 133

5.7 Measuring System for Secondary Gamma-ray Production Cross Section at FNS • Y. Ikeda et al . 136

5.8 Determination of D-T Neutron Source Strenghth and Position by Foil Activation Technique < Y. Ikeda et al. > ..139

5.9 A Plan of Neutronics Experiments for Next Fusion Engineering Facilities Using FNS < H. Maekawa > 142

5.10 An Intense DT Neutron Source and Test Conditions for Fusion Nuclear Technology Research •• Y. Oyama et al 145

6. Radiation Shielding < T. Suzuki > 143 6.1 Experiment and Analysis on the Behavior of 14 MeV

Neutrons Incident to a Large Cavity < H. Nakashima et al. -149 6.2 Measurement and Analysis on Reaction-Rate Distribution

in a Large Iron Cylindrical Assembly < K. Oishi et al. > •••152 6.3 A Comparative Study of Shielding Design Between Advanced

Marine Reactors in Concept < T. Ise et al . > 155 6.A Design Method of Compensational Shield for Shield

Irregularities of Reprocessing Plant ••' A. Yamaji et al . ..158 6.5 Physics and Shielding Design of High Quality Neutron Beam

Hole in a Research Reactor — Comparison Between Tangential and Radial Beams — < T. Ise et al.. > 161

6.6 Exposure Buildup Factors for Slant Penetration through Slab Shields from Point Isotopic Gamma-Ray Source < Y. Kanai et al . > 164

7. Reactor and Nuclear Instrumentation < N.Wakayama > 167 7.1 Development of Nuclear Instrumentation System for HTTR

< N. Wakayama et al. > 168

]AE R 1・M 88-221

5.5 Absolute Cross Section Measurements for ~7Al(n ， u)2 匂 N Ll，

9oZr(n，2n)69Zr and ~JNb(n ， 2n)g 川 NbActivLltion cross

Section Llt Neutron Energy Range from 13.3 to 15.0 ~1己 V

by ト~Ll ns of Associated a-Particle Counting Method

Y. 1 ked3 e t u 1 . ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・lJ()

5.6 Th己 Formulationof Systematics for (n，p)， (n，np)， (n，，，)

und (n，2n) Reaction Cross Sections Based on the Data

トlellsuredat FNS ・じ. Kllnl111 et <11. ・・・・・・・・・・・・・・・・・・・・・・・・ 1)3

5. 7 ~leasllring System for Secondury Cumma-ray I'roduction

じ1"OSS Section at FNS . Y. lkedu eL a1. ...................136

5.8 Determination of D-T Neutron Sourc巴 Strenghthand

Position by Foil Activation Technique < Y. lkeda et al. > ..139

5.9 A Plan of Nelltronics Experiments [or Next Fusion

Engineering Facilities Using FNS ど H. トlaekawu > ••..••...... 142

5.10 An lntense DT Neutron Source and Test Conditions for

Fusion Nuclear Technology Research . Y. Oyama et al. ・・・・ 145

6. Radiation Shielding <: T. 日117.11ki > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 148

6.1 Experiment <1nd An江 lysison the llL'havior of )!j ~leV

Neutrons Incident to a Large Cavity ~ H. Nukushima et al. . 149

6.2 Meusurement and An品 lysison Reaction-Rateυistribution

in a Large Jron Cylindrical Assembly ぞ K. Oishi et al. > ・・・ 152

6.3 A Comparative Study of Shlelding Design Betweell Advanced

ト~rine Reactors in Concept ~ T. Ise et al. > ・・・・・・・・・・・・・・・ 155

6.1， Design Method of Compensational Shield for Shield lrregularities of Reprocessing Plant < A. Yamaji et al. ・・ 158

6.5 Physics and Shielding Design of High Quality Neutron Heum

Hole in a Research Reactor

ーー ComparisonBetween Tangential and Radial Beams --

ベ T. Ise et a1.. > •••.•.••...••..••.••••••..•• •••••.• ••••.. .161

6.6 Exposure BlIildup Factors for Slant Penetration through

Slab Shields from Point Isotopic Gamma-Ray Source

< Y. Kanai et al. > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 164

7. Reactor and Nuclear lnstrumentation < N.Wakayama > ・・・・・・・・・・・・ 167

7.1 Development of Nuclear Instrume~tation System for HTTR

く N. Wakavama et al. > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 168

J A E R I - M 8 8 - 2 2 1

7.2 High-Temperature Out-Pile Tests of N-Type Thermocouples < M. Yamada et al. > -J72

7.3 Development of Fuel Failure Detection System for High Temperature Gas Cooled Reactors < S. Asami et al. > ^75

7.A Development of Nondestructive Measuring Techniques of Transuranic Elements in Waste Drums < H. Clotoh et al. > .... 1 7g

7.5 Development of an In-situ Nondestructive Measurement System for Measuring Radioactivities Inside Contaminated Pipe •- M. Katagiri et al. • 281

7.6 Formation of Magnetic Gratings by Electron/Laser Beam Irradiation < K. Ara et al. > ^g^

7.7 Determination Methods of Gamma-ray Peak Area < K. Teranishi et al. > ,gy

8. Reactor Control, Diagnosis and Robotics < Y. Shinohara > jgQ 8.1 Dynamics Analysis Code for Nuclear Ship Propulsion Reactor

< K. Nabeshima et al. > jg^ 8.2 Nonlinear Simulation Method in the CAD System for Control

System Analysis of Nuclear Plant ••' J. Shimazaki et al. > .. . X93 8.3 Study of Optimal Reactor Control Using an Al Method

< Y. Shinohara > 105 8.4 A Design Study on Computer Control of HTTR Plant

< J. Shimazaki et al. > ^gy

8.5 Development of Computer Code STAR-II for Dynamics Analysis and Diagnosis of Nuclear Reactor System < K. Hayashi et al. > ^99

8.6 Development of Nonstationary Reactor Noise Signal Recording System < K. Hayashi > 201

8.7 Study on the Goodness of System Identification Using Multivariate AR Modeling < K. Hayashi at al. > 204

8.8 A Method of Nonstationary Noise Analysis Using Instantaneous AR Spectrum and Its Application to Borssele Reactor Noise Analysis < K. Hayashi et al. > 205

8.9 Study on Sodium Boiling Detection Method < Y. Shinohara et al. > 207

8.10 Theoretical Studies of Manipulator Inverse Kinematics < S. Sasaki et al. > 209

]AERI-M 88・221

7.2 High-Temperature Out-Pi1e Tests of N-Type Thermocoup1es

く M. Yamada et a1. > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 172

7.3 Deve10pment of Fuel Fai1ure Detection System for High

Temperature Gas Coo1ed Reactors < S. Asami et al. > ・・・・・・・・ 175

7.4 Development of Nond己structiveMeasuring Techniques of

Transuranic E1ements in Waste Drums に H. (;otoh et al. > ・・・・ 178

7.5 Deve10pment of an ln-situ Nondestructive Measurement

System for ト~asuring Radioactivities Inside Contaminated

Pipe 、ト1. Katagiri et a1. 、. ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 181

7.6 Formation of Magnetic Gratings by E1ectron/Laser Beam

lrradiation < K. Ara et a:. > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 184

7.7 Determination Methods of Gamma-ray Peak Area

< K. Teranishi et a1. > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 187

8. Reac tor ControJ.. Diagnosis and Robotics < Y. Shinohara > ・・・・・・ 190

8.1 Dynamics Ana1ysis Code for Nuc1ear Ship Propu1sion Reactor

< K. Naheshima et a1. > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 191

8.2 Nonlinear Simu1ation Method ill the CAD System for Contr01

Systern Ana1ysis of Nuclear P1ant J. Shimazaki et a1. > ・・・ 193

8.3 Study of Optima1 Reactor. Contro1 Usipg an Al Method

く Y. Shinohara > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 195

8.4 A Design Study on Computer Contro1 of HTTR Plant

< J. Shimazaki et a1. > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 197

8.5 Development of Computer Code STA良ー11 for Dynamics

Ana1ysis and Diagnosis of Nuc1ear Reactor System

く K. Hayashi et a1. > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 199

8.6 Deve10pment of Nonstationary Reactor Noise Signa1

Recording System < K. Hayashi > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 201

8.7 Study on the Goodness of System ldentification Using

Mu1tivariate AR Mode1ing < K. Hayashi et a1. > ・・・・・・・・・・・・・ 204

8.8 A Method of Nonstationary Noise Analysis Using

lnstantaneous AR Spectrum and Its App1ication to

Borsse1e Reactor Noise Ana1ysis < K. Hayashi et a1. > ・・・・・・ 205

8.9 Study on Sodium Boiling Detection Method

< Y. Shinohara et a1. 、・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 207

8.10 Theoretical Studies of Manipu1ator lnverse Kinematics

く S. Sasaki et a1. > •••••••••••••••••••••••••••••••••••••• 209

JAERI-M 88-221

8 .11 Genera l Formula t ion of a Man ipu la to r Linkage Mechanism

< S. Sasak i > 211

8.12 Development of Robotic Remote Handling Technology < H. Usui et al. > 212

9. Facility Operation and Technique Development < M. Nakano > .... 2\k

9. 1 Operation Report of FCA < K. Satoli et al . > 215 9.2 Operation Report of FNS < J. Kusano et al. > ••• 217 9.3 Tritium Quantification by Measurement of Characteristic

X-Ray < C. Kutsukake et al. > 219 9.4 An Improvement of Beam Transmission Performance in FNS

Accelerator * J. Kusano et al. > 222 9.5 Operation Report of VHTRC < S. Fujisaki et al. > 224 9.6 Development of VHTRC Operating Data Processing System

< S. Fu j isaki et al. > 225

10. Activities of the Research Committee on Reactor Physics ' Y. Kaneko et al. > 228

Publication List 231 Author Index 238 Appendix I Department of Reactor Engineering

Organization Chart 242 Appendix II Abbreviations 243

]AERI-M 88・221

8.11 General Forrnulation of a Nanipulator Linkage Nechanisrn

< S. Sasaki > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 211

8.12 Deve10prnent of Robotic Remote Handling Tecllnology

< H. Usui et a1. > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 212

9. Facility Operation and Technique Developrnent 、M. Nnkano > ・・・・ 214

9.1 Operation Report of FCA < K. Satoh et al. ;> ・・・・・・・・・・・・・・・・ 215

9.2 Operation Report o[ FNS < J. IくlIsuno ~~t al. > .............・・ 217

9.3 Tritiurn Quantificution byト1easurernentof Characteristic

X-Ray く C ‘ Kutsukake et 01. > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 219

9.4 An Irnprovement of Beam Transrnission Performance in FNS

Accelerator ど J. Kusano et al. > ........................句・・ 222

9.5 Opera tion Report of VHTRC < S. Fuj isaki et a 1. > ・・・・・・・・・・・ 224

9.6 Development of VHTRC Operating Data ProcessiDg System

< S. Fuj isaki et a1. > •••••••••••••.••••.•••..••..•.•••.•.• 225

10. Activities of the Research Committee ，)11 Reactor Physics

Y. Kaneko et 31. > ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 228

Publication List .................................................・ 231

Author Index .............................................・・・・・・・・・ 238

Appendix 1 Departrnent of Reactur Engineering

Organization Chart ・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・ 242

Appendix 11 Abbreviations .•••..••....•••..•••..••.•••••..•••. ...243

JAERI-M 88-221

Foreword

The annual research activities of the Department of Reactor Engineering, Japan Atomic Energy Research Institute, during the period of April 1987 - March 1988, i.e. fiscal year 1987 in Japan are presented in this report. The research activities of the Department has extended over a broad area including reactor physics, fusion reactor physics, shielding and reactor instrumentation and control. At the start of the present period, the design stuay works were newly added for the assessment of the advanced reactor systems.

The total number of people working in the Department at the end of the period was 117 of which 104 were regular members. JAERI funded expenditures during the period amounted to about 910 million yen, excluding nuclear fuel costs and personnel expenses.

In addition, a considerable amount of funding was provided by research contracts from outside organizations; Science and Technology Agency (STA) for reactor decommissioning technology and non-destructive measurement technology of transuranic elements, and Power Reactor and Nuclear Fuel Development Corporation (PNC) for fast reactor physics, shielding and instrumentation.

The research activities were conducted in 7 laboratories and one newly organized group:

Reactor System Laboratory, Fast Reactor Physics Laboratory, Thermal Reactor Physics Laboratory, Reactor Instrumentation Laboratory, Reactor Control Laboratory, Shielding Laboratory, Fusion Reactor Physics Laboratory, and Advanced Reactor Assessment Team

with the suport of Reactor Physics Facility Operation Division and the Research Committee on Reactor Physics.

The major research and development projects which the research programs in the Department are related to closely are:

(1) Development of very high temperature gas-cooled reactors (VHTR),

(2) Engineering research for fusion reactors, (3) Development of high conversion light water reactors,

JAERI-M 88・221

Foreword

The annual research activities of the Department of Reactor

Engineering， Japan Atomic Energy Research Institute， during the period

of April 1987 -~~rch 1988， i.e. fiscal year 1987 in Japan are presented

1n this report. The research activities of the Department has eKtended

over a broad area including reactor physics， fusion reactor physics，

shielding and reactor instrumentation and control. At the start of the

present period， the design stuay works were newly added for the assess-

ment 0f the aevanced reactor systems.

Thεtotal number of people working in the Department at the end of

the period was 117 of which 104 were regular members. JAERI funded

expenditures during the period amounted to about 910 million yen，

excluding nuclear fuel costs and personnel expenses.

In addition， a considerable amount of funding was provided by

research cOlltracts from outside organizations; Science and Technology

Agency (STA) for reactor decommissioning technology and non-destructive

measurement techllology of transuranic elements， and ?ower Reactor and

Nuclear Fuel Development Corporation (PNC) for fast reactor physics，

shielding and instrumentation.

The research activities were conducted in 7 laboratories and one

newly organized group:

Reactor System Laboratory，

Fast Reactor Physics Laboratory，

Thermal Reactor Physics Laboratory，

Reactor Instrumentation Laboratory，

ReactつrControl Laboratory，

Shielding Laboratory，

Fusion Reactor Physics Laboratory， and

Advanced Reactor Assessment Team

!vith the suport of Reac tor Physics Facility Opera tion Division and

the Research Committee on Reactor Physics.

The major research and development projects which the research

programs in the Department are related to closely are:

(1) Development of very high temperature gas-cooled reactors

(VHTR) ，

(2) Engineering research for fusion reactors.

(3) Development of high conversion light water reactors.

J A E R I - M 8 8 - 2 2 1

(4) Assessment of the advanced reactor systems. Efforts continued in the field of reactor physics, reactor

instrumentation and reactor control of VHTR R&D. Critical experiment was successfully proceeded on the Very High Temperature Reactor Critical Assembly with 2 - 6% enriched uranium fuel elements.

The criticality predictions by the SRAC code agreed very closely with the experimental results. Irradiation tests of in-core chambers on the Fort Saint Vrain Reactor were planned under a joint research program between US-DOE and JAERI. On the other hand, performance tests of high temperature fission counters on AVR started under a cooperative program between KFA and JAERI. In the study of reactor control, characteristics on load-following VHTR was investigated.

Concerning the fusion reactor physics, the second stage FNS experiments on the lithium oxide system of closed geometry was performed under a collaborative research program between US-DOE and JAERI. The program has provided important experimental data for validating calculation methods of fusion blanket design. Absolute cross section measurements for the structural materials of the fusion reactors have been conducted. Benchmark test of the JENDL-3T nuclear data library has also been extensively done.

Addressing R and D of LMFBR, measurements of the Doppller effect in the high temperature range have been planned using the core oscillation technique. The correlation analysis of the supersonic waves due to the sodium coolant boiling has been done under a IAEA coordinated research program.

The research activity on a high conversion light water reactor (HCLWR) has been extended with purpose to investigate a possibility and to establish the reactor concept as an advanced light water reactor for the next generation. A basic idea for a flat core has been proposed, in which the higher values can be achieved for both the conversion ratio and fuel burn-up under the condition of the negative void coefficient.

In the R and D project for decommissioning of nuclear reactor, most of developments are approaching their final stages. The mock-up test of the remote-controlled robotic manipulator system has been performed for cutting off the parts of reactor core internals using a plasma torch.

]AERI-M 88・221

(4) Assessmer.t of t.he advanced reactor systems.

Efforts continued in the fie1d of reactor physics， reactor

instrumentation and reactor control of VHTR R&D. Critical experiment

wa~ successfu11y proceeded on the Very High Temperature Reactor Critical

Assembly with 2 -6% enriched uranium f~e1 elements.

The crit1cality predlctions by the SRAC code agreed very closely

with the experimenta1 results. Irradlatlon tests of ln-core chambers

on the Fort Salnt Vrain Reactor were planned under a joint research

program bet~ýeen US-DOE and JAERI. On the other hand， performance tests

of high temperature fission counters on AVR started under a cooperative

program between KFA and JAERI. In the study of reactor control，

characteristics on load-following VHTR t17as investigated.

Concerning the fusionreactor physics， the second stage FNS exper-

iment己 onthe lithium oxide system of closed geometry was performed

under a collaborative research program between US-DOE and JAERI. The

program has provided lmportant experimental data for validating calcu-

lation methods of fusion b1anket design. Abs01ute cross section

measurements for the structural materials of the fusion reactors have

been conducted. Benchmark test of the JENDL-3T nuc1ear datD. 11brary

has a1so been extensively done.

Addressing R and D of LMFBR， measurements of the Doppller effect in

the high temperature range have been planned using the core oscillation

technique. The corre1ation analysis of the supersonic waves due to

the sodium coolant boiling has heen done under a IAEA coordinated

research program.

The research activity on a high conversion 1ight water reactor

(HCLWR) has been extended with purpose to investigate a possibility and

to estab1ish the reactor concept as an advanced light water reactor for

the next generation. A basic idea for a flat core has been proposed，

in which the higher values can be achieved for both the conversion

ratio and fue1 burn-up under the condition of the negative void

coefficient.

In the R and D project for decommissioning of nuclear reactor，

most of deve10pments are approaching their final stages. The mock-up

test of the remotc-contro11ed robotic manip~1ator system has been

performed for cutting off thc parts of reactor core internals using a

plasma torch.

J A E R I - M 8 8 - 2 2 1

In the design studies of the advanced reactor system, some basic ideas have been concieved. A concept of the system integrated pressurized water reactor (SPWR) is proposed as a medium size power reactor with highly passive safety features. Comprehensive neutronics calculations have been made aiming at very high fuel burn-up for MOX fuel in LWR. A metallic fuel fast reactor has been considered as a burner of transuranic radioactive wastes. Several other significant achievements have been made. These include the development of a nondestructive measuring technique of transuranic elements in waste drums, a transmutation study of transuranic wastes by using spallation reaction, analysis of the Chernobyl reactor accident, completion of the DOIC code system for calculation of the inventory of radio-active materials, evaluation study of radiation shielding calculation accuracy through participation in the NEACRP benchmark, development of a generalized group cross section library system and the standardized intelligent shielding analysis code package.

Yoshihiko Kaneko, Director Department of Reactor Engineering

]AERI-M 88-221

In the design studies of the advanced reactor system， some basic

ideas have been concieved. A concept of the system integrated pres-

surized wate~ reactor (SPWR) i5 proposed as a medium size power

reactor with highly passive safety features，' Comprehensive neutronics

calculations have been made aiming at very high fuel burn-up for MOX

fuel in LWR. A metallic fuel fast reactor has been considered as a

burner of transuranicτadioactive wastes. Several other significant

e¥chievements have been made. These inclucle the develCJprnent of a non-

destructive measuring technique of transuranic elements in waste drums，

a tran丹市u':ationstudy of transurar.ic wastes by using spallation

reaction， analysis 0ど theChernobyl reactor accident， completion of

the DOIC code system for calculation of the inventory of radio-active

materials， evaJ.uation study or radiation shielding calculatioll

accuracy through participation in the NEACRP benchrnark， development

of a generalized group cross section library system and the standardiz-

ed intelligent shielding analysi5 code package.

Yoshihiko Kaneko， Director

JAERI-M 88-221

1. Nuc lea r Data and Group Cons tan t s

- Benchmark T e s t s of JENDL-3T -

The c o m p i l a t i o n of the Japanese Eva lua ted N u c l e a r Data L i b r a r y ,

v e r s i o n 3 (JENDL-3) is now in p rog re s s a i m i n g the f i n a l c o m p i l a t i o n

in 1988. A t e m p o r a r y f i l e JENDL-3T which c o n t a i n s da t a for some

p r i m a r i l y i m p o r t a n t n u c l i d e s was prepared to a s s e s s the adequacy of

JENDL-3 d a t a for use in n u c l e a r r e a c t o r d e s i g n and a p p l i c a t i o n s .

V a r i o u s benchmark t e s t s of JENDL-3T have been pe r fo rmed for t h e r m a l

r e a c t o r , h igh c o n v e r s i o n l i g h t w a t e r r e a c t o r , f a s t r e a c t o r , s h i e l d i n g

and fus ion n e u t r o n i c s .

The r e s u l t s of benchmark c a l c u l a t i o n s us ing JENDL-3T a r e s u m

m a r i z e d a s f o l l o w s ; In the t h e r m a l and high conve r s ion l i g h t w a t e r

r e a c t o r s , the £ e / / s o b t a i n e d w i t h the JENDL-3T d a t a g ive b e t t e r a g r e e

ment w i t h the e x p e r i m e n t s than those ob t a ined by JENTJL-2. In the f a s t

r e a c t o r benchmarks , the fee// for JENDL-3T a re o v e r e s t i m a t e d for u r a n i u m

fuel co res and u n d e r p r e d i c t e d for p l u t o n i u m fuel c o r e s . The r e a c t i o n

r a t e r a t i o s of U-238 c a p t u r e to Pu-239 f i s s i o n and U-238 f i s s i o n to

Pu-239 f i s s i o n a re c o n s i d e r a b l y o v e r e s t i m a t e d . On the o the r hand.

Doppler and soc ium void r e a c t i v i t i e s , and r e a c t i o n r a t e d i s t r i b u t i o n

of Pu-239 f i s s i o n a r e s i g n i f i c a n t l y improved by us ing the JENDL-3T

d a t a . In the s h i e l d i n g benchmark t e s t s , i t i s shown tha t the t o t a l

c r o s s s e c t i o n s for 0 , N, Na and Fe have been found to be improved from

a n a l y s i s of BROOMSTICK e x p e r i m e n t . The i ron 24 KeV resonance d a t a

of JENDL-3T a r e b e t t e r than the ENDF/B-IV d a t a , and i n e l a s t i c s c a t

t e r i n g c r o s s s e c t i o n s is r a t h e r too h igh in the energy range from 2.0

to 5.0 MeV. In the fus ion n e u t r o n i c s c a l c u l a t i o n s , the t r i t i u m p r o

d u c t i o n r a t e s c a l c u l a t e d for a t y p i c a l fus ion b l a n k e t w i t h o u t Be and

Pb a r e in good ag reemen t w i t h the e x p e r i m e n t s , but r e e v a l u a t i o n work

for Be and Pb n u c i e a r d a t a i s recommended.

The d a t a in JENDL-3T could be r e v i s e d p a r t l y on the b a s i s of the

p r e s e n t benchmark t e s t i n g r e s u l t s , and t h i s w i l l be r e f l e c t e d to the

f i n a l v e r s i o n of JENDL-3.

( Hidekf TAKANO)

jAER I・M 88・221

1. Nuclear Data and Group Constants

-Benchmark Tests of JENDL-3Tー

The compi lat ion of the Japanese Evaluat自d Nuclear Data Library.

version 3 (JENDL-3) is now in progress aiming the final compilation

in 1988. A temporary fi le JENDL-3T which contains data for 品om自

primari ly important nucl ides was prepared to assess the adequacy of

JENDL-3 da ta for us自 in nuclear reactor design and applications.

Var ious benchmark t自sts of J&'I'DL-3T have been performed for therm且 l

reactor. high conversion 1 ight water reactor， fast reactor. shielding

and fusion n自utronics.

The resul ts of benchma北 calculations using J巳:-IDL-3T are sum-

marized as folln'."s; ln the thermal and high conversion 1 ight wa.ter

reactors. the k.1fS obtained with the JENDL-3T data give better agre自-

ment wi th the exper iments than those obtained by JENDL-2. In the Cas t

reactor benchmarks， the k.1I for JENDL-3T are overes t imated Cor uranium

fue I cores 且nd underpredicted Cor plutonium fuel cores. The reaction

rate rati05 of U-238 capture to Pu-239 fission and U-238 fission to

Pu-239 fission are considerably overestimated. On the other hand.

Doppler and sor:ium void reactivities. and reaction rate distribution

of Pu-239 fission are significantly improved by using the JENDL-3T

data. ln the shielding benchmark tests. i t is shown that the total

cross sections for 0，うi. ~a and Fe have been found to be improved from

analys is of BROOMSTICK exper iment. The i ron 24 KeV resonance data

of JENDL-3T are better than the ENDF/B-IV data. 且nd inelastic scat-

tering cross sections is rather too high in the energy range from 2.0

to 5.0 MeV. ln the fusion neutronics calculations. the tritium pro-

duction rates calculated for a typical fusion blanket without Be and

Pb are in good agreement with the experiments. but reevaluation work

for Be and Pb nuciear data is recommended.

The data in JENDL-3T could be revised part!y on the basis of the

present benchmark testing results. and this will be refl自cted to the

final version of JE~DL-3.

( Hideki TAKANO)

JAERI-M 88-221

1.1 Benchmark T e s t of JENDL-3T for Thermal Reac tor and

High Convers ion L igh t Water Reac to r

H. Takano and K. Kaneko'

A t e m p o r a r y n u c l e a r d a t a f i l e JENDL-3T 1 ' has been g e n e r a t e d for t e s t i n g an e v a l u a t e d d a t a f i l e of JENDL-3. To a s s e s s the adequacy of JENDL-3T d a t a for use in n u c l e a r r e a c t o r d e s i g n and a p p l i c a t i o n s , ben chmark c a l c u l a t i o n s have been pe r fo rmed for t h e r m a l and h igh conve r s ion l i g h t w a t e r r e a c t o r s ( H C L W R ) .

The s e l e c t e d benchmark cores a re a number of c r i t i c a l e x p e r i m e n t s w i t h d i f f e r e n t f u e l s of U-235, U-233 and P u - 2 3 9 3 ' ' 3 ' , two w a t e r - m o d e r a t e d l a t t i c e (TRX- l and 2 ) 2 ' , two heavy w a t e r - m o d e r a t e d c o r e s (ETA-1 and 2)*' and a l a rge number of u n i f o r m w a t e r - m o d e r a t e d l a t t i c e s c o l l e c t e d by S t r a w b r i d g e and B a r r y 5 ' . The PROTEUS cores ' ' ' a r e s e l e c t e d for the HCLWR benchmarks .

Benchmark e x p e r i m e n t s were ana lysed w i t h the SRAC code s y s t e m " u s i n g two c r o s s s e c t i o n l i b r a r i e s SRACLIB-JENDL2 and -JENDL3T based on JENDL-2 and JENDL-3T d a t a , r e s p e c t i v e l y . These l i b r a r i e s c o n t a i n 74-group c o n s t a n t s for f a s t energy r e g i o n and 48-group c o n s t a n t s for t h e r m a l energy r e g i o n . In resonance energy r e g i o n , a u l t r a - f i n e group l i b r a r y was p r e p a r e d for some i m p o r t a n t heavy r e s o n a n t n u c l i d e s . C e l l s p e c t r u m c a l c u l a t i o n s were pe r fo rmed by the c o l l i s i o n p r o b a b i l i t y m e t h o d . C r i t i c a l i t y c a l c u l a t i o n s were pe r fo rmed w i t h P i - S j a p p r o x i m a t i o n by a o n e - d i m e n s i o n a l S „ - t r a n s p o r t code ANISN.

F i g u r e 1.1.1 shows the k,// ob t a ined for ORNL and McNeany-Jenk ins c o r e s as a f u n c t i o n of the a t o m i c r a t i o of H / U - 2 3 5 . The r e s u l t s unde r e s t i m a t e d w i t h i n c r e a s e of H/U-23S r a t i o . The kt/f c a l c u l a t e d for 116 c a s e s of S t r a w b r i d g e and Bar ry were as f o l l o w s : The averaged k,// o b t a i n e d for U 0 2 - r o d l a t t i c e cases was 0.991 for JENDL-3T and 0.983 for JENDL-2. The averaged k,,f for U - m e t a l rods a r e 0.992 for JENDL-3T and 0 .989 for JENDL-2. I t is the main reason for these d i f f e r e n c e s t ha t the i»-value of U-235 for JENDL-3T is 0.24 X l a r g e r than JENDL-2 d a t a a t the 2200 m / s e c as seen in Table 1 . 1 . 1 . I n t e g r a l l a t t i c e p a r a m e t e r s were c a l c u l a t e d for the TRX-1 and 2 c o r e s , and Pu and Sag Tor JENDL-3T were o v e r e s t i m a t e d by 6 - 9 % as shown in Table 1 . 1 . 2 .

F i g u r e 1.1.2 shows the m u l t i p l i c a t i o n f a c t o r s c a l c u l a t e d for McNeany-Jenk ins co res a s a func t ion of the a t o m i c number r a t i o H/LT-233. The r e s u l t s o b t a i n e d by JENDL-3T d a t a become about 1.0 X s m a l l e r than those by JENDL-2 d a t a and g ive good ag reemen t w i t h e x p e r i m e n t s .

The i n t e g r a l p a r a m e t e r s floi, CR, CR" and S02 c a l c u l a t e d for the ETA-1 and 2 co re s u s i n g JENDL-3T da t a were s i g n i f i c a n t l y improved over c o r r e s p o n d i n g a n a l y s e s u s i n g JENDL-2 d a t a 9 ' . Th i s i s because t a h t Th-232 c a p t u r e c r o s s s e c t i o n s of JENDL-3T a r e s i g n i f i c a n t l y l a r g e r than JENDL-2 d a t a in the resonance r e g i o n below 200 eV.

The c a l c u l a t e d ktljs a r e shown as a func t ion of the a t o m i c r a t i o of H/Pu-239 in F i g . 1 . 1 . 3 . The ke// o b t a i n e d u s i n g JENDL-3T d a t a a r e about 0.6 % s m a l l e r than those us ing JENDL-2 d a t a , and the o v e r p r e -d i c t i o n by JENDL-2 i s improved .

The r e s u l t s c a l c u l a t e d for the PROTEUS cores 1 - 3 a r e shown as a f u n c t i o n of the coo lan t void f r a c t i o n (X) in F i g s . 1 . 1 . 4 and 5. The km u s i n g JENDL-3T g ives very good ag reemen t w i t h e x p e r i m e n t s in the zero void s t a t e as seen in F i g . 1 . 1 . 4 . However , the km of JENDL-3T depend s t r o n g e r than tha t of JENDL-2 on coo l an t voidage s t a t e s . F i g u r e 1.1.5 shows the compar i son of r e a c t i o n r a t e r a t i o of U-238 c a p t u r e to Pu-239 f i s s i o n (C8 /F9) . The r e s u l t s u s i n g JENDL-3T a r e about 2 X l a r g e r than those us ing JENDL-2.

« Japan I n f o r m a t i o n S e r v i c e , C o . , L t d . , Tokyo

jAERI-M 88・221

1.1 Benchmark Test of JENDL-3T for Thermal Reactor and

High Conversion Light Water Reactor

H. Takano and K. Kaneko'

A t emporary nuc 1 ear da ta f i 1 e JENDL-3T1) has be en g自nerated Cor testing an evaluated data file of JENDL-3. To assess the ad自quacyoC JEl~DL-3T data for use in nuclear re且ctord自signand applications. ben-chmark calculations have been performed for thermal and high conversion 1 ight wa ter re且ctors (HCLWR).

The selected benchmark cores ar自 anumber of cr i t i cal exper iments with different fuels of U-235， U-233 and Pu-23921.幻 two'.lVa ter吋 noderated lattice(TRX-l and 2)幻 two heavy wat自r-moderatedcores (ETA-l and 2)') and a larg自 number c.f uni form water-moderated lattices collected by Strawbridge and Barry5). The PROTEUS cores川

are selected for the HCLWR benchmarks. Benchmark 自xperiments wer自 analysedwi th the SRAC code system7

using two cross sec'ion libraries SRACLIB-JENDL2 and -JENDL3T based on JENDL-2 and JENDL-3T data. respectively. These libraries contain 74-group constants for Cast energy region and 48-group constants Cor thermal energy region. In resonance energy region. a ultra-Cine group library was prepared for some important heavy resonant nuclides. Cell spectrum calculations were performed by the collision probability method. Criticality calculations were perCormed with P1-Ss approxi-mation by a one-dimensional SII-transport code ANISN.

Figure 1.1.1 shows the k." obtained Cor ORNL and McNeany-Jenkins cores as a function of the atomic ratio of H/U-235.τ'he resul ts und-erest imated wi th increase oC H/U-235 rat io・TheIr.." calculat自d Cor 116 cases of Strawbridge and Barry were as follows: The av自rag自dIr.." obtained for UOz-rod lattice cases was 0.991 for JENDL-3T and 0.983 for JENDL-2. The averaged k.ff for U-me tal rods are 0.992 for J&'ffiL-3T and 0.989 for JENDL-2. It i5 the main rea50n for these differences that the II-value of U-235 for JEND:L-3T i5 0.24 " larger than JENDL-2 data at the 2200 m/sec a5 seen in Table 1.1.1. Integral lattice parameters were calculated for the TRX-l and 2 cor自5.and Pzs and 828 [or JENDL-3T were overes t ima ted by 6 -9 ~ as shown in Table 1. 1. 2.

Figure 1.1.2 shows the multiplication factors calculated for McNeany-Jenkins cores as a function of the atomic number ratio H/U-233. The results obtained by J&'ffiL-3T data become about 1.0 " smaller than those by JENDL-2 data and give good agreement with experlments.

The integral parameters 内 2. CR. CR' and 802 calculated for the ETA-l and 2 cores using JEJfiDL-3T data were significant

JAER1-M 88-221

The benchmark c a l c u l a t i o n r e s u l t s us ing JENDL-3T a re summar i zed as f o l l o w s : In t h e r m a l r e a c t o r benchmark t e s t s , the ke//s ob ta ined w i t h the JENDL-3T d a t a gave b e t t e r ag reemen t w i t h the e x p e r i m e n t s than those o b t a i n e d by JENDL-2, though i t was observed t ha t they depend c o n s i d e r a b l y on the a t o m i c r a t i o of H/L".

R e f e r e n c e s 1) JENDL C o m p i l a t i o n Group(Nuc 1 ear Data C e n t e r , JAERI ) : JENDL-3T,

P r i v a t e c o m m u n i c a t i o n (1987 ) . 2) "Cross S e c t i o n E v a l u a t i o n Working Group Benchmark

Speci f i c a t i o n s , "ENDF-202(BNL-19302) .Brookhaven N a t i o n a l Lab. 3) V: t feany S. and J e n k i n s D. : Nuc l . S c i . Eng . , 65, 441 (1978) . 4) Hardy J . et a l . : N u c l . S c i . Eng . , 55, 401 (1974 ) . 5) S t r a w b r i d g e L.E. and Bar ry R . F . : N u c l . S c i . Eng . , 23, 58 (1965) . 6) Chawla R. et a l . : N u c l . T e c h n o l . , 67, 360(1984) . 7) T s u c h i h a s h i K. et a l . : JAERI 1285 (1983) and 1302 (1987 ) . 8) Takano H. et a l . : JAERI-M 82-072(1982) .

Tab le 1 .1 .1 Compar ison of v va lues (2200m/s )

n u c l i d e JENDL-2 JENDL-3T ENDF/B-IV ENDF/B-V U-235 2.4286 2.4345 2.4188 2.4367 Pu-239 2.8806 2.8843 2.8706 U-233 2.4930 2.4930 2.4980

F i g . 1 . 1 . 1 *•"' for J M U fuet cores <ANISN,P,S,)

- l . O l

0.98 E

-1 t~n—TT

JENOL-2 JENDL-3T

10 J I0> 'H/"'Pu

F i g . 1 . 1 . 3 k * " f o r " * P u f v J e l c o r e a <ANISN. PiS,)

F i g . 1 . 1 . 2 k >" f o r J M u f u e L core* (ANISN. P|S t)

1.01 .

i" 1.00

—9-— exp. — * — JENDL-2 - ~ t - - JENOL-3 T

^ N ^ V ,

_i ' ' 20 *0 60 SO 100

Void (X)

Fig. 1. 1 • 4 Companion of k. for PROTEUS corei

]AERI-M 88-221

The benchmark calculat ion resul ts using JENDL-3T are summarized as follows: In thermal reactor benchmark tηsts， the keffs obtained with the JE:-;DL-3T data gave better agreement wi th the experiments than those obtained by JE~DL-2 ， though i t was observed that they depend considerably on the atomic ratio of H/じ目

References 1) JE..'¥TDL Compilation Group(Nuclear Data. Center， JAERI): JENDL-3T，

Private communication (1987). 2) "Cross Sec t ion Evaluat ion Working Group 8enchmark

Spe c i f i ca t i ons ， '・ENDF-202(8NL-19302) ，8rookhaven Nationa.l Lab. 3)ν:Neany S. and J自nkinsD. Nucl. Sci. Eng.， 65， -141 (1978). 4) Hardy J. et al.: Nuc1. Sci. Eng.. 55，401 (1974). 5) Strawbridge L.E. and 8arry R.F. トfuc1. Sci， Eng.. 23. 58 (1965). 6) Chawla R. 日tal.:Nucl. Technol.. 67. 360(1984). 7) Tsuchihashi K. et a1.: JAERI 1285 (1983) and 1302 (1987). 8) Takano H. e t且1.: JAERI-M 82-072(1982).

Table 1.1.1 Comparison of 11 values (2200m/s)

nuclide U-235 Pu-239 U-233

JENDL-2 2.4286 2.8806 2.4930

i・ 111 守一一二:':'~-.hζIF7!?-L 一一一…「一 3

;jjJJJ:J 096L'l ; a 4 2 2 j i

Fig.1. 1. 1 k.押 for2判Jfuel c:ores (ANISN. P.S.l

、1.01z

--，・・・・・・・・・・・・・・・・--

~\ ・』ごたごこ---・ブ1ー「一-ー・・・・・・・・------

・4位、叱'; '1 ・一。ー JENDL-2

一一一一一一引い…ー… JEドlDL・3T.

， .. -------------------~--・・・・・_.・・・・・ー・ー・・ー・・

10' 'H/2lOpu

F i g . 1 . 1 . 3 k...，向rUtpu fuel c:ores (ANI5N. P.S.l

JENDL-3T ENDF/B-IV ENDF/B-V 2.4345 2.4188 2.4367 2.8843 2.8706 2.4930 2.4980

0.96 10'

'H/2>>tJ 10' 。

F i g . 1 . 1 . 2 k... for "'u fue l c:ores CAN 1 SN. p.S.】

一+ー EXP.ー-ー-JENOL・2・......・ JENDL・3T

ぞ1.00u 四一一一一、

T111111AVl--tEi

嗣個個闘

、+0.98

o 20 40 60 80 100 Vald (揮3

Fig.1.1.4 c。帽.rlaan af k. far PROTEUS c:are.

J A E R l - M 88-221

J 1.02

1 1 ' • 1 1 1 1

. — o — EXP. — 4 — JEN0L-2 - - < - - JENDL-3T

* ^*— •* _^f^

" • ^^ ^ ** ' ^ ^

t i i i i i

20 40 60 80 100 Void (X)

F i g . 1 . 1 . 5 U,"<n.r),'Pu",Cn. f) for PROTEUS cores

T a b l e 1.1.2 The C /E-va lue s for l a t t i c e c e l l p a r a m e t e r s

a s s e m b l y p a r a m e t e r

C" 028 $25 "^28 C" 002 S25 Soz CR 002 $23

J END L-2 JENDL-3T 1.057 1.081 1.022 1.012 1.052 1.095 1.02 1 1.031 1.040 1.063 1.007 0.997 1.824 1.058 1.011 1.015 0.977 1.011 1.064 1.060 0.810 0.846 0.919 0.974 0.958 0.982 1.055 1.054 0.944 1.007 0.895 0.938

028 825 $28 C" 002 $32 CR 823 $02 CR"

U-238 c a p t u r e epi t h e r m a l to t h e r m a l U-235 f i s s i o n epi t h e r m a l to t h e r m a l U-238 f i s s i o n to U-235 f i s s i o n U-238 c a p t u r e to U-235 f i s s i o n Th-232 c a p t u r e e p i t h e r m a l to t h e r m a l Th-232 f i s s i o n to U-235 f i s s i o n Th-232 c a p t u r e to U-235 f i s s i o n U-233 f i s s i o n e o i t h e r m a l to t h e r m a l Th-232 f i s s i o n .0 U-233 f i s s i o n Th-232 c a p t u r e to U-233 f i s s i o n

]AERI-M 88・221

1. 08 +

，，

a' ，，

1. 06 "“+ー-EXP. -ー-JENOL.2 -ー』・ JENOL・3T

ぞ1.02u

0.96 o 20 ‘o 60 80 100

Vo Id (潟】

Fig .1.1.5 U，so(n，1'l/Pu"'(n.ヂ1for PROTEUS cores

Table 1.1.2 The C/E-values for lattice cell parameters

assembJy parameter ρ26

TRX-l 825 826 C' ρ26

TRX-2 825

"26 c"

P02 ETA-l 825

802 CR

。，uA

JENDL-2 1.057 1. 022 1. 052 1，02.1 1. 040 1. 007 1.日241.011 0.977 1. 064 0.810 1).919 0.958 1.055 0.944 0.895

JENDL-3T 1.081 1. 012 1. 095 1.031 1.日630.997 1.058 1. 015 1.011 1. 060 0.846 0.974 0.982 1. 054 1. 007 0.938

P26 U-238 capture epi thermal to thermal 825 U-235 f i ss i on ep i therma I to therma I 828 U-238 fission to U-235 fission C' U-238 capture to V-235 fission ρ02 Th-232 capture epi thermal to thermal 8n Th-232 fission to U-235 fission CR Th-232 capture to V-235 fission 823 U-233 f i ss i 00 (>ηi thcrma I to therma I 8O2 Th-232 fission ;0 U-233 fission CR' Th-232 capture to U-233 fission

JAERI-M 88-221

1.2 Benchmark Tes t of JENDL-3T on Fas t Reac to r s

H. Takano and K. Kaneko"

Fas t r e a c t o r benchmark c a l c u l a t i o n s have been per formed to a s s e s s the adequacy of a t empora ry nuc l ea r da t a f i l e JEXDL-3T 1 1 for use in n u c l e a r r e a c t o r d e s i g n and a p p l i c a t i o n s .

Benchmark co re s c o n s i s t of 17 benchmark a s s e m b l i e s c o l l e c t e d for ENDF/B-IV d a t a t e s t i n g 2 ' , the JOYO and MONJU mock-up co res (FCA-V-2 and F C A - V I - 2 ) , MORZART cores(MZA and MZB) and the JUPITER r e f e r e n c e core (ZPPR-9) . These have a wide v a r i e t y from 12 to 1600 l i t e r of core s i z e s , from zero to e igh t c o n c e n t r a t i o n r a t i o s of f e r t i l e to f i s s i l e in c o r e , and c o n s i s t of 15 p l u t o n i u m and 7 uran ium fuel cores as shown in Table 1 . 2 . 1 .

The c a l c u l a t i o n s for kr/f and c e n t r a l r e a c t i o n r a t e r a t i o s were pe r fo rmed w i t h o n e - d i m e n s i o n a l d i f f u s i o n and t r a n s p o r t codes . R e a c t i o n r a t e d i s t r i b u t i o n s , Doppler and Na-void r e a c t i v i t i e s for ZPPR-9 and FCA-VI-2 co re s were c a l c u l a t e d by t w o - d i m e n s i o n a l d i f f u s i o n t h e o r y . For these c a l c u l a t i o n s , a JFS3-JENDL3T l i b r a r y w i t h 70-group s t r u c t u r e was produced by the TIMS-PGG process ing 1 c o d e . 3 1

The kt[f o b t a i n e d w i t h the JENDL-3T da t a a re o v e r e s t i m a t e d for u r a n i u m cores and a r e u n d e r e s t i m a t e d for p l u t o n i u m cores as shown in F i g . 1 . 2 . 1 . The o v e r e s t i m a t e for u ran ium cores is due to a l a rge u ( U - 2 3 5 ) - v a l u e e v a l u a t e d for JENDL-3T, and the u n d e r e s t i m a t e for p l u ton ium co re s i s due to s m a l l e r Pu-239 f i s s i o n c ro s s s e c t i o n s as shown in F i g . 1 . 2 . 2 . The JENDL-3T da t a a re 5 % s m a l l e r than the JEN'DL-2 in the energy range from 10 KeV to 1 MeV. This causes 1.7 % r e d u c t i o n of k,fj for P u - c o r e s and about 4 % i nc rease for c e n t r a ! r e a c t i o n r a t e r a t i o of U-238 c a p t u r e to Pu-239 f i s s i o n ( C 8 / F 9 ) .

The c e n t r a l r e a c t i o n r a t e r a t i o C8/F9 or C8/F5 ob ta ined by JENDL-3T a r e l a r g e r than those for JENDL-2. The r e s u l t s ob t a ined for C8/F9 a r e compared in F i g . 1 . 2 . 3 .

F i g u r e 1.2.4 shows the d e v i a t i o n for U-238 c a p t u r e c r o s s s e c t i o n s of JENDL-3T from JENTJL-2 d a t a . There is c o n s i d e r a b l e d i s c r e p a n c y in the energy range above 100 KeV. This causes 0.3 - 0 .8 r e d u c t i o n of fee// and 3 % i n c r e a s e of C8/F9 .

In JENDL-3T, f i s s i o n s p e c t r u m was ev a lu a t ed on the b a s i s of M a d l a n d - N i x f o r m u l a , and i t i s ha rde r than tha t of JENDL-2. The k,ff

c a l c u l a t e d w i t h t h i s ha rde r s p e c t r u m becomes 0.5 % l a r g e r than the r e s u l t of JENTJL-2, and the t h r e s h o l d r e a c t i o n r a t e r a t i o F8/F5 a l s o becomes l a r g e r by 5 %.

The F9/F5 c a l c u l a t e d by JENDL-3T improves the u n d e r p r e d i c t i on observed for the r e s u l t s o b t a i n e d w i t h JENDL-2. This may be p a r t l y because the f i s s i o n c r o s s s e c t i o n s for Pu-239 and U-235 were e v a l u a t e d on the b a s i s of a s i m u l t a n e o u s e v a l u a t i o n method .

T w o - d i m e n s i o n a l benchmark c a l c u l a t i o n s were pe r fo rmed for ZPPR-9 and FCA-VI-2 a s s e m b l i e s to a s s e s s Doppler r e a c t i v i t y , sodium r e a c t i v i t y w o r t h and r e a c t i o n r a t e d i s t r i b u t i o n . The o v e r e s t i m a t i o n for sod ium void w o r t h o b t a i n e d by JENDL-2 is r e m a r k a b l y improved by JENDL-3T as seen in F i g . 1 . 2 . 5 . The \ a t U 0 2 Doppler w o r t h c a l c u l a t e d w i t h JENDL-3T i n c r e a s e d by about 6 % in the compar i son w i t h those for JENDL-2 and is in good ag reemen t w i t h the e x p e r i m e n t s as shown in Table 1 .2 .2 . The r e a c t i o n r a t e d i s t r i b u t i o n in the ou te r core r eg ion is a l s o improved by about 1.0 % as observed from F i g . 1 . 2 . 6 .

The benchmark c a l c u l a t i o n r e s u l t s us ing JENDL-3T a r e s u m m a r i z e d as f o l l o w s : In f a s t r e a c t o r benchmarks , the k,n c a l c u l a t e d w i t h the JENDL-3T d a t a was o v e r e s t i m a t e d for U-cores and underpred i c ted for P u - c o r e s . The r e a c t i o n r a t e r a t i o s of C8/F9 and F8/F9 were o v e r e s t i -

» Japan I n f o r m a t i o n S e r v i c e , C o . , L t d . , Tokyo

]AERI-M 88・221

1.2 Benchmark Tesヒ ofJENDL-3τon Fast Reactors

H. Takano and K. Kuncko'

F且st reactor benchmark calculations have been performed to assess the adequacy of a temporary nuclear duta file ，IE:¥DL-3T1

) for use in

nuclear reactor design and appl ications. Benchmark cores consist of 17 benchmark ussemblies collected for

E~DF/B-IV data tes t ing.2). the JOYO and ~1O~J t.; mock-up corcs (上'C.¥-V-2and FCA-VI-2). :VIORZART cores(~IZA and Y1ZB) and the Jt:PITER ref自rencecore (ZPPR-9). These have a wide variety from 12 to .1600 lit自rof core sizes. from zero to eight concentrution ratios of fertile to fissil日

in core. and consist of 15 plutonium and 7 '-lranium ruel cores as shown in Table 1.2.1.

The calculations for k.1f and central reaction rate ratios were performed with one-dimensional diffusion and transport codes. Reaction rate dist.ributions. Doppler and Na-void reactivities for ZPPR-9 and FCA-VI-2 cores were calculated by two-dimensional diffusion theory. For these calculations. a JFS3-JENDL3T library with 70-group structure was produced by the TIMS-PGG process i ng code. 3)

The k.ff obtained wi th the JE~DL-3T data 且re overestimated for uranium cores and are underestimated for plutonium cores as shown in F i g . 1 . 2 . 1 . Th e 0 v e r e 5 t i ma t e f 0 r u r an i um c 0 r e 5 i s d u e t 0 a 1 a r g e v(U-235)-value ev且 luated for JENDL-3T. and the underestimate for plu-tonium cores is due to smaller Pu-239 fissioll cross sections as shown

in Fig.l.2.2. The JE~DL-3T dat且 are 5 % smaller than the JE~DL-2 in the energy range from 10 KeV to 1 MeV. This causes 1.7 % reduction of k.f! for Pu-cores and about 4 % increase for central reaction rate ratio of U-238 capture to Pu-239 fission (C8/F9).

The central reaction rate ratio C8/F9 or C8/F5 obtained by JE~DL-3T are larger than those for JE:--iDL-2. The resul ts obtair.ed for C8/F9 are compared in Fig.1.2.3.

Figure 1.2.4 shows the deviation for U-238 capture cross sections of JE~DL-3T from ~lE~、一DL-2 data. There is considerable discrepancy in the energy range abo¥'e 100 KeV. This causes 0.3 - 0.8 reduction of k.n and 3 % increase of C8/F9.

1n JE~DL-3T. fission spectrum was evaluated on the basis of Madland-Nix formula. and i t is harder than that of JENDL-2. The k." calculated with this harder spectrum becomes 0.5 % larger than the resul t of JEN百L-2. and the threshold reaction rate ratio F8/F5 also becomes larger by 5 %.

The F9/F5 calculated by JENDL-3T improves the underprediction observed for the results obtained with JENDL-2. This may b

on the basis of a simultaneous evaluation method. Two-d imens ional benchmark cal cula t ions were per formed for ZPPR-9

and FCA-V1-2 asser.1blies to assess Doppler react:vity. sodium reac-tivity worth and reaction rate distribution. The overestimation for sodium void wonh obtained by JENDL-2 is r自markably improv自d by JEXDL-3T as seen in Fig.l.2.5. The :'-l'atu02 Doppler worth calculated wi th JE:¥"'DL-3T increased by about 6 % in the compar ison wi ，h those for JEXDL-2 and is in good agreement with the experiments as shown in Table 1.2.2. The reaction rate distribution in the outer core region is also improved by about 1.0 % as observed fr口m Fig.l.2.6.

The benchmark calculation results using JE.¥"DL-3T are summarized as follows: 1n fast reactor benchmarks. the k町 J caicuiated wi th the JENDL-3T data was overestimated for (j-cores and underpredicted for Pu-cores. The reaction rate ratios of C8/F9 and F8/F9 were overesti-

• Japan Information Service. Co. .Ltd.. Tokyo

。。

J A E R I - M 8 8 - 2 2 1

mated for the JENDL-3T data. On the other hand, Doppler and sodium void reactivities, and reaction rate distribution obtained for the JENDL-2 data were significantly improved by using the JENDL-3T data.

The present benchmark tests of JENDL-3T showed that nuclear data to be reevaluated are u, fission cross section and fission spectrum for U-235, fission cross section and fission spectrum for Pu-239. and capture and inelastic scattering cross section for U-238 until the final compilation of JENDL-3.

References 1) JENDL C o m p i l a t i o n Group(Nuc1 ear Data C e n t e r , J A E R I ) : JENDL-3T,

P r i v a t e c o m m u n i c a t i o n (1987) . 2) "Cross S e c t i o n E v a l u a t i o n Working Group Benchmark

Speci f i c a t i o n s , "ENDF-202(Bi\L-l 9302 ) .Brookhaven N a t i o n a l Lab. 3) Takano. H et a l . : JAERI-M 82-072(1982) .

Table 1 .2 .1 F a s t c r i t i c a l benchmark cores

A s s e m b l y fuel vo lume (1) N8/N9 VERA-11A Pu 12 0.05 ZEBRA-3 Pu 50 8.5 SNEAK-7A Pu 110 3.0 FCA-5-2 Pu 200 2 .3 ZPR-3-53 Pu 220 1.6 SNEAK-7B Pu 310 7.0 ZPR-3-50 Pu 340 4 .5 ZPR-3-48 Pu 410 4 .5 ZPR-3-49 Pu 450 4 .5 ZPR-3-56B Pu 510 4 .5 MZA Pu 570 3.9 FCA-6-2- Pu 630 6.6 MZB Pu 1800 5.8 ZPPR-2 Pu 2400 5.5 ZPR-6-7 Pu 3100 6.5 ZPPR-9. PU 4600 9.4 VERA-IB u 30 0.07 ZPR-3-6F u 50 1.1 ZPR-3-12 U 100 3 .8 ZPR-3-11 U 140 7.5 ZEBRA-2 U 430 6.2 ZPR-6-6A u 4000 5.0 « Two-dimens ional benchmark core

Table 1 .2 .2 N"atU02 Doppler r e a c t i v i t y c a l c u l a t e d for ZPPR-9 assembly

T e m p e r a t u r e C a l c u l a t i o n / E x p e r i m e n t (k) JENDL-2 JENDL-

298 - 487 0.879 0.938 298 - 644 0.886 0.947 298 - 794 0.858 0 .918 298 - 935 0.896 0.959 298 -1087 0.888 0.951

jAERI-M 88・221

mated for the JENDL-3T data. On the other hand. Doppler and sodium void reactivities. and reaction rale dislribution obtained for the JENDL-2 data were signific且ntly improved by using the JENDL-3T data.

The present benchmark tests of JENDL-3T showed that nuclear data to be reevaluated are νfission cross section and fission spectrum for U-235. fission cross section and fission spectrum for Pu-239. and capture and inelaslic scaltering cross section forじー238unti 1 the final compi lat ion of JENDL-3.

Referenccs 1) JENDL Compilation Group(Nuclellr Data Center. JAERI): JENDL-3T.

Private communication (1987). 2) "Cross Section Evaluation Working Group Benchmark

Specifications."ENDF-202(BI'iL-19302) .Broukhaven National Lab. 3) Takano. H et al.: JAERI-M 82-072(1982).

Table 1.2.1 Fast critical benchmark cores

Assembly fuel volume (1) N8/N90r N5 VERA-IIA PU 12 0.05 ZEBRA-3 PU 50 8.5 SNEAK-7 A PU 110 3.0 FCA-5-2 PU 200 2.3 ZPR-3-53 PU 220 1.6 SNEAK-7B PU 310 7.0 ZPR-3-50 PU 340 4.5 ZPR-3-48 PU 410 4.5 ZPR-3-49 PU 450 4.5 ZPR-3-56B PU 510 4.5 MZA PU 570 3.9 FCA-6-2・ PU 630 6.6 MZB PU 1800 5.8 ZPPR-2 PU 2400 5.5 ZPR-6-7 PU 3100 6.5 ZPPR-9・ PU 4600 9.4 VERA-IB U 30 0.07 ZPR-3-6F U 50 1.1 ZPR-3-12 U 100 3.8 ZPR-3-11 U 140 7.5 ZEBRA-2 U 430 6.2 ZPR-6-6A U 4000 5.0 • Two-dimensional benchmark core

Table 1.2.2 NatC02 Doppler reactiyity calculated for ZPPR-9 assembJy

Tempera t ure Ca 1 cula t ion/Exper imen t (k) JENDL-2 JENDL-3T

298 -487 0.879 0.938 298 -644 0.886 0.947 298 -794 0.858 0.918 298 -935 0.896 0.959 298 -1087 0.888 0.951

J A E R I - M 8 8 - 2 2 1

10' 10 5" Core Volume (liter)

F i ^ . l . ^ . l k,ff for fd9t critical cores

• 20

• 10

S 0 C o S -io

i—i iinim i 11nnil—r run., 1 ; tn.1:'? ) intiji) I nwii I imm 1 - ! n : : ! 3

IV-1 i l . . i i L J ! ^

•- f-!- 1 r r

^ * % 5

,| U J.

JENDL-3T JENOL-2

- 4 0 F 1 I llMITf 1 I " • 1 1 ' 1 • • • '

.;•' io° 10' IO 1 10' io' io* io* 10' Energy (eV)

F i t : . 1 . 2 . 2 Deviat ion for o>("'Pu) of JENDL-3T from JENDL-2

1 i .; 1 2 . 3 C / E v « l u e j o f <A"/oVi

• 50

• 40

> 6 MO

; 1 J Mini I ll i l iai I m i i IIIIUI I imill i in;

JENDL-3T JEN0L-2

— I — - - ! • — - ) • -I i r i < I

IT 1 0 P i n '"'• • I ' m " •" • . • !••»•• I I I I I ;

1 0 ' 10' 10' 10' 10' 10' 10' 10* 10' Energy (eVl

F i p . 1 . 2 . I }e»ta-.lon for «,( , HU) of JEN0L-3T from JENDL-J

0 . 0 1 0 0 . 0 2 0 0 . 0 3 0 0 . 0 4 0 0 . 0 5 0 0 . 0 Void Region ( I )

" I l l i l l l l

fa " i ' • • • ' • • ' • " • ' ' • • ' C i 2 0 . 0 4 0 . 0 6 0 . 0 8 0 . 0 1 0 0 . 0 1 2 0 . 0

D i s t a n c e from Core Center (cm) F i g - 1 • 2 . a Companion of Na-vold reac t iv i ty at the ZPPR-9 corr

F i K . 1 • 2 . (i °Vu f t i l l o n rat* d i s tr ibut ion at the ZPPR-9 core

- 1 0 -

]AERI-M 88・221

0.97両

o Io・-，-.

IF岨和凶Aft-Ene'gy (eV)

"'i~. I.:!. 1 )..IO'lon for 6，('町】 ofJENDl.・3Tfr開淀川DL・2

'-10・-10

四日開日

ze-H噂桐谷刷白

k." for fdS! crlttcdl cores

日 I. ~. 3 C〆E'141 ues of <J.M1Jr>

Fi I!. I . t. r

l. 05 ωu 、.... 1.00

一10-

J A E R I - M 8 8 - 2 2 1

1.3 Problem of Total Cross Section for 0, N, Na, Fe, SUS in JENDL-3T by the Analysis of BROOMSTICK Experiment

A. Hasegawa

Total cross-section check in MeV range (lMeV-lOMeV) of JENDL-3T1* has been performed through the analysis of Broomstick experiments.

This series of experiments are uncollided spectrum measurements performed by Straker2'"6^ using the Tower Shielding Facility II (TSF-II) at ORNL to investigate minima in the total cross sections in MeV range for the typical shielding materials, 0, N, Na, Fe and SUS310.

In the experiment, the sample was a cylinder form of 4 inch diameter and placed so that its axis coincided with the axis of the neutron beam. In order to reduce the effect of neutron inscattering in the sample, the distance from the neutron source (TSF-II) to the sample was 50 feet and the detector was 50 feet from the sample.

As this experiment was performed in good geometry as stated above, no transport calculation Is needed for the analysis. Calculation proceeds as follows: first, to determine a transmitted uncollidt't spectrum and next, to fold thus calculated spectrum with the resolution function of NE-213. Source neutron spectrum and the detector resolution functions are given in a tabulated form.

A code system WINDOW'' has been developed to access the ENDF/B format library and to calculate the value directly comparable to the experimental one. The same calculational procedure is applied for ENDF/B-IV, JENDL-2 and -3PR1 as well as JENDL-3T.

Results and discussion Average values and standard deviations over the energy range in C/E

basis are given in Table 1.3.1. Oxygen:

Agreement is better for ENDF/B-IV. JENDL-3T result is systematically lower than that of ENDF/B-IV by 6-7 %. From the sensitivity calculation, about 5% change (decrease) in the total cross-section will compensate the difference. Nitrogen:

The results for ENDF/B-IV and JENUL-3T are comparable.

- 1 1 -

JAERI-M 88・221

1.3 Problem of Total Cross Section for 0， N， Na， Fe， SUS

in JENDL-3T by the Ana1ysis of BROOMSTICK Experiment

A. Hasegawa

Total cross-sectlon check ln MeV range (lMeV・10MeV) of JENDL-3T1)

has been performed through the analysls of Broornstick experirnents.

This series of experlments are uncollided spectru悶 measurements

perforrned by Straker2)・6)using the Tower Shielding Facility 11 (TSF-1I)

at ORNL to investigate min1.rna ln the total cross sections in MeV range

for the typical shielding materials. O. N. Na. Fe and SUS310.

1n the experirnent. the sample was a cylinder form of 4 inch diameter

and placed so that its axis coincided胃iththe axis of the neutron beam.

1n order to reduce the effect of neutron inscattering in the sample.τhe

distance from the neutron source (TSF-11) to the sample was 50 feet and

the detector was 50 feet from the sample.

As this experirnent was performed in good geometry as stated above.

no transport calculation is needed for the analysis. Calculation proceeds

as follows: f1rst. to determine a transrnitted uncollide1 spectrum and

next. to fold thus calculated spectrum wlth the resolut1on function of

NE-213. Source neutron spectrum and the detector resolutlon functians are

given in a tabulated form.

A code system WINDOW7) has been developed to access the ENDF/B forrnat

library and to calculate the value directly comparable to the

experirnental one. The same calculational procedure 1s applied for

ENDF/B・1V，JENDL-2 and -3PRl as well as JENDL-3T.

Rcsults and discussion

Average values and standard deviations over the energy range ln C/~

basis are given in Table 1.3.1.

Oxygcn:

Agreernent Is better for EYDF/B-1V. JENDL-3T result Is systematically

lower than that of EYDF/B-1V by 6-7 %. From the sensltlvity calculatlon，

about 5% change (decrease) ln the total cross-section will compensate the

difference.

Kitrogcn:

The results for ENDF/B-IV and JENDL-3T are cornparable.

JAERI-M 88-221

Sodiua: C/L is overestimated by about 20 % for all files. There are no

difference between JENDL-2 and -3T. Clear difference is observed between JENDls and EtVDF/B-IV in the energy range cf 6MeV to 10 MeV, where range ENDF/B-IV is better. Iron:

The average value of C/E is nearly the same between all of the files, but the value of standard deviation for E.VDF/B-IV is about the half of the others. Figure 1.3.1 shows the problem of JEffDL-3T cross section. Some clear tendency is seen in JENDL-3T, i.e., below 3 MeV total cross section of JENDL-3T seems to be overestimated and in the high energy range 8-10 MeV underestimated since spectrum and the cross sections have inversely proportional relation. Requests for the re-evaluation is asked. SUS310:

Rough composition is as follows, Fe(51 w/ 0), Ni(21 w/ 0), Cr(25 w/ 0), Si, Mn, C(rest). Overall agreement is better for ENDF/B-IV. For JENDL-3T below 3 MeV underestimation of spectrum is found, this is the same as Fe case. The difference between JENDL-3T and -3PR1 comes from the cross section of Cr and Ni, since there is no difference between those of Fe. JENDL-3T data are found to be improved.

From this analysis, it is demonstrated that the total cross-section in MeV range for these nuclides in JENDL-3T is not superior to ENDF/B-IV yet. Re-evaluation work is requested.

References 1) JENDL compilation group (Nuclear Data Center, JAERI):JENDL-3T

private communication (1987). 2) Maerker R.E. 3) Maerker R.E. 4) Maerker R.E. 5) Maerker R.E. 6) Maerker R.E.

ORNL-TM-3867 (Revised), (1972), ORNL-TM-3868 (Revised), (1972). ORNL-TM-3869 (Revised), (1972). ORNL-TM-3870 (Revised), (1972). ORNL-TM-3871 (Revised), (1972).

7) Hasegawa A. : to be published.

- 1 2 -

]AER!-M 88・221

Sodiu.:

C/L is overestimated by about 20 % for a11 fl1es. There are no

difference between JENDL帽 2and 圃 3T.C1ear dlfference Is observed between

JENDLs and ENDF/B-IV in the energy range cf 6MeV to 10 ~reV ， where rang-e

ENDF/B-IV 1s better.

The average value of C/E Is nearly the same between a11 of the fl1es，

but the value of standard deviation for EXDf/B-IV 1s about the ha1f of

the others. Fl忠Jre1.3.1 shows the prob1em of JENDL-3T cross section.

Some clear tendency 1s seen in JENDL-3T， i.e.， be10胃 3MeV tota1 cross

section of JENDL-3T seems to be over~stimated and 1n the high energy

range 8-10 MeV underestimated since spectrum and the cross sections have

inverse1y proportiona1 re1ation. Requests for the r・e-eva1uatlonis asked.

SUS310:

Rough composition 1s as fo11ows， Fe(51w/o)， Ni(21W/o)' Cr(25W/o)， Si，

Nn， C(rest). Overa1l agreement is better for ENDF/B-IV. For JENDL-3T

be10w 3河eVunderestimat10n of spectrum 1s found， this 1s the same as Fe

case. The difference between JENDL-3T and -3PR1 comes from the cross

section of Cr and Ni， since there 1s no difference between those of Fe.

JENDL-3T dat.a are found to be 1mproved.

From this analysis， it 1s demonstrated that the tota1 cross-section

in ilfeV range for these nuclides 1n JENDL-3T 1s not superior to ENDF/B-IV

yet. Re-evaluation work 1s requested.

References

1) JENDL compilat10n group (Nuclear Data Center， JAERI):JENDL-3T

private communication (1987).

2) ilfaerker R.E. ORNL-TM-3867 (Revised)， (1972).

3) Maerker R.E. ORNL-TM-3868 (Revised)， (1972).

4) Maerker R.E. ORNL-T河ー3869(Revised). (1972).

5) Maerker R.E. ORNL-T河ー3870 (Rev1sed)， (1972).

6) Maerker R.E. ORNL-T河-3871 (Revised). (1972).

7) Hasegawa A. to be published.

88-221

o u a « O

OS" i.

wnyi33ds

in (-CM ""

v "••'

«•*• —' > r.

in •

» > »

,-* /S ^ .,-, 0 2

OC 0 2

"" "" 0 ~

"** ""* "-* *"" > a>

US s S

_ > s I

IIIS i

51.11.llco 01 C/~ ，.111"拘 rcu'Ih'onlll"lJck C!xl.orll1wnte T.hl. 1.3.1

TH I CK ; CM 20.3 FE-O ClE DAT JENDL司 3T口

同.同-

1~~.2 cm

C/F. ・Id.)

0.68日 (0.218)

0.7~6 (0.213)

2MaV . -8MoV )

thlcknの師同

JF.Nl>I，.3T

I';NIJI'/II ~

(onorgy rnnr，n・

。腿ygnn

ClED円TENOFB4

ClED円IJENDL-2

， / ，

'‘ ロ凶.白一

口出・一-

εコ区ト

U止の

Nllro匝cn thlckllOBft: 91.-t-t cm

JliNlll. 3T 1 . 13~ (0.297)

ENlJf/1l ~ 1.138 (0.298)

{・norgyr.II，': aOOk.V -IOMoV )

」〉伺河

-'r-∞∞aNN

ロ凶・口【

τ口一"

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-LのZ江区ト

Smliufn lhicknoS5! 60.6 crn

.11>1'11)1. 3T 1. 2~3 (0.189)

JENIlI，-2 1. 2~2 (0.188)

F.NI】t'/ Il-~ 1.217 (0.2151

(oner，y r・nc・OOOk.V'-IIM.V ) tA)

error min

ロ凶・田

ロ凶.F

20.3 cm 30.5 cm

0.9~8 (0.160) 0.937 (0.25~)

0.959 (0‘163) 0.935 (0.255)

0.959 (0.163) 0.938 (0.254)

0.95~ (0.086) 0.986 (0.176)

1 2M.V -IIMoV I.M.V -8MoV

lhickn・8S!.IF.NIII.-3T

JF.NIJI..31'1

.IENIJI， Z

F.NIJf 111-4

t ・ ne~ R' Y ，.nI5:

」ト心

口凶・曲目官.00nu

au .， n

0.40 0.60

ENERGY

Fig. 1.3.1

20.3 cm

J '12 (0.276)

1.;04 (0.308)

1.020 (0.254)

1.2M.V -IIM.V)

SUS310 Ihlckll・-・:JEN¥lI.-3T

JENIJl，.31'1

ENJJf/s.4

t・"・ror・"'・: C/E proCiies or the transmitted spcctrulIl

thick) (20.3 cm for Fe s且mple n.b. Ih. lIumb.r 111 Ih・pu・nlh・.1・1・

・・t・nd.rdd・，1.11011.

J A E R I - M 8 8 - 2 2 1

1.4 Assessment of Natural Iron Cross Sections of JENDL-3T through ASPIS Deep Penetration Shielding Experiment

A. Hasegawa

The author performed an assessment of JEXDL-3T natural iron data 1' to the fission reactor shielding applications by analyzing the benchmark-quality experiment of deep penetration measured at Winfrith ASPIS shielding facility '. To investigate natural Fe cross-section of JENDL-3T, ENDF/B-IV data are also used for the reference.

Experimental configuration is as follows:a natural uranium converter plate, driven by the source reactor JfESTOR, was used as the source. The plate provided a large thin disc source of fission neutrons at the interface of a graphite moderator and extensive iron shield (= 140 era thickness) for the experiment.

Analyses are made for the axial attenuation measurements with three threshold detectors and one low-energy activation detector, and for the spectrometer measurements at four selected positions in the iron shield. We closely follow the calculation model reported by M.D.Carter et al. '.

Cross-sections were prepared by PROF-GROUCH-G/B code 3' with BERMUDA 121 energy group structure. Anisotropy was considered up to Pg. Weighting functions used were 1/E above 0.32241 eV and Maxwellian below it. As to the natural Fe cross-sections in the test region, which is thought as nearly pure material, fully shielded cross-sections including higher P^ matrices(1/0^ weight i.e., 0^=0.) were calculated and used.

Calculations were performed by DOT3.54' with P5-S3 R-Z:53x92 meshes. 1) Results for axial attenuation Measurements

The results are given in Fig.1.4.1. A. S-32(sensitive energy range: 2-4 MeV)

For both of ENDF/B-IV and JENDL-3T C/E values are greatly underestimated along with the penetration depth; it seems that the inelastic cross-section from 2 to 4 MeV is too high.

B. Rh-103 (sensitive energy range: 100-700 keV) Froa C/E behaviors no particular conclusion is drawn.

C. In-115 (sensitive energy range: 340 keV-1.3 MeV)

- 1 4 -

]AERI-M 88・221

1.4 Assessment of Natural Iron Cross Sections of JENDL-3T

through ASPIS Deep Penetration Shielding Experiment

A. Hasegawa

The author performed an assessment of JE~DL-3T natural lron data1)

to the fission reactor shieldlng appllcatlons by analyzing the benchmark-

quality e~perlment of deep penetratlon measured at Winfrith ASPIS

shielding faci11ty2). To investigate natural Fe cross-section of JE~DL-3T. ENDF/B-IY data are a1so used for the reference.

Experimenta1 cor.figuration 1s as fo110ws:a natural uranium converter

p1ate. driven by the source reactor NESTOR. 胃asused as the source. The

plate provided a 1arge thin disc source of fission neutrons at the

interface of a graphite moderator and extens1ve iron shield (= 140 crn

thickness) for the experiment.

Ana1yses are made for the axlal attenuat10n measurements w1th three

threshold detectors and one low-energy activation detector. and for the

spectrometer measurements at four selected posltlons 1n the lron shleld.

We c1ose1y fo11ow the calcu1ation mode1 reported by M.D.Carter et a1.2). 3) Cross-sections were prepared by PROF-GROUCH-G/B code"" with BER.J)IUDA

121 energy group structure. Anisotropy was cons1dered up to P5・

胃eighting functions used were 1/E above 0.32241 eV and Maxwel11an below

1t. As to the natural Fe cross-sections 1n the test reg1on. which 1s

thought a~ nearly pure mater1al. fullY shielded cross-sections including

higher Pl matrices(l/dt weight 1.e.. 00=0.) were calculated and used.

Calculations were performed by DOT3.54)胃ithP5・88R-Z:53x92 meshes.

1) Rcsults for axfal attenuation .casurc.ents

The resu1ts are given in Fig.1.4.1.

A. 8-32(sensitive energy range: 2-4 MeV)

For both of ENDF/B-IV and JENDL-3T C/E va1ues are greatly

underestimated along with the penetration depth: it seems that

the inelastic cross-section from 2 to 4 MeV 1s too high.

B. Rh-103 (sensitive energy range: 100-700 keV)

From C/E behaviors no particular concluslon 15 drawn.

C. !n-115 (sensitive ener6，Y range: 340 keV-1.3河eV)

J A E R I - M 8 8 - 2 2 1

A clear difference is observed between ENDF/B-IV and JENDL-3T. C/E of ENDF/B is almost stable but JE.N'DL profile is underestimated considerably. This is attributed to the differences in the share of elastic and inelastic cross sections in 600 keV to 1.2 MeV, because the total cross sections are nearly identical. Re-evaluation is requested for JENDL data as to this point. This indication is the same as that to JENDL-3PR1, i.e., no improvement has been made by the transition from JENDL-3PR1 to 3T.

D. Au-197 Cd covered (resonance detector -sensitive energy: 4.9 eV) C/E value of ENDF/B-IV data shows good results, no special tendency is seen with the penetration depth but JENDL-3T result has a tendency of underestimation along with the penetration depth. This tendency is amended a little for JENDL-3T than that for JENDL-3PR1 because of the improvements on the thermal cross section. But some problems other than the thermal cross section are still left.

2) Results for flux spectrometer ieasure«cnts The results are shown in Fig.1.4.2. For JENDL-3T data, as mentioned in the In-115 case, clear underesti

mation in fluxes from 600-keV to 1.2 MeV is seen. The agreements by the JENDL-3T data near the 24 keV Fe resonance range

is better than that by ENDF/B-IV data, i.e., the 24 keV resonance data of JENDL-3T are better than those of ENDF/B-IV.

Underestimation of fluxes in MeV range (2-5 MeV) appearing In both of the data suggests that the inelastic cross section is rather too high.

References 1) JENDL compilation group (Nuclear Data Center, JAERI): JENDL-3T

private comnunication (1987). 2) Carter M.D., McCracken A.K. and Packwood A. : "The Winfrith Iron

Benchmark Experiment, A Compilation of Previously Published Results for use in the International Comparison of Shielding Data Sets Sponsored by NEA," AEE Winfrith (1982).

3) Hasegawa A. et al., to be published. 4) Rhoades W.A. and Mynatt F.R. : ORNL-TM-4280(1973).

JAERI-M 88・221

A clear difference 1s observed between ENDF/B-IV and JENDL・3T. C/E

of ENDF/B 1s almost stable but JENDL prof11e 1s underestimated

considerably. This Is attrlbuted to the differences In the share

of elastic and lnelast1c cross sectlons 1n 600 keV to 1.2 MeV，

because the total cross sectlons are nearly 1dentlcal. Re-evaluatlon

1s requested for JENDL data as to this p01nt. This 1nd1cat1on

1s the same as that to JENDL-3PR1， 1.e.， no improvement has been

made by the transitlon from JENDL-3PR1 to 3T.

D. AU-197 Cd covered (resonance detector -sensitive energy: 4.9 eV)

C/E value of ENDF/B-IV data shows good results， no special tendency

is seen with the penetratlon depth but JENDL-3T result has a

tendency of underestlmatlon along with the penetration depth. Thls

tendency 1s amended a 11ttle for JENDL・3T than that for JENDL-3PR1

because of the improvements on the thermal cross sectlon. But sCJme

problems other than the thermal cross sect10n are st111 left.

2) Rcsults for flux spectro・eter.easure.cnts

The results are shown 1n Fig.1.4.2.

For JENDL-3T data， as ment10ned ln the 1n-115 case， clear underesti-

mation in fluxes from 600，keV to 1.2 MeV 1s seen.

The agreements by the JENDL-3T data near the 24 keV Fe resonance range

1s better than that by ENDF/B-IV data， i.e.， the 24 keV resonance data of

JENDL・3Tare better than those of ENDF/B-IV.

Underestimation of fluxes in MeV range (2・5河eV)appearing 1n both of

the data suggests that the 1ne1astic cross sectlon 1s rather too h1gh.

References

1) JENDL compilation group (Nuclear Data Center， JAERI): JENDL・.3T

private commun1cation (1987).

2) Carter M.D.， McCracken A.K. and Packwood A. "The 胃infrithIron

Benchmark Experlment， A Co圃pi1ationof Prev10us1y Published Resu1ts

for use in the Internatlonal Comparlson of Shielding Data Sets

Sponsored by NEA，" AEE胃infri th (1982).

3) Hasegawa A. et a1.， to be pub1ished.

4) Rhoades W.A. and Mynatt F.R. ORNL-TM-4280(1973).

J A E R l - M 88-221

10 ?J_

QE'ECTOR RESPONSE

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detector responses

I . 4 . 1 Fil5.

f SPEC 門jZ: 85.7

J-CRL-DRT -‘02 8-4

_ Cf'lL.O白r.._: .406 J ・3

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1d!?Ib;J， ~1~ I ~ 、 10lj7~T f-/-~~I 1 ild!?11a1 110.jE 1 p¥ ~ iljl1111 三10'三 1ll11

R DAI< Comparison of calculnled flux profiles wilh

unfoldod spcctrum measurement at 85. cm posi t:on

一16-

1.4.2 Fig..

J A E R I - M 8 8 - 2 2 1

1.5 Benchmark Test of JENDL-3T for Materials Used in Fusion Reacti

T. Mori and M. Nakagawa

Benchmark test of JENDL-3T has been performed for materials used in fusion reactors. Experiments usad for this test include the neutron angular spectra measured at LLNL and the beryllium multiplication factors measured at Kurchatov laboratory." In the LLNL pulsed sphere experiments, leakage neutron angular spectra at 30* and 120* were measured for lithium-6, -7, carbon, oxygen, iron, lead, polyethylene, water and beryllium with one or a few kinds of thickness. The calculations have been performed by using a Monte Carlo

3) code MORSE-DD and the double differential form cross sections of 125 group produced from JENDL-3T and JENDL-3/PR1.

The results of comparison between the measurements and the calculations are discussed below. Polyethylene and wate: Quite good agreement is seen for the leakage spectra from the materials including hydrogen. Li thium-6: Discrepancy is observed in the range 1 3 - 6 MeV and the discrepancy observed in the result with JENDL-3/PR1 is considerably decreased for JENDL-eT below 6 MeV. Li thium-7: Slight underestimation is seen in the range 1 0 - 7 MeV and below 5 MeV as shown in Fig.1.5.1 though JENDL-3/PR1 gives larger one. Such discrepancy is increased for the spectra at 120* but an agreement is good for the case with 1.6 M.F.P. thickness. Carbon: Satisfactory agreement is obtained for both angular spectra. Oxygen For the spectra at the angle 30*, JENDL-3T reduces the discrepancy observed in the result by JENDL-3/PR1 though that of the fine structure at about 4 MeV due to discrete level inelastic scattering is still remained. Iron: Underestimation is observed in the range 1 2 - 7 MeV and it increases with thickness of the assembly. Lead: Discrepancy appears below 13 MeV as shown in Fig.1.5.2, which will cause disagreement in predicting a multiplication effect. Beryllium The discrepancy observed in the result by JENDL-3/PR1 decreases but still remains in the range 1 0 - 6 MeV as shown in Fig.1.5.3.

From this benchmark test, we can conclude that JENDL-3T can improve

- 1 7 -

jAER[-M 88・221

1.5 Benchmark Test of JENDL-3T for Materials Used in Fusion Reactl

T.'対ori and M. Nakagawa

Benchmark test of JENDL-3T has been performed for materials used

in fusion reactors. Experiments usad for this test include the 1)

neutron angular spectra measured at LLNL.' and the beryllium 2 )

multiplication factors measured at Kurchatov laboratory.w， [n the

LL~L pulsed sphere experiments， leakage neutron angular spectra at 30・and 120・weremeasured for lithium・6，-7， carbon， oxygen， iron， lead，

polyethylene， water and beryllium with one or a few kinds of

thickness. The calculations have been performed by uSing a Monte CarJo 3>

code MORSE-DDu

， and the double differential form cross sections of 125

group produced from JENDL-3T and JENDL-3/PR1.

The results of comparison between the measurements and the

calculations are discussed below.

Polvethvlene and wat ~: Quite good agreement is seen for the

leakage spectra from the materials including hydrogen.

Li主主主皿ニ阜 Discrepancyis observed in the range 13・6MeV and the

discrepancy observed in the result with JENDL-3/PRl is considerably

decreased for JENDL-eT below 6 MeV.

Lよthi1坦ニ1:Slight underestimation is seen in the range 10・ 7MeV and

below 5河eVas shown in Fig.l.5.1 though JENDL-3/PRl gives larger

one. Such discrepancy is increased for the spectra at 120. but an

agreement is good for the case with 1.6 M.F.P. thickness.

E金王並立旦 Satisfactoryagreement is obtained for bQth angular spectra.

' Q茎yg豆旦 Forthe spectra at the angle 30.. JENDL-3T reduces the

discrepancy observed in the result by JENDL-3/PRl though that of the

fine structure at about 4 MeV due to discrete level inelastic

scattering is stil1 remained.

i工on: Underestimation is observed in the range 12 -7河eVandit

increases with thickness of the assembly.

L皇室g: Discrepancy appears below 13 河eVas shown in Fig.l.5.2. which

will cause disagreement in predicting a multiplication effect.

E金ul.li旦mThe discrepancy observed in the result by JE~DL-3/PRl

decreases but still remains in the range 10 -6 MeV as shown in

Fig.1.5.3.

From this benchmark test， we can conclude that JENDL-3T can improve

-17ー

JAERI-M 88-221

the prediction accuracy of neutron spectra compared with that by JENDL-3/PR1 for most materials mentioned above although discrepancies still remain for some materials.

In addition to the spectrum analysis, we tested the neutron multiplication experiment for beryllium of which accurate prediction is important in calculating tritium breeding ratio. The measured and calculated values are compared in Table 1.5.1. In the calculations, the isotropic neutron source produced by (d,t) reaction was assumed since the detailed information about source neutrons was not available. We can see that the multiplication factors?total leakage in this experiments) can be well predicted for all the assemblies.

References 1) Wong C. et al.:"Livermore Pulsed Sphere Program, Program Summary

through July 1971",UCRL-51144(Rev.182)(1971),and "Measurements and Calculations of the Leakage Multiplications from Hollow Beryllium Spheres",UCRL-91774(1985).

238 2) Zagryadskij V.A. et al.: "Measurement of Neutron Leakage from U,

232 Th and Be Spherical Assemblies with a Central 14-MeV Source",

INDC(CCP)-272/G(1987), 3) Nakagawa M. and Mori T.:"MORSE-DD",JAERI-M 84-126(1984) .

Table 1.5.1 Comparison of measured and calculated multiplication effects of beryllium spheres

Thickness(cm) (H.F.P.) Multi.a N b Tc

1.5(0.27) Expt. 1.1441.036 .320i.035 J3/PR1 1.10 (0.96) d .232 .866 •13/T 1.11 (0.97) .245 .859

5.0(0.90) Expt. 1.36510.40 .762Ji.039 J3/PH1. 1.345 (0.99) .730 .615 J3/T 1.364 (1.00) .760 .604

8.0(11.4) Expt. 1.5291.043 1.0661.. 042 J3/PR1 1.524 (1.00) 1.073 .451 J3/T 1.553 (1.02) 1.114 .439

a mul tiplicationUotal leakage) secondary neutron component

c transmitted component of 14 MeV neutron(calculated value) ratio of calculation to experiment

- 1 8 -

JAERI-M BB・221

the prediction accuracy of neutron spectra compared with that by

JENOL-3/PRl for most materials mentioned above although discrepancies

still remain for some materials.

In addition to the spectrum analysis， we tested the neutron

multiplication experiment for beryllium of which accurate prediction

is important in calculating tritium breeding ratio. The measured and

calculated values are compared in Table 1.5.1. In lhe calculations，

the isotropic neulron source produced by (d，t) reaction was assumed

since the detailed informalion about source neutrons was not

available. We can see that the multiplication factors(total leakage

in this experiments) can be well predicted for all the assemblies.

References

1) Wong C. et al. : "Livermore Pulsed Sphere Program， program Summary

through July 1971"，UCRL-51144(Rev.I&2)(197lJ，and "Measurements and

Calculations of the Leakage Multiplications from Hollow Beryllium

Spheres"，UCRL・91774(985). 238

2) Zagryadskij V.A. et al.: "Measurement of Neutron Leakage from ~VVU ，

232 Th and Be Spherical Assemblies with a Central 14-MeV Source"，

INOC(CCP)ー2721G(1987入

3) !lJakagawa ~. and ~ori T.:"MORSE-DO"，JAERI-M 84-126(1984).

Table 1.5.1 Cロmparisonof measured and calculated

multiplication effects of berylllum spheres

Thickness(cm) (M.F.P.l Multi.a b N

戸、伊目ゐ

1.5 【0.27)

Expt. 1.144土.036 . 320:t..035

J3/PRl 1.10 (0.96)d .232 .866

J3fT 1.11 (0.97) .245 .859

5.0(0.90)

Expt. 1.365主0.40 . 762.t..039

J3IPRl 1.345 (0.99) .730 .615

J31T 1.364 (1.00) .760 .604

8.0(11.4)

Expt. 1. 529主.043 1. 066t.. 042

J3/PRl 1.524 (1.00) 1.073 .451

J3/T 1.553 (1.02) 1.114 .439

a ・ultiplication(totalleaka，e) b secondary neutron c。圃ponent

c trans.itted co・ponentof 14 MeV neutron(calculated value>

d ratio of calculation to experi・ent-18-

JAERI-M 88-221

_ / •Vb4-

7 Fig.1.5.1 Leakage spectrum from Li sphere(0.7MFP) at 30°

0.0 0.2 0 .4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 Energy < eV ) ( * 1 0 ' )

Fig.1.5.2 Leakage spectrum from lead sphered.4MFP) at 30°

Fig.1.5.3 Leakage spectrum from beryllium(0.8MFP) at 30

]AERI-M 88・221

10' 3『

“申よW

。』ω3024句

2h¥2-“町、ヒ

Eh“zz

Fig.l.5.2 Leakage spectrum from lead

sphere(l.4MFP) at 30・

l・:・《・¥0匂 j

Fig.l.5.1 Leakage spectrum from 7Li

sphere(O.7MFP) at 30。

。2¥0・

0.0 l. 6 . e ( ~! 0' )

Fig.l.5.3 Leakage spectrum from beryllium(0.8MFP) at 300

J A E R I - M 8 8 - 2 2 1

1.6 Integral Test of JENDL-3T Through Benchmark Experiments Using FNS

H. Maekawa, K. Kosako, Y. Ikeda, Y. Oyaraa, s. Yamaguchi, K. Tsuda C. Konno and T. Nakamura

The evaluated nuclear data file, JENDL-3T'" was examined through the analysis of fusion blanket benchmark experiments at FNS of JAERI. These

1) 2) integral experiments were performed on Li 20 , C , Li20-C and Be-sandwich Li 20 assemblies. The configurations of assemblies are summarized in Table 1.6.1. Lithium-oxide, graphite and/or beryllium blocks were stacked in a frame made of thin-wall aluminium tubes to form a cylinder in the same manner for each of the four assemblies. The diameter was 63 cm. The D-T neutron target was located at 20 cm from the front surface of the assembly on the central axis. Measured quantities and their methods are summarized in Table 1.6.2. The measurements were performed along the central axis. Neutron yields were determined by means of the associated a-particle detection method.

3) In the present analysis the D0T3.5 code was used with the P 5-S 1 6

approximation. The 125-group cross section set FSX125/J3T was obtained from the JENDL-3T file using the processing code PROF-GROUCH-G/B. The

4) same type group constants based on JENDL-3PR1, -3PR2, ENDF/B-IV and -V (only for carbon) were used for comparison. The GRTUNCL code was used to calculate the first collision source for the succeeding DOT calculations.

Figure 1.6.1 shows the calculation to experiment ratio (C/E) distribution of tritium production rate (TPR) of 6Li (T 6) in the Li 20 assembly. The C/E distribution of T 6 in Be-sandwich Li 20 assembly is shown in Fig. 1.6.2.

From the comparison between calculated and experimental results, the following facts are pointed out for JENDL-3T: (1) A good agreement within 5 % 10 % is observed for T 6 and T 7 in the

Li 20 assembly. In the Be-sandwich Li 20 assembly, the calculation

t JENDL-3T is a temporary file for testing the evaluated data for JENDL-3. The data in JENDL-3T will be partly revised in JENDL-3.

]AERI-M 88・221

1.6 Integral Test of JENDL-3T Through Benchmark Experiments Using FNS

日. Maekawa， K. Kosako， Y. Ikeda， Y. Oyama， s. Yamaguchi， K. Tsuda

C. Konno and T. Nakamura

The evaluated nuclear data file， JENDL-3T十 wasexamined through the

analysis of fusion blanket benchmark experiments at FNS of JAERI. These 1) ~2)

1ntegra1 exper1ments were performed on L120.'， C-'， L120-C and

Be-sandwich Li20 assemblies. The configurations of assemblies are

summarized in Table 1.6.1. Lithium-oxide， graphite and/or bery11ium

b10cks were stacked 1n a frame made of thin-wa11 a1uminium tubes to form

a cy1inder in the same manner for each of the four assemb1ies. The

d1ameter was 63 cm. The D-T neutron target was 10cated at 20 cm from the

front surface of the assemb1y on the centra1 ax1s. Measured quantities

and the1r methods are summar1zed 1n Tab1e 1.6.2. The measurements were

performed a10ng the central ax1s. Neutron yie1ds were determined by

means of the associated a-partic1e detection method. 3)

In the present ana1ysis the DOT3.5 code-' was used with the PS-SI6

approx1mat1on. The 125-group cross sect10n set FSX125/J3T was obtained

from the JENDL-3T f11e us1ng the process1ng code PROF-GROUCH司 G/B. The 4)

same type group constants" based on JENDL-3PR1，ー3PR2，ENDF/B-IV and -V

(on1y for carbon) were used for comparison. The GRTUNCL code was used to

ca1cu1ate the first col11sion source for the succeed1ng DOT ca1cu1a-

t10ns.

Figure 1.6.1 shows the ca1culation to experiment ratio (C/E)

distribution of tritium product10n rate (TPR) of 6L1 (T6) 1n the L120

assembly. The C/E distribution of T6 in Be-sandwich L120 assembly is

shown in Fig. 1.6.2.

From the comparison between calcu1ated and exper1menta1 resu1ts，

the follow1ng facts are pointed out for JENDL-3T:

(1) A good agでeementwithin 5 ~ 10 % is observed for T6 and T7 in the

L120 assembly. In the Be-sandw1ch L120 assemb1y， the calcu1at1on

十 JENDL-3Tis a temporary file for testing the evaluated data for

JENDL-3. The data in JENDL-3T wi11 be part1y revised in JENDL-3.

-20ー

JAERI-M 88-221

does, however, not reproduce well the measured data near the Be-region.

(2) Most parr, of C/E values deviate from unity by less than 10 %. (3) The C/E values based on JENDL-3T for the reactions having high

thresholds are lower than those based on JENDL-3PR2 by a few peircents. On the other hand, the C/E values for reactions sensitive to low energy neutrons, such as T 6 and the 2 3 Slf fission rate, are higher by a few percents.

It can be concluded that the nuclear data of 6Li, 7Li, 1 2C and 1 6 0 in JENDL-3T are now good enough in accuracy of 5 -v, 10 % for the calculation of integral values such as T 6 and T 7 in a typical fusion blanket without beryllium. As beryllium is the most promising neutron multiplier, reevaluation for Be data in JENDL-3T is recommended for the application to fusion blanket neutronics.

References 1) Maekawa H., et al.: "Fusion Blanket Benchmark Experiments on a 60 cm

-thick Lithium-Oxide Cylindrical Assembly," JAERI-M 86-182 (1986). 2) Maekawa H., et al.: "Benchmark Experiments on a 60 cm-Thick Graphite

Cylindrical Assembly," JAERI-M 88-034 (1988). 3) Rhoades W. A., Mynatt F. R.: "The D0T-III Two Dimensional Discrete

Ordinates Transport Code," 0RNL/TM-4280 (1979). 4) Kosako K., et al.: "Neutron Cross Section Libraries for Analysis of

Table 1.6.1 Configurations of experimental assemblies.

Assembly Material

Region [cm] 0 - 5 . 0 8 5 .08-10 .16 10 .16-40 .64 4 0 . 6 4 - 6 0 . 9 6

Li20 < Li 20 > C •? C > Li20-C <? Li 20 X C > Be-sandwich ^Li20>^ Be X Li 20 • >

JAERI-M 88-221

does， however， not reproduce well the measured data near the

Be-reg10n.

(2) Most part of C/E values deviate from un1ty by less than 10 %.

(3) The C/E values based on JENDL-3T for the reactions having high

thresholds are lower than those based on JENDL-3PR2 by a few

percents. On the other hand， the C/E values for react10ns sens1t1ve

to low energy neutrons， such as T6 and the 235U f1ss10n rate， are

h1gher by a few percents.

It can be concluded that the nuclear data of 6L1， 7L1， 12C and 160

1n JENDL-3T are now good enough 1n accuracy of 5 ~ 10 % for the

calculation of 1ntegral values such as T6 and T7 1n a typ1cal fusion

blanket without beryllium. As beryllium 1s the most promis1ng neutron

multipl:ler， reevaluation for Be data 1n JENDL-3T 1s recommended for the

appl1cation to fusion blanket neutronics.

References

1) Maekawa H.， et a1.: "Fusion Blanket Benchmark Experiments on a 60 cm

-thick Lith1um-Oxide Cy11ndr1cal Assembly，" JAERI-M 86-182 (1986).

2) Maekawa H.， et al.: "Benchmark Experiments on a 60 cm-Thick Graphite

Cylindrical Assembly，" JAERI-M 88-034 (1988).

3) Rhoades W. A.， Mynatt F. R.: "The DOT-III Two D1mensional Discrete

Ordinates Transport Code，" ORNL/TM-4280 (1979).

4) Kosako K.， et a1.: "Neutron Cross Section Libraries for Analysis of

Table 1.6.1 ConfigGrations of experimental assemblies.

Assembly Material

Region [cmJ 0-5.0815.08-10.16110.16-40.64140.64-60.96

ぞ L120

ラヲ

ヤぐー一一 c一一一歩"'

Li20-C く Li20

Be-sandwich ~Li20~モーーー Be-→~ Li20ー一一ーーー→，

-・且qzu

Table 1.6.2 Measured quantities and their methods for integral experiments.

(1) Tritium production rates of 6L1 and 7Lt • Liquid scintillation met hud with 6Li 2 0 and

7 L 1 2 0 pellets • Self-irradiation method with LIF TI.Ds

for Be-sandwich assembly • 6Li and 7Li glass scintillators for T 6

• NF213 scintillator for T 7

(2) Fission rates • Micro-fission chambers (mfc) ( 2 3 5 U , 2 3 B U ,

2 3 7 N P , 2 3 2 T h ) • Solid-state track detectors (SSTD) with

2 3 5 U , 2 3 8 U , and 2 3 2 T h foils except for Be-sandwich assembly

(3) Reaction rates • Foil activation method -- with Al, In, and Ni foils

for L i 2 0 assembly -- with Al, Au, In, Nb, Ni, and Zr foils - with Al, An, Co, Fe, In, Mn, Nb, Ni,

Ti, Zn and Zr for l.i20-C and Be-sandwich assemblies

(4) Response of PIN diodes (5) Response of TI.Ds (measured in l.i 20 and C)

• TLD-600, -700, -100 LiF • UD-110S CaSO,, • M g 2 S j O „ , Sr 2SiO, 1 ( Ba 2SiO,,

(6) In-system neut^n spectra • Small sphere NF.213 spectrometer

LI20 Sl.b ( LI-6ln.>] ) 1.2

0. 8 20 30 40 SO 60 70

Distance f r t a the t « r | e t (cnl

Fig. 1.6.1 Comparison of C/E values for tritium production rate of 6Li in Li 20 assembly.

20 40 60 BO D i s t a n c e froi the t a r g e t (ci)

Fig. 1.6.2 Comparison of C/E values for T 6

in Be-sandwich assembly.

li2日SI.b1 L1-filn.・1】

--..ー JEtIl.-3I司1¥1---j!J-- JEtIl.-31帯一・-El{f/fHV

‘・ー←・ JEtIl.-3T:畑、・一一~ 骨.A.__..

-e，' &・・4-_--::ご二ι二~.

'" 『a・・ーー・ 4ド・--_.....-~-ー‘

、・・ー・:、グ~，~.

Measured quantities and their methods for integral experiments.

0.8 20

"-' ¥1.0 巳3

」〉何

白∞ENN-

Comparison of C/E v~l~e~ for tritium production rate of 6li in li20 ass佃 bly.

target

f rロ・

40 50 60 0101・00.Ir.. Ih. tor." 10闘l

0.7 20

Distl!nce

Fig. 1.6.1

• 1 w

(1) 'fritiul1I productiol¥ rates (}f 6Lt i'l¥d 7¥.1 • Liq¥lid sclntil1atiol¥ methud wilh bLi 20 al1u

7LI20 pellets • 5elf-lrradlatlon method wi th LiF TLl>s

for se-sandwich asse川uly• 6Li and 7Ll gJass scinLl11alors for T6

・NE213scintillaLor for T7 (2) Fission rates

• ~llcro-fission chambers (mfc) (235U， 23BU， 217Np， 232Th)

• 50lict -state track uelectors (55TO) wllh 235U， 238U， and 2321h fol1s

except for Be-s日I¥dwichassemuly

(3) Reactiol¥ rates • Foil activatioll method

ー withAl， 11¥， alld Nl foils

for Li20 asse"~ly w1th Al， Au， ln， Nb， Nl， and Zr [011s with Al， A¥I， Co， Fe， 11¥， HI¥， Nb， Ni，

Tl， Zn al¥d Zr for 1.120-C and se-s<ll¥dwich assembl jし s

( 1，) Re胃l川 nseof PIN dtodes (う) Response of T1.Ds (lIIeasllred i 11 1.120 alld C)

• TI.D-600， -700， -100 ------- LiF

• 111>-1105 -ーー』ーーーーーーーーーー一ーー--CaSO" • ~lg25i 0匂， 5r25iO~ ， Ra2S10~

(6) ll¥-system l¥eutP "n spectra

• 5mall sphere Nt':213 spectrometer

Table 1.6.2

Comparison of C/E values for T6 in Be-sandwich ass棚 bly.

Fig. 1.6.2

J A E R I - M 8 8 - 2 2 1

1.7 Angular Neutron Flux Measurement and Nuclear Data Test on Slabs of Fusion Blanket Materials

Y. Oyama, K. Kosako, S. Yamaguchi and H. Maekawa

In order to examine the validity of the nuclear data file JENDL-3T , the angular neutron flux experiments ' have been analyzed using Monte Carlo codes.

On fusion blanket materials, angular neutron flux spectra had been measured for DT neutrons using a time- of-flight method with an NE213 liquid scintillator. The angular neutron flux spectrum leaking from a slab is a basic quantity to calculate the neutronic parameters and provides the information of the bulk neuron transport. Since the single scattering is dominant for a thin slab, the angular flux relates directly to the double-differential cross section. On the other hand, the effect due to the uncertainties of the removal and elastic scattering cross sections are enhanced on leaked flux through a neutron slowing-down for a thick slab. Thus, the overall validity of the nuclear data can be tested through the systematics of the measured spectra with thickness and angle.

The angular neutron flux spectra are obtained in the range of 15 MeV down to below 100 keV by the experimental arrangement shown in Fig. 1.7.1. The slab assemblies are made in a pseudo-cylindrical slab shape by stacking rectangular blocks with the area-equivalent radius of 315 mm. The slab thicknesses are 50, 200 and 400 mm corresponding upto 5 mean free pach for 14.8 MeV neutrons. The assembly is placed at 200 mm from the DT source at FNS. The angles of the measured flux are 0, 12.2, 24.9, 41.8 and 66.8 degrees. The measured fluxes are normalized to one source neutron emitted from the target.

The Monte Carlo calculations were performed by the continuous energy code MCNP for the Li, C and Li 20 slabs and the multi-group code M0RSE-DD 6 ) for the Be slab. The MCNP library is FSXLIB 7 )

processed by ACER in the NJ0Y code and the MORSE-DD library is DDXJ3T in a double-differential-cross-section form of 125 groups processed by PROF-DD. The angular fluxes were estimated by point detector estimators. The measured source spectrum was taken into the input source condition for the calculation.

Figure 1.7.2 shows the measured and calculated spectra for the

- 2 3 -

J AER 1・ M 88・221

1.7 Angular Neutron Flux Measurement and Nuclear Data Test 1 )

on Slabs of Fus10n Blanket Mater1als

Y. Oyama， K. Kosako， S. Yamaguch1 and H. Maekawa

In order to exam1ne the valid1ty of the nuclear data file 2)_ __J_____3)，4)

JENDL-3T-'， the angular neutron flux experiments -" .， have been

analyzed us1ng Monte Carlo codes.

On fus10n blanket materials， angular neutron flux spectra had

been measured for OT neutrons using a time-of-flight 田ethodwith an

NE213 liquid sc1ntillator. The angular neutron flux spectrum leak1ng

from a slab is a bas1c quantity to calculate the neutronic parame:ers

and prov1des the informat1on of the bulk neuron transport. Since the

single scatter1ng is dom1nant for a thin slab， the angular flux

relates d1rectly to che double-cl1fferential cross sect1on. On the

other hand， the effect due to the uncertaint1es of the removal and

elastic scatter1ng cross sect10ns are enhanced on leaked flux through

a neutron slow1ng-down for a th1ck slab. Thus， the overall validity of

the nuclear data can be tested through the systemat1cs of the measured

spectra w1th th1ckness and angle.

The angular neutron flux spectra are obtained in the range of 15

MeV down to below 100 keV by the exper1mental arrangement shown in

Fig. 1.7.1. The slab assemu11es are made 1n a pseudo-cylindr1cal slab

shape by stacking rectangular blocks w1th the area-equivalent rad1us

of 315 mm. The slab th1cknesses are 50， 200 and 400 田辺 corresponding

upto 5 mean free pach for 14.8 MeV neutrons. The assembly is placed at

200 m皿 fromthe DT source at FNS. The angles of the measured flux are

0， 12.2， 24.9， 41.8 and 66.8 degrees. The measured fluxes are

normalized to one source neutron emitted from the target.

The Monte Carlo calculations were performed by the continuous 5)

energy code MCNP-' for the L1， C and L120 slabs and the multi-group

6) ~~_ _L_ "_ ""'-_ u，..，.:t"n ，，~'--_____ ~ _ "， roVT T 'n7} code 珂ORSE-OD-' for the Be slab. The MCNP library is FSXLIB

processed by ACER in the NJOY code and the MORSE-DD library 1s DDXJ3T

in a double-d1fferent1al-cross-section form of 125 groups processed by

PROF-DD. The angular fluxes were estimated by point detector

estimators. The measured source spectrum was taken into the input

source condition for the calculation.

Figure 1.7.2 shows the measured and calculated spectra for the

J A E R 1 - M 8 8 - 2 2 1

beryllium slab. The flux integrated over the interested energy regions were also compared in the form of the calculated-co-measured value (C/E) ratio as shown in Fig. 1.7.3 for the lithium-oxide slab. From the present comparisons, it is pointed out for the application of JENDL-3T to a fusion neutronics calculation that:

1) Carbon data overestimate the neutrons scattered by the second and third level inelastic scattering, but after penetration over 2 rafp, the predicted spectra become good.

2) Beryllium data underestimate the flux of 2-LO MeV range. This may cause the underprediction of tritium production rate from 7Li in beryllium containing blanket.

3) Lithium data is sufficent for high energy part of estimated spectrum, but it is necessary to check the total cross section.

4) Selection of oxygen data, JENDL-3T or -3PR1, has no difference for the lithium-oxide case and the calculation showed excelent agreement with the experiment. Hence, the present data is good enough.

References

1) Oyama Y., et al.:"Angular Neutron Flux Measurements and Nuclear Data Test on Slabs of Fusion Blanket Materials," Proc. Int. Conf. of Nuclear Data for Science and Technology, Mito, May (1988)

2) JENDL Compilation Group (Nuclear Data Center, JAERI):" JENDL-3T," private communication (1987). JENDL-3T is a temporary file for testing the evaluated data which are for JENDL-3. The data in JENDL-3T will be partly revised in JENDL-3.

3) Oyama Y. and Maekawa H.: Nucl. Sci. Eng., 9_7_, 220 (1987). 4) Oyama Y., Yamaguchi S. and Maekawa H.: J. Nucl. Sci. Technol., _25£5], 419 (1988).

5) Los Alamos Radiation Transport Group (X-6):"A General Monce Carlo Code for Neutron and Photon Transport," LA-7396-M (1981).

6) Nakagawa M. and Mori T.: "MORSE-DD, A Monte Carlo Code Using Multi-Group Double Differential Form Cross Section," JAERI-M 84-126 (1984).

7) Kosako K., et al.: " Neutron Cross Section Libraries for Analysis of Fusion Neutronics Experiments," JAERI-M 88-076 (1988) (In Japanese).

]AERI-M 88・221

beryllium slab. The flux integrated over the interested energy regions

were also compared in the for皿 ofthe calculated-to-measured value

(C/E) ratio as shown in Fig. 1. 7.3 for the 1ithium-oxide slab. From

the present comparisons， it is pointed out for the application of

JENDL-3T to a fusion neutronics calculation that:

1) Carbon data overestimate the neutrons scattered by the second and

third level inelastic scattering， but after penetrat10n over 2

mfp， the predicted spectra become good.

2) Beryllium data underestimate the flux of 2・10MeV range. This may

cause the underprediction of tritium product10n rate from 7L1 1n

beryl11um containing blanket.

3) Lithium data is sufficent for high energy part of estimated

spectrum， but it is necessary to check the total cross section.

4) Selection of oxygen data， JENDL-3T or -3PRl， has no difference

for the lithium-oxide case and the calculation showed excelent

agreement with the experiment. Hence， the present data 1s

good enough.

References

1) Oyama Y.， et al. : "Angular Neutron Flux Measurements and Nuclear

Data Test on Slabs of Fusion Blanket Materials，" Proc. Int. Conf.

of Nuclear Data for Science and Technology， Mito， May (1988)

2) JENDL Compilation Group (Nuclear Data Center， JAERI):" JENDL・3T，"

private communication (1987). JENDL-3T is a temporary f11e for

testing the evaluated data"which are for JENDL-3. The data in

JENDL-3T will be partly revised in JENDL・3.

3) Oyama Y. and Maekawa H.: Nucl. Sci. Eng.， 2L， 220 (1987).

4) Oyama Y.， Yamaguchi S. and Maekawa H.: J. Nucl. Sci. Technol.，

.35]， 419 (1988).

5) Los Alamos Radiation TranSpOI"t Group (X-6):" A General Monte Car 10

Code for Neutron and Photon Transport，" LA-7396-M (1981).

6) Nakagawa M. and Mori T.: "MORSE但 DD，A Monte Carlo Code Using

Multi-Group Double Differential Form Cross Section，" JAERI-M 84-126

(1984) .

7) Kosako K.， et al.: " Neutron Cross Section Libraries for Analysis

of Fusion Neutronics Experiments，" JAERI-M 88-076 (1988) (In

Japanese).

D-l utgtl

Asstntbty

Fig. 1.7.1 Experimental arrangement for angular neutron flux measurement

11.1.1. - SUB mi A«|l. - 24.0 4.,

•f Eipillpcnl - I jfJn-3i

I JQU. 3TIH

10 II) ' 10" 10" NOJinON EMilXJY (MoV)

Fig. 1.7.2 Measured and calculated spectra for the beryllium slab

JEWDL-3T

20 40 60

Angle I degree I

F i g . 1.7.3 Comparison of the c a l c u l a t e d - t o - m e a s u r e d

v a l u e (C/E) r a t i o f o r the ene rgy r e g i o n -

wise f lux for the l i t h i u m - o x i d e s l a b

』〉伺刃】

'g∞∞'MMF

ぞ一一宗主¥一、

刷、ヘ一一(《巨ふへい

~¥wfL~vr竺一¥

古山理川浩一一一ヘ

A. 0・t.ctOl¥

、h刷

、-nv ‘ 、品u

，‘ 、''

ト~I←J一寸

/-・ーー・・¥

0.1 ，.・由刷

Ass・mbfvExper1mental arrangement for angular F1g. 1. 7.1

60 柑20 。。泊 40ω

Asg le (degree I 60 40 20

neutron flux measurement

__I!t'寸可 1.-........・γ，-:yイ.1'

1 ()-2l-〓吋tftL-l一二一一一I 0-3'-_.JIW川 I_'~AI....L←ー ...L_~~.... ，

IU-' 100 IU' IU2

'-I]J 1111.剃日企IlGrIIloVI

1~1 ・..・ ωB嗣1

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-12111!rE • J日U.苫可11

UL2口町¥』戸田』

×コピ

EZ一¥H邑¥』回

¥E〕

広qdd望叫

Comparison of the calculated-to-measured L 7.3 l'ig.

value (C/E) rat10 for the euergy region-

wise flu瓦 forthe lithlum-oxlde slab Measured and calculated spectra

for the berylllum slab

1.7.7. Fig.

JAERI-M 88-221

1.8 Compilation of MCNP Data Library Based on JENDL-3T

K. Sakural, N. Sasaraoto, K. Kosako, T. Ishikawa"1. 0. Sato" 2, Y. Oyama, II. Narita, H. Maekawa and K. L'cki"3

For the purpose of benchmark test of JENDL-3T, a pointwise neutron cross section library was compiled based on JENDL-ST1**-', which contains 39 nuclides from 11-1 to Ara-241 which are often used for shielding calculation2'. Compilation was carried out at MCNP Data Library Compilation Subworking Group (SWG) in Research Committee on Reactor Physics of JAERI. A code system for cross section library compilation being used is the nuclear data processing system .YJOY-83 (modified version of NJOY) coupled with a newly developed library compilation code MACROS.

The compilation method employed was found to generate adequately the pointwise cross section library through analyses of quite simple neutron penetration problems with MCNP 3 >. Then, in order to certify reliability of the cross section library based on JENDL-3T, various benchmark experiments were analysed with MCKP. The benchmark experiments we analysed are 4' : 1) KFK Iron Experiment, 2) ORNL Sodium Experiment, 3) LLL Sphere Experiment and 4) WINFRITH Iron Experiment.

Comparisons of some MCNP spectra based on JENDL-3T with the measured ones are presented in Figs.1.8.1 through 1.8.4 for analyses of KFK iron experiment, WINFR.ITH iron experiment and LLL carbon and iron experiments, respectively. They are also compared with the spectra calculated using the ENDF/B-IV based library. Only the calculational results are given in the present report without further investigation on their differences. The comparisons show that the calcualtions using the library based on JENDL-3T give apparent differences from both other calculations and measurements. This means the analysis of benchmark experiments using pointwise data are useful for evaluation of JENDL-3T.

The present library includes only neutron transport and secondray

*1 Century Research Center Corp. Ltd »2 Mitsubishi Research Institute Inc. »3 Ship Research Institute •4 JENDL-3T is a temporary file for testing the evaluated data which

are for JEXDL-3. The data in JE.VDL-3T will be partly revised in JEN'DL-3.

]AERI-M 88‘221

1.8 Compilation of HCNP Data Library Based on JENDL-3T

K. Sakurai. N. Sasamoto. K. Kosako. T. Ishlkawa- 1• O. Sato- 2 •

Y. Oyama. 11. ~arìta. 11.河ackawaanu K. Lckl・3

For the purpose of benchmark test of JENDL-3T. a pointwise ncutron

cross scctlon 11brary was compiled based on JENDL-3T1)・4 which contains

39 nuclldes from 11-1 to A~-241 which are oftcn uscd for shicldlng

calculatlon2). Compllatlon was carrled out at MCNP Data Llbrary

Compilatlon Subworking Group (SWG) ln Research Committee on Reactor Physlcs

of JAERI. A code system for cross sectlon library compilation being uscd

Is the nuclear data processlng system汗JOY-83(modlfied verslon of NJOY)

coupled wlth a newly developed library compilation code i1.-¥CROS.

The compilation method employed was found to generate adequatelY

the polntwise cross section 11brary through analyses of quitc simple

neutron penetratlon problems wlth MCNp3). Then. in ordcr to certify

rcllability of the cross section library bascd on JENDL-3T. various

benchmark experlments were analysed with ~CNP. The bcnchmark experlmcnts we

analysed are4) 1) KFK Iron Experiment. 2) ORNL Sodium Expcriment. 3) LLL

Sphere Expcrimcnt and 4) WINFRITH Iron Expcrfmcnt.

Comparisons of some河CNPspectra based on JENDL田 3Twlth thc mcasured

oncs are presented in Figs.l.8.1 through 1.8.4 for analyses of KFK iron

experiment. WINFRITH lron experlment and LLL carbon and iron

experiments. respectively. Thcy are also compared with the spectra

calculated using the ENDF/B-IV based library. Only the calculatlonal

results are glven ln the present report without further investigation on

their differcnces. The comparisons show that the calcualtions using thc

library based on JENDL-3T glve apparent differenccs from both othcr

calculations and measurements. This means the analys1s of bcnchmark

experiments using pointwise data are useful for evaluation of JENDL-3T.

The present library includes only neutron transport and secondray

.1 Century Research Center Corp. Ltd

・2河itsubishiResearch Institute Inc.

・3Ship Research Institute

・4JENDL-3T is a temporary file for testing the evaluated data which

are for JEXDL-3. The data in JE~DL-3T will be partly revIsed 1n JEXDL-3.

J A E R I - M 8 8 - 2 2 1

gamma-ray production cross sections, we have a plan to generate and compile the final version of JEN'DL-3 based ] ibrary by adding S(u,S ) data and gamma-ray transport cross sections to it, at the time when JENDI.-3 will be released for general use.

References 1) JENDL Compilation Group (Nuclear Data Center, JAERI) : JENDL-3T,

private communication (1987). 2) Sakurai K. et al. : "Compilation of MCNP Data Library Based on

JENDL-3T and Its Test through Analysis of Benchmark Experiment." to be submitted to 7th International Conference on Radiation Shielding (1988).

3) MCNP : tfonte Carlo Neutron and Photon Transport Code, CCC-200 (1983).

4) Tanaka S. et al. : "Shielding Benchmark Problems." JAERI-M 7843 (1978) (in Japanese).

- 2 7 -

]AERI-M 88-221

gaRlma-ray production cross scctions. we h月vca plan to gcncratc and

compile thc final vcrsion of JEXDL-3 hascd ] lbrary by addlng S ( U， S) data

and gamma-ray transport cross scctions to it. at thc timc whcn JEXDI.-3

will bc released for gcneral use.

Rcferences

1) JENDL Compl1品tionGroup (Nuclear Data Ccnter. JAERI) JENDL-3T，

prlvate communlcatlon (1987).

2) Sakural K. et al. "Compl1atlon of河CNPData Llbrary Bascd on

JENDL-3T and Its Test through Analysis of Benchmark Experimcnt." to

be submltted to 7th International Confercnce on Radiation Shieldlng

(1988) .

3) MCNP Monte Carlo Neutron and Photon Transport Code， CCC-200

(1983) .

4) Tanaka S. et al. "Shieldlng Benchmark Problems，" JAERI-M 7843

(1978) (In Japanese).

J A E R 1 - M 8 8 - 2 2 1

K I K F« Ola. 1 0 c m

• Measured ~ ENOF/B-IVILANtl - ENDF/B-IVtSWGI • JEN0L-3T

I 0 3 I 0 6 10'

Neutron Energy (eV)

I I 10'

'jj i i i | 1 r—r - r r

Mr.NP ColculaUd (al 83,41cm depthl

JEN0L-3T ENDF-B/1V

o Eipfrlmfnl

ICT't l ' I I I . ' I I I I i i I I 10"' 10"' 10"

Neulron Energy (MeVl

Fig.1.8.1 Leakage neutron spectra in KfK iron experiment

Fig.1.8.2 Neutron spectra in WINFRITI! iron experiment

Carbon 2.9 m l . p 3 0 *

• Measured — ENDF/8-IV(LANL) — ENW7B-IV(SWG) • JEN0L-3T

A 6 B 10 12 14 IG

Neutron Energy (MeV)

Iron 4 . 8 m f p 3 0 °

• » • Measured ENOF/B-ENDF/B-

V t U N U V(SWGI

— JENDI-3T

^ % f si L 4 6 8 10 12 14

Neulron Energy ( M e V )

Fig.1.8.3 Leakage neutron spectra in I.LL enrbon experiment

Fig.1.8.4 Leakage neutron spectra in LLL iron experiment

- 2 8 -

JAERI-M 88・221

r-r一「

J1!idh ~.~O .~:

1.2!.豆主主旦皇陛且!i10

K 1 K F. 目10‘OcmM開制Itd

・-ENOFI目・IVILANLI-ENOF/8.IVI SWG 1 ・JVlOL・3T

-君、担当aEこE=JEZJEf

lげ'ー10‘

107 106

Neulron Energy (eV)

10! IO

Neutron spectru in

WIWRITI! Iron expcrlmcnt

FJg.1.8.2 Fig.l.8.1 Lenkugc ncutron spcctrn

In KfK iron cxperiment

Iron 4.8m fp 30・+ Meosu，.d . _. ENOF I目・ IV(lANLI - ENDFI日開IV(SWG)

... JENOl駒3T

E o ‘・2 国-

z 10-' ω 色』

‘・コお¥ 〉・22 、、5162

“ コtι

Carbon 2.9 m.l. p. 30.

争 M.osur.d一-ENOF/B.IV (LANLJ - ENOFI日・IVISWG) ... JENOL -3T

1j 'jt

LL.ーよ--L10-1 100 101

Neulron Enerqy (MeV)

‘' ::z .， i ldv

、、コー圭

16 4 6 8 10 12 14

Neulron Enerqy ( MeV ， 2

ドig.l.8.4 Lenkage neutron spectrn

in LLL iron experimpnt

-28ー

Fig.l.8.3 Leakage neutron spcctrn

in LLL c九rhoncxperimf'nt

J A E R J - M 8 8 - 2 2 1

1.9 Development of the JSSTDL-295 Group Cross Section Library System for Shielding Calculation

A. Hasegawa

The author developed the JSSTDL-295 (JENDL-Shielding STanDard Library-295) group cross section library system , which is composed of a common 295 group cross-section library and its utility codes, by defining the concept to be furnished as the standard shielding library. After the completion of the specifications of the library, we compiled 25 major nuclides for shielding applications into a library to use in the shielding benchmark test for JENDL-3T 2 ).

Because this library system has a character assigned as a common library in the shielding community for the standard use, it is recommended to prepare this type of library in no time delay at every new version release of the JENDL-file by JNDC (Japanese Nuclear Data Committee). 1. Library specifications a. For what codes: Though so many transport codes are used in

Japan, those commonly used in the shielding community of Japan are selected as the library to be prepared. Those are Sn code like ANISN 3' or D0T 4 ) and Monte Carlo code MORSE-CG 5 ).

b. Energy group structure: Surveys are made to cover almost all energy group structures currently used in Japan. From all group structures listed in Table 1.9.1. a common 295-energy group structure is defined by regrouping, i.e., all the energy groups within the difference of 0,3 % between the adjacent groups are bunched to one representative energy.

Recurrent use of this library for the coarse group library production is guarantee! by the utility code in this system due to the following reason. This library possesses all weighting fluxes for self shielding factors as well as for the infinite dilution cross-section. Therefore we can produce libraries for any energy group structures of the libraries listed in Table 1.9.1 (except thermal library part for GAM, WIMS and MGCL). c. Weighting functions: For thermal energyffron 10 eV up to 0.3224 eV)

Maxwellian spectrun is used, above this energy 1/E spectrun is used. For

)AER 1・M 88・221

1.9 Development of the JSSTDL-295 Group Cross Section Library

System for Shie1ding Calcu1ation

A. Hasegawa

The author deve10ped the JSSTDL-295 (JE~DL-Shielèing STanDard

Library-295) group cross section 11brary system1). wh1ch 1s composed of

a comrnon 295 group cross-sect!on 11brary and its uti1ity codes. by

deflning the concept to be furnished as the standard shie1ding 11brary.

After the comp1etlon of the speclflcatlons of the library. we cornpi1ed 25

major nuc11des for shie1ding app11cations into a 1ibrary to use 1n the

shie1ding benchmark test for JENDL事::lT2)

Because this 1ibrary system has a character assigned as a common

library in the shie1ding comm'm1ty for the standard use， it is

recommended to prepare this type of 1ibrary in no time de1ay at every

new version re1ease of the JENDL-f11e by JNDC (Japanese Nuc1ear Data

Committee) .

1. Library spec1f1cat10ns

a. For what codes: Though so many transport codes are used 1n

Japan. those common1y used 1n the shle1dlng commun1ty of Japan are

selected a~ the 11brary to be prepared. Those are Sn code 11ke ANISN3) or

DOT4) and Monte Car10 code MORSE-CG5).

b. Energy group structure: Surveys are made to cover a1most a11 energy

group structures current1y used 1n Japan. From a11 group structures

1isted 1n Tab1e 1.9.1， a COffimon 295・energygroup structure 1s defined by

regrouping， 1.e.， a11 the energy groups within the d1fference of 0.3 %

between the adjacent groups are bunched to one representative energy.

Recurrent use G~ this 1ibrary for the coarse group 11brary

production 1s guaranteei by the uti1lty code 1n this system due to the

following reason. This library possesses a11宵e1ghtingf1uxes for self

shielding factors as we11 as for the infin1te dl1ution cross-section.

Therefore we can produce 1ibraries for any energy group structures of the

libraries listed in Table 1.9.1 (except thermal 1ibrary part for GAM.

WDIS and ~GCL).

c. Weighting functions: For therma1 energy(from 10・5eyup to 0.3224 eV)

Maxwellian spectrum is used， above this energy 1/E spectru. is used. For

J A E R 1 - M 8 8 - 2 2 1

the self-shielding- factor calculation, usual 1/((T£*(J )/E spectrum is used. d. Self-shielding factors: Following 9 points are used as back ground

cross section grid points: 0., 0.1778, 1.0, 10., 100., 10 3, 10 4, .105 and 10 6. Prepared temperatures are for 300, 600, 900 and 2100 K. Self-shielding tables are calculated for total, fission, capture(for (n,gamma) only) and elastic.

e. Scattering matrices: Up to P 5 matrices are considered for the anisotropy. And two independent scattering matrices are stored, i.e., one for elastic scattering and the other is for non-elastic. This separate storing is due to the convenience to the self-shielding factor calculation for elastic matrices. 2. Utility codes developed a. CONDNSJ : for the group collapsing of the 295-group common library

to the user requested one. b. MACROJ : generating of region dependent macroscopic cross-section

file directly used in ANISN or DOT. In this code,effective cross sections are calculated by the interpolation of the self-shielding factors in the JSSTDL library.

c. CON'VJSS: conversion code from BCD to Binary or vice versa for the JSSTDL-295 common library. This code is developed for the transportability of the library (which is a binary file) to the different computer system.

Interrelation between these codes and library is given in Fig.1.9.1. 3. Production of the JSSTDL-295/JENDL-3T conon group library

We produced a JSSTDL-295 common group library from JENDL-3T file for the calculation of the benchmark test of the JENDL-3T. Processed nuclides and their code numbers are listed in Table 1.9.2. At the same time for the sake of users of JSD-100 group structure library, we also produced a JSSTDL-100/JENDL-3T coarse group library by CONDNSJ.

References 1) Hasegawa A. : to be published as JAERI-M report. 2) JENDL compilation group (N'uclear Data Center, JAERI): JEXDL-3T

private communication (1987). 3) Engle W.W.Jr. : K-1693 (ORNL) (1967). 4) Rhoades W.A. and Mynatt F.R. : ORNL-TM-4280 (1973). 5) Straker E.A., Stevens P.N., et al.: ORNL-TM-2242 (1968).

- 3 0 -

]AERI・M 88・221

the self-shieldlng factor calculation. usual 1/(ðt~句 )/E spectrum is used

d. Self-shielding factors: Following 9 points are used as back ground 円-1 ，，，5

cross section grid points: 0.. 0.1778. 1.0， 10.， 100.， 100， 10~. .100 and

106. Prepared temperatures are for 300， 600. 900 and 2100 K. Se1f-

shie1ding tables are calculated for total. f1ssion. capture(for (n.garnma)

only) and elastlc.

e. Scattering matrlces: Up to PS matrlces are considered for the

anisotropy. And two lndependent scatter1ng matrices are stored. 1.e.. one

for elastic scatter1ng and the other 1s for non-elast1c. This separatc

storing 1s due to the convenience to the self-shielding factor

calculation for elast1c matr1ces.

2. Utility codcs dcvcloped

a. CONDNSJ for the group collapsing of the 295-group common library

to the user requested one.

b. ~~CROJ generating of region dependent macroscopic cross-section

file directly used 1n AN1SN or DOT. 1n this code.effective

cross sections are calculated by the 1ntecpolation of the

self-sh1elding factors 1n the JSSTDL 11brary.

c. CONVJSS: conversion code from BCD to B1nary or vice versa for the

JSSTDL-295 cornmon 11brary. Th1s code 1s developed for the

transportability of the library (which is a binary fl1e)

to the different computer systern.

1nterrelation bet胃eenthese codes and library 1s given 1n Fig.l.9.1.

3. Production of thc JSSTDL-295/JENDL-3T co..on group library

胃eproduc巴da JSSTDL-295 common group library frorn JENDL-3T file for

the calculation of the benchmark test of the JENDL・・3T.Processed nuclides

and their code nurnbers are 11sted 1n Table 1.9.2. At the same tirne for

the sake of users of JSD-100 group structure library. we a1so produced a

JSSTDL-IOO/JENDL-3T coarse group library by CONDNSJ.

References

1) Hasegawa A. to be published as JAERI四月 report.

2) JE1DL compilation group (Nuclear Data Center， JAERI): JENDL-3T

private communication (1987).

3) Engle胃.W.Jr. K-1693 (ORNL) (1967).

4) Rhoades W.A. and Mynatt F.R. ORNL-TM-4280 (1973).

5) Straker E.A.， Stevens P.N.， et al.: ORNL-T河ー2242(1968).

JAERI-M 88-221

J S S T O L - 2 9 5

CONVJSS

CONONSJ

MACRQJ

COARSE GROUP LIBRARY

b i n a r y f i l e

R e j i o n d e p e n d e n t MACRO f i l e

f o r A N I S N , DOT and flGRSE

F i g . 1.9.1 I n t e r r e l a t i o n between the JSSTDL-295 l i b r a r y and

i t s u t i l i t y codes

Tablt 1.9.1 Lisc of group scructurfs currtacly used ia ca« labia 1.9.2 List of nuclidts availnolt i.l shialdiny or raactor design fields ia Japan JSSTDL-295/J3T librarv

Lib. nama aumbar of groups

J S D - 1 0 0 100

JSD-10O0 100

SERMUDA-121 121

FNS-125 125

VITAMI>fE-C 17L

- J (E+C) ns G I C X - 4 2 42

ABBN-2S 25

JFS-iVa" 70

GAM-123 ( f a s c on l y ) 92

M G C L - 1 3 7 { f a s t on iy> 91

WIMS-63 ( f a s c on l y ) 28

auc l id i -9od i aue i td i

I I I S L i - I 3031

: 375 L i - 7 3032

] t » S 5 . - 3 3041

i SOS a -;o 3091

i I I S B - ' .1 2032

5 . 0 5 c - ;2 30* 1

r I SOS A I - : T 3131

i 2405 Cr 3240

3 3SSS Ma-SJ 3211

10 2105 X t 3210

LI :>os C« ) 2 > 0

12 . 2 0 3 Mo 3420

13 U S H - 1 SOU

14 ;:s H - : 3012

IS :<s H i - . 1022

19 105 0 - i f 3011

n :«o5 r> 3210

l i 1109 . f . - :3 3111

19 1.05 Si 3140

:o 1203 ?» 3120

2 ! 9299 V -!25 3924

:z 92JS •J - m 3926

33 3495 p-j-239 :943

:4 J . 0 3 ? - ; - :43 1944

:! 3 . 1 9 ? - i - : . i 3949

- 3 1 -

JAERI-M 88・221

ca刊V}SS巴斗CONDNSJ

3 MACROJ

file Recion dependent 阿ACRO

for ANlSN. DOT and 月Qj(SE

rnterrelation between the JSSTDL-295 Iibrary and

its utility codes

1. 9.1

Tabi・1.9. List ofπuc 1 id・S ''''&1 I c.o l・1:1

JSSTDL・:95，J3T 1 i 'orary

List QC iroup structur・scu.rr・ntlyus.d in tb・5bi・ldinll0< <・&ctord・silfIlci・ldsin Jap&D

Tabl・l.9.1

巴wnb・ro( it"oups

-az・

，.，‘，azeda--

J.、，

Lib ‘o.am.

JSD-IOO

JSD-IOOO

己ERMUDA-121

FNS-125

VIT.んMlNE-C

-J (E+C)

G1CX-42

A1lBN-2S

JFS-N.官

GA凶M-!23 (fast 0111y)

!.1GCL-l37{fast 0111y)

Wl!.1S-69 (fast 0111y)

J A E R I - M 88-221

2. Theoretical Method and Code Development

Continuous efforts have been devoted to the theoretical m o d e l of nuclear spallation reaction. The dependence of m a s s - n u m b e r distribution of spallation products on the level density p a r a m e t e r a was investigated using the U n o & Y a m a d a ' s m a s s f o r m u l a and the Le Contour's equation for the value of a. The particle evaporation was calculated according to the W e i s s k o p f ' s statistical m o d e l , where the level density param e t e r a plays the most important role. On the other hand, an i m p r o v e m e n t w o r k are under way for the one-point depletion code D C H A I N and its nuclear data library to be able to treat all the build-up and decay s c h e m e s including spallation products. For this purpose, the new decay data for spallation products due to the pro t on—i nduced T R U tr a n s m u t a t i o n were computed by using a code based on the gross theory of beta decay. The spallation yields were also calculated for typical target nuclei using the spallation reaction s i m u l a t i o n code N U C L E U S . In addition, a co m p u t e r code s i m u l a t i n g the intranuclear cascade reaction has been m o d i f i e d to eli m i n a t e the discrepancies b e t w e e n the calculated and experimental results for both the spallation neutron spectra and the m a s s yields of residual nuclei. H e r e , the effects of the nuclear-cluster collision are included in the intranuclear cascade under consideration of the Pauli's exclusion principle.

D e v e l o p m e n t and i m p r o v e m e n t of c o m p u t a t i o n codes continue to be carried out. A vectorized M o n t e Carlo code G M V P has been developed, where the stack-driven a l g o r i t h m is adopted for the vectorization. Calculationa1 tasks in r a n d o m w a l k s i m u l a t i o n are divided into free flight analysis, next zone search and collision analysis. By this vector calculation, the factor of speed—up 8 ~ 5 has been achieved. The vectorization rate of G M V P is 90 ~ 96 percent, depending on p r o b l e m s . Standardization and consolidation is now going on for the shielding code a i m i n g at the easy and a u t o m a t i c use of the transport codes and the proper selection of nuclear data. In this w o r k , use is m a d e of the advanced computational technology such as "artificial intelligence", "computer aided design" or "computer graphics". (Yukio Ishiguro)

- 3 2 -

JAERI-M 88・221

Continuous efforts have been d申voted to the theoretical

model of nuclear spallation reac~ion. The dependence of

mass-number distribution of spallation products on the level

density parameter a was investigated using the Uno & Vamada's mass formula and the Le Conteur's equation for the value of

a. The particle evaporation was calculated a.ccording to the

Weisskopf's statistical model， where the level density para-

m喧 ter a plays the most important role. On the other hand，

an improvement work are under way for the one-point depletion

code DCHA1N and its nuclear data library to be able to treat

al1 the build-up and decay schemes including spallation

products. For this purpose， the new decay data for spallation

products due to the proton-induced TRU transmutation were

computed by us ing a code based on the gross theory of beta

dccay. The spallation yields were also calculated for typical

target nuclei using the spallation reaction simulation code

NUCLEUS. [n addition， a computer code simulating the intra-

nuclear cascade reaction has been modified to eliminate the

discrepancies between the calculated and experimental results

for both the spallation neutron spectra and the mass yields

of residual nuclei. Here， the effects of the nuclear-cluster

collision are included in the intranuclear cascade under con-

sideration of the Pa.uli's exclusion principle.

Deve lopment and improvement of computat ion codes cont i-

nue to be car..ied out. A vectorized Monte Carlo code GMVP

has been developed， where the stack-driven algorithm is

adopted for the vectorization. Calculational tasks in random

walk simulation are divided into fr自e flight analysis， next

zone search and collision analysis. By this veci:or

calculation， the factor of speed-up 8 ~ 9 has been achieved.

The vectorization rate of GMVP is 90 - 96 percent， depending

OD problems. Standardization and consolidation is now going

o n f 0 r t h e s h i e 1 d i n g C 0 d e a i m i n g a t t h e e a s y a n d a u t 0 ma t i c

use of the transport codes and the proper selection of nuclear

data. 1n t.his work， use is made of the advanced computational

technology such as "artificial intelligence ぺ ~computer aided

design" or .computer graphics". ( Y u k i 0 1 s h i gu r 0 )

J A E R I - M 88-221

2.1 Dependence of Spallation Produce Distributions on the Level Density Parameter

T. Nishida, Y. Nakahara and T. Tsutsui

After proton-induced nuclear spallation reactions, a variety of nuclei, especially many neutron-deficient nuclides with mass numbers greater than 80, are produced." From the transmutation point of view, the reliable estimate of spallation product yields is very important and its reliability critically depends on the calculative model used in the Monte Carlo simulation NUCLEUS code.2' In the present report examinations are carried out to know how the spallation product yield varies on the variation of the level density parameter a in the evaporation stage of nuclear reaction induced by a 500 MeV proton bombarding' on a Np-237 nucl eus (Transuranic Waste). In the present calculations the Uno & Yamada's new mass formula3)'4) has been employed.

In Fig.2.1.1 the mass distribution of spallation products of both residual nuclides and some particles from a Np-237 nucleus bombarded by protons of 500 MeV are examined in more detail by evaluating the contribution of level density parameter a to the evaporation calculation. A sharp peak of highly excited residual nuclides appears in the mass number region near the target nucleus just after the nucleon intranuclear cascade in the fast reaction has ceased as seen in Fig. 2.1.1 (a), and then the peak decays through the slow reaction process to the broad distribution as shown in Fig. 2.1.1(b). These products are redistributed to two components of fission products and non-fission ones (almost particle evaporation) respectively as seen in Fig. 2.1.1 (c) and (d). The particle evaporation is calculated according to the Weisskopf's statistical model, in which the level density parameter a

plays the most important role. The value of a had been determined to be A/10 and A/20, where A implies the mass number, in fitting the measured data by Dostrovsky et al., Barashenkov et al . and Chen et al. In NUCLEUS the Le Conteur's equation was employed to obtain the value of a as follows,

a—gil + y —g )

where B and y are 8 MeV and 1.5 respectively. This equation gives A/7 . 7~A/7 . 4 to the value of a for the nuclides

with the mass number greater than 200. The number of particles

]AERI-M 88・221

2.1 Dependence of Spallation Product Distributions on the

Level Density Parameter

T. Nishida， Y. Nakahara and T. Tsutsui

After proton-induced nuclear spallation reactions， a variety of

ouclei， especially many neutron-defici自nt nucl ides wi th mass numbers

greater than 80， are produced.1l From the transmutation point of view，

the reliable estimate of spallation product yields is very impol.tant

and its r司liabilitycritically depeods on the calculative model u泊 ed in

tヒe Monte Carlo simulation NUCLEUS code.2l In the pr自sent r自port

examinatioo.s are carried out to know how the spallation product yield

V日 ies on the variatioo of the 自vel density parameterαin the

e¥aporation stage of nuclear reaction ioduced by a 500 MeV protoo

bombardio.g 00. a Np-237 nucleus(Transuranic Waste). 10. the present

calculations the Uoo & Yamada's new mass formula3J • 叫 has be eo. emp 1 oyed.

In Fig.2.1.1 the mass distribution of spallation products of both

residual nuclides and some particles from a Np-237 nucleus bombarded

by protons of 500 MeV are examined in more detai 1 by evaluating the

contribution of level deo.sity parameter a to the evaporation

calculation. A sharp peak of highly excited residual nuclid自sappears

io. the m且sso.umber region near the target nucleus just after the nucleon

iotranuclear c且scade in the fast reaction hιs ceased a::; se自n io Fig ・

2.1.1 (a)， and theo the peak decays through the slow reaction process

to the broad distribution as shown io. Fig. 2.1.1(b). These products

are redistributed to two componeo.ts of fission products and non-fission

ones (almost particle evaporation) respectively as seen in Fig. 2.1.1

(c) ao.d (d). The particle evaporation is calculated according to the

Weisskopf's statistic且 1model. in which the 1自vel density parameter a

plays the most important role. The value of a had been determined

to be AIlO and A/20， where A implies the mass number， in fitting the

measured data by Dostrovsky et al.， Barashenkov et al. and Chen et al.

In NUCLEUS the Le Conteur・sequation was employed to obtain the value

of a as follows，

A，". (A-2Z)2 a-E(l + Y ~ー}

where B and y are 8 MeV and 1.5 respectively.

This equat ion gi ves AI7. 7-A/7. 4 to th由 valueof a for the nuclides

wi th the mass number greater th且n 200. The number of particles

-33ー

J A E R I - M 88-221

evaporated from the non-fission component of products is calculated including ..4/30 and A/5 to these parameter values. Table 2.1.1 summarizes ratios of the number of each particle for five parameter values to one calculated by the equation, where a figure in the parenthesis represents the number of evaporated particles. It is apparent that the yields of neutron and proton decrease by about thirty percents as a decreases to A/20 - A/30 but increase by ten percents with a = A/5. For other particles, the inverse tendency is seen and their yields have the wider tolerances than in cases of proton and neutron. Figure 2.1.2 shows the variation of the non-fission component in the mass distribution of products from a J 3 7 N p nucleus bombarded by protons with the energy of 500 M e V in changing the a value. Although the spires vary relatively rapidly by magnitude from 0.6 to 0.1, the basic pattern of distribution is invariant for each case.

Re ferences

1) N i s h i d a T. and Nakahara Y. : Kernf c^hnik, 50, 193 (1987 ) .

2) N i s h i d a T . , Nakahara Y. and T s u t s u i T. : " Deve lopment of a

N u c l e a r S p a l l a t i o n S i m u l a t i o n Code and C a l c u l a t i o n s of P r i m a r y

S p a l l a t i o n P r o d u c t , " JAERI-M 86-116 (1986) ( i n J a p a n e s e ) .

3) Uno M. and Yamada M. : " A t o m i c Mass F o r m u l a w i t h Cons t an t

S h e l l T e r m s , " P r o g . T h e o r . Phys . , J_5 . ,No.4 , 1322(1981) .

4) N i s h i d a T . , Nakahara Y. and T s u t s u i T. : " A n a l y s i s of the

Mass F o r m u l a Dependence of S p a l l a t i o n Product D i s t r i b u t i o n , "

JAERI-M 87-088 ( 1 9 8 7 ) .

Table 2 . 1 . 1

R a t i o s of p a r t i c l e s e m i t t e d from a Neptunium-237 nucleus bombarded by protons w i t h 500 MeV in the n o n - f i s s i o n component

Level Densi ty A/30 A/20 A/10 Le Countour A/5 Para, a (A/7.7 - A/7.4)

Proton 0-70 0.71 0.89 1. (1.572) 1.12

Neutron 0.68 0.77 0.95 1. (7.412) 1.08 Deutron 2.30 1.9S 1.17 1. (0.233) 0.47

Triton 4.61 3.47 1.52 1. (0.085) 0.39 Xeliua 3 11.92 6.72 1.61 1. (0.0036) 0.17 Alpha 2.68 2.24 1.17 1. (0.121) 0.37

Kucleons/P 0.9S 0.96 0.98 1.(10.200) 1.01

- 3 4 -

JAERI-M 88・221

evaporated from the non-rission component of products is calculated

including A/30 且nd A/5 to these par且meter values. Table 2.1.1

summari z自s ratios of the number of each particl自 for f i ve parame ter

values to one calculated by the equat ion. where a f igure in the

parenthesis represents the number of evaporated particles. It is

apparent that th自 yields of neutron and proton decrease by about thirty

perc自ntsas a decreases to A/20 -A/30 but increase by t自npercents wi th

a = A/5. For other particles. the inverse 自ndency is seen and their

yi自 ldshave the wider toierances than in cases of protoロ andneutron.

Figure 2.1.2 shows the variation of the non-fission component in the

mass distribution of products from a 237Np nucleus bombarded by protons

wi th the energy of 500 MeV in changing th自 αvalue. Al though the spi res

vary relatively rapidly by magnitude from 0.6 to 0.1. the basic pattern

of distribution is invariant for each case.

References

1) Nishida T. and Nakahara Y. Kernt8<..hnik. 50.193 (1987).

2) Nishida T.. Nakah且raY. and TsutS'li T. " Deve lopment of a

Nucle.e.r Spallation Simulation Code and Calculations of Primary

Spallation Product." JAERI-M 86-116 (1986) (in Japanes自).

3) Uno M. and Yamada M. ，. Atomic Mass Formula with Constant

Shell Terms." Prog. Theor. Phys.， ll.No.4. 1322(1981).

4) Nishida T.. Nakahara Y. and Tsutsui T. "Analysis of the

Mass Formula Dependence of Spallation Product Distribution."

JAERI -M 87-088 (1987).

Table 2.1.1

Ratios of particles emi tted from a Neptunium-237 nucleus bombarded by protons with 500 MeV in the non-fission component

Leve I Dens i h A/30 A/20 AI10 Le Countour A/5

Para. a (AI1 . 7 -AI7. 4)

Proton 0.70 0.71 0.89 1. (1.572) 1.12

Neutron 0.68 0.77 0.95 1. (7.412) 1.08

Deulron 2.30 1.95 1.17 1. (0.233) 0.41

Tr i ton 4.61 3.47 1.52 1. (0.085) 0.39

Heliu 3 11.92 6.72 1.61 1. (0.0036) 0.17

Alpha 2.68 2.24 1.17 1. (0.121) 0.37

lIucleons/P 0.95 0.96 0.98 1. 00.200) 1.01

SOT lOO. o 150 -

F i g . 2 . 1 . 1 Mass y i e l d d i s t r i b u t i o n s of products a f t e r the f i r s t s t e p (a) and the second s tep ( b ) , and the l e t t e r ' s f i s s i o n component ( c ) and n o n - f i s s i o n component (d) for a Neptunium-237 nucleus bombarded by 500 MeV protons

L e C o a l e a r

"117 I d a . A iio.

too. p iSo. 2 M . !S

F i g . 2 . 1 . 2 The n o n - f i s s i o n component of mass y i e l d d i s t r i b u t i o n in the nuclear s p a l l a t i o n r e a c t i o n of a Neptunium-237 nucleus bombarded by 500 MeV protons Level d e n s i t y parameter :

a) A / 30 . b) A / 20, c) A / 10. d) Le Couteur and • ) A / 5

0.2 。・2

-aJ'E .. ‘

ミZEJ匂g

F1“自白

E-、.wgHdν28

.，> 1I'SUP

‘、"・・Mdus--

0.' 0.¥

Ali . ，..ノao

0.0 O.

0.1 11

五 JJo.0。0.0

2S0. 200. 同，50.，Oo 5O. R ，SQ. 両O.百:0・6

』〉何均一-三∞∞sNN-

• A/l0 a

0.0 O.

d.) Non-liuicn

a ・'"~ 11

_l 2nd SI!p

'.11・2 「ーー

0.0 O. 50.

Fig ・ 2.1 . 1

Mass yield distributions of produc!s after the first stop (且)and the second slep (b)， and the latter's Cission c白mponent(c) 且ndnon-C ission component (d) for a Neptunium-237 nucleus bombarded by 500 Me V pro tons

200 A 150. ‘ue. 50.

iaE110唱

。。aW 0

a 150 I~O.

.) A I 5

Fig. 2.1.2 The non-fission component oC m且S5 yield .distribution in the nuclear spallation reaction oC a Neptunium-237 nucl・UIIbombarded by 500 MeV protons Lev・1d・nsi ty param・tera) A I 30， b) A / 20， c) A / 10， d) Le Cout.ur and

J A E R 1 - M 8 8 - 2 2 1

2.2 Evolution Analysis of Spallation Products in Their Buildup and Decay Processes

Y . Nakahara, T. Nishida and T. Tsutsui

For the T R U W (Trans Uranic W a s t e ) incineration study it is important to obtain the basic knowledge concerning the transmutation of radioactive wastes w i t h long lifetime to the ones w i t h short lifetime or stable ones, using high energy protons generated by a linear accelerator n . In the present report the time evolution calculation of buildup and decay of all the nuclides produced in the target irradiated continuously by proton beams are intended. H o w e v e r there are no code which can m a k e the calculation possible due to the enormous consumption of computing time and computer m e m o r y , and the insufficient compilation of nuclear data in the energy range higher than 20 M e V for nuclides.

W e use a calculative model based on the one point approximation for these processes and have slightly modified its formula given in the previous annual report, as represented in E q . ( 1 ) . This equation includes a n e w term and is rearranged as

i "V" / * \ -? ft n

J, ; - E li.j*jXj{t) + £ ki.r*Httt,j-Xj-(t) + £ ai.j*p'..jXiit)

+ riF(t) + BiG(.t) - (it + 4pos.i + <fif,ac,i)Xi(t), (1) where i = (A,Z), j = (A',Z') and /' = (A'-1,Z") with mass number A and atomic number Z,

Xi(t) : the concentration of nuclide i at the time t, Xi : the decay constant of nuclide i,

lij : the branching ratio for the production of nuclide t from the decay of nuclide j ,

F{t) : fission rate at the time t. n fission yield for nuclide i,

G { t ) : spallation rate at the time t, Pi : spallation yield for nuclide i,

kij the branching ratio for the production of nuclide i from the reaction of nuclide / w i t h neutron flux,

ff,.; : the production rate of nuclide i due to the spallation of nuclide j ,

$y : neutron flux ,

- 3 6 -

1AERI-M 88・221

2.2 Evolution Analysis of Spallation products in Their

Buildup and Decay Processes

Y. Nakahara， T. Nishida and T. Tsutsui

For the TRUW (Trans Uranic Waste) iIlcineration study it is

important to obtain the basic knowledg自 concerning the transmutat ion

oC radioactive wastes with long liCetime to the ones with short lifetim自

or stable ones， using high energy protons generated by a linear

accelerator 1) In the present report the time evolution calculation

oC buildup and decay of all the nuclides produced in the target ir・radiated

continuously by proton beams are intended. However there are no code

which can make the calculation possible due to the enormous consumption

oC computing time and computer memory， and the insuCCici叩 t

compilation of nuclear data in the energy range higher than 20 MeV for

nuclides.

We use a calculative model based on the one point approximation

for these processes and have slightly modiCied its Cormula giv自n in the

previous annual report， as repres自nted in Eq.(l).

includes a new term and is rearranged as

This equation

dX‘(t)苧寸「ーん I川ばJ(t)+JZlhaj eNUC J XJ(t)+z hJφpa..jXj( t)

+ riF(t) + /JιG( 0 ー (Ai +φpa..ι+φNac.i)X‘(0， (1)

where (A.Z). j (A' ，Z') and j' (A・一1，Z ・) wi th mass number A

and atomic number Z，

Xi(t) the concentration oC nuclide at the time t，

Ai the decay constant of nuclide i，

li， i the branching ratio fロrthe production of nuclide

from the decay oC nuclide j，

F(t) Cission rate at th自 time t.

Y‘ fission yield Cor nuclide i.

G(t) spallation rate at the time t，

/J. spallation yield Cor nuclide i.

k..i the branching ratio for the production oC nuclide

Crom the reaction of nuclide j with neutron flux.

g ‘・ theproduction rate of nuclide due to the spa1.1ation

of nuclide j.

ON neutron Clux •

一部一

J A E R I - M 8 8 - 2 2 1

$P : incident proton flux , oc,i neutron capture cross section of nuclide i,

i)j,i total spallation cross section of nuclide i.

( i = 1. 2 n) This equation can be approximated for the target nuclide as

**lil± ( j r + • J* e. r+» w*,.j. )XT(t) (2)

and can be easily solved as the following' exponential function

XT(t) =XT(0)exp(-(XT + <ti*oe.T + *pO,.T) O (3) where

X r { t ) : the concentration of target nuclide at time £, .XY(O) : the concentration of target nuclide at t = 0,

XT the decay constant of target nuclide, ac_T • neutron capture cross section of target nuclide, as,r •" total spallation cross section of target nuclide.

Then the functions of F(t) and G(t) are given using the analytic function <XY(0 as follows,

F(t)-ffF0yXr(O)exp(-(AT + QNOI.T + QPOS.T) t) (4)

G{t)-os.T<t>pXr(0)exp(-(AT + $N<>c.T+<t>p<>*.T)<-) (5)

where Op is neutron fission cross section of target nuclide. The one-point depletion code D C H A I N and its suppl emenatal

nuclear data library 3 ) had been developed in JAERI on the basis of the Bateman method for the calculations of decay and built-up of fission products in a nuclear reactor. This method assumes that any complicated decay process of produced nuclides can be disintegrated to linearized decay chains. On the basis of the formulation as described above, this code is being improved to be able to calculate all the build-up and decay schemes including spallation products. The nuclear data in the energy range higher than 20 M e V are being compiled also. In the present work the preliminary computations were curried out to prepare unknown some input data for running the time evolution code. The beta decay constants of spallation products from the T R U transmutation have been computed using the gross theory code of beta decay. 4 1 The spallation yields 8 were calculated for target nuclei such as Np-237. Pu-238, Pu-239, Pu-240, Am-241. Am-242. Cm-242 and Cm-244

- 3 7 -

JAE R 1・M 88・221

op incident proton flux .

d.・‘ neutroncapture cross section of nuclide i.

d..i total spallation cross section of nuclide i.

1. 2..... n)

This equation can be approximated for the targ自 tnuclide as

dXr(t) ーす7ー園田 (Ar + φpd..r+ONI1..r )Xr( t) ( 2)

and can be easily solved as the following 自xponential function

Xr( t) =Xr( 0) exp(ー(Ar + φ"，d..T+φpd.. T) t ) (3)

Xr(t) the concentration of target nuclide at time t.

XT(O) the concentration of target nuclide at t O.

AT the decay constant of target nuclide.

d..T neutron capture cross section of target nuclide.

ds.T total spallation cross section of target nuclide.

Then the functions of F(t) and G(t) are given using the analytic

function Xr(t) as follows.

F(t)・dFO"，Xr(O)exp(ーOT + φ"，O..T+φpO.. T) t )

G(t)-I1..TφpXr( 0) exp(一(AT + φ"，d..T+φpd s. T) t )

where dF is neutron fission cross section of target nuclide.

The one-point deplet ion code DCHAIN幻 and i ts suppl emenatal

nuclear data libraryωhad been developed in JAERI on the basis of the

Bateman method for the calculations of decay and built-up of fi5sion

products in a nuclear reactor. This method assumes that any

complicated decay process of produced nuclides can be disintegrated to

linearized decay chains. On the basis of the formulation as described

above. this code is being improved to be able to calcu!ate all the

build-up and decay schemes including spallation products. The nuclear

data in the energy range higher than 20 MeV are be ing compi 1 ed aI 50.

In the present work the preliminary computations were curried out to

prepare unknown some input data for running the time evolution code.

The beta decay constants of spallation products from the TRU

transmutation have been computed using the gross theory code of beta

decay.4' The spallation yields 8 were calculated for target nuclei such

as Np-237. Pu-238. Pu-239. Pu-240. Am-241. Am-242. Cm-242 and Cm-244

"' 内

jAF.RI-M 48-221

in the i n c i d e n t pro ton energy ra is ;? . 0.1 GeV to 1.5 GeV. ' j s in? the

s p a l l a t i o n r e a c t i o n s i m u l a t i o n code .VUCLEUS''

R e f e r e n c e

1) M i s h i d a T. . Makahara Y. and Tsu t su i T. " Development of a

Muclaa r S p a l l a t i o n R e a c t i o n S i m u l a t i o n Coda and C a l s u l a t i a n s

of P r i m a r y S p a l l a t i o n P r o d u c t s , " JAERI-M 38-116 (1936)

( i n J a p a n e s e ) .

2) Tasaka K . : " DCHAIN2: A Computer Coda for C a l c u l a t i o n of

T r a n s m u t a t i o n of N u c l i d e s . " JAERl-M 8727 ( 1 9 3 0 ) .

3) Tasaka K . , I h a r a H . , A k i y a m a M. ,Yosh ida T . , M a t s u m o t o 2. and

N a k a s h i m a R. : "JNDC .Vuclaar Data L i b r a r y of F i s s i o n P r o d u c t s , "

JAERI 1287(1983).

4) Yosh ida T. : " GROSS-M and GROSS-P. Codas for P r a d i i t i a n of

Be ta -Daoay P r o p a r t i a s and tha Eva lua t ion of t h e i r A p p l i c a b i l i t y

to Dacay Haat C a l c u l a t i o n s . " JAERI-M 36-116 ( 1 9 8 8 ) ( i a Japanaaa)

- 3 8 -

J A F. R [-'vl ヨヨ -~21

in the incident proton 弓π9rgy ra..，g8目ヨ G3V ゥ:; GdV， 'Hing the

spallat ion raa.c: iつn3 i mu; a.: i on code :>il.jCLEGS' I

Referenca

1) Nishid且 T" :-Ta.ka.ha.r旦 y， and T9UtSU! T 喝 Devaloprnant of a.

Nuclea.r Spa.llation Reaction Simulation Code and Cal士ulations

口rPrimarySp且llationProduct3，咽 JAER[-:vr35-115 (，936)

(in Japanese)

2) Tasaka K.: .. DCHAIN2: A Computor Coda ~or Ca.lcul且tlヨn01

Transmutat i司nof :-Tucl ides，時 JAER 1 -~r 3 7 ~ 7 (1ヨ80)

3) Tas且kaK.， lha.ra. H. ，Akiy且maM. ，Yoshida T. ，:vratヨumotaZ 主nd

Nakashima R. 時 JNDC:-Tuclear Data Liht・aryof Fi~ 圭 L 白n Products，"

J~I 1287(1983)

4) Yoshida T. 喝 GROss-:vr旦ndGROSS-P， Codes far Pr 過di~tion o(

Be ta-Deoay Prロparties 且nd tha Eva.lullt ion 01' thei r Appl icabi 1 i ty

to Dacay Heat Ca~culations ，" JAERI-M 86-116 (1986)(io J亀pao~sa)

J A E R I - M 8 8 - 2 2 1

H.Takada. Y.Nakahara and K.Ishibashi"

There are some discrepancies between calculated and experimental results for both the spallation neutron spectra and the mass yields of residual nuclei. 1 ) J ) We have made some modifications in the intranuclear cascade model to eliminate the discrepancies. They are the introduction of the nuc1eon-cluster collision, the correction of the angular distribution of the delta particle and the incorporation of the exciton mode 1.

Nuclear reactions in the intranuclear cascade process are treated as a sequence of nuc 1 eon-nuc leon two body collision. In case of the nucleon energy comes below 200 MeV, the de Broglie wave length becomes longer than 2 fm which is equal to the mean distance between the two nucleons in a nucleus. Consequently, the approximation of the two body collision is not valid in this energy range. In addition, it is suggested that the mean free path evaluated from the imaginary part of the optical model potential is several times longer3' than that given by the intranuclear cascade model. The nucleon-cluster collision is effective to take these aspects into consideration. It is treated in this work as follows. The cluster is composed of two or four nucleons which are selected from uniform random number in accordance with the fraction of the nucleons in a target nucleus. The probabilities of the nucleon-cluster collision are parame trically given as a function of the energy. They are listed in Table 2.3.1. When a collision occurs, the Pauli's exclusion effect is checked with respect to each nucleon. If even one nucleon in the cluster cannot have energy above Fermi level, the collision is inhibited. Since the forbidden events occurs more frequently in the nucleon-cluster collision than in the nucleon-nucleon one, the inclusion of the nucleon-cluster collision has the effect that lengthen the mean free path of the incoming particle.

The inelastic collision plays an important role in the high energy region above 600 M e V because of its large cross section. The inelastic collision is treated via the formation of delta particle" that is heavier

»: Faculty of Engineering, Kyushu University.

- 3 9 -

]AERI-M 88・221

H.Takada. Y.Nakahara and K.Ishibashi'

Ther自 are some discrepancies between calculat自d and experimental

results for both the spallation neutron spectra and the mass yields of

residual nuclei.I).2】 Wehave made some modifications in the intranuclear

cascade model to eliminate the discrepancies. They are the introduction

of the nucleon-cluster collision， the correction of the angular

distribution of the delta particle and the incorporation of the exciton

mode 1.

Nuclear reactions in the intranuclear cascade process are treated

as a sequence of nucleon-nucleon two body collision. In case of the

nucleon energy comes below 200 MeV， th告 deBrogl ie wav自 leugthbecomes

longer than 2 fm which is equal to !he mean distance between the two

nucleons in a nucleus. Consequently， the approximation of the two body

collision is not valid in this energy range. 1n addition， it is suggested

that the mean fre自 pathevaluated from the imaginary part of the optical

model potential is several times longer3J than that given by the

intranuclear cascade model. The nucleon-cluster collision is effective

to take these aspects into cons iderat ion. 1 t is treated in this work

as follows. The c¥uster is composed of two or four nucleons which are

selected from uniform random number in accordance with the fraction

of the nucleons in a target nucleus.τもe probabilities of the

nucleon-c¥uster collision are parametrically given as a Clmction of the

energy. Theyare listed in Table 2.3.1. When a collision occurs， the

Pauli 's exc¥usion eCfect is checked with respect to each nucle口n. If even

one nucleon in the cluster cannot have energy above Fermi level， the

collision is inhibited. Since the Corbidden events occurs more

frequently in the nucleon-cluster collision tha~ in the nucleon-nucleon

one， the inclusion oC the nucleon-cluster collision has the effect that

lengthen the mean free path of the incoming particle.

The inelastic collision plays an important role in the high energy

region above 600 MeV because of its large cross section. The inelastic

collision is treated via the formation oC delta particleU that is heavier

-: Faculty of Engineering， Kyushu University.

J A F R I - M * 8 - > 2 !

than the nucleons by a b c u : 300 M e V . If the incident energy is as high as 800 M e V , a nuc1 eon w h i c h produced delta partible can be recoiled w i t h sufficiently high energy. It is necessary to understand the behavior of the delta particle to r e m e d y the discrepancy in the high energy region. The angular distribution of the delta particle, h o w e v e r , is n o : studied in detail. W e a s s u m e that the delta particle is fo r m e d only in the bac k w a r d d i r e c t i o n of 135-130 deg. in the c.m. s y s t e m .

It is suggested that the preequi1.bri u m process should be incorporated b e t w e e n the intranuclear cascade and the particle evaporation processes according to the analysis of the neutron energy-s p e c t r a . 5 1 In this w o r k . G u d i m a ' s cascade-exc i ton m o d e l 6 1 has been introduced into the high energy transport code ( H E T C ) based upon the M o n t e C a r l o a l g o r i t h m reported in R e f . 7. This incorporation changes the code f r o m t w o - s t e p to three-step. The particles f r o m neutron to l i t h i u m can be e m i t t e d until the excitation energy per exciton c o m e s b e l o w 3.5 M e V in this process.

Figure 2.3.1 s h o w s the results of m a s s yields for 1 G e V proton incidence on gold and Fig. 2.3.2 show s those of spallation neutron spectra for 800 M e V proton incidence on u r a n i u m . O ne can see f r o m these figures that the three-step calculation improves the tw o - s t e p one well und obtains good a g r e e m e n t s w i t h the experimental data.

This study has been p e r f o r m e d in collaboration w i t h K y u s h u Uni vers i ty.

R e f e r e n c e s

1) Howe S.D. et al.: "Neutron spectra from 800 MeV (p,xn) reactions on target of Al, Cu, In, Pb, and U," LA-UR-85-3360 (1935).

2) Kaufman S.B. and Steinberg E.P.: Phys. Rev. C, 11, 167 (1980). 3) Meyer H.O. and Schwandt P.: Phys. Lett., 107B, 353 (1981). 4) Steinheimer R.M. and Lindenbaum S.J.: Phys. Rev., J[£5, 1874

(1957). 5) Tsukada K. and Nakahara Y. : Atomkernenergi e Kerntech. , £4, 186

(1984). 6) Gudima K.K.. Mashnik S.G. and Toneev V.D.: Nucl. Phys.. A401.

329 (1983). 7) Nakahara Y. and Nishida T. : "Monte Carlo Algorithms for

J.¥ F R r -¥.{ 弐月 ~ ~ ~

than the nucleons by abcu: 300 ~\'!eV. lf :he inciオ9:1: eロP.!"gy ‘s 旦雪 :11.;h

as 800 ~reV. 且 nucleon which produced delt且 ~ar t 1ご!e こ且立 :Je ~~C 口~ ed '.V i t h

sufficiently high energy. It IS rlecessa~y to underヨ:and :he behavior of

the delta particle to remed}' the discrepancy ln :he hi.;h 台nergy reglon.

The angul且rdistribution of the delta particle. howp.v弓r. is no: studied

in detail We assume that the delta particle ヨ九rmed only ロ the

backw且rddirection oi 135-180 deg. in the c.m. systi!m

It is suggested that the preequil.brlum ;lroce:iS should be

i ncorpora t ed be tween the 且tranucl宮且r cascade and the particle

evaporati0n processes 且ccording to the an皐 Iysis of the n自utron energy

spectra.Sl In this work. Gudim晶 5 ごas己主d舎-exciton mode16l has been

introduced into the high ecergy :r品目sport code (HETC) based upon the

~ronte Carlo algorithm reponed in Ref. I. This incorporation changes

the code irom two-step to three-s:ep. The particles from neutron to

lithium can be emitted until the excitation energy per exciton comes

below 3.5 MeV in this process.

Figure 2.3.1 shows the results of mass yields for 1 GeV proton

incidence 00 gold aod Fig. 2.3.2 shows those of spallation neutroロ

spectra fo!' 800 MeV pr:oton incidence on ur孟nium. One can see fr:om thes自

iigures that the three-step calculation improves th自 two-stepone well

ltnd obtains good agreements with the 3xperimental data.

This study has been per:formed in collaboration with Kyushu

University.

Refer:ences

1) Howe S.D. et al.: ftNeutrロnspectra from 800 MeV (p.xn) react ions

on target oi Al， Cu. 1n. Pb. and U.ft LA-UR-85-3360 (1985).

2) Kaufman S.B. and Steinberg E.P.: Phys. Rev. C.~. 167 (1980).

3) Meyer H.O. and Schwandt P.: Phys. Lett.. 1旦7B. 353 (1981).

4) Steinheimer R.M. and Lindenbaum S.J.: Pb.ys. Rev.. 1旦~. 1874

(1957) .

5) Tsuk且daK. and Nakahara Y.: Atomkernenergie Kerntech.. 主主. 186

(1984) .

6) Gudima K.K.. Mashnik S.G. and Toneev V .D.: Nucl. Phys.. 主主立1..

329 (1983).

7} Nakahara Y. and Nishida T.: "Monte Carlo Algorithms for

-40ー

J A E R I - M 83-221

S i m u l a t i n g P a r t i c l e E m i s s i o n from P r e e q u i I i b r i u m S t a t e s d u r i n g

N u c l e a r S p a l l a t i o n R e a c t i o n s , " JAERI-M 86-074 (1986) .

Table 2 . 3 . 1 P r o b a b i l i t y for n u c l e o n -

c l u s t e r col 1 i s i o n a s

a func t ion of energy

Enerxy Nuaber of nucleons in i cluster <MeV) 1 2 4

<130 0.1 0.3 0.6 130-200 0.3 0.3 0.4 200-300 0.5 0.3 0.2

>300 0.7 0.3 O.C

10' r J 3

to o cc O

1 0 ' :

: 1 1 1 1 1

• x EXPERIMENT (KAUFMAN) . a HETC 2STEP O 3STEP CALCULATION .

" 8: - i* SB-. cfiP SB-

I a | o o

0 1- a : D # <E '•

" a T ° &> q -" i i I" • a

I >—±—i—i—L.

J U _ H _

120 140 160 180 MASS NUMBER

rrm| 1—rn ruii 1—rTTnrq 1—n-rnT:

EXPERIMENT HETC 2STEP

10' ,-i

rZ 1Q L »oo««v • on V

_ l •'

10 10 10 10' NEUTRON ENERGY (HeV)

Fig. 2.3.1 Mass Yields for 1 GeV Fig. 2.3.2 Spallation neutron spectra proton incidence on for 800 M e V proton gold. incidence on uranium.

JAERI-M 88・221

during States

Nuclear Spallation Rea.ctions，" JAERI-M 86-074 (1986).

Preequilibrium from Emiss ion Particle Simula.t ing

Ta.ble 2.3.1 Prob畠bilityfor nucloon-

cluster collision as

a function of energy

Nu・bel"of間 cleonsin & cluster 2 4

EnerlY 〈円eV)

戸炉』守

M--BO'--

VAnpp』戸、

ppL-H阿H内4u

一•• 司・

「，...-， r-rr廿甘可

0.4 0.2

130・200

200・300

'oow.v， .n u'

司・・・

(〉ωZ¥」的

¥AE)ZCH

・5凶

換炉。けい

t昭夫町堂守岡工

ふ仇m

x EXPERIMEtlT (KAUF憎めロHETC2STEP o 3STEP CALCIλATION

(aE)zo--FUHU目的{〕宍)

品品L....Jー.w....u.

NEUTRON ENERGY (門eV)

10-2 200

Fig. 2.3.2 Spallatior. neutron spectra 2.3.1 Mass Yields Cor 1 GeV Fig.

Cor 800 MeV proton

incidence on uranium.

proton incidence 00

J A E R I - M 8 8 - 2 2 1

2.4 Development of a Vectorized Monte Carlo Code GMVP

T. Mori, M, Nakagavva and M. Sasaki"

A vectorized Monte Carlo code G M V P ( Multi-Group Monte Carlo Code for Vector Processors ) has beeii developed for the FACOM VP-100 vector computer. G M V P has the following functions: (1) Problem to be solved : fixed source and eigenvalue (ktt!) problems. (2) Cross section : multi-group double-differential form or Pi expan

sion type cross sections. (3) Description of geometry : combinatorial geometry in which a cal

culation system is represented by unions, intersections and differences of unit bodies. Any unit body usable in MORSE-DD 1 1 is available in the present code.

(4) Tallies flux, reaction rate, kt// based on track length and collision est imators .

(5) Variance reduction technique : Russian roulette kill and splitting based on cell importance and weight window which are the simple and effective technique for any type of problems. In G M V P , the stack-driven algorithm"1 is adopted to vectorize

Monte Cairo calculations. Calculational t.isks in random walk simulation are subdivided into free flight analysis, next zone search and collision analysis. Each of the three tasks has its own stack for queueing up pointers of particles and includes its own control logic for dispersing particles on completion to their next tasks. The computation proceeds by selecting the task for which the largest number of particles are queueing, excecuting that task, and then dispersing the particle pointers to other stacks. All particle data reside in a large particle bank, and some of them are gathered into working arrays and put back into the bank at the beginning and the end of esJ. task. In an eigenvalue problem, another bank is used to store fission neutrons as source particles in the next batch.

In the free flight analysis and the next zone search, the code carries out each calculation task at the same time for particles either in one geometric zone (one-zone logic) or in all zones of the system (all-zone logic), by user's choice. The collision analysis is carried

• Japan Information Service Co., Ltd. (Tokyo)

- 4 2 -

JAERI-M 1:l8..221

2.4 Development of a Vectorized ~onte Carlo Code GMVP

T. :¥Ior i. M. ~akagawa and :¥1. Sasak i・

A vectoriz骨d :Vlonte Carlo code GMVP ( Multi-Group :Vlonte Carlo

Code for Ve<::tor Processors ) has beel. developed for the FACOM VP--100

vector computer. G:VIVP has the following functions:

(1) Problem to b自 solved fixed source and eigenvulue (k.l/) problems.

(2) Cross section multi-group double-differential form or P1 expan..

sion type cross sections.

(3) Description of geometry combinatorial geometry in which 貸 cal-

culation system is represented by unions. intersec~ions and diffe-

rences of unit bodie5. Any unit body usable in Y10RSE-0011 is

available in the present code

(4) Tallies flux， reaction rate， k，Jj based on track length and col-

1 i s i on e s t i ma t 0 r 5 .

(5) Variance reduction technique Russian roulette kil 1 and splitting

based on cell importance and weight window which are the simple

and effective technique for any type of problems.

ln GMVP， the stack-driven algorithrn21 is adopted to vectorize

;¥1ont.e Calro calculations. Calculational t.o.sks in random walk sirnula-

tion are subdivided into free flight 且nalysis目:lext zone search 且nd

collision analysis. Each of the three tasks has its own stack for

queueing up pointers of particles and includes i ts own control logic

for dispersing particles on completion to their next tasks. The com-

putat ion proceeds by se 1 ec t ing the task for whi ch the larges t nurnber

of particles are queueing， excecuting that task， and then dispersing

the particle pointers to other stacks. All particle data reside in a

large particle bank，旦ndsome of them are gathered into working arrays

ana put back into the bank at the beginning and the end of ee、 task.

ln an eigenvalue problem， another bank is used to store fission neutrons

as source particles in the next batch.

In the free fl ight analysis and the next zone search. the code

carries out each calculation task at the same time for particles either

in one geometric zone (one-zone logic) or in all zones of the system

(all-zone logic). by user・5 choice. The collision analY5i5 i5 carried

• Japan [nformation Service Co.， Ltd. (Tokyo)

J A E R I - M 88-221

out by the all-zone logic. For sampling from a probability-distribution, the BMC sampling technique3', which can be easily vectorized, is adopted instead of a conventional method by using a cumulative probability table.

To examine the effect of vector izat ion by GMVP, we calculated two sample problems shown in Table 2.4.1, and the results were compared with those by MORSE-DD and KEN04 4 1. Figure 2.4.2 show speed-up factors in sample problem 1 by GMVP compared with MORSE-DD. The results of sample problem 2 are summarized in Table 2.4.2. By a vector calculation, the factor of speed-up 8~9 is achieved compared with a scalar calculation by GMVP. However, it is 4~18 compared with MORSE-DD. As for a kt// problem (sample problem 2,) the calculation speed of GMVP is 3~9 faster than that of KEN04. A vector izat ion rate of GMVP is 90~96 percent, depending on problems.

Efforts are now concentrated on vectorization of the continuous energy Monte Carlo method and the lattice geometry.

References 1) M. Nakagawaand T. M o r i : "MORSE-DD: a Month C a r l o Code Us ing

M u l t i - g r o u p D o u b l e - d i f f e r e n t i a l Form Cross S e c t i o n s , " JAERI-M

84-126 (1984) .

2) W . R . M a r t i n e et a l . : I n t . J . Supercompute r Appl i c a t ion, 1, 11

( 1 9 8 7 ) .

3) F . B. Brown: T r a n s . A m . N u c l . S o c 38 , 354 ( 1 9 8 1 ) .

4) L. M. P e t r i e and N. F . C r o s s : 0RNL-4938 ( 1 9 7 5 ) .

Table 2 . 4 . 1 S p e c i f i c a t i o n of sample problem

: P rob lem 1 : P r o b l e m 2

P r o b l e m type : F ixed source of 14 MeV : k,// G e o m e t r y : LizO c y l i n d e r c o n s i s t i n g : Pu fuel s t o r a g e rack

: of 18, 98 and 1002 zones : of 42 zones , kt//=0.5 : (R=30,50 cm, H=40,80 cm) : ( see F i g . 2 . 4 . 1)

Cros s s e c t i o n : d o u b l e - d i f f e r e n t i a l form : Pi expans ion : 65 g r o u p s , 20 angle b ins : 137 g r o u p s , up s c a t t e r

- 4 3 -

]AERI-M 88・221

out by the all-zon由 logic. For sampling from a probability

distribution. the BMC sampl ing technique3¥ which can be easily

vectorized， is adopted instead of a co'aventional method by using a

c um u 1 a t i v e p r 0 ba b i 1 i t y t a b 1 e .

To examine the efCect of vectorization by GMVP， we calculated

two sample problems shown in Tabl自 2.4.1. and the results were com-

pared wi th those by MORSE-DD and KEN044I . Figure 2.4.2 show speed-up

factors in sample problem 1 by GMVP compared with MORSE-DD. The

results of sample problem 2 are summarized in Tabl自 2.4.2. Bya vector

calculation， the factor of speed-up 8"'9 is achieved compared with a

scalar calculation by GMVP. However， it is 4"'18 compared with

MORSE-DD. As for a k.11 problem (sample problem 2，) the calculatio口

speed of GMVP is 3"'9 fast自rthan that of KEN04. A vectorization ratp

of GMVP is 90"'96 percent. depending on problems.

Efforts are now conceロtrated on vec tor i zat ion of the cont inuous

energy Monte Carlo method and the lattice f，'eometry.

References

1) M. Nakagawaand T. Mori: "MORSE-DD: a MontbCarlo Code Using

Multi-group Double-differential Form Cross Sections，" JAERI-M

84-126 (1984).

2) W. R. Martine et al.: Int. J. SupercomputerApplication， 1，11

(1987).

3) F. B. Brown: Trans. Am.Nucl. Soc.， 38， 354 (1981).

4) L. M. Petrie and N. F. Cross: ORNL-4938(1975).

Table 2.4.1 Specirication of sampie problem

Problem 1 Problem2

Problem type Fixed source of 14 MeV k.11 Geometry Li20 cylinder consisting: PU fuel storage rack

of 18， 98 and 1002 zones of 42 zones， k~II=O.5 (R=30，50 cm， H=40.80 cm) (see Fig.2.4.1)

Cross sect ion doubl e-di fferent ial form P1 expans ion 65 groups， 20 angle bins 137 groups. up scatter

J A E u I - M 8 8 - 2 2 1

Table 2 . 4 . 2 Speed-up factor in sample problem 2

P a r t i c l e s / b a t c h one-zone log ic

300 3000 6000

GMVP( s c a l a r ) GMVP( v e c t o r )

( R a t e to KEN04)

1.0* 1.0 1.0 2.25 6.25 7.35

(2 .79 ) ( 7 . 6 5 ) (8 .63 )

al1-zone 1 ogi c 300 3000 6000

0.89 0.82 0.82 2.35 6.03 6.91 (2.90) (7.37) (3.12)

V e c t o r i 2 a t i o n ( % ) 81.2 94.5 96.0 81.6 95. i 96.0

( KEN04 ca l c u l a t ion was c a r r i e d out by J . K a t a k u r a ( J A E R I ) . ) • N o r m a l i z e d to u n i t y .

M Fuel (Pu)

r I Concrete

HI I ron

• Water

V e r t i c a l Plan Hor izon ta l Plan

F i g . 2 . 4 . 1 C o n f i g u l a t i o n of s amp le p rob lem 2

18 regions (one zone) (all zone)

90 regions (one zone) (ell zone)

10.0 15.0 Pertlcles / betch ( 10* )

F i g . 2 . 4 . 2 Compar i son w i t h MORSE-DD for sample p r o b l e m 1

( CPU by MORSE-DD / CPU by GMVP )

- 4 4 -

JAEI、I-M 88-221

Table 2.4.2 Speed-up factor in sample problem 2

one-zone logic all-zone logic Particles/batch 300 3000 6000 300 3000 6000

GMVP(scalar) 1.0・ 1.0 1.0 0.89 0.82 0.82 GMVP( v自ctor) 2.25 6.25 7.35 2.35 6.03 6.91 (Ra t e t 0 KEN04) ( 2 . 79) (7. 65) ( 8 . 63) ( 2 . 90) (7. 37) ( 8 . 12)

Vectorization(%) 81.2 94.5 96.0 81.6 95.i 96.0

( KEN04calculation was carried out by J. Katakura (JAERl).) • Norm且 lized to unity.

圏 Fuel (PU)

口 Concrete

[コ Waler

Fig.2.4.1 Configulation of sample problem 2

Horizontal Plan Vertical

川園-----~一一-一.-.-:=てさ二-----..一日:'_-::-_~・-…ー一一…---....--- .....，.._.......-υ〆t ・・・

....................・・・・・・〆.".・・..‘・・・・・・・...・・4・・一_...._.._---ー一----------一一.................・・・・・・・ー・・._../-，

......~

一一一一一一一一一一ι 二三ふ♂.-:-::::二.._-......

，，'乙.----一-一一・・.. ;~，._;，〆円一一一-一一.._-_.......__...........---_...・・・ 7 ・・・・ー一ー・・一一一-------"'-----_.-一-

jl-…… zone】(all zone) L・・H-9" reg10帽 (onezone) -ー (allzon・3

5.0 10.0 15.0 20.0 p・T包IclesI b・tch【 10')

【 1e

~ 16 w

"' ~ 14 コ

を12ξ

~ 10 u

‘ aヨ司U副

Fig.2.':'.2 Comparison with ~ORSE-DD for sample problem 1

( CPU by NJORSE-DD / CPU by G:VIVP )

JAERI-M 88-221

2.5 Development of the Standardized Intelligent Shielding Analysis Code Package: INTEL-BERMUDA

A.Hasegawa, H. N&kashima, S.Tanaka and T.Suzuki

The area in need of the shielding calculation covers all of the nuclear application fields. For the designers of irradiation facilities or reactors accurate and economical shielding calculation is highly requested. But the shielding calculation requires advanced knowledge of the calculation codes and nuclear data. For the persons not familiar with these properties, shielding calculation is very difficult one to attain the accurate results from highly sophisticated codes.

The project of standardization and consolidation of the shielding codes is now going on in JAERI , aiming at the easy and automatic use of the sophisticated transport codes and proper selection of nuclear data which are abundant in the sources, applying the advanced computational technology such as A/I(Ar t if icial Intelligence), CAD(Computer Aided Design) or CG(Computer Graphics), using the EWS(Engineer ing Work Station) and the mainframe.

The system to be developed is defined from conceptual design works. It is constructed from the following three parts: 1) The preprocessing in ^rder to generate the input data for the

transport code, this is performed by EWS. The system to be solved is defined using CAD option by using interactive conversational mode with computer. A/I option are fully applied to the automatic spatial mesh generation considering mean free path of each zones for each group and automatic energy group selection/generation using point total cross—section data accounting the minima of each resonances of the constituents. 2) In the system usable transport codes are BERMUDA D , ANISN 2>

and DOT 3), among which only multidimensional codes are calculated by mainframe. Not yet sufficient computing powers are available for the full scale shielding calculation" by EWS. Today we must perform such an enormous calculation represented as requiring several hours in CPU time, more than 1-volume of disk (1000 M bytes) spaces and more than 300 thousand I/O operations, i.e., consuming immense system resources. We cannot afford such a huge calculation without a large mainframe or vector processor. 3) The post processing, which is performed by EWS, acts as the solved flux processor to show the contour map, bird eye map in three dimensional representations with full colour options for dose, reaction rate and fluxes. Here user-friendly aspects of the EWS are highly applied.

- 4 5 -

]AER[-M 88・221

2.5 Development of the Standardized Intelligent Shielding

Analysis Code Package: INTEL-BER}ruDA

A.Hasegawa， H.;-rakashima. S.Tanaka. and T.Suzuki

The area in need of tha shialding calculation covars all of the

nucle~r application Cields. For tho designers of irradiation f且ci 1 i t i・sor raactors accurata and aconomical shielding calcula.tion is hiihly

requasted. But the shieldin!f calculation r・quires 且dvancedknowl・dg・。Cthe calculation codes and nuclaar data. For山 p・rsonsnot Cami 1 iar u. with these properties. shielding calculation is very dj(ficult on8 to

a.ttain tha accurate rasults from highlY sophisticated codes.

Tha projact of sta.ndardization and consolida.tion of the shielding

codes i s now go i ng on in JAERI，且iming 'I.t tha aasy and automa.tic usa

of tha sophisticated transport codes and proper selectiロnof nu~la a. r data

which are abundant in the sources，且pplying the advancad computational

tacMロlogy such as AII(Artificial Intelligence)， CAD(Computer Aided

Des ign) or CG(Computer Graphics)， us ing the EWS(Engineeri且g Wor註

Station) and the mainframe.

Tha systern to be devaloped is defined frorn conceptual design works.

It is constructed frorn the following thrae parts:

1) The preprocess ing in "，rder to genera.te the input data for the

transport code. this is perforrnad by EWS. The systern tロ besolved is

dafined using CAD option by using intera.ctiva conversational moda with

computer. AII option are fully applied to the automa.tic spa.tial mesh

generation considering mean Craa path of aach zonas Cor aach group a.nd

automa.tic energy group selection/generation using point total

cross-section data accounting the minirna. of each rasonancas of the

cons t i tuents.

2) In the system usable transport codes ara BERMUDA 1). ANISN 2)

and DOT 3)， among which only multidimensiona.l codas are ca.lcula.ted

by mainfrarne. Not yet suCCicient cロmputingpowers ara available Cor

the Cu[[ scale shielding calculation' by EWS. Tロdaywe mus t per f orm

such an enormous calculation represented as requiring several hours in

CPU time， more than l-volume oC disk (1000 M bytes) spaces and more

than 300 thousand 1I0 opera t i ons， i. e.， cロロsuming immensa systilm

resources. 、"e cannot aCCord such a huge calculation without a large

mainCrame or vec tor processor.

3) The post processing， which is perCormed by EWS. acts as tha solved

flux processor to show the contour map. bird eye map in three dimen-

sional representations with full colour options Cロrdose. reactiou rate

and fluxes. Here user-friendly aspects of the EWS are highly applied.

JAERI-M 88-221

Defined system concept is shown in Fig. 2.5.1. Cross-section data base of the INTEL-BERMUDA system

Two types of cross section libraries are used in this system. 1) JSSTDL-295 common group cross section library for shielding appli

cations (group constants) This library is used as the starting group cross section library

for the succeeding calculation. The number of energy group is 295 and this group structure covers almost all group cross sections utilized in our country. From this library, a library fitted for the users needs is generated by the collapsing code by the directions of the optimal group structure decision mechanism at the preprocessing stage using A/I or user's indication to the calculation expressed as 'uso VITAMINE-J structure'.

Because this library furnishes self-shielding factor tables, effective cross sections can be calculated properly. 2) Total cross-section library (point-wise data basis)

Resonance reconstructed cross sections in point data basis generated with 0.1 % relative errors are prepared for all nuclides available in the JSSTDL-295 common group library.

This library is used at the automatic group structure decision process. First region dependent macroscopic cross sections in point data basis are generated, after checking the minima of the macroscopic cross sections a proper group structures is determined surveying all region macroscopic cross section data.

In some cases those generated point basis macroscopic data are used in the group cross section generation step for the rigorous calculation of the region dependent group constants.

For the first year work, the cross section data bases above stated and a proto type system only applied for DOT3.5 has been developed using mainframe and the performance test is undergoing.

References 1) Su2uki T. et a l . : JAERI-M 82-190 (1982). 2) Engle W. W. Jr. : K-1693 (ORNL) (1967). 3) Rhoades W. A. and Mynatt F. R. : ORNL-TM-4280 (1973).

- 4 6 -

JAER[-M 88・221

Defined system concept is shown in Fig. 2.j.1.

Cross-section data base oC the INTEL-BER:¥<IUDA system

Two types of c，oss seむtion lib，aries are used in this system.

1) JSSTDL-295 common group cross sectioo library for shielding appli-

cations (group coostants)

This library is used as the starting group cross section library

for the succeediog calculation. The nwnber of eoergy group is 295 and

this group structure covers almost &11 group cross s ・ctions utiliz ・din our country. From this library， a library Citt由dCor the users n.・dsis generated by the co[lapsing code by the directioロs of the opt imal

group structure decision mechaoism at the pr自processiog s tage u5ing

A/I or user・s indtcation to the calculation expressed as ・usoVITAlAlNE-J structure'.

Because this library Curnishes self-shielding factor tables.

effective cross sections cao be calculated properly.

2) Total cross-section library (point-wise data basis)

Resonance recoロstructed cross sections in point data basis geoer-

ated with 0.1 ~ relative errors are prepared for al1 nuclides available

io the JSSTDL-295 conunon group library.

This library is used at the automatic group structure d・ci5ionprocess. First region dependeot macroscopic cross s・ctions in point

data basis are generated， after checking the minima of th. macroscopic

cross sectioos a proper group structures is d・termined surveying all

region macroscopic cross section data.

ln 5ロmecases those generated point basis macroscopic data are used

in the group cross sectioo generation step for the rigorous caLcul且tiロa

oC the region dependent group constants.

For the first year work. the cross section data bases above stated

and a prロtotype system only applied Cor DOT3.5 has been deveLoped using

mainframe and the performance test is u且dergoins.

ReCerences

1) Suzuki T. et a1. JAERI-M 82-190 (1982).

2) Engle W. W. Jr. K-1693 (ORNL) (1967).

3) Rhoades W. A. and Mynatt F. R. ORNL-TM-4280 (1973).

J A E R I - M 8 8 - 2 2 1

MAIN FRAME

TRANSPORT CALCULATION:

Flux Solver

I n t e g r a l Transpor t = > BERMUDA

Sn = = > ANISN

Sn =*=> DOT3.5

Pre-process ing Post-process ing

Input data processor

CAD : g e o m e t r i c a l da t a input

A/I : a u t o m a t i c mesh g e n e r a t i o n

a u t o m a t i c energy group

s t r u c t u r e d e c i s i o n

Solved f lux p r o c e s s o r

for Dose . R e a c t i o n Rate

CAD/CG

Contour map

Bird eye map

Colour mapping

= > I n t e r a c t i ve o p e r a t i o n

ENGINEERING WORK STATION

DATA BASE

Common group c ross s e c t i o n

= = > JSSTDL-29S L i b r a r y

for reg ion dependent macro c a l .

To ta l c ross s e c t i o n

===> TOTAL-LIB (po in t w i se b a s i s )

for a u t o m a t i c energy group s t r u c t u r e

dec is ion

for d i r e c t beam c a l c u l a t i o n

far r eg ion dependent w e i g h t i n g f luxes

for group c r o s s - s e c t i o n g e n e r a t i o n

D e t e c t o r c ross s e c t i o n

===> DETX-295 L i b r a r y

for Reac t ion R a t e . Dose. e t c .

F i z . 2 .5 .1 Defined svs ;em concept of INTEL-BERMUDA

•47-

JAERI・M 88・221

~(A 1 ~ FRA，¥，IE

TRA:¥SPORT CAUTLλTIOI ~:

fl~:\ 501"e:

lntegral Transport ===> BERML"DA

Sn => Al'.J'lS:-l

Sn =-==> ooT3.5

~ 九---ーーPre-processing Post-processing

ー『ー・ーー--------ー---- - --圃・・・・M・・・・・・・開._.岡・h ・・岡隆司・・・・・回・・・圃・・・圃・・開-_.

I且putdata processor Solved flux processor

for Dose. Reaction Rate

CAD geometrical data input CAD/CG

Contour map

AIl automatic mesh gen&ration Bird ey・map

automatlc energy group Co 1 our mapp i ng

structure decision ===>Interactive operation

E~GINEER[NG WORK STATION

DATA BASE

Conunon group cross sectioロ

====> JSSTDL-295 Library

for region dependeロtmacro caし

70tal cross section

====> TOTAL-LIB (point wise basis)

for automatic energy group structure

decision

for direct oeam calculation

'.Ill" region dependent weigbting fluxes

for group cross-section generation

Detector cross section

====> DETX-195 Library

for Reaction Rate. Dose. etc

fig. 2.5.1 Dei:ned 5ys.em concept of l:--'iEL-BER:MじDA

JAERI-M 88-221

3 . Reactor Physics Experiment and Analysis

A c t i v i t i e s on reactor physics exper iment dur ing the p re sen t per iod are mainly re la ted to High Temperature Engineering Test Reactor (HTTR) and High Conversion Light Water Reactor (HCLWR).

In o rde r to e v a l u a t e the n e u t r o n i c des ign accuracy of the HTTR, measurements were made on the r e a c t i v i t y worth of the experimental control rod and that of the experimental burnable poison rods in the VHTRC-1 core loaded wi th 4%-enriched uranium f u e l . As for power d i s t r i b u t i o n , the s p a t i a l d i s t r i b u t i o n of " - ' C u ( n , l O r e a c t i o n was measured in the VHTRC-1 core and in the VHTRC-3 core fueled wi th 6%-enriched uranium.

A simple method was developed for r e a c t i v i t y determination based on the i n t eg ra l version of the pulsed neutron a r e a - r a t i o method. The simple method was applied to the s u b c r i t i c a l i t y measurement of down to -40$ a t the VHTRC.

A s e r i e s of e x p e r i m e n t s have been c a r r i e d out a t the FCA to o b t a i n n e u t r o n i c c h a r a c t e r i s t i c s on t he HCLWR co re . The e x p e r i m e n t a l program c o n s i s t s of t he two phases : Phase-1 wi th enr iched uranium fueled core (FCA XIV) and Phase-2 with plutonium fueled core (FCA XV).

Experimental and ana ly t i c a l works were carr ied out on the react ion r a t e r a t i o s and r e a c t i v i t y worth of absorber m a t e r i a l a t the FCA-HCLWR co re s of XIV-1, XIV-1(45V) and XIV-2. The a d j o i n t weighted i n f i n i t e mul t ip l i ca t ion factor k + was measured a t the FCA XIV-1 and XIV-2 cores to conf i rm the r e l i a b l i l i t y of the k po va lues ob ta ined from the m a t e r i a l buckling measurement.

The f i r s t co re of t he Phase-2 exper imen t , the FCA XV-1 c o r e , went c r i t i c a l in May 1987. The e x p e r i m e n t a l va lues of k r r and K ^ were compared with the ca lcula ted values based on the SRAC code system and the JENDL-2 nuclear data f i l e .

(M. Nakano)

- 4 8 -

]AERI-M 88・221

3. Reactor Physics Experiment and Analysis

Activities on reactor physics experiment during the present period

are mainly related to High Temperature Engineering Test Reactor (HTTR) and

High Conversion Light Water Reactor (HCLWR).

1n order to evaluate the neutronic design accuracy of the HTTR，

measurements were made on the reactivity worth of the experimental contro1

rod and that of the experimental burnab1e poison rods in the VHTRC-l core

loaded with 4%-enriched uranium fuel. As for power distribution， the

spatial distribution of 63Cu(n，ず)reaction was measured in the VHTRC-l

core and in the VHTRC-3 core fueled with 6%-enriched uranium.

A simple method was developed for reactivity determination based on

the integral version of the pulsed neutron area-ratio method. The simple

method was applied to the subc.riticality measurement of down to -40$ at

the VHTRC.

A series of experiments have been carried out at the FCA to obtain

neutronic characterist工cson the HCLWR core. The experimental program

consists of the two phases Phase-l with enriched uranium fueled core

(FCA XIV) and Phase-2 with plutonium fueled core (FCA XV).

Experimental and analytical works were carried out on the reaction

rate ratios and reactivity worth of absorber material at the FCA-HCLWR

cores of XIV-1， XIV-1(45V) and XIV-2. The adjoint weighted infinite

multiplication factor k+m was measured at the FCA XIV-l and XIV-2 cores to

confirm the reliablility of the k∞ values obtained from the material

buc.kling measurement.

The first core of the Phase-2 experiment， the FCA XV-l core， went

critical in May 1987. The experimental values of keff and Koo were

compared with the calculated values based on the SRAC code system and the

JENDL-2 nuclear data file.

(M. Nakano)

-48ー

J A E R I - M 8 8 - 2 2 1

3.1 Measurement of Reactivity Worth of Experimental Control Rod in VHTRC-1 Core with Integral Version of PNS Method

I. Kanno, Y. Akino, M. Takeuchi , T. Ono and H. Yoshifuji

In the design of the HTTR. the control rod worth and the reactor shut down margin must be determined accurately. For this purpose, we measured the worth of an experimental control rod in VHTRC-l core with multipoints space integral method with pulsed neutron source (PNS). The experimental results were analyzed by a two-dimensional S„ method and by a Monte Carlo Method.

An experimental control rod called BW-40 type consisted of a stack of neutron absorbing hollow pellets (outer diameter 89.6mm, inner diameter 60.0mm, height 25.2mm and boron content 38.3wtX) and stainless steel tube (SUS304, outer diameter 97.5mm, thickness 2.0mm, length 1148mm). This experimental control rod had almost the same characteristics in neutronics.

In the following experiments, 274 fuel rods (enrichment 4X) and 4 BF 3 counters were installed in the VHTRC-1 core. The PNS was attached to the center of the end plane of the fixed half assembly. First, we measured the subcritical i ty of the core without experimental control rod. The subcriticality of this core was 0.217$. Next, the BW-40 type experimental control rod was inserted in a control block, 9.5cm radially distant from the core center as indicated in Fig.3.1.1. The BF3 counters were placed on the central axis of the fuel blocks 15cm and 40cm distant from the midplane in the facing movable and fixed half assemblies. The experimental control rod worth was measured placing the BF3 counters in 12 fuel blocks and 48 measured values were obtained. The results are shown in Fig.3.1.1. The minimum and the maximum values of the experimental results were 13.8$ and 28.0$. Analyzing these values by a space integral method with the weight of the volume which detector was responsible, we obtained the worth of BW-40 type experimental control rod as 18.8 ±0.2$.

We carried out the calculations of the experimental control rod worth with the two-dimensional S„ method and Monte Carlo method to compare with the experimental one. In the S„ method, cross sections of the fuel block were prepared by collision probability method" into 10 group constants, first. The experimental control rod worth was calculated with these group constants by TWOTRAN code (X-Y, S 4

- 4 9 -

JAERI-M 88・221

3.1 Measurement of Reactivity Worth of Experirnental Control

Rod in VHTRC-l Core with Integral Version of PNS Method

1. Kanno. Y. Akino. M. Takeuchし T. Ono and H. Yoshit'uji

In the design of the HTTR. the control rod worth and the reactor

shu t down marg i n mus t be de t自rmi ned accura t骨ly. For this purpose. we

measur自d the worth of an experimental control rod in VHTRC-l core with

mul t ipoints space integral method wi th pulsed neutron source (PN8).

The experimental resul ts w自re analyzed by a two-dimensional 5" method

and by a Monte Carlo Method.

An experimental control rod called BW-40 type consisted of a stack

of neutron absorbing hollow pellets (outer diameter 89.6mm. inner

diameter 60.0mm. height 25.2mm and boron content 38.3wt"> and stainless

steel tube (8U5304. outer diameter 97.5mm. thickness 2.0mm. length

1148mm). This experimental control rod had almost the same characte-

ristics in neutronics.

In the following experiments. 274 fuel rods (enrichment 4") and 4

BF3 counters were installed in the VHTRC-l core. The PNS was atts.::hed

to the center of the end plane of the fixed half assembly. First. we

measured the subcr;ticality of the core without experimental control

rod. The subcriticality of this core was 0.217$. Next. the BW-40 type

experimental control rod was inserted in a control block. 9. 5cm radially

distant from the core center as indicated in Fig.3.1.1. The BF3 counters

were placed on the central axis of the fuel blocks 15cm and 40cm distar.t

from the midplane in the facing movable and fixed half assemblies.

The experimental control rod worth was measured placing the BF3 counters

in 12 fuel blocks and 48 measured values were obtained. The results

are shown in Fig.3.1.1. The minimum and the maximum values of the

experimental results were 13.8$ and 28.0$. Analyzing these values by

a space integral method with the weight of the volume which detector

was responsible. we obtained the worth of BW-40 type experimental con-

trol rod as 18.8 土0.2$.

We carried out the calculat ions of the exper imental control rod

;vorth wi th the two-dimensional 5" method and Monte Carlo method to

c(¥mpare wi th the experimental one. In the S同 method. cross sections

of the fuel block were prepared by collision probability methodl) into

10 group constants. first. The experimental control rod worth was

calculated wi th these group constants by TWOTRAl"l code (X-Y. 8.

JAERI-M 88-221

a p p r o x i m a t i o n ) t a k i n g i n t o account the p r e c i s e geometo ry of g r a p h i t e

block w i t h BW-40 c o n t r o l rod . In the Monte C a r l o method , the VIM

c o d e 2 1 was employed .

The c a l c u l a t e d c o n t r o l rod wor ths a re shown in Table 3 . 1 . 1 w i t h

e x p e r i m e n t a l one. These va lues showed good agreement w i t h e x p e r i m e n t a l

one in IX by two d i m e n s i o n a l S„ method and in 2% by Monte C a r l o method .

R e f e r e n c e s

1) Tsuch ihash i K. et a l . : "Rev i sed SRAC Code S y s t e m " , JAERI 1302

(1986 ) .

2) B l o m q u i s t R. N . , Le 11 R. M. and Ge lba rd E. M. : "A Review

of the Theory and A p p l i c a t i o n of Monte C a r l o Me thods" , p . 3 1 ,

ORNL/RSIC-44 ( 1 9 8 0 ) .

Table 3 . 1 . 1 E x p e r i m e n t a l and c a l c u l a t e d va lues of BW-40

type e x p e r i m e n t a l c o n t r o l rod w o r t h .

Exper imen ta 1

Value ($)

C a l c u l a t e d Va lues ($) Exper imen ta 1

Value ($) 2 -Dimens ion Sn Monte C a r l o

1 8 . 8 ± 0 . 2 18.7 19.2

0ai = 0.00725 Movable Side Fixed Side

VHTRC-1 Core

F i g . 3 . 1 . 1 S p a t i a l dependence of the BW-40 type e x p e r i m e n t a l

c o n t r o l rod w o r t h . The value w i t h / w i t h o u t p a r e n t h e s i s

i n d i c a t e s the r e a c t i v i t y w o r t h measured a t 15cm/40cm

from the m i d p l a n e .

- 5 0 -

JAER I・M 88・221

approximation) taking into account the precise geometory of graphite

block with BW-40 control rod. [n the Monte Carlo method， the VIM

cod自Z) was employed.

The calculated control rod worths are shown in Table 3.1.1 with

自xperimentalone. These values showed good agreement with experimental

one in l~ by two dimensional S" method and in 2% by Monte Carlo method.

References

1) Tsuchihashi K. 白ta I . "Re v i s自d SRAC Code Sys t自m開， JAERI 1302

(1986).

2) Blomquist R. N.， Lell R. M. 且nd Gelbard E. M. "A Review

of the Theory and Application of Mont自 Car1 0 Me thods

ORNLlRSIC一44 (1980).

Table 3.1.1 Experimental and calculated values of BW-40

type experimental control rod worth.

Exper i men ta I

Value ($)

18.8土 0.2

MovabJe Side

Calculated Values ($)

2-D i mens i on Sn

. . ExporimootaJ

ControJ Rod

β." = 0.00725

VHTRC-l Core

Monte Carlo

Fixed Side

Fig.3.1.1 Spatia.l dcpendence of the BW-40 type experimental

control rod worth. The value with/without parenthesis

indicates the reactivity worth measured at 15cm/40crr.

from the midplane.

一回一

J A E R I - M 8 8 - 2 2 1

3.2 Measurement of Reactivity Worth of Burnable Poison Rod in VHTRC-1 Core

F.Akino, M.Takeuchi and Y.Kaneko

As the core design for the HTTR progresses", evaluation of design accuracy has become increasingly important. A high accuracy design of the HTTR requires adequate group constants and a calculation method accurately describing neutron transport.

The HTTR is designed to accommodate burnable poison rods for reactivity compensation. Accordingly, experimental burnable poison rods equivalent to those designed for the HTTR were prepared for the mock up experiment at VHTRC. They were made by inserting 20 absorbing pellets (total length o( absorber:72 cm) in a hollow graphite rod. The absorbing pellets were made of B 4C particles dispersed in graphite powder. Three types of the experimental burnable poison rods were prepared. They are different in boron content and grain size of B<C.

The specifications of the absorbing pellets are shown in Table 3.2.1. The experimental burnable poison rod was inserted into the corner of the prismatic fuel block of central column of the VHTRC-1 core. This core was made by loading fuel rods which contained fuel compacts of 4X enriched uranium coated particles. The reactivity worth of experimental burnable poison rod was measured by the combination of the period and fuel rods substitution method.

An analysis was made by using the SRAC code system 2 1 with ENDF/B-IV nuclear data library. The double heterogeneity of fuel compact and coated particle, and of absorbing pellet and 560/tm B«C particle were taken into account in the cell calculation by the collision probability method. The modeling of the central column inserting the experimental burnable poison rod is shown in Fig.3.2.1. The core calculation was performed by the three-dimensional 24 groups diffusion theory. Thermal energy range from O.OeV to 1.1254eV were divided into 13 groups and fast energy range from 1.1254eV to lOMeV were divided into 11 groups .

The results are shown in Table 3.2.1. The measured and calculated values are found to agree with each other within the experimental uncertainty of o%. These results indicate that the reactivity of burnable poison rod can be estimated fairly good by the present method and data.

- 5 1 -

]AERI-M 88・221

3.2 Measurement of Reactivity Worth of Burnable Poison Rod

in VHTRC-l Core

F.Akino. Nl.Takeuchi and Y.Kaneko

As the core design for the HTIR progresses!l. 自valuation of design

accuracy has becom自 increasingly important. A high accuracy design of

the HTIR requires adequa.te group constants and a calculation method

accurately describing neutron transport.

The HTIR is designed to accommodate burnable poison rods for reac-

t i v i ty compensat ion. Accordingly. experimental burnabl e poi son rods

equivalent to those designed for th自 HTIRwere prepared for the mock

up exper iment at VHTRC. They were made by insert ing 20 absorbing

pellets (total length of absorber:72 cm) in a hollow graphite rod. The

absorbing pe 11 e ts were made of B‘C particles dispersed in graphite

powder. Three types of the experimental burnable poison rods were

prepared. They are different in boron content and grain size of B.C.

The specifications of th自 absorbingpell自tsare shown in Table 3.2.1.

The experimental burnable poison rod was inserted into the corner of

the prismatic fuel block of central column of the VHTRC-l core. Tbis

core was made by loading fu自1rods which contained Cuel compacts of 4~

enriched uranium coated particles. The reactivity worth of experimental

burnable poison rod was measured by the combination of the period and

fuel rods substitution method.

An analysis was made by using the SRAC code system 幻 with ENDF/B-IV

nuclear data 1 ibrary. The double heterogenei ty of fueJ compact and

coa.ted particle. and oC absorbing pellet and 560llm B.C particle w申re

taken into account in the cell calculation by the collision probability

method. The modeling of the central column inserting the experimental

burnable poison rod is shown in Fig.3.2.1. The core calculation was

performed by the three-dimensional 24 groups diffusion theory. Thermal

energy range from O.OeV to 1.1254eV were divided into 13 groups and Cast

energy range from 1.1254eV to lOMeV were divided into 11 groups

The results are shown in Table 3.2.1. The measured and calculated

values are found to agree wi th each other wi thin the experimental

uncertainty of 5~. These results indicate that the reactivity of burnable

poison rod can be estimated fairly good by the present method and data.

----whuw

JAERI-M 88-221

Re f e r ences

l )Sanokawa K. a n d S a i t o S . : J . A t . Energy Soc. Japan , 29, 603(1987)

2 )Tsuch ihash i K. et a l . : JAERI-1285( 1985)

Table 3 . 2 . 1 R e a c t i v i t y wor th of an e x p e r i m e n t a l burnab le po ison

rod in the c e n t r a l column of VHTRC-l core

Type of D i a m e t e r of Absorb ing p e l l e t R e a c t i v i t y w o r t h

bu rnab le B«C Boron D i a m e t e r Measured C a l c u l a t e d

po i son rod p a r t i c l e c o n t e n t

(/im) (wtX) (mm) ($) ($)

N-8 5 8.2 7.9 1.63±0.08 1.66

W-2.5 S 2.5 11.9 l . 78±0 .09 1.82

N-16 560 16.0 8.0 1.67±0.08 1.76

Absorbinj pellet

Experimental jeoietry Calculation feoaetry

F i g . 3 . 2 . 1 Mode l ing of the c e n t r a l co lumn w i t h the e x p e r i m e n t a l bur

n a b l e po ison rod

- 5 2 -

JAER!-M 88・221

References

l)Sanokawa K. and Saito S. :J. At. Energy Soc. Japan， 29， 603(1987)

2)Tsuchihashi K. et al. :JAERI-1285(l985)

Table 3.2.1 Reactivity worth of ao experimental burnable poison

rod in th骨 C自ntralcolumo or VHTRC-l core

一一一一一一一ーーーー一一一- 一一

Type of Diameter or Absorbing p申IIe t Reactivity worth

burnable B.C Borロロ Diameter Measured Calculated

poison rod particle content

(μm) (wt") (mm) ($) ($)

N-8 5 8.2 7.9 1. 63主0.08 1.66

W-2.5 5 2.5 11.9 1. 78主0.09 1.82

N-16 560 16.0 8.0 1. 67之0.08 1. 76

Experi.en!.1 ，eo.etry Calculallon ceo.elry

Fig.3.2.1 Modeliog or the central columo with the experimental bur-

oable poison rod

J A E R 1 - M 8 8 - 2 2 1

3.3 Measurements of 6 3Cu(n,y) Reaction Rate Distribution in VHTRC-3 Core

T. Yaroane, F. Akino, H. Yasuda, M. Takeuchi, T. Ono, K. Kitadate, H. Yoshifuji and Y. Kaneko

Distributions of 6 3Cu(n,/) reaction rates were measured in VHTRC-3 core at room temperature and 200*C by the foil activation technique to evaluate the accuracy of neutronic design calculation on power distribution in the HTTR.

The VHTRC-3 core was assembled to study the fundamental neutronic characteristics of the HTTR core, as well as the previously reported core VHTRC-l 1'' . The VHTRC-3 core was loaded mainly with 6%-enriched uranium fuel and partially with 42-enriched one. Figure 3.3.1 shows the fuel loading pattern at 200*C. Copper foils and wires of natural abundance were used as activation samples for the measurements in the radial and axial directions, respectively: the foils were irradiated at every fuel-rod positions of S-3, S-4 and C-6 blocks in Fig.3.3.1, and the wires at the positions of 50 mm intervals along the graphite rods indicated by "A", "B" and "C" in the same figure. The measuring points were chosen in consideration of the symmetry of fuel loading pattern. The radioactivities of irradiated samples induced by 3Cu(n,/) *Cu reaction were measured by counting / rays mainly coming from 8 annihilation with a well-type Nal(Tl) detector.

3) Analysis was carried out using the SRAC code system . Twenty four group constants were obtained from cell calculation by the collision probability method and core calculation was performed by the three-dimensional diffusion theory.

The measured distribution was compared with the calculated one. Figure 3.3.2 shows the typical comparison in the radial direction where both results are normalized by using the average value of twelve points in the 3-3 block. The agreement is fairly good both at room temperature and at 200*C, although the deviations of calculated values from measured ones show a tendency to slight increase near the boundary between the core and reflector, and between the fuel regions with different enrichment. As regards the distribution in the axial direction, the calculated results agreed with the experimental ones within the experimental uncertainty.

- 5 3 -

jAER1-M 88・221

3.3 Measurements of 63Cu(n，i) Reaction Rate Distribution in VHTRC-) Cor巴

T. Yamane， F. Aklno， H. Yasuda. M. Takeuchi. T. Ono. K. Kitadate.

H. Yoshifuji and Y. Kaneko

Distribulions of 63Cu(n.y) reaclion rates were measured [n VHTRC-3

core at room temperature and 200.C by the loil activation tcchnique to

evaluale the accuracy of neulronic design calculation on power dislribu-

tion in the HTTR.

The VHTRC-3 core was assembled to study the fundamental neulronic

characleristics of the HTTR core， as well as the previously reported core 1)，2)

VHTRC-l" '~'. The VHTRC-3 core was loaded mainly with 6χ-enriched uranium

fuel and partially with 4χ-enriched one. Figure 3.3.1 shows the fuel

loading pattern at 200・C. Copper foils and wires of natural abundance

were used as activation samples for the measurements in the radial and

axial directions， respectively: the foi 15 were irradiated at every fuel-

rod positions of S-3， S-4 and C-6 blocks in Fig.3.3.1， and the wires al

the positions of 50 mm intcrvals along lhe graphile rods indicated by

"Att，日" and "C" in the same figure. The measuring poinls were chosen in

consideration of the symmelry of fuel loading pallern. The radioac-63_. _.640

tivities of irradiated samples induced by --Cu(n. ず )-~Cu reaction were +

measured by counling y rays mainly coming from B annihila~ion with a

well-type Nal(Tl) delector. 3)

Analysis was carried out using the SRAC code system-'. Twenty four

group constants were obtained from cell calculalion by the collision

probability method and core calculation was performed by the three-

dimensional diffusion theory.

The measured distribution was compared with the calculated one.

Figure 3.3.2 shows the lypical comparison in the radial direction where

both results are norma1ized by using the average value of twelve points

in the S-3 block. The agreement is fairly good both at room temperature

and at 200・C，although the deviations of calculated values from measureu

ones show a tendency to slight increase near the boundary between the

core and reflector， and between the fue1 regions wi th di fferent

enrichment. As regards the distribution in the axial direction， the

calculated resu1ts agreed with the experimental ones within the ex-

perimental uncertainty.

J A E R I - M 8 8 - 2 2 1

References 1) Yamane T. et al.: "Reactor Engineering Department Annual Report,"

JAERI-M 86-125, 62 (1986) 2) Yamane T. et al.: "Reactor Engineering Department Annual Report,"

JAERI-M 87-126, 47 (1987) 3) Tsuchihashi K. et al.: "Revised SRAC Code System," JAER1 1302 (1986)

Pos. A

S-4 block

• Fuel rod (4%EU) • Fuel rod (6%EU) • Sofety rod hole • Control rod hole © Heater

S -3 block

C-6 block

Fig.3.3.1 Fuel loading pattern at 200*C

- 1 . 2 -

i,.o o <u cr

At room temperature • Expt. * Calc.

6% -4% Fuel enrichment

20 30 40 50

Distance from core axis (cm)

- 1 . 2

s I .O

At 200 C • Expf. A Cole.

a * 4 8

6% _ | _ 4 %

Fuel enrichment

20 30 40 50

Oistance from core axis (cm)

Fig.3.3.2 Typical results of 6 3Cu(n,V) reaction rate distribution in the radial direction

]AERI-M 88・221

Referel'lces

et al.: "Reactor Engineering Department Annual Report，" 1) Yamane T.

JAERI-M 86-125， 62 (1986)

Report，" et al.: "Reactor Engineering Department Annual

JAER(-M 87-126， 47 (1987)

3) Tsuchihashi K. et al.: "Revised SRAC Code System，" JAERI

2) Yamane T.

1302 (1986)

• Fuel rod (4% EUI ・Fuelrod (6 % EU )

A 50fety rod hole

• Control rod hole

o Heoter

-3 block

Pos. A

loading pattern at 200.C Fuel Fig.3.3.1.

Expt. Colc.

At 2000C

• a z -1.2 由>

At room temperoture • Expt.

Colc. A • 4 1.2 一也、こち一白』} 。

• -t d a

6% +4%

Fuel enrichment

.室u o U a:

-a ・4

ー十川

• 1.0 cozuo由巳

r 20 羽 40 50

Distonce from core oxis (cml

了」ー一一司」

Distonce trorn core axis (cm)

0.8 L ~~ o

Typical results of 63Cu(n，i) reaction rate distributio~ Fig.3.3.2

in the radial direction

JAERI-M 88-221

3.4 A Simple Method for Reactivity Determination Based on Integral Version of Pulsed Neutron Area-Ratio Method

T. Yamane, F. Akino and Y. Kaneko

A simple method for reactivity determination was formulated, which was based on the integral version of the pulsed neutron area-ratio method. The simple method determines the static reactivity from one-point measurement by using a correction factor. The present method was examined in the measurements of subcriticality ranging from -3 to -40 dollars at VHTRC.

The reactivity measurement by the integral version is practically carried out by observing prompt and delayed neutron area, A (r.) and

-* -» P ! A (r.), at several spatial points r. with appropriate neutron detectors. In applying the formula given by Kosaly and Fisher , one must average

-» -> the weighting function W(r) over an effective volume V. allotted to r.:

i i the static reactivity in dollar unit is expressed by

1 -» ' P /fi = [ I W . A (r. )]/[ [S..A. (r. )] (1) s eff .^ pi p I J.J di d i

where I is the number of measuring points. The averaged weighting functions W . and W,. are defined by - P l r d 2> -» -» , W . = J„.W(r)A (r)dr / A (r.) (2) qi Vi q q l wi th the defini tions

W(r) = ]"x(E)<f+ (r,E)dE (3) and

A (r) = J"\>£f(r,E)4> (r.E)dE . (4) Here the subscript q denotes p or d corresponding to prompt or delayed mode, respectively, and the function <P is time- and direction-integrated prompt or delayed flux, $ the static adjoint eigenfunction of fundamental mode and X the fission spectrum. The expression of Eq.(l) is straightforward, but it is necessary for 4> and 4>, to be known. It would

p d be expected that the shape of <f> is well approximated by static eigenfunction of fundamental mode, <t> , and if I is large, the function <P

os os might be substituted even for <!> . But if the measuring points are a few, such substitution would introduce a significant systematic error into the resulted reactivity value, particularly in highly subcritical and/or singular systems where the effects of spatial higher modes and kinetic distortion of neutron distribution would be strongly enhanced.

Here let us consider a limit in reducing the number of measuring points. By setting 1=1 and Y = V , Eq.(l) becomes

- 5 5 -

1AERI-M 88・221

3.4 A Simple Method for Reactivity Determination Based

on Integral Version of Pulsed Neutron Area-Ratio Method

T. Yamane. F. Akino and Y. Kaneko

A simple method for rcactivi ty determination was formulated. which

was based on the integral versinn of the pulsed neutron area-ratio

method. The simple method determines the static reactivity from nnc-point

measurement by using a correction factor. The present method was examined

in the measurements of subcriticali ty ranging from -3 to -40 dollars at

VHTRC. 1)

The reactivity measurement by the integral version is practically →

can-i ed ou t by observ ing promp t and de l ayed neu t ron area ，AP (r i)and →

Ad {r i}， a t severa l spa t l a l poin ts r i w i th appropr i ate neu t ron de tec tors. 11

[n applying the formula given by Kosaly and Fisher". one must average

→ → the we i gh t ing fline t i on W(r}over an ef fee t i ve vol lime V i a l lot t ed t or l:

the static reactivity in dollar unit is expressed by

1 →→ p ~ / B " == [ 1. W_， A_ (r， )] / [ 1. W ~， A， (r， )] (l ) s eff i=l pl p i i=i di d i

where 号 thenumber of measuring points. The averaged weighting func-

tions wpi aqd wdL are白 finedby

= 1..， W(r)A (r)dr / A (r， ) qi 'Vi .." "'q '" _. . "q" i (2)

with the definitions →，曲+→

W(r) = J;'~X(E)φ( r. EldE -u os

and →，田→→

A (r) = Jn¥lε(r.EIや (r.E)dE • (41 q 1- JO........f............'Tq'..'

Here the subscript q denotes p or d corresponding to prompt or delayed

mode. respectively， and the functionφis time-and direction-integrated q

+ prompt or delayed flux，φthe static adjoint eigenfunction of fundamen-

os tal mode and X the fission spectrum. The expression of Eq.(ll is

straightforward. but it is necessary fOl'φand φto be known. [t would p -..- 'd

be expected that the shape ofφis well approximated by static eigen-d

function of fundamental mode.φ. and i f [ is large， the funct ionφ oS' os

might be substituted even forφBut if the measuring points are a few， P

such substitution would introduce a significant syst巳maticerror into the

resulted reactivity value. particularly in high!y subcritical and/or

singular systems where the effects of spatial higher modes and kinetic

distortion of neutron distribution would be stronglY t:nhanced.

Here let us consider a limit in reducing the number of measuring

points. By setting [=1 and V.= V_. Eq.(11 becomes c

J A E R I - M 8 8 - 2 2 1

s eff p d n where the quantity V is the whole volume of fuel region and

f = ( L W(r)A (r)dr / L W(r)A(r).dr ) / (A /A.)-» . (6) Vc p Vc d p d n This simple method determines the static reactivity from one-point measurement. The quantity f is a correction factor to the value of area-ratio, *A,/A,):f , observed at a position r. and the value of f must be estimated entirely by calculation. The denominator of the r.h.s. of Eq.(6) is the calculated area-ratio at the measuring point, and it should be noted that the simple method does not limit the measuring point to be in the fuel region if a suitable neutron detection-efficiency of detector is used for estimating that area-ratio.

The present method was applied to subcriticality measurements at VHTRC. Four different subcritical states were defined by loading the core with different numbers of fuel rods, 96, 144, 192 and 240. The arrangement of PNS and neutron detectors is illustrated in Fig.3.4.1.

In the calculation of the correction factor, the group constants 2) were obtained with the SRAC code system using the nuclear data based on

the ENDF/B-IV. The neutron fluxes, 4> ,<)> , 4> and 4> were calculated by os os p d

the three-dimensional diffusion theory. The calculated correction factors for each subcritical state are shown in Fig.3.4.2.

The experiments were analyzed by the simple method and the results are shown in Fig.3.4.3. The reactivity values obtained at different measuring points are in good agreement each other, although the observed area-ratio indicates strong spatial dependence, for instance, its value varies from -26 $ to -56 $ in the case of 96 fuel rod loading. The reactivity values by the simple method also showed good agreement with those by the conventional integral version analysis; the difference of reactivity values between two analyses were, for example, less than 2.52 even for the most highly subcritical state.

The results of experimental examination leads us to the conclusion: (1) The present method gives almost constant value of reactivity at any measuring point. The measuring point is not restricted in the fuel region. (2) The present method gives the same results as the integral version in spite of one-point measurement. (3) In calculating 4> and 4>., much care should be given to the treatment

P d of energy and spatial distributions of source neutrons. (4) As for the integral version, the averaging of weighting function should be carefully performed, if the measuring points are a few.

- 5 6 -

JAERl・M 88・221

ρ/8 = f ・ (~/A ， )~XP (5) s' -eff . "p"o'rt where the quantity V is the whole volume of fuel region and

，→ー→，→→→f = ( J.. _ W ( r ) A_ (r) d r / L W ( r ) A ( r) ，d r ) / (A / A. )→ (6 ) ， V c .. ，. . "p .， . -. . ， v c .. ，. ...，. . d .• • • . • . 'p' "d • r 1

This simple method determines the static reactivity from one-point

measurement. The quantity f is a correction factor to the value of area-金xp _L___.._-' _， _ ___，目→

ratio， (AjA.，)斗 observedat a position r， and the value of f must be p'.ld/rl ""..，...."'.......'" ........ ~ t' v~.....&v.. 1. 1

estimated entirely by calculation. The denominator of the r.h.s. of

Eq.(6) is the calculated area-ratio at the measuring point， and it should

be noted t~at the simple method does not limit the measuring point to be

in the fuel region if a suitable neutron detection-efficiency of detector

is useq for estimating that area-ratio.

The present method was app1ied to subcriticality measurements at

VHTRC. Four different subcritical states were defined by loading the core

with different numbers of fuel rods， 96， 144， 192 and 240. The arrange-

ment of PNS and neutron detectors is i1lustrated in Fig.3.4.1.

In the ca1culation of the correction factor， the group constants 2)

were obtained with the SRAC code system-' using the nuclear data based on +

the ENDF/B-IV. The neutron fluxes 中 φ 中 andφwere calculated by 'os ・os' p -，，-'d

the three-dimensional diffusion theory. The ca1culated correction factors

for each subcritical state are shown in Fig.3.4.2.

The experiments were ana1yzed by the simp1e method and the results

are shown in Fig.3.4.3. The reactivity values obtained at different

measuring points are in good agreement each other， although the observed

area-ratio indicates strong spatial dependence， for instance， i ts value

varies from -26 $ to -56 $ in the case of 96 fue1 rod loading. The reac-

tivity values by the simple method a1so showed good agreement with those

by the conventiona1 integral version ana1ysis; the difference of reac-

tivity va1ues between two analyses were， for example， less than 2.5χeven

for the most highly subcritical state.

The results of experimental examination leads us to the conclusion:

(1) The present method gives almost constant va1ue of reactivity at any

measuring point. The measuring pOint is not restricted in the fuel

reglon.

(2) The present method gives the same results as the integral version in

spite of one四 pointmeasurement.

(3) In ca1culatingφand φmuch car~ shou1d be given to the tr告atmentp -..- 'd ・of energy and spatial distributions of source neutrons.

(4) As for the integral version，

J A E R I - M 8 8 - 2 2 1

References 1) Kosaly G. and Fisher J.: J. Nucl. Energy, 2j>, 17 (1972) 2) Tsuchihashi K. et al.: "Revised SRAC Code System," JAERI 1302 (1986)

A : Arrangement-A B : Arrongement - B

(F ) : In fixed half (M): In movable half

[U :PNS ®,CSI : B F 3 counters

Z F= 80/105 cm Z M= 135/160 cm

Fig.3.4.1 Arrangement of pulsed neutron source and detectors

~i |—I I | i T~~1 r~ I I I I | i i I I | I I I

o 0 0 O O O Of

4 4 » — "

^f <#•>»*

J&&""*"*"** • • • • ' A A A A

+ 8 + + *tS* °C(M( 96 fu«l rodt )

A * « ° . C 0 6 < \AA f u » l r o d t > * «° *C08( 192 fusl rodi > o +C10( 240 fuel rods )

i i 1 i i i i I i 50 100 ISO

AXIAL DISTANCE (cm)

_J_ 200

Fig.3.4.2 Correction factor in arrangement-A

11„^™-,.^ ! ° : Arrongement - A Uncorrected { A : A m * m t n t . g

rnrr„i,A S • : Arrangement- A Corrected ( A : A r r a n a e n , e n t _ B

- 40 -

10 0 0 192 fuel rods

• •~~~~~* • — — 8 8-240 fuel rods

9 9- • • a • -i ' i i i i

-» • i I

80 105 135 160 Axial Distance (cm)

Fig.3.4.3 Results by simple method

- 5 7 -

JAERI-M BB・221

References

1) Kosaly G. and Fisher J.: J. Nucl. Energy. 2旦. 17 (972)

2) Tsuchihashi K. et al.: "Revised SRAC Code System，" JAERI 1302 (1986)

80/105 cm ZF=

ZM= 135/160 cm

国 :PNS

o，図 :BF3 counters

A : Arrangement同 A8 : Arrangement -B (FJ : 10 fixed ho/f 1M) : In movable half

。守N|

i引 on占-fFl図|図 l

い3cm→t-73cm-l

120cm-+ー120cm斗Arrangement of pulsed neutron source and detectors Fig.3.4.1

.f-ir J

A州内ロ

--aL .. ‘、

b削財

。“.・0

。、

，‘、

-也、

・0、

、・

lili--』「

句、JW

内〆』

、ι4

• • 192 fue I rods o 0__ ........ ~....，.-. ，

. . ー・・ー一-~~一一-z:10

向。0000 0 。。‘'0 _0

白00・-・・・・・・・・・・.0 ・-

a。:..・.....66・a・.666"4 ..

do-:.I・._"，60-

。ョ:"6・- ・'-'-'占争-I-H-++ + +

++++Festf 目+-r6;0

++:.:;80 。C04tgEFu・1rod・3

0.5ト+ ..:~Ö . C061 IH fu・Irod.，て6T?z。・C081192 fu.1 rod.，

• ; O +Cl01 2~0 fu・1rod.， 8 g

恥I.S.

o: o ト-u

ま1.0z o ートー色SUJ 0: a: o 色，

240 fue 1 rods

ーー-ー・---ー・-ー・80 105 135

Ax ial Disfonce (cm) 160

o Correction factor Fig.3.4.2

Results by simple method Fig.3.4.3

in arrangement-A

JAERI-M 88-221

3.5 Measurements and Analyses on K , c and K, , a t the FCA-XV-1 Core eff inf

T. Osugi, A. Ohno, H. Oigawa, K. Satoh, K. Hayasaka and M. Nagatani*

A s e r i e s of e x p e r i m e n t s have been c a r r i e d out a t FCA to ob t a in

n e u t r o n i c c h a r a c t e r i s t i c s on the High Convers ic Light Water Reactor

(HCLWR) core. The experimental program cons i s t s of two phases : Phase-1 ' 2) with e n r i c h e d uranium fueled core and Phase-2 ' wi th plutonium fueled

core .

In the Phase-2 experiment, particular emphasis is put on the measurements of conversion ratio, moderator void worth and plutonium isotopic composition effect, in addition to the HCLWR characteristics studied in the Phase-1 experiment. The heterogeneity effects of the plate-type and pin-type fuels are studied at the core in which the mixed oxide fuel are loaded at the central test zone.

The first core of the Phase-2 experiment, the FCA-XV-1 core, went critical in May 1987. Futl enrichment of the test zone cell "Pu08" is 8.11 %Pu f i s s/(Pu+U) and moderator-to-fuel volume ratio is 0.6. Plutonium isotopic composition in plate is 2 3 9 P u / 2 A 0 P u / 2 4 1 P u = 91.6/8.0/0.4.

Comparisons of the experimental and calculated values are made for the effective multiplication factor kg££ and the infinite multiplication factor k,-„r. The calculations were made with use of the SRAC code inr system ' and 86-energy group cross-section set based on the JENDL-2 data file*>.

Table 3.5.1 presents the experimental and calculated k e££ values. The calculated k„££ is in good agreement with the experimental one.

Table 3.5.2 presents the experimental and calculated k- £ values for the FCA-XV-1 test zone cell. The experimental k^n£ value was deduced from the measured material buckling B and the theoretical migration area M . ' The calculated k- £ is also in good agreement with the experimental one.

Table 3.5.3 summarizes the calculated-to-experimental (C/E) values for k ££ and k- £ in the enriched uranium fueled FCA-XIV cores and the

* Visiting scientist from I.S.L., Inc., Tokyo

jAERI-M 88-221

円しl

r白』h

gι FL

vd 司

↑L n

、内』-EJ -

T. Osugi， A. Ohno， H. Oigawa， K. Satoh， K. Hayasaka and M. Nagatani発

A series of experiments have been carried out at FCA to obtain

neutronic characteristics on the High Conversic Light Water Reactor

(HCLWR) core. The experimental program consists of two phases Phase-11)

with enriched uranium fueled core and Phase-22) with plutonium fueled

1n the Phase-2 experiment， particular emphasis is put on the

measurements of conversion ratio， moderator void worth and plutonium

isotopic composition effect， in addition to the HCLWR characteristics

studied iu the Phase-l experiment. The heterogeneity effects of the

plate-type and νtn-type fuels are studied at the core in which the mixed

oxide fuel are loaded at the central test zone.

The first core of the Phase-2 experiment， the FCA-XV-1 core， went

critical in May 1987. Fuじ1enrichment of the test zone cell "Pu08" is

8.11 %Pufiss/(pu+U) and moderator-to-fuel volume ratio is 0.6. Plutonium

isotopic composition in plate is 239pu/240pu/241pu 9 1.6/~.O/O.4.

Comparisons of the experimenta1 and calculated values are made for

the effective multiplication factor keff and the infinite multiplication

factor k，_&. The calculations were made with use of the SRAC code l.n工

system3) and 86-energy group cross-section set based on the JENDL-2 data

file斗)

Table 3.5.1 presents the experimental and calculated keff values.

The calculated keff is in good agreement with the e且perimentalone.

Table 3.5.2 presents the experimenta1 and calcu1ated kinf values for

the FCA-XV-l test zone cell. The experimental kinf value was deduced from

the measured material buckling Bm 2 and the tteoretical migration area

M2.5) The ca1culated kinf is a1so in good agreement with the experimental

Table 3.5.3 summarizes the calculated-to-experimental (C/E) values

:or keff and kinf in the enriched uranium fueled FCA-XIV cores and the

持 Visitingscientist from I.S.L.， Inc.， Tokyo

一回一

JAERI-M 88-221

plutonium fueled XV-1 core. From these comparisons i t can be pointed out that the present ca lcu la t ion (SRAC with JENDL-2) gives larger C/E values by about 1% for k f£ and k- f in the plutonium fueled HCLWR core than those in the enriched uranium fueled HCLWR core.

References 1) Osugi T. e t a l . : "3.6 FCA Phase-1 Experiment on HCLWR," JAERI-M 87-

126, p.57 (1987). 2) Osugi T. e t a l . : "3.10 FCA Phase-2 Exper imenta l Program on HCLWR,"

JAERI-M 87-126, p.69 (1987). 3) Tsuchihashi K. et a l . : "Revised SRAC Code System," JAERI 1302 (1986). A) Nakagawa T. (Ed.) : "Summary of JENDL-2 General Purpose F i le , " JAERI-M

84-103 (1984). 5) Osugi T. e t a l . : "3.7 C r i t i c a l i t y and K i n f Measurement a t the FCA XIV

and XIV-1(45V) Cores," JAERI-M 87-126, p.60 (1987).

Table 3. 5.1 Comparison of the experimental and calculated k . r f

values in the FCA-XV-1 core

Calculation (C) Experiment (E) C/E

Standard calculat ion c *' 1.0040 - -Correction factors =

Energy co l laps ing^' 0.9990 Transport < c } 1.0020

Corrected k . f r 1.0050 1.0055 ± 0.0002 1.000

(a) 3D-XYZ model. 10-energy group diffusion calculation (b) From 86- to 10-energy group correction (c) 2D-RZ model. 10-energy group S8-P0 transport calculation

- 5 9 -

JAERI-M 88・221

plutonium fueled XV-l core. From these comparisons it can be pointed out

that the present calculation (SRAC with JENDL-2) gi ves 1arger C/E va1ues by

about 1% for keff and kinf in the plutonium fueled HCLWR core than those

in the enriched uranium fueled HCLWR core.

References

1) Osugi T. et a1. "3.6 FCA Phase-l Experiment on HCLWR，" JAERI-M 87-

126， p.57 (1987).

2) Osugi T. et a1. "3.10 FCA Phase-2 Experimental Program on HCLWR，" JAERI-M 87-126， p.69 (1987).

3) Tsuchihashi K. et a1. "Revised SRAC Code System，" JAERI 1302 (1986).

4) Nakagawa T. (Ed.) "Summary of JENDL-2 General Purpose Fi1e，" JAERI司 M

84-103 (1984).

5) Osugi T. et a1. "3.7 Criticality and Kinf Measurement at the FCA XIV

and XIV-l(45V) Cores，" JAERI-M 87-126， p.oO (1987).

Table 3. 5. 1 Compadson of t..:le expedrr削 taland calculated k.rr

values in the FCA-XV-l core

Calculation (C) Ex戸rirr健nt(E)

Standard calculation<a】 1.0040

Correction factorョ:

Energy collapsing<‘〉 0.9990

Transport < < ) 1.0020

Corrected k. r r 1. 0050 1. 0055 :!: O. 0002

(a) 3D-XYZ即 del. 10-energy group diffusion calculation

(b) From 86-to 10-energy group correction

(c) 2D-RZ mode1. lO-energy group S8-PO trans戸町 calculation

J A E R I - M 8 8 - 2 2 1

Table 3 .5 .2 Comparison of the experimental and calculated k ( « f values for the FCA-XV-1 t e s t zone ce l l

Experiment ( E ) < 0 Calculation ( C ) C k ) C/E

B 2 (10" 4 cm" 2 ) 10.67 ± 0.49

B 2 (iO~ 4cm~ 2) 11.22 i 0.65

B 2 (10" 4 cm" 2 ) 7.83 ±0 .09

B 2 (10" 4 cnT 2 ) 29. 72 ± 0. 82 c '> 30.24 1.017 n

kin( 1.1718 ± 0.0049 c n 1.1748 1.003

(a) Errors quoted are estimated within 95% confidence level (b) 86-energy group cell calculation (c) B„ 2 = B, 2 + B, 2 + B, 2

(d) k,., = 1 + B B2M 2 (M2 = 57.80cm2)

Table 3. 5.3 Comparisons of the calculated-to-experimental (C/E) values for k,f f and k („ f in the enriched uranium fueled FCA-XIV cores and the plutonium fueled FCA-XV-1 core

C/E value in Core/Test zone c e l l

Item XIV-1/EU06A XIV-1(45V)/EU06A(45V) HV-2/EU05 XV-l/Pu08

k . r r

k | n f

0.986 0.987

0.992 0.989 1.000 0.989 0.989 1.003

JAERI・M 88・221

Iable 3.5.2 Co即 arisonof the ek~ri隅ntal and calculated k.Ar values

for吋leFCA-}{V-1 test zone cell

ExperIl田nt(E) (・〉 Calculation (C)(‘ C/E

。，-a

--E--aゐ

10.67 :1: 0.49

11.22土 0.65

丹，包内，M

9&ヲゐ

7.83:1: 0.09

29.72:!: 0.82(<) 30.24 1. 017

kinf 1.1718:1: 0.0049(d) 1. 1748 1. 003

(a) Errors quoted a目白ti胴 tedwithin 95% confid回 celevel

(b) 86-energy group cel1 calculation

(c) B..2 = B.2 + B，2 + B.1

(d) k. n r = 1 + B..2W (M2 = 57. 8Ocm2)

Iable 3.5.3 Comparisons of the calculated-to-experimental (C/E) val11e$ for

k. r r and k. n r in the enriched uraniurn fueled FCA-氾V∞民5and

the plutoniurn fueled FCA-}{v-l∞陀

C/E value in白re/Iestzone ce11

Item XIV-l/EU06A XIV-l(45V)/EU06A(4SV) 氾V-2/日J05 XV喧 1/Pu08

k. r r 0.986 0.992 0.989 1. 000

k. n r 0.987 0.989 0.989 1. 003

JAERI-M 88-221

3.6 Absorber Material Reactivi ty Worth in the FCA-XIV Cores

S. Okajima, T. Osugi, T. Sakurai, W. Satoh* and Y. Tahara**

Measurements of absorber r e a c t i v i t y worth were made a t the FCA-HCLWR c o r e s of XIV-1, XIV-1 (45V) and XIV-2 1 } . The hafnium p l a t e and the B^C p la tes with d i f fe ren t i so top ic compositions of natural (20%), 40%, 60% and 90% B contents were chosen as the candidates of absorber mater ia l s for HCLWR concept.

F igure 3.6.1 shows a p l a t e c o n f i g u r a t i o n of a fue l c e l l a s s o c i a t e d wi th the abso rbe r m a t e r i a l to s i m u l a t e the c o n t r o l rod. R e a c t i v i t y worths of fully inser ted control rods with d i f ferent con ten t s /mate r i a l s were measured a t the c e n t e r a x i s of core us ing the modif ied source m u l t i p l i c a t i o n method ' • ' . Sample r e a c t i v i t y wor ths of t he a b s o r b e r m a t e r i a l s a t t he core c e n t e r wi th the same u n i t c e l l as the s i m u l a t e d c o n t r o l r o d s were a l s o m e a s u r e d to e x a m i n e t h e v a l i d i t y of t h e ca l cu la t iona l method and data.

The measured rod worth r a t i o s of the abso rbe r m a t e r i a l s to the n a t u r a l B^C rod a r e shown in F ig . 3.6.2 as a func t ion of B c o n t e n t . In case of the B^C rod wor th s , the r a t i o i n c r e a s e s wi th i n c r e a s i n g B c o n t e n t : for the 90% 1 0 B rod, t he worth r a t i o s a r e 1.41, 1.61 and 1.38 in the XIV-1, XIV-1(45V) and XIV-2 c o r e s , r e s p e c t i v e l y . The hafnium rod worth r a t i o to n a t u r a l B^C rod worth i s 0.74 in the XIV-1 c o r e , 0.71 in the XIV-1(45V) core and 0.77 in the XIV-2 core.

An ana lys i s of absorber r e a c t i v i t y worth was car r ied out with use of the SRAC code system and an 86 group cross sect ion se t based on the JENDL-2 library" 4 ' ' . Simulated control rod worths were obtained from eigenvalue c a l c u l a t i o n s for whole r e a c t o r us ing the two-d imens iona l c y l i n d r i c a l diffusion theory. Sample r e a c t i v i t y worths were ca lcula ted with the exact p e r t u r b a t i o n theory based on the two-d iment iona l cy l i nd r i ca l diffusion t heo ry . The c a l c u l a t e d va lues of both s i m u l a t e d c o n t r o l rod wor ths and sample r e a c t i v i t y worths were corrected by the t ranspor t e f fec t .

To i n v e s t i g a t e the dependence of absorber m a t e r i a l wor ths on B content and neutron spectrum, both simulated control rod worths and sample r e a c t i v i t y worths are normalized to the sample r e a c t i v i t y worth of na tura l B/C a t the co re c e n t e r in co r re spond ing c o r e s . Comparisons of the

* On leave from Japan Information Service Co. Ltd. ** On leave from Mitsubishi Atomic Power Indus t r i e s , Inc .

- 6 1 -

jAERI-M 88・221

3.6 Absorber Þ~terial Reactivity Worth in the FCA-XIV Cores

S. Okajima， T. Osugi， T. Sakurai， W. Satoh* and Y. Tahara*者

Measurements of absorber reactivity worth were made at the FCA-HCLWR

cores of XIV-l， X1V-1 (4SV) and XIV-2 1). The hafnium p1ate and the B4C

p1ates with differe~t isotopic compositions of natura1 (20%)， 40%， 60% and

90% 10B contents were chosen as the candidates of absorber materials for

HCLWR concept.

Figure 3.6.1 shows a p1ate configuration of a fue1 cell associated

with the absorber material to simulate the control rod. Reactivity

worths of fully inserted control rods with different contents/materials

were measured at the center axis of core using the modified source

multiplication method2，3). Sample reactivity worths of the absorber

materials at the core center with the same unit cell as the simulated

control rods were a1so measured to examine the validity of the

calculational method and data.

The measured rod worth ratios of the absorber materia1s to the

natural B4C rod are shown in Fig. 3.6.2 as a function of 10B content ・ 1n

10 case of the B4C rod worths， the ratio increases with increasing ~vB

content for the 90% 108 rod， the worth ratios are 1.41， 1.61 and 1.38 in

the X1V-l， X1V-l(4SV) and X1V-2 cores， respective1y. The hafnium rod

worth ratio to natura1 B4C rod worth is 0.74 in the XIV-l core， 0.71 in

the XIV-l(45V) core and 0.77 in the XIV-2 corc.

An ana1ysis of absorber reactivity worth was carried out with use of

the SRAC code system and an 86 group cross section set based on the JENDL-

2 1ibrary4). Simu1ated contro1 rod worths were obtained from eigenvalue

calculations for whole reactor using the two-dimensional cylindrical

diffusion theory. Sample reactivity worths were calculated with the exact

perturbation theory based on the two-dimentional cylindr工caldiffusion

theory. The calculated values of both simulated contro1 rod worths and

sample reactivity worths were corrected by the transport effect.

To investigate the dependence of absorber materia1 worths on 10B

content and neutron spectrum， both simulated control rod worths and sample

reactivity worths are normalized to the sample reactivity worth of natural

B4C at the core center in corresponding cores. Comparisons of the

普 Onleave from Jap叩 InformationService Co. Ltd.

努箸 On 工eavefrom Mitsubishi Atomic Power Industries， Inc.

JAERI-M 88-221

c a l c u l a t e d and measured normal ized worths a r e shown in Fig .3 .6 .3 . The caluculated values of normalized sample r e a c t i v i t y worths agree well with the measured ones within experimental e r ro r s for d i f fe ren t °B contents in t h r e e k inds of c o r e s . On the o t h e r hand, t he c a l c u l a t i o n u n d e r p r e d i c t s the normalized hafnium r e a c t i v i t y worths by about 10% in XIV cores. The ca lcula ted to experimental (C/E) value for the normalized B^C rod worth i s about 1.07 in XIV-1 c o r e , 1.03 in XIV-1(45V) and 1.10 in XIV-2 c o r e . We did not obse rve the C/E dependence on * B c o n t e n t s in XIV c o r e s , but the C/E values increase as the neutron spectrum becomes soft .

References 1) Osugi T. e t a l . : "FCA phase-1 Exper iments on HCLWR," 3.6 in JAERI-M

87-126 (1987). 2) Mukaiyama T. e t a l . : "Reactivity Measurement in a Fa r -Subcr i t i ca l Fast

System ( I I ) , " JAERI-M 6067 (1975). 3) Mizoo N. : " T h e o r e t i c a l and Exper imen ta l S t u d i e s on Measurement of

Large Negative R e a c t i v i t i e s , " JAERI-M 7753 (1978). 4) Tsuchihashi K. e t a l . : "Revised SRAC code System", JAERI 1302 (1986).

XIV-1 ond XIV-1 (45VI XIV-2

Absorber Material -

20% Enriched Uranium metal

(1 = 1.5875 mm)

Natural Uranium metal

(t*1.5875mml

Depleted Uranium Oxide

(t * 6.35mm)

Aluminum Oxide

(I * 1.5875 mml

Polystyrene 0% Void

45% Void for XIV-1 (45V! core

(1=3.175 mm)

I 0%Void (or XIV-1 and XIV-2 cores) >4F,V. UniH fnr YIV- ! IdSU ! rnr* I

(t:thickness of plate)

F ig . 3 .6 .1 Plate configuration of a fuel cell associated with the absorber material

- 6 2 -

jAERI-M BB・221

ca1cu1ated and measured norma1ized worths are shown in Fig.3.6.3. The

ca1uculated va1ues of norma1ized samp1e reactivity worths agree we11 with

the measured ones within experimental errors for different 10B contents in

three kinds of cores. On the other hand， the ca1cu1ation underpredicts

the norma1ized hafnium reactivity worths by about 10% in XIV cores. The

ca1culated to experimental (C/E) va1ue for the norma1ized B4C rod worth is

about 1.07 in XIV-1 core， 1.03 in XIV-l(45V) and 1.10 in XIV-2 core. We

did not observe the C/E dependence on 10B contents in XIV cores， but the

C/E values increase as the neutron spectrum becomes soft.

References

1) Osugi T. et a1. "FCA phase-1 Experiments on HCLWR，" 3.6 in JAERI-M

87-126 (1987).

2) Mukaiyama T. et a1. I~eactivity Measurement in a Far-Subcritica1 Fast

System (II)，" JAERI-M 6067 (1975).

3) Mizoo N. "Theoretical and Experimental Studies on Measurement of

Large Negative Reactivities，" JAERI-M 7753 (1978).

4) Tsuchihashi K. et a1. "Revised SRAC code System"， JAERI 1302 (1986).

XIV-l and XIY-l (45YXIY-2

Absorber Malerial

1:… Uranium melal

(1 = 1.5875 mm I -5 」n

n川口U

川刷

'm，r

nMHMHMHHHHU

Ilmim Oxide

(t z 1.5875mml

11目目r目院P内州削川附Dω叫卵削l旬榊刷附ys阿51オ( o% vw刷州剖刷idfor XIV-1 End XIV-2 com ) 45"1. Yoid lor XIY-l (45Y l core

(t =3.175mml

( I : lh ickness 01 plate I

Fi g. 3.6.1 PI日teconligurotion of 0 fuel cell ossoci日tedwith 愉eobsorber maleriol

o XIV-I O XIV-I (45VI a XN-Z

XIV-1 (45V)

—Cole.

0.5 C/R full insertion

JL 20 40 60 80 100 Hf , 0 B enrichment (atom 7.) of B4C/Hf

Fig. 3.6.2 Experimental results of various obsorber material worth ratios, (rel.to not. B4C worth)

O XIV-I O XIV-1 (45V) A XIV-2

Central reactivity worth i 1 1. 1 ' « "

1 - 1 1 -

- O XIV-1 O XIV-1 (45V) " A XIV-Z

-C/R full

insertion

0 20 40 60 80 100 Hf W B enrichment (otom %) of B4C /Hf

Fig. 3.6.3 Comparison of normalized reactivity worth between calculated (C) ond measured IE! in FCA XIV cores

1.1 o XIV-I 。XIY-1145YId. XIY-2

1.01 g-g一一-~

Centrol reoclivily wor1b

'「qd nu

o XIY-I o XIY-l145YI (; XIV-2

」〉伺刃

7玄∞∞・NN-

0.8ト

o XIY-l o XIY-1I45YI寸

(; XIY-2 1.2

XIY・1XIY咽 2

ー-Colc.

~ 1.0 g a

。。。。。

品1.1

w ¥1.0 、J (; 。

C/R full inserlion

20 40 60 80 ∞ IOS enrichmenl lolom'.) of B4C IHf

0.9 C/R full insertion

20 40 60 80 108 enrichment (olom -，.) of 84C' Hf

Fig. 3.6.3 Compロrisonof normolized reoctivity worlh bet悦朗

CロIculoled(Cl ond rre刷 f凶 IElin FCA XIV cores Fi g. 3.6.2 Experimenlol re釦 Itsof voriousぬsorbermoler iol worth

rotios. (re¥.to no1. B4C worth)

JAERI-M 88-221

3.7 Measurement of Reaction Rate Ratios at FCA-XIV Cores

M. Obu, T. Nemoto, T. Sakurai, S. Iij ima and Y. Tahara*

Reaction r a t e r a t i o s to " 5 U fiss ion were measured in three uranium-fue led c o r e s of FCA (XIV-1(45V), XIV-1 and XIV-2 c o r e s ) which s i m u l a t e d the HCLWR core spec t rum. The c o r e s possess the c e n t r a l t e s t zones wi th d i f fe ren t fuel enrichments and moderator to fuel volume r a t i o s . ' P la te c o n f i g u r a t i o n s in the t e s t zone c e l l of t he FCA-XIV c o r e s a r e shown in F ig .3 .7 .1 . For t he FCA XIV-1(45V) not i l l u s t r a t e d in F i g . 3 . 7 . 1 , the c e l l arrangement i s i d e n t i c a l with that of the FCA XIV-1, except for the use of foamy p o l y s t y r e n e p l a t e s (45% void) as the modera tor p l a t e . The c a l c u l a t i o n i n d i c a t e s t h a t neut ron s p e c t r a in the FCA XIV-1(45V), XIV-1 and XIV-2 successively change from hard to soft with varying the atomic number densi ty r a t i o (H/U).

The f i s s ion r a t e s of 2 3 9 P u ( F 4 9 ) and 2 3 8 U ( F 2 8 ) to 2 3 5 U f i ss ion ( F 2 5 ) were measured by cy l i nd r i ca l micro f i s s i o n c o u n t e r s w i th t h i n aluminum wall . The counter i s 6mm in diameter and 32mm in e f fec t ive length. The counter was inser ted a t the center c e l l posi t ion through the experimental ho l e . The 2 3 8 U c a p t u r e r a t e (C ) t o 2 3 5 U f i s s i o n r a t e was measured by using me ta l l i c depleted and 93% enriched uranium f o i l s . The f o i l s have a d i sk shape of 12.7mm in d i a m e t e r . The t h i c k n e s s of d e p l e t e d and 93% e n r i c h e d uranium f o i l s were of 0.127mm and 0.025mm r e s p e c t i v e l y . The f o i l s were mounted in the d e p l e t e d UO2 p l a t e a t the c e n t e r c e l l . F o i l

a c t i v i t i e s were de te rmined by us ing the T - r ay s p e c t r o s c o p y system 235 connec ted w i t h a G e - d e t e c t o r . F i s s i o n in i J J U was measured by f i s s i o n

p r o d u c t a c t i v i t i e s from 1 A 3 C e (293 .2keV) , 133I(529.8keV) and 97 238 7 Zr(743.4keV)), while capture in i J O U was measured by the activity from 239Np(277.6keV). For 2 3 8 U capture rate, a neutron self-shielding effect inside the thick depleted UO2 plate was estimated by measuring small foils which were arranged in the plate.

Reaction rate ratios were analyzed by the cell calculations under the condition that the fundamental mode spectrum was well established in the central part of the test zone ' where the measurements were mode. The

* On leave from Mitsubishi Atomic Power Industries, Inc.

jAERI-M 88・221

3.7 Measurement of Reaction Rate Ratios at FCA-XIV Cores

M. Obu， T. Nemoto， T. Sakurai， S. Iijima and Y. Tahara持

Reaction rate ratios to 235U fission were measured in three uranium-

fueled cores of FCA (XIV-l(45V)， XIV-l and XIV-2 cores) which simulated

the HCLWR core spectrum. The cores possess the central test zones with

different fuel enrichments and moderator to fuel volume ratios.l) Plate

configurations in the test zone cell of the FCA-XIV cores are shown in

Fig.3.7.1. For the FCA XIV-l(45V) not illustrated in Fig.3.7.1， the cell

arrangement is identical with that of the FCA XIV-l， except for the use of

foamy p01ystyrene p1ates (45% void) as the moderator p1ate. The

calculation indicates that neutron spectra in the FCA XIV-1(45V)， XIV-l

and XIV-2 successively change from hard to soft with varying the atomic

number density ratio (H/U).

The fission rates of 239pu (F49) and 238U (F28) to 235U fission (F25)

were measured by cy1indrical micro fission counters with thin aluminum

wall. The counter is 6mm in diameter and 32mm in effective 1ength. The

counter was inserted at the center cell position through the experimental

hole. The 238U capture rate (C28) to 235U fission rate was measured by

using meta11ic dep1eted and 93% enriched uranium foi1s. The foi1s have a

~isk shape of 12.7mm in diameter. The thickness of depleted and 93%

enriched uranium foils were of O.127mm and O.025mm respecti ve1y. The

foi1s were mounted in the dep1eted U02 p1at~ at the center ~e11. Foi1

activities were determined by using the r-ray spectroscopy system 235 connected with a Ge-detector. Fission in .....J..JU was measured by fission

product activities from 143Ce (293.2keV)， 133I(529.8keV) and 977_，，，，. 'l '.'.~I1\\ . ....~， ~ _~_...._~ "_ 238 Zr(743.4keV))， while capture in .....JoU was measured by the activity from

239，，_，....，.， .::.，._"， '1:'__ 238 Np(277.6keV). For .......OU capture rate， a neutron se1f-shielding effect

inside the thick dep1eted U02 p1ate was estimated by measuring sma11 foi1s

which were arranged in the p1ate.

氏eactionrate ratios were ana1yzed by the cell calculations under the

condition that the fundamental mode spectrum was we11 estab1ished in the

central part of the test zone2) where the measurements were mode. The

長 Onleave from Mitsubishi Atomic Power Industries， Inc.

JAERI-M 88-221

SRAC code system- 5 ' with the l i b r a r y based on the nuclear data f i l e JENDL-2 was used for the c a l c u l a t i o n s . The c e l l c a l c u l a t i o n s were made by one-dimensional i n f i n i t e s lab model and 86-energy group constants .

Central react ion r a t e r a t i o s and C/Es in FCA XIV cores are summalized in Table 3.7.1. The c a l c u l a t i o n s a r e shown to unde rp red i c t ed the experimental values of F 9 / F 2 5 in the three cores. The r e s u l t s of C/E show t h a t for i m p o r t a n t pa ramete r C /F , the ca lcu la t ions overpredict the experimental values by 8 to 10%.

References 1) Osugi T., Ohno A., Okajima S e t a l . : "FCA Phase-1 Experiment on

HCLWR," 3.6, JAERI-M 87-126 (1987). 2) (5bu M., Nemoto T., I i j i m a S. e t a l . : "Measurement of Reac t ion Rates

in t he FCA XIV-1 Core ," i b id .3 .9 (1987). 3) T s u c h i h a s h i K. e t a l . : "SRAC : JAERI Thermal Reac tor Standard Code

system for Reactor Design and Analysis ," JAERI 1285 (1983).

Table 3.7.1 Reaction rate rat ios and C/E in FCA-HCLWR cores

Core Reaction rate ratio

Experimental Calculated SRAC

FCA p * » / f 2 6 »> 2. 083 ± 3. 0% 1.894 0.910 XIV-1 p l ' / P " «> 0.00798 ± 5. 0% 0. 00881 1.104 (45V) p2 f / p 2 5 b > 0. 05489 ±2.2% 0. 05917 1.078

FCA F«»/p2« •) 2. 353 ± 3.1% 2.248 0.955 XIV-1 p»»/p*« •> 0. 00568 ± 6. 0% 0. 00635 1.118 (reference) C 2 7 F 2 5 k ) 0. 03768 ± 2. 8% 0.04144 1.100

Fa p»/p*s »> 2. 282 ± 3.1% 2.272 0.996 HV-2 pj«/p25 .) 0. 00288 ± 9.1% 0. 00348 1.208

C " / F 2 S > } 0. 02413 i 2. i% 0. 02623 1.087

a) : Measured by micro f iss ion counter b) : Measured by f o i l s in depleted UQt plate

- 6 5 -

]AERI胴 M 88・221

SRAC code system3) with the library based on the nuclear data file JENDL-2

was used for the calculations. The cell calculations were made by one-

dimensional infinite slab model and 86-energy group constants.

Central reaction rate ratios and C/Es in FCA XIV cores are summalized

in Table 3.7.1. The calculations are shown to underpredicted the

experimental values of F49/F25 in the three cores. The results of C/E show 28/"，25 that for important parameter C"o/F"".J， the calculations overpredict the

experimental values by 8 to 10%.

References

1) Osugi T.， Ohno A.. Okajima S et a1. "FCA Phase-1 Experiment on

HCLWR，" 3.6， JAERI-M 87-126 (1987).

2) 百buM.， Nemoto T.， Iijima S. et a1. "Measurement of Reaction Rates

in the FCA XIV-1 Core，" ibid.3.9 (1987).

3) Tsuchihashi K. et a1. "SRAC JAERI Thermal Reactor Standard Code

system for Reactor Design 3nd AIlalysis，" JAERI 1285 (1983).

Iable 3. 7. 1 Reaction rate ratios and C/E in FCA-HCLWR cores

仁:Ore Reaction Experir臨 ltal Calωlated C!E rate ratio SRAC

FCA F4・IF26 .) 2.083土 3.0% 1.894 0.910

XIY-l F" IF' & ・2 0.00798:t 5.0% 0.00881 1. 104

(45V) C'・IF%5 b) 0.05489土 2.27. 0.05917 1. 078

FCA F‘'IF25 ・3 2.353土 3.1% 2.248 0.955

XIY-} F'・IF'5 .) 0.00568 :t 6.0χ 0.00635 1.118

(refe四 n但) C2・/F%5 .) 0.03768:t 2.8% 0.04144 1. 100

FCA F‘， IF2 & .) 2.282:t 3.1% 2.272 0.996

XIY-2 F2・IF2S .) 0.00288:t 9.1% 0.00348 1.208

C2・IFZ5、】 0.02413土 2.4% 0.02623 1. 087

a) He田町をdby micro fission counter

b) : He田 uredby foils in depleted l幻2plate

J A E R I - M 8 8 - 2 2 1

f- • 50 .8 m ^ .. . *>

- i • !: : ] -

• - • • !

! • •

_' _ _ i . • : * : # : • ' i

- '. - l;;tt¥;; -. i_ ;.-.:;'v -

_:!!:_•_:• I- 1 • r FCA XIV-I cell

r 5 0 . 8 ^

FCA XIV-2 eel

20% enriched U metal

Natural U metal

Depleted U02

A/,0 2^3 Polystyrene

Fig .3 .7 .1 P la te conflgla t lon of t e s t zone c e l l of FCA XIV-1 and XIV-2

- 6 6 -

]AERI-M 88・221

50.8 mm 「一一5oj--1ー

11111111111111111

111- 1111111111111111

FCA XIV-I cell FCA XIV-2 cell

~ ~日 enrichedU metal U metal U02

~ Polystyren

Fig.3.7.1 Plate configlation of test zone cell of FCA XIV-l

and XIV-2

JAER1-M 88-221

3.8 Measurement of Adjoint Flux Weighted In f in i t e Mul t ip l ica t ion Factor Using Central Cell React ivi ty Worth a t the FCA-XIV-1 and -XIV-2 Cores

T.Sakurai , S.Okajisna and T.Osugi

To v e r i f y the r e l i a b i l i t y of the e x p e r i m e n t a l r e s u l t s of i n f i n i t e m u l t i p l i c a t i o n f a c t o r K^ , t h e a d j o i n t f l u x w e i g h t e d i n f i n i t e m u l t i p l i c a t i o n factor K +

m * ' was measured using the r e a c t i v i t y worth of the cen t r a l uni t c e l l a t FCA-XIV-1 and -XIV-2 c o r e s 2 ) . The measured K +

m value was used to o b t a i n the KQQ va lue by a d i f f e r e n t method from the c o n v e n t i o n a l ones such as measurement of g e o m e t r i c a l buck l ings in t h e medium.

The K m va lue was measured by the method which was developed in the research reac tor PROTEUS at the Paul Scherrer I n s t i t u t e in Switzer land 3 ^. When the t e s t zone of the reac tor i s large enough for the fundamental mode to be es tab l i shed in the cen t r a l uni t c e l l , K +

m i s defined as Eq.(l) which can be transformed in to Eq.(2).

«<t¥F <t>» K\ : C 1 )

«9i + Hrf» where (j> . <b : neutron flux and ad jo in t flux in the c e n t r a l un i t c e l l

F : operator represent ing neutron production ( f i s s ion) H '• o p e r a t o r r e p r e s e n t i n g n e u t r o n d i s a p p e a r a n c e

(absorption and s c a t t e r i n g ) << >> : i n t eg ra t ion with respect to whole energy and volume in

the c e n t r a l un i t c e l l

V m = C 2 ) P c . I 1

1 ( 0 c r / S C f ) V R ,

The K<JO value was obtained by combination of the reactivity worth of the central unit cell {P «.ii), the reactivity worth of 2 5 2 C f neutron source (Per ) as fission neutron importance, the intensity of 2^ 2Cf neutron

•67-

JAERI-M 88・221

3‘8 Measurement of Adjoint F1UK Weighted Infinite Nultiplication Factor

USing Central Cell Reactivity Worth at the FCA-XIV-l and -XIV-2 Cores

T.Sakurai. S.Okajima and T.Osugi

To ¥'erify the reliability of the experimental resulls of infinite

mult~plication factor K伺 .the adjoint flux weighted infinite

multiplication factor K+m1) was measured using the reactivity worth of the

central unit cell at FCA-XIV-l and -XIV-2 cores2). The measured K・4・m value

was used to obtain the K∞ value by a different method from the

conventional ones such as measurement of geometrical bucklings in the

medium.

The K+ m val ue was measured oy the method which was developed in the

research reactor PROTEUS at the Paul Scherrer Institute in Switzerland3).

When the test zone of the reactor is large enough for the fundamental mode

to be established in the central unit cell， K+m is defined as Eq.(l) which

can be transformed into E司.(2).

K+，. 、，，

、3''

、/-、J

一，φ

a--FA守

《↑一《

o ， o・:F

くく >> :

neutron flux and adJoint flux in the central unit cell

operator repregenting neutron production (fi99ion)

operator representing neutron disappearance

(absorption and scattering)

integration with respect to whole energy and volume in

the central unit cell

( 2 ) K+岨

ρc • I 1

-4 〈ρcr/ S c r ) i7 R r

The Koo value was obtained by combination of the reactivity worth of the

central unit cell (ρ<. 11)， the reactivity worth of 2S2Cf neutron source 252 (ρcr ) as fission neutron importance， the intensity of 'J'Cf neutron

JAERI-M 88-221

source ( S c r )i s p a t i a l l y in tegrated f i ss ion r a t e in the c e l l (Rj) and the number of n e u t r o n s per f i s s i o n ( "j7 ). The r e a c t i v i t y worth o f / 0 < . i i and Pet were measured by the ca l ib ra ted control rod. The Re was determined from the absolute f i ss ion r a t e measurements with f i s s ion f o i l s of enriched uranium and d e p l e t e d uranium, and from the c a l c u l a t e d f i s s i o n r a t e d i s t r i b u t i o n s in t he c e l l . The Rr and Per were no rma l i zed to the same neutron power l eve l . The V was obtained from c e l l ca lcu la t ion .

The SRAC code system ' and a 86-energy group cross sec t ions se t based on the JENDL-2 l i b r a ry were used for the c e l l ca l cu la t ion . The conversion f a c t o r f de f ined as t he r a t i o of Kco to K +

m was o b t a i n e d from 86-energy group R-Z diffusion core ca lcu la t ion .

The measured K +

m v a l u e s and the c a l c u l a t e d f a c t o r f a r e shown in Table.3.8.1. The Koo value was obtained experimental ly as ; Kco = K m* f The e r r o r s in K +

m values, for which only the random e r r o r s in the measured values were considered, a re within 1%. Comparison of KM values obtained from the p r e s e n t method and buck l ing method i s a l s o shown in Tab le .3 .8 .1 . The Kco value by present method agrees well with the KM value by buckling method for FCA-XIV-2 core. On the other hand, s i ng i f i c an t discrepancy i s found between t h e s e v a l u e s for FCA-XIV-1 c o r e . The cause of t h i s discrepancy i s being inves t iga ted .

References 1) Nakano M. e t a l . : " In t e rp re t a t i on of the Central Cel l React ivi ty Worth and Experimental Determinations of a Charac te r i s t i c Value of the Reactor C e l l Compos i t ion K +

m , " J . Nucl. S c i . Technol . , 10 [2 ] .69 (1973). 2) Osugi T. e t a l . : " R e s u l t s and Analyses for FCA p h a s e - I Exper iment on High Conversion Light Water Reactor," NEACRP-A-843, September,1987. 3) S e i l e r R. e t a l : " I n v e s t i g a t i o n of the Void C o e f f i c i e n t and Other I n t e g r a l P a r a m e t e r s in the PROTEUS-LWHCR Phase I I P rograme ," Nucl. Technol . ,80[2] .311(1988) . 4) T s u c h i h a s h i K. e t a l . : "Revised SRAC Code System, "JAERI 1302 (1986).

- 6 8 -

]AERI-M 88・221

source (S c t ). spatially integrated fission rate in the cell (Rf) and the

number of neutruns per fission 官).The reactivity worth oIρe・II and

ρcr were measured by the ca1ibrated contr01 rod. The Rf was determined

from the abso1ute fission rate measurements with fission foi1s of enriched

uranium and dep1eted uranium， and from the ca1cu1ated fission rate

discributions in the ce11. The Rf and ρcr were norma1ized to the same

neutron power level. The V was obtained from cell calculation.

The SRAC code system4) and a 86-energy group cross sections set based

on the JENDL-2 1ibrary were used for the cel1 ca1culation. The conversion

fac tor f defined as the ratio of K∞ to K+ m was obtained from 86-energy

group R-Z diffusion core ca1cu1ation.

The measured K+m va1ues and the ca1culated factor f are shown in

Table.3.8.1. The Kωva1ue was obtained experimenta11y as; Kω=K+m-f

The errors in K+m va1ues， for which on1y the random errors in the measured

岨 1ues were considered， are within 1%. Comparison of K伺 va1ues obtained

from the present method and buckling method is a1so shown in Table.3.8.1.

The K~ value by present method agrees well with the Kωvalue by buckling

method for FCA-XIY-2 core. On the other hand， singificant discrepancy is

found between these va1ues for FCA-XIY-1 core. The cause of this

discrepancy is being investigated.

References

1) Nakano M. et a1. "Interpretation of the Central Ce11 Reactivity Worth

and Experimental Determinations of a Characteristic Ya1ue of the Reactol'

Cell Composition K+ m'" J. Nuc1. Sci. Techno1.， 10 [2].69 (1973).

2) Osugi T. et a1. "Resu1ts and Analyses for FCA phase-I Experiment on

High Conversion Light Water Reactor，" NEACRP-A-843， SeptembeI， 1987.

3) Sei1er R. et a1 "Investigation of the Yoid Coeffic1ent and Other

Integral Parameters 1n the PROTEUS-LWHCR Phase II Programe，" Nuc1.

Techn01.，80[2].311(1988).

4) Tsuchihashi K. et a1. "Revised SRAC Code System， "JAERI 1302 (I986).

J A E R 1 - M 8 8 - 2 2 1

Table 3. 8. 1 Comparison of the measured K*„ and Km values for FCA-XIV-1 and -XIV-2 cores

XIV-1 XIV-2

Ko,, (present method) * )

^ ( b u c k l i n g method)"

1.209 ± 0.018

1.205 ± 0.018

1.176 ± 0.004

1.222 ± 0.013

1.211 ± 0.013

1.206 ± 0.007

a) Koc = f • K*„ b) K w = 1 + B 2- M 2 ( B 2 : material buckling , M 2 : migration area )

JAERI-M 88-221

Iable 3. 8. 1 Comparison of the鵬 asuredK+.. and Koo values for FCA-XIV-l and -XIV-2 cores

況Y-1 XIY-2

K令市 1. 209 :!: 0.018 1.222土 0.013

f 0.997 0.991

K同 (presentmethod)・1 1. 205 :!: 0.018 1. 211 :!: 0.013

K∞(buckling腿 thod)‘〉 1. 176 :!: O. 004 1. 206 :!: O. 007

a) K伺= f. K+描

b) Kω= 1 + B 2. M 2 (B 2 : matet'Ial buckling ，M Z : migration area )

JAERI-M 33-22!

'i. Advanced Reactor System Design Studies

Scudies of several advanced reactor concepts have been performed to expand the reacor application in more safe and economic manner.

From the view point of effective utilization of fuel resources. High Conversion Light Water Reactor(HCLWR) and new concepts of FBR have been assessed. In the field of HCLWR, a new concept of flat core with blanket was proposed to achieve higher conversion ratio with improvement of void coefficient into more negative direction. The plutonium enrichment is reduced to 11%. A burnup calculational method in blanket was developed and dependence of plutonium production in blanket on burnup was analyzed. Effect of spectrum shift on burnup was studied for both cases of insertion and extraction conditions of fertile rods. Treatment of the effect of a large resonance of Pu-24Q on reactivity was also investigated. To see the dependence of HCLTO's characteristics on cross section sets and analysis method, an international benchmark problem was analyzed in eight countries and reviewed.

A new concept of high breeding and high burnup FBR with FP gas purge/tube-in-shell type plutonium alloy fuel assemblies was proposed. Its breeding ratio is more than 1.8 and the doubling time is less than seven years because of its very hard neutron spectrum. Even by uranium metal fuel, plutonium production capability is very high. Feasibility of conventional type FBR of nitride and carbide fuel were also assessed.

From the view point of further improvement of safety, a small simplified integrated reactor without control rods, SPWR, was conceptually designed. In its reator vessel, the core was enveloped by a large poinson tank from which borated water comes into the core by automatic opening of hydraulic pressure valves when a main pump loses its function. Another concepts of the inherently safe reactor, ISER, was also studied.

From another aspect of safety, feasibility of concepts of the TRU burning reactor with fuel only by TRU was invetigated as an upgraded technology of high level waste treatment. Unknown material properties of TRU fuel were metallurgically estimated. Build-up of fission products and minor actinide in PWR was analyzed and effect of uncertainty of nuclear data of TRU on reactor design was examined.

From the view point of improvement in economics of LWR, the nuclear scenario to achieve very high burnup of more than 100 GWD/t was investigated and a story by application of spectrum shift for a wide range of fuel/water ratio by special absorption rods of depleted uranium was proposed.

(Toru Hiraoka)

J A E R 1 -.¥! 占ヨ ~ ~ 1

~. Advanced Reactor Svst竺mDesi号nStCldi己g

Studies of severa1 advanced r自actorconcepts have been performed to expand the reacor application 1n more safe and自conomicmanner.

From the v1ew point of effectlve uti1ization of fuel resources. High Convers1on Light Water Reactor(HCLWR) and new concepts of FBR have been assessed. In the f 1e1d of HCLWR， a new concept of f lat core w1th blanket was proposed to achieve higher conversion ratio w i t h improv邑ment of void coefficient into more negative direction. The plutonium enr ichment 1s reduce(1 to 11%. A burnup calculational m白thod ln blank自t was deve loped and dependence of p1utonium product lon in blank自t on burnup was ana lyzed. Effect of spectrum shift on burnup was studied for both cas号5 of insertion and extraction conditions of fertile rcds. Treatment of th自 effect of a large r邑sonance of Pu-240 on reactivity was a150 investigated. To see the d自P由nd邑n巴自白f HCLWR' 5 characteristlcs on cross section 5ets and ana1ysis method， an internationa1 benchmark problem was analyzed in e1ght countri邑5and rev1ewed.

A new concept of h1gh breed1ng and h1gh burnup FBR w1th FP gas purge/tube-1n-shell type plutonium alloy fuel assernblie5 wa5 proposed. Its breed1ng rat10 15 mor自 than 1. 8 and th自 doubling time 1s 1由ss than seven years becaus告 of 1 ts very hard n日utronspectrulr，. Even by uran1um metal fuel， plutonium production capabil1ty 1s very high. Feas1bllity of c:onv自nt1ona1 type FBR of nitride and carbide fuel were a1so assessed.

From the v iew point of further improvement of safety， a small simp1ified integrated reactor without control rods， SPWR， was conceptua11y desi宮ned. 工n its reator vesse1， the core was enveloped by a large poinson tank from whic:h borated water c:omes into the core by automatic opening of hydraulic pressure valve5 when a main pump loses its fun巳tion. Another concept5 of the inherent1y 5afe reac:tor， ISER， was a1so studi由d.

From another aspect of 5a王ety，feasibility of concepts 0:1: the TRU burning reactor wi th fuel on1y by TRU was invetigated as an upgraded techno10gy of high 1eve1 waste treatment. Unknown materia1 propertie.s of TRU fuel were metaUurgically estimated. Build-up of fission products and minor actinide in PWR was analyzed and effect of uncertainty of nuclear data of TRU on r自actordesign was examined.

From the view point of improvement in ec:onomics of LWR， the nuclear scenario to ac:hieve very high burnup of mor自 than 100 GWD/

(Toru Hiraoka)

JAERI-M 88-221

4.1 The Concept of High Conversion Light Water Reactor with Flat Core and Its Applications

Y. Ishiguro and K. Okumura

Particularly important for the design study of high conversion light water reactors (HCLWRs) using M O X are the problems of the positive void reactivity coefficient and the large plutonium inventory. The former problem will degrade the function and performance, while the latter will restrict the number of H C L W R s which are to be introduced.

A new concept is proposed for the H C L W R to achieve both of high conversion and high burnup under the condition of negative void reactivity coefficient. This H C L W R has a flat core with thick axial blankets - so-called pancake type. With the flat core, the problem of the positive void reactivity coefficient of H C L W R can be mitigated by neutron leakage, and a fuel assembly of very tight lattice pitch can be used. The leakage neutrons are positively utilized in the axial blanket to enhance the conversion ratio.

In this report, the H C L W R s of the core height ranging from 50 to 60 cm are discussed, while the blanket thickness is assumed to be 30 cm. The volume ratio of moderator to fuel (Vm/V/) under study ranges from 0.5 tc 1.4. At first, in order to grasp the overall core physics characteristics of the flat core with the blankets, the burnup analyses based on a one— dimens ional diffusion code have been performed in infinite slab geometry assuming a radial buckling. The effective multi-group cross sections with burnup dependence have been calculated by the SRAC code .

On attaching the axial blankets, the effect of neutron reflection is enhanced predominantly for the flat core and the resulting ft,// increases, when compared with the bare core. This increase in k t i ;

means that the plutonium enrichment can be largely reduced from that determined by the cell calculation, say, from 15.6% to about 1 IS for the case of V m/V/=0.5. Moreover, upgraded burnup characteristics can be obtained by the accumulation of fissile plutonium in the blankets. This is just a new fact that has been gained through the study of the H C L W R with the flat core. It is this fact to make it possible to enhance both of the conversion ratio and burnup under the condition of negative void coefficient.

- 7 1 -

]AER1-M 88・221

4.1 The Concept of High Conv巴rsionLight Water R巴actorwith

Flat Core and Its Applications

Y. Ishiguro and K. Okumura

Particularly important for the design study of high conversion

1 ight water reactors (HCLWRs) usinlI MOX are the problems of the

positive void reactivity co白Uicientand the large plutonium inventory.

The former problem will degrade the function and performanc由. whi le

the latter will restrict the number of HCLWRs which are to be

introduced.

A new concept is proposed for the HCLWR to achieve both of high

conversion and high burnup under the condition of negative void reac-

tivity coefficient. This HCLWR has a flat core wi.th thick axial

blankets -so-cal1ed pancak自 type. Wi th th自 flat cor自. the problem of

the positive void r日activitycoefficient of HCLWR can be mitigated by

neutron leakage. and a fuel assembly of v自ry tight lattice pitch can

be used. The leakage neutrons are positiv自 ly utilized in the axial

blanket to enhance the conversion ratio.

1n this report， the HCLWRs of the core he ight rang ing f rom 50 to

60 cm are discussed， whi le the blanket thickness is assumed to be 30

cm. The volume ratio of moderator to fuel (Vm/Vf) under study ranges

from 0.5 tc 1.4. At first. in order to grasp the overall core physics

characteristics of thta flat core with the blankets. the burnup analyses

based on a one-dimens:onal diffus;on code have been performed in infi-

nite slab geometry assuming a radial buckling. The effective multi-

group cross s.ections with burnup dependence have been calculated by

the SRAC code.

On attaching the axial blankets. the effect of neutron reflection

is enhanced predominantly for the flat core and the resulting k.ff

increases. when compared wi th the bare core. This increase in k.f.'

means that the plutonium enrichment can be largely reduced from that

determined by the cel1 calculation. say， from 15.6% to about 11" for

the case of V m/Vj=O. 5. Moreover. upgraded burnup character i雪 tics can

be obtained by the accumulation of fissile plutonium in the blankets.

This is just a new fact that has been gained through the study of the

HCLWR with the flat core. [t is this fact to make it possible to

enhance both of the conversion ratio and burnup under the condi tion of

negative void coeCCicient

J A E R 1 - M 8 8 - 2 2 1

In the pancake type HCLVVRs, power peaking factor defined as the peak value of the power distribution divided by the average power in the core is very small due to the neutron reflection, compared with the case of a usual cosine distribution. This fact means that the average linear heat rating can be increased for a given radius of fuel pin or the radius of fuel pin may be reduced for a given average linear heat rating. Hence, the smallness of power peaking factor is one of the features on the H C L W R with the flat core and the axial blanket.

Since our concerns are the flat core with short height, the transport effect is considered to be important. Hence, a validation check has been made to the method adopted for the present analysis by using the one dimensional transport code ANISN and the Monte Carlo code VIM. The transport effect is considerably large and ranges from about 0.4% to 0.8%( Aklk) , depending on Vm/V/, plutonium enrichment and fertile material. Therefore, we can expect a much longer burnup time, compared with the present results based on the diffusion calculation. In another way, a more improved conversion ratio will be obtainable by reducing the plutonium enrichment, thus the plutonium inventory. Moreover, the void reactivity coefficient evaluated by combined use of the cell code and the one-dimensional diffusion code is shown to be conservative, compared with the V I M code.

Table 4.1.1 shows the typical burnup characteristics of pancake type H C L W R . Even for such a tight pitch lattice, a non-positive void coefficient can be obtained by the proper choice of the plutoniuxn enrichment. Hence, with the concept of the H C L W R with the flat core, a value near unity can be expected for the plutonium surviving rate (PSR) defined as the total plutonium at EOC(end of cycle) divided by that at BOC(beginning of cycle). A high burnup more than 45000 MWd/tonne can be anticipated if the transport effect is properly taken into account.

One of the shortest ways to apply the concept of the flat core is its direct use in an L W R of small or intermediate scale. On the other hand, the needs for large power output can be coped with by stacking the two flat cores with blankets. A variety of applications of the concept of the flat core can be considered for an L W R system of next generat ion.

- 7 2 -

JAERI-M 88・221

In the pancake type HCLWRs， power peaking factor defined as the

peak value of the power distribution divid自d by the av自rage power in

the core is very sma11 due to the neutron reflectio:l. compared with

the case of a udual cos in自 distribution. This fact means that the

av自rage linear heat rating can be increased for a given radius oC Cuel

pin or the radius of fuel pin may be redur.ed for a given average lin自ar

h自at rating. Hence. the smallロess of power peaking factor is one of

the features on the HCLWR with the flat core and the axial blanket.

Since our concerns are the flat core with short height. the

transport effect is consider自d to be important. Hence， a validation

check has been made to the method adopted for the present analysis by

using the one dimensional transport code AL'Ij'ISN and the Monte Carlo code

VIM. The transpor t日ffect is considerably larg合 andranges from about

0.4% to 0.8%(Jk/k)， depending on Vm/V/， plutonium enrichment and fer-

tile material. Therefore， we can expect 8 much longer burnup time，

compared with the present results based on the diCfusion calculation.

In another way， a more improved conversion ratio will be obtainable

by reducing the plutonium enrichment. thus the plutonium inventory.

Moreover， the void reactivity coefficient evaluated by cロmbineduse of

the cell code and the one-dimensional diffusion code is shown to be

conservative. compared with the VIM code.

Table 4.1.1 shows the typica1 burnup characteristics of pa.ncake

type HCLWR. Even for such a tight pitch lattice. a non-positive void

coefficient can be obtained by the proper choice of the plutonium

enrichment. Hence. 宵 iththe concept of th由日CLWRwith the flat core，

a value near uni ty can be expected for the plutonium surviving rate

(PSR) defined as the total plutonium at EOC(end of cycle) divided by

that at BOC(beginning of cycle). A high burnup more than 45000

MWd/tonne can be anticipated if the transport effect is properly taken

into account.

One of the shortest ways to apply the concept of the flat core is

its direct use in an LWR of small or intermediate scale. On the other

hand. the needs for large power output can be coped with by stackiag

the two f1at cores with blankets. A variety of applications (lf the

concept of the flat core can be considered for ac LWR system of next

generatlon.

J A E R 1 - M 8 8 - 2 2 1

Table 4.1.1 Burnup Characteristics of Pancake Type HCLWR Evaluated by 1-0 Calculation

Core Thermal Output(MWt) 1042 1714 1714 Equivalent Core Diaitieter(cm) 400 469 450 Fissile Core Height(cm) 50 50 60 Blanket Thickness(cm) 30 30 30 V m/Vf(in Assembly) 0.5 0.7 0.9 Fuel Pin Diameter(cm) 0.95 0.83 0.83 Loading Fissile Pu Enrichment(wt.X) 11.8 12.2 11.3 Blanket Material Metal-U Metal-U Metal-U Average Linear Heat Rating[BOC]

Core (W/cm) 161 165 167 Blanket (W/cm) 7.5 8.1 7.8

Axial Power Form FactorCBOC] 1.24 1.24 1.27 Conversion RatioCEOC] 1.00 0.95 0.90 Fissile Surviving Rate [EOC] 0.92 0.91 0.89 Total Pu Surviving Rate [EOC] 0.98 0.96 0.93 Discharge Burn-up(GWd/t) [3 cycle] 50 45 42 Burnup/Net Consumption of 5622 4401 3585 Fissi !«.• lnventory(GWd/t-f iss.) Fuel Cycle Duration(FulI Power Day) 517 381 344 Moderator Void Reactivity +9.5 -3.6 -31.3 Coefficient(pcm/Xvoid)[EOC]

- 7 3 -

JAERI-M 88・221

Table 4.1.1 Burnup Characteristics of Pancake Type HCLWR Evaluated by 1-0 Calculation

Core Thermal Output(門Wt) 1042 1714 1714

Equivalent Core Oiameter(cm) 400 469 450

Fissi le Core Height(cm) 50 50 60

Blanket Thickness(cm) 30 30 30

Vm/Vt(in Assembly) 0.5 0.7 0.9

Fuel Pin Oiameter(cm) 0.95 0.83 0.83

Loading Fissi le PU Enrichment(wt.%) 11.8 12.2 11.3

B lanket門aterial Metal-U Meta l-U Meta 1 -U Average Linear Heat Rating[BOC] Core (¥J/cm) 161 165 167

Blanket (¥J/cm) 7.5 8.1 7.8

Axial Power Form Factor[BOC] 1.24 1. 24 1.27

Conversion Ratio[EOC] 1.00 0.95 0.90

Fissi le Surviving Rate [EOC] 0.92 0.91 0.89

Total PU Surviving Rate [EOC] 0.98 0.96 0.93

Discharge Burr.-up(GWd/t) [3 cycle] 50 45 42

Burnup/Net Consumption of 5622 4401 3585

Fissi I~ Inventory(GWd/t-fiss.) Fuel Cycle Ourati山 lipowe「Day〉| 517 381 344

Moderator Void Reactivity +9.5 -3.6 -31. 3

Coefficient(pcm/%void)[EOC]

J A E R I - M 8 8 - 2 2 1

4.2 Note on Burnup Calculation of Pu Isotopes in Blanket Used in HCLWR

K. Okumura and Y. Ishiguro

In high conversion light water reac tor(HCLWR) with hard neutron spectrum, the conversion ratio is expected to increase by use of radial or axial blanket. We have studied burnup calculation method for blanket region how to take into consideration of the blanket effects at the evaluation of burnup performance of HCLWR.

In our design code", core burnup calculations are performed based on "table-look-up method of macroscopic cross sections "(refer to the section 4.3). In this method, the accuracy of core burnup calculations depends on how suitable burnup scheme is employed at the step of cell burnup calculations. In the case of blanket burnup, the power prediction becomes particularly important. The cell burnup calculations are usually performed under the assumption of constant power level throughout burnup duration, while i'. really increases as fissile Pu accumulates. Hence, we have investigated the dependence of production of each Pu isotopes on power mode.

Two kinds of power models were employed for this study. One was conventional constant power model throughout burnup duration. Based on our experience of H C L W R analyses, we employed three different linear heat rating for blanket pin. i.e., 1/10 (17W/cm), 1/20 (8.5W/cm) and 1/50 (3.4W/cm) of typical average linear heat rating (170W/cm) of M O X fuel pin. The other power model is constant flux one, where a constant flux level corresponding to initial power level is assumed throughout burnup duration. That is to say, power level increases in proportion to the total fission cross section of blanket. This function was newly incorporated into our burnup code. We employed 3.4W/cm or 8.5W/cm as initial linear heat rating for this model.

We performed cell burnup calculations up to 10GWd/t for pin cells of depleted UOj blanket using these different power modes, where the volume ratio of moderator to pe I le t (Vm/Vp) ranges from 0.7 to 1.1. Figure 4.2.1 shows the power mode and burnup dependence of production of each Pu isotopes in the case that Vm/Vp is 0.7. Productions of Pu-239 and Pu-240 depend on neither power models nor levels at all. while those of Pu-241 and Pu-242 considerably depend on them. These power dependence becomes larger in the case of small Vm/Vp, where

- 7 4 -

JAERI-M 88・221

4.2 Note on Burnup Calculation of PU Isotopes 1n Blanket

Used in HCLWR

K. Okumura and Y. Ishiguro

10 high cooversion light water reactor{HCL¥VR) with hard neutroo

spectrum. the conversioo ratio is expected to increase by use of radial

or axial blanket. We hav由 studiedburnup calculation method for blanket

region how to take into considerat ion of the blaoke t 骨ff喧cts at the

evaluat ion of burnup performance of HCLWR.

1n our design code 1l• core buroup calculations are performed based

00 "table-look-up method of macroscopic cross sections "(reCer to the

section 4.3). 10 this method. the accuracy of core burnup calculatioos

depends 00 how suitable burnup scheme is employed at the step oC cel1

buroup calculations. ln th申 caseoC blanket burnup. the power predic-

tion becomes particularly important. The cell burnup calculations

ara u邑ually performed under the assumption oC constant power level

throughout burnup durat ion. whi le i ~ 1"eally increases as C iss i le PU

accumulates. Hence. we have in"'estigated the depeodence of production

oC each PU isotopes on power mode.

Two kinds oC power models were employed for this study. One

was conventional constant power model throughout burnup duration.

8ased on our experience oC HCLWR analyses. we employed three diCCereot

linear heat rating for b!anket pin. i.e.. 1110 (17W/cm). 1120 (8.5W/cm)

and 1/50 (3. 4W/cm) of typical average 1 inear heat rat ing (170W/cm) of

MOX fue 1 pin. The other power mode 1 is constant C lux one. where a

constant flux level corresponding to initial power level is assumed

throughout burnup dur.ation. That is to say. power level increases in

proportion to the total Cission cross section oC blanket. This function

was newly incorporated into our burnup code. We employed 3. 4W/cm or

8.5W/cm as initial linear heat rating for this model.

We performed cell burnup calculations up to 10GWd/t for pin cells

of depleted U02 blanket using th自5e d i f r e r en t pロwer modes. where the

volwne ratio oC moderator to pellet(γmlVp) ranges Crom 0.7 to 1.1.

Figure 4.2.1 shows the power mode and burnup dependence of production

of each PU isatopes in the case that VmlVp is 0.7. Productions oC

Pu-239 and Pu-240 depend on neither power models nor levels at &11.

W b.i le those of Pu-241 and Pu-242 cons idnably depend on them. These

power dependence becomes l&rger in the case oC small VmlVp. where

-74ー

J A E R I - M 88-221

the m a x i m u m descrepancy was about 30tf on the Pu-241 production and about 25X on Pu-242, respectively. These are caused by competition of 0-decay of Pu-241(14 years of half life time) and neutron capture reaction of Pu-240 at lower flux level. Inspite of large descrepancy on the production of Pu-241 and Pu-242, we can usually evaluate overall conversion ratio accurately, because the production of Pu-239 and Pu-240 being independent on power levels occupy about 90X of total Pu and because the fractional production rate in blanket region is relatively small compared to core region. It is however necessary to choose a suitable power model and level for blanket burnup when its effects •are large like the case of a flat core with thick blanket.

Reference 1) T s u c h i h a s h i K . et a l .

( 1 9 8 6 ) . M

— is e u £ is

i i o.a i

!« r i o.«

"Rev i sed SRAC Code S y s t e m JAERI 1302

/ 7 7^

/ -c - fowtf fin l .5v/c*

/ r— fawcr flu 3.4w/c» •— Fig* f i * t.Sw/c»

' ' 1 1 1 2.0 4.0 SO 8.0 10.0 12.0 Burn-up (GW<J/t)

' 0.0 2.0 4.0 (.0 S.O 10.0 12.0 8urn-up (GWd/f)

E ? 2-0

_.. d A \ s U s

/ - 0 - f«««r f)# iT.Wc*

/ 9 -~-

t**mr f\n l.fcf'C* r i » fit t .S /e* ria* fit l.t*/ca

1 1 1 ' 0.0 2.0 4.0 ( .0 8.0 10.0 12.0

Burn-up (GWd/t)

—<^- f***r f i t l7.w/ea m ?•«*.*• f\t I.Sw/et

" i r - * t*«r fit 3 . W C I -o— f;« ri* t.Sv/c* i

—<^- f***r f i t l7.w/ea m ?•«*.*• f\t I.Sw/et

" i r - * t*«r fit 3 . W C I -o— f;« ri* t.Sv/c*

0.0 2.0 4.0 «.0 ( .0 10.0 12.0 Burn-up (GWd/t)

Fig.4.2.1 Dependence of Pu production in blanket of HCLWR on burnup in power models

- 7 5 -

JAERI-M 88・221

and production Pu-241 the on 30" about was descr唱pancymaXlmum the

These are caused by competition of respectively. about 25X 00 Pu-242.

capt¥lre neutron and t ime) 1 i f e hal f oC years Pu-241(14 oC 6-decay

!arge descrepancy

on the production of Pu-241 and Pu-242. we can usually evaluat申 overall

conversion ratio accurately， because the production oC Pu-239 and Pu-240

Inspi te of level. f lux lower Pu-240 at reaction of

and PU total of 90" about occupy !日velspow自r00 independent being

is relatively reglon in blll.nket production rate Cractione.l th申because

eUects

choose to

necessary

when power

• ar骨 large like the case oC a Clat core with thick blanket.

burnup

however

blanket

reglon.

level and

to ∞mpared

sui table

Reference

~ JAERI 1302

。。

Eu-C』ea¥TO--h2ωeω。。守刷'コ仏

-‘ 1

同Hリ

，，

OJ0 ・wG 。，E

叩• s

a回。z 。AM

。。

~Revised SRAC Code System

i レ/

ピーや- 1相官， Fi.悶 ."/C. 1: 国噌戸，...er Fi. I.$.'c .. ー喧..- Po....er '1. 3.4"/c.-ー〈炉・ fI.膚 'i.r I.$，，/C. ー喧r- rt¥lS Fi I 1.唱団1..

e t a 1. 1) Tsuchihashi K.

色。

【 UI

5 ιl5 、.。ぞM

' 9 ~2 -とωm

~ O.・0

~~.・‘N

b。‘。」。z。.0，

(1986).

内回・

-Eu-e』ea

、・・0--h=帥C山

WON-wN'2t

邑。‘.0 50 8.0 咽 .0 12.。6urn-up (GWd/t)

民。

』。a、、・・0--h-一叫Emwo-ぜN'2t

。.-。s

Oependence of PU production :n blanket of HCLWR on bu~nup in power models

fig.4.2.1

J A E R I - M 8 8 - 2 2 1

4.3 Burnup Calculation Method for Spectral Shift Reactor Using Fertile Rod

K. Okumura , Y. Ishiguro and K. Kaneko

A new concept of H C L W R 1 1 was proposed by Framatome, which is a mechanical spectral shift reactor using fertile rods, In this report, we propose a new core burnup calculation method applicable to it.

This type of H C L W R s has many merits compared to normal H C L W R s without spectral shift"'2'. Few codes have however, been developed for core burnup calculations for it. This is because of the complication such that fuel lattice geometry and cell average fuel enrichment change by withdrawing fertile rods from fuel assemblies. The fertile rod withdrawal results in sudden changes of neutron spectrum and macroscopic cross sections in arbitrary time and position during burnup duration. Therefore, burnup analyses for this type of H C L W R s have been made only with cell burnup calculation based on one point reactor approximation. For further detailed analyses, core burnup calculation is indispensable.

In our design code system SRAC 3 1, core burnup calculations are made on following steps: First, a cell burnup calculation is performed to each fuel type lattice in a core and macroscopic cross sections are generated at each burnup step and tabulated to the burnup degree. In the next step, a core burnup calculation based on diffusion theory is performed using the macroscopic cross sections interpolated to burnup degree distributing in the core. We call this conventional core burnup calculation method "table-look-up method of macroscopic cross sect ions".

W e revised it to be applicable to spectral shift reactors. In the new method, two types of cell burnup calculations are performed independently in the conditions that fertile rods are withdrawn or inserted in a fuel lattice using cell models like shown in Fig.4.3.1. In this way. at first, we prepare two kinds of macroscopic cross section tables, that is, "table for inserted condition of fertile rods" and "table for withdrawn one of fertile rods". In the former table, burnup degree at each burnup step is necessary to be distinctively recorded by material(enriched M O X and fertile), while in the latter table, burnup degree of only M O X is recorded. In the core calculation, the cross sections of the burnup nodes where the fertile rods are inserted are

- 7 6 -

JAERI-M 88・221

4.3 Burnup Calculation Method for Spectral Shift Reactor

Using Fertile Rod

K. Okumura • Y. Ishiguro and K. Kaneko

A new concept of HCLWRυwas proposed by Framatome. which is

a mechanical spectral shift reactor using fertile rods. [n this report.

we propose a new core burnup calculation method applicable to it.

This type of HCLWRs has many merits compared to normal HCLWRs

without spectral shift1).2). Few codes have however. been developed

for core burnup calculat ions fo[' i t. This is because oC tha compl ica-

tion such that fuel lattice geometry and cell average fuel enriclunent

change by withdrawing fertile rods from fuel assemblies. The fertile

rod wi thdrawal resul ts in sudden changes of neutron spectrum and

macroscopic cross sections in arbitrary time and position during burnup

duration. ThereCore. burnup analyses for this type oC HCLWRs have been

made only wi th c骨 11 burnup calculation based on one point reactor

approximation. For further detai led analyses. core burnup calculation

is indispensable.

(n our design code system SRAC3). core burnup calculations are

made on following steps: First. a cell burnup calculation is performad

to each fuel type lattice in a core and macroscopic cross sections are

generated at each burnup step and tabulated to the burnup degree. In

the next step. a core burnup calculation based on diCfusion theory is

performed us iロg the macroscopic cross sections interpolated to burnup

degree distributing in the core. We cal1 this conventional core burnup

calculation method "table-look-up methロd oC macroscopic cross

sections".

We revised it to be applicable to spectral shiCt reactors. In

the new method. two types of cell burnup calculations are performed

independent ly in the condi t ions that Cert i 1 e rods are wi thdrawn or

inserted in a Cuel lattice using cell models like shown in Fig.4.3.1.

In this way. at first. we prepare two kinds of macroscopic cross section

tables. that is. "table for inserted condition of fertile rods" and

"table for withdrawn one of fertile rods ¥I n the former table. burnup

degree at each burnup step is necessary to be distinctively recorded by

matGrial(enriched MOX and fertile). while in the latter table. burnup

degree of ooly MOX is recorded. In the core calculation. the cross

sections of the burnup nodes where the fertile rods are inserted are

J A E R I - M 8 8 - 2 2 1

interpolated to the total burnup degree of M O X and fertile regions with the former table. When fertile rods are withdrawn, the table is exchanged to the latter one, and the cross sections are interpolated to the burnup degree of M O X region.

Strictly speaking, in this treatment, the higher conversion in M O X pins due to hardened neutron spectrum by fertile rod insertion can not be reflected to the macroscopic cross sections of burnup nodes after fertile rods are withdrawn. In order to confirm the accuracy of this method, we compared the results of km and conversion ratio for the infinite lattice shown in Fig.4.3.1 with the rigorous results by a restart calculation. In the restart calculation, the cell burnup calculation starts, under the condition that fertile rods are withdrawn, using the fuel composition calculated at the end of burnup in the condition that fertile rods are inserted. It should be noted that the restart calculation is usually impossible in the actual core calculation because of too many burnup nodes with different burnup degree. As shown in Fig.4.3.2 and Fig.4.3.3, discrepancy of hm from the rigorous one is less than 0.5X. It corresponds to discrepancy of about 1.5GWd/t of discharge burnup degree. On average conversion ratio, the discrepancy is about 0.6%. These discrepancies are acceptable in the present feasibility studies.

'.Ye revised burnup modules in the SRAC system on the basis of this method. Now, we can perform core bunup calculations for spectral shift reactors using fertile rods.

References

1) Hittner D. et al. : "Preliminary Results of the Feasibility Study on the Convertible Spectral Shift Reactor Concept ,"Nuc1.Technol . , 80,181(1988).

2) SAJl , E. et al. "Feasibility studies on high conversion pressurized water reactors with semitight core configurations," Nucl.Technol.,80.18(1988) .

3) Tsuchihashi K., et al. : "Reviced SRAC Code System ." JAERI 1302 (1986).

- 7 7 -

JAERI-M 88・221

interpo!ated to the tota! burnup degree of ~OX and fert i!e regions wi th

the forrner table. When ferti!e rods are withdrawn. the 且h1 e i 5

exchanged to the latter one. and the cross sections are interpo!ated to

the burnup d白gr自由口 fMOX region.

Strictly speaking， in this treatment， the higher conversion in

MOX pins du自 to hard申ned neutron spectrurn by fertile rod insertioc

can no t b司 ref!ected to the macroscopic cross sectioロsof burnup nロdes

after fertile rods ar自 withdrawn. In order to conf i rm th自 accuracy

of this rnethod， w日 compared the r申sults of h. and .;r;n'lersion r且ti 0

for the infinite lattice shown in Fig.4目 3.1 with the rigorous resu!ts

by a restart calculation. In the restart calculation. the cell burnup

calculation starts， under the condition that fertil自 rodsare wi thdrawn，

using the fuel cornposition calculated at the end of burnup in the con-

dition that fertile rods are inserted. It should be noted that the

restart calculation is usually impossible in the actual core calculation

bec且useof too many burnup nodes with different burm!p degree. As shown

in Fig.4.3.2 and Fig ・4.3.3，discrepancy of h. from the rigorous on自 is

less than 0.5%. It corresponds to discrepancy of about 1.5GWd/t of

discharge burnup d自gree. On average conversion ratio， the discrepancy

is about 0.6%. These discrepancies are acceptable in the present

feasibility studies.

'，'/9 revised burnup modules in the SRAC system on the basis of this

method. Now. we can perform core bunup calculations for spectral shift

reactors using fertiLe rods.

References

1) Hittner D. et al. "Prelirninary Results of the Feasibility Study

on the Convertible SpectraL Shift Reactor Concept ."NucI.Technol..

80.181(1988).

2) SAJI. E. et al. "Feasibility studies on high conversion pres-

surized water reactors with semitight core conrigurations."

Nuc 1. Technol . ，80. 18( 1988).

3) Tsuchihashi K.. et al. "Reviced SRAC Code System "JAERI 1302

(1986).

J A E R I - M 8 8 - 2 2 1

MOX : F e r t i l e = 7:1 (Vm/Vr= 1.1->1.43 . E=6.4°/O->?.3%)

e e ^x e ^

Fig.1.5.1 Geo«etry of cell burnup cilculation for spectral shift type HClvR ( l e f t " f e r t i l e rod inserted , right : withdrawn)

Restort Calcu.

0.4 6 *

Burnup(GWd/t)

F i t .4 .3 .2 Coiparision of K* for spectral shift type HCIVR between rigorous(restart) calculation and proposed calculation •ethod

10 20 30 40 50

Burnup(GWd/t)

Fig.4.3.3 Conparision of conversion ratio for spectral shi f t type HCLVR between rigoroua(restart) calculation and proposed calculation aethod

- 7 8 -

jAERI・M 88・221

円OX rert i i e = 7: 1 (Vm/Vr=-1.1→1.43 • E=-6.4%→ア.3%)

0.90 。 10

，‘.

何日・

‘，

ee---HW

F AM---

，. l

，、

肉“

Fe "uJnu

• . ，内K

・・h----

・句・・Ht

Mvr，，‘

v，u，

，・&v--

・-t・ai・LnH-o

• . ‘e

叫• .

刊• •. Fr

B(MOX)=2BGWd/t .•..••.• B(Fert札e)=7GWd/t

20 30 40

8urnup(GWd/t)

Fi，.4.3.2 Co.parisi。向。fk.，. for spectral sh.ft type HCL~R between ri&orous(restart) calculation and proposed calculation .ethod

0.85ト......・....由。d¥¥¥ .......•.•

千e('も@い..-

。0.80

ei 0.75 帥

‘ー‘' 〉ei 0.70 u

同点、•.

、.，‘ d

FIll--』

‘叫

向。

，d(ACR:fCRdU戸t)=O.6"

10 20 30 40 50

8urnup(GWd/t)

Fi，.4.3.3 Co.parision of conversion ratio for spectral shitt type HCLVR between r i 10 rouG( re宮tart)calculation and proposed calculation・ethod

J A E R 1 - M 8 8 - 2 2 1

4.4 NEACRP HCLWR Cell Burnup Benchmark Calculation

H. Akie, Y. Ishiguro, and H. Takano

A benchmark problem, proposed by Japan, for HCLWR cell burnup calculation was approved at the 29th meeting of the Nuclear Energy Agency Commi t tee on Reactor Physics (NEACRP), September, 1986. n This benchmark problem aims to extract the problems included in the data and methods for HCLWRs, and to accelerate the developing work of the data and methods. Summary of the Benchmark Results

A report, related to the sixteen preliminary solutions given by thirteen organizations, was submitted at the 30th NEACRP meeting, September, 1987.2> Fairly large discrepancies were found in the preliminary results of kas, conversion ratios and void reactivities. These differences were caused mainly by the differences of the reaction rates of U-238 and Pu-239. At the same time, those of Pu-240, Pu-241, Pu-242 and fission products were also important in some cases.

After the meeting, six new results were submitted and several contributions have been revised or removed. As a result, fifteen organizations from eight countries submitted the twenty sets of benchmark results. The differences of the reaction rates of Pu-240, Pu-241 and Pu-242 decreased in comparison with the preliminary results. On the other hand, the reaction rates of U-238, Pu-239 and fission products still show large discrepancies. As a result of these facts, agreement of the burnup dependence of km becomes better than that in the previous result, but the large differences are found especially in conversion ratio. The typical results are shown in Figs. 4.4.1~4.4.3. Except for the result of KfK(1985 lib.), the largest difference in km is about 3% through burnup. In addition, the difference becomes as large as 754 for the voided case. The discrepancy in conversion ratio is up to nearly 10%, and is almost the same as in the preliminary result. Specialists' Meeting on the Benchmark

A specialists' meeting on this benchmark was held to clarify the physics problems of HCLWRs, at the NEA Data Bank, France from 19 to 21 April, 1988. The general conclusions resulting from the meeting are summarized as follows: There are still unacceptable differences, especially in the calculation

]AERI-M 88・221

4.4 NEACRP HCLWR Cell Burnup Benchmark Calculation

H. Akie. Y. lshiguro. and H. Takano

A benchmark problem. proposed by Japll.n. for HCLWR cell burnup

calculation was approved at the 29th meeting of the Nuclear Energy

Agency Committee on Reactor Physics (NEACRP). September. 1986.1) This

benchmark problem aims to extract the prnblems included in the data

and methods for HCLWRs. and to accelerat自 the developing work of the

data and methods.

包巴旦主yof the Benchmark Results

A report. related to the sixteen preliminary solutions given by

thirteen organiz8otions. w80s submitted 80t the 30th NEACRP meeting，

September. 1987.2) Fairly large discrepancies were found in the preli-

minary results of k~s. conversion ratios and void re8octivities. Thes自

differences were c80used m80inly by the differences of the reaction rates

of U-238 and Pu-239. At the s80me t ime. thos日 ofPu-240. Pu-241. Pu-242

and fission products were also important in some cases.

After the meeting. six new results were submitted and sever且 l

contributions h80ve been revised or remo'!ed. As a result. fifteen

organizatiuns Irnm eight countrie" submitted the twenty sets of ben-

chm80rk resul ts. The differences of the reaction rates of Pu-240， Pu-241

and Pu-242 decreased in comparison with the preliminary results. On

the other h8ond. the reaction rates of U-238. Pu-239 and fission products

still show large discrep岨ncies. As a result of these facts， agreement

of the burnup dependence of k. becomes better than that in the previous

result. but the large differences are found especially in conversion

ratio. The typical results are shown in Figs. 4.4.1"'4.4.3. Except

for the result of KfK(1985 lib.). the largest difference in k. is about

3~6 through burnup. [n addition. the difference becomes as large as 7~

for the voided case. The discrepancy in conversion ratio is up to nearly

10". and is almost the same as in the prel iminary resul t.

Special ists' Meeting on the 8enchmark

.'¥ speci8olists' meeting on this benchm且rkwas held to clarify the

physics problems of HCLWRs. at the );EA Data Bank. France from 19 to

21 Ap!"il. 1988. The general conclusions resulting from the meeting

are summarized as follows:

There are stilJ unacceptable differenccs. especially in the c8olculation

J A =-R I -M 88- 22 !

of void coefficients and conversion ratios Th e reasons for the 'lev; ations are related primarily to problems .p.

resonance shielding over the whole energy region. • spectral calculation, including1 adjustment of f.ssion >pncir.i ,ind scattering matrices. insufficient quality of '.he evaluated nuclear 1at.i ised

It was also proposed in the meeting to perform Monte Carlo calculations, in o.-der to obtain reference solutions for the benchmark problems. The fol.owing problems were proposed l The PROTEUS-LWHCR Phase I double cell experiment. 2. The two NEACRP benchmark problems at zero and 50 GWd/t. In the

case of 50 GVVd/t burnup. an isotope compos i t i on wi l ] be specified. The proposed deadline for these calculations is the end of 1988.

References 1) Ishiguro Y. ei al.. . "Proposal of Benchmarks on Data and Methods

to Calculate Reactor Characteristics in High Conversion L"'ht Water Reactors." N'EACRP-A-789 (1986).

2) Akie H. et at., "Preliminary Report of HCLVVR Cell Burnup Benchmark Calculations. " NEACRP-A-849 (1987).

1.14 1.06

0.0 10.0 20.0 30.0 40.0 Burnup (GWd/t)

Fig.4.4.1 Burnup dependence of ka» : Vm/Vf-1.1

J :¥ r;-R 1 -:>'1 8巧-22!

。f ソ円 id (;oufficiGnt.; ョnd COnVt1rヨ100 r，1.・.勺.5 Thρ :-.)泊弓つnミ「勺 :hρ 1ρv白

且t1つn5 ara rel叫 :ari pr:mari iy [0 pr:lh:のrns .::

re50nance 5hl~lil~耳つ VAr :he whola 司円丹 r~y rθ耳:on，

<;pecrral calcul，u;っn， includ:ng' dri;ils:ml"nt ')1 :'， s~: つつ ;0ρc t r."1 ，lnd

5C且:t e r 1 n.;ロiatrlcOS，

lnsufiicia:tt qu且 11ty οf t he ，) V 且 lu 日 ~t)d nuc:。且「寸吐・ '1 J':'の司

1 t waョ a I so pr口po沿ed i n t he :nのの tI n可 :n pρr f勺 r:n :vI'ont.1 Cllrlo

calculations， ln 0プder tn obt且:n r P. r.1 r司n口の弓巾:¥Jt;司ns :or the benchmark

prohla:ns， Thp. fnl，owing prohlems wert~ prllplls命d

The PROTECS--LWHCR Pha5e 1 doub 1 ~ ce:: 吟lCper:mi!n t

2， The two ~E.、'.C' :tP tlenchm畠 rk prob I em3 a t zero and 50 GWd/ t. In the

case of 30 GWd!t burnup，且o ヨo:ope :ompositionwill be speciried.

The ?roposed deadl ine fの了 these calculations is the end of 1988.

References

1) [shiguro Y. et 且i "Proposal of Benchmarks on Data and :Vlethod昌

to Caicu!ate Re且ctorCharacteristics in High Conversiun L'，'ht W且 tor

R"actors，時ト:EACRP-A-789 (1986)，

2) Ak i e H. e t a 1 ，， 'Preliminary Report 01' HCLWR C白 11 Burnup

Benchmark Caiculations，・:-iEACRP-A-8 H (1987)

8 」ピ1.06

1. 00 0.0

-….z・・・・阿API-CRC....y....， NAIG -・輯・.... PNC

一--PS! CBOXERl _. -ー--PS I WANDEI

、、. 、. 、. 、、. 、

1. 06 一-e--ANSTO 一ー骨ー-HITACHI (B4) ぺ….， H /TACHI 叫】~ 1.04 -~ーー IKE...-$.... JAERI (SRAC】

...:11…ー JAERI(VI附 j1.02 -一+ー.-Kf'K (1985 L lb. )

:-.-JC-._ I(f'K (newest)

10.0 20.0 30.0 40.0 0.92

Burnup CGWd/七)

Fig.4.4.1 Burnup depend~nce of k曲 Vm/Vf・1.1

一回一

J A E R I - M 8 8 - 2 2 1

r ANSTO

-H HITACHI (B4) •)<••••• HITACHI (J2) •o-- IKE

O 0 . 8 4 I " " " 3 JAERKSRAC) ••-••X JAERKVIM) — * K f K ( l 9 8 5 l i b . ) — K — KfK(newest )

•z MAPI-CRC C .o 0.80

•••V—- NAIG ••*—• PNC —» PS!(BOXER)

- - * - - PSHDANDE) « STUDSVIK 1 TUBS(DATU8S4)

- - X - - TU3S(DATU8S5) -• VA.TECH(VIM)

W1NFRITH

0.0 10.0 20.0 30.0 40.0 Burnup (GWd/t)

Fig.4.4.2 Burnup dependence of conversion ratio : Vm/Vf-1.1

8 XL 1 .08

1 .04 0.0 20.0 40.0 60.0 80.0

Void fraction (%) 100.0

Fig. 4.4.3 Dependence of km on void fraction : Vm/Vf =»!.!, OGWd/t

jAE R 1・恥I88・221

ー--ANSTO

0.86 L - HI~ACHI (Bω .....)(..... HITACHI【J2】

ーー←ー IKE

吋 …・明・・… JAER! CVIM】

言ー・←肝K<l985llb・}~ 0.82

c 20.80

ω 」

ω0.78 〉

乙。に30.76

0.0 10.0

ーー唖一一 STUDSVIK 一一吾一一 TUBS <DA TUBS4l --;r-- TU3S <DA TUBS5) -ー-VA. TECHCVIMl ・・・・・ WINFRITH

20.0 30.0

8urnup CGWd/t)

40.0 50.0

Fig.4.4.2 Burnup dependence of cor可vers10n rat10 Vm/Vf・1.1

‘‘ 、

‘‘ .‘ 、、

lヨ 10

8 .::t:.

、、. 、. 、-、. 、、. 、、、

、、、、‘、、、、~一、.

、._.一ー一一一'_.ー一一 ι ム〉〆一一一一一-....

1. 04 0.0 20.0 40.0 60.0 80.0

Void frac七ion (%)

Fig.4.4.3 Dependence of k= on void fraction Vm/Vf2 1.1. OGWd/t

J A E R I - M 8 8 - 2 2 1

4.5 Treatment of the 1.06eV Resonance of Pu-240 in Tight Lattice Cell

H. Akie, H. Takano and J. Saito*

In the neutronics calculations of HCLWRs with tight lattices, resonance absorption calculation is one of the important problems because HCLWRs have an intermediate neutron spectrum between thermal and fast reactors. It is known that the large resonance of Pu-240 at 1.06eV plays an important role in reactivity change such as coolant void reactivity". In this note, the effect of the treatment of the resonances on nuclear characteristics is examined by performing cell calculations by the SRAC system 2'. The SRACLIB-JENDL2 group cross section library based on the JENDL-2 nuclear data is used in the present calculations. The tight lattice model used in the calculation is the same specification as in the NEACRP benchmark calculations 3 1.

In the treatment of this large 1.06eV resonance of Pu-240, there are two primary problems. One is energy group structure for this resonance. The SRAC library which includes infinitely dilute cross sections and self-shielding factors has a slightly broad energy group structure. The W I M S - D library4' has a fine group structure and has not self-shielding factors. A new SRAC hermal library based on the JENDL-2 data has been produced to study the difference between the SRAC and the W I M S - D methods. This group structure is fine and similar to that of the WIMS-D. This library has not the self-shielding factors for the 1.06eV resonance of Pu-240. The calculations with the original and the new thermal libraries were made. In order to use the new energy structure for the 1.06eV resonance, thermal cut off energy(Ec) was selected to be 3.93eV in these calculations. The energy region below Ec is treated as thermal energy region, and the effects of upscattering and chemical binding are taken into account. The SRAC system can select this energy from between 0.414 and 3.93eV.

The selection of this thermal cut off energy is the second problem. In the SRAC system there is an option to treat resonances with ultra-fine energy mesh by using the PEACO routine, but in order to use the PEACO option the 1.06eV resonance is to be included in the resonance energy range (Ec should be less than leV). The 1.06eV resonance is on the boundary between thermal and resonance energy region

Japan Information Service, Ltd.

- 8 2 -

]AERI-M 88・221

4.5 Treatment of the 1.06eV Resonance of Pu-240 in Tight Lattice Cell

H. Akie. H. Takano and J. Saito.

In the neutronics calculations o( HCLWRs with tight lattices.

r自sonanc8 absorption calculation is one of the important problems

because HCLWRs have an intermediate neutron spectrum between thermal

and fas t re且ctors. [t is known that the la.rge resonance of Pu-240 at

1.06eV pla.ys an important role in reactivity change such as coolant void

reactivityll. In this note. the effect of the treatment of the reso-

nances on nuclear characteristics is examined by performing cell caト

culation5 oy the SRAC sysrem31. The SRACLIB-JE;';DL2 group cross sec-

tion library based on the J&~DL-2 nuclear dat且 is used in the present

calculations. The tight lattice model used in the calculation is the

sam自 specificationas in the NEACRP benchrnark calculations31.

In the treatment of this large 1.06eV resonance of Pu-240. there

且re two primary problems. One is energy group structur自 for this

r申sonllnce. The SRAC library which includes infinit自 Iy dilute cross

sections and selC-shielding factors has a slightly broad enet'gy group

structure. The WIMS-D library~1 has a fine group structure and has

not 5el f-5hielding factors. A new SRAC hermal 1 ibrary based on the

JR.....DL-2 data has been produced to study the di fference between the SRAC

and the WIMS-D methods. This group structure is fine and similar to

that of the WIMS-D. This 1 ibrary has not the se 1 f-shielding factors

for the 1.06eV resonance of Pu-240. The calculations with the original

and the new thermal libraries were made. In order to use the new

anergy structure for the 1.06eV resonance. thermal cut off energy(Ec)

was selected to be 3.93eV in these calculations. The energy region

b骨low Ec is treated as thermal energy region. and the effects of

upscat ter ing and chemi cal binding are taken into account. The SRAC

system can select this energy from between 0.414 and 3.93eV.

The selection of this thermal cut off energy is the second problem.

ln the SRAC system there is an opt ion to treat resonances wi th

ultra-fi:J.9 energy mesh by using the PEACO routine. but in order to

use the PEACO option the 1.06eV resonance is to be included in the

resonance energy range (Ec should be less than leV). The 1.06eV reso-

nance is on the boundary between thermal and resonance energy region

禽 Japanlnformation Service. Ltd.

J A E R I - M 8 8 - 2 2 1

and there may be the effect of upscattering and chemical binding around leV. But. if the PEACO is used for the 1.06eV resonance, upscattering and chemical binding effects are to be neglected around leV. The calculation with Ec=0.683eV was also made by using the PEACO routine for the 1.06eV resonance.

The results calculated with the two different group structure libraries are compared in Table 4.5.1 for the case of the moderator/fuel volume rat io(Vm/Vf ) = 1. 1 . In this table, Case 1 and Case 2 show the results obtained with the original and new energy group structure. The difference between Cases 1 and 2 is very smal 1 . It is also observed that the results for Cases 1 and 2 are in good agreement with those calculated by the continuous energy Monte Carlo code V I M J | . The cross section library of VIM has been produced81 on the basis of the JENDL-2 nuclear data. In this library, the energy pointwise cross sections of the 1.06eV resonance of Pu-240 were generated. The?e results show that the absorption rate of Pu-240 1.06eV resonance can be obtained accurately with the broad energy group 'tructure and the table-look-up method for self-shielding factor.

The calculated results for the case of Ec=0.683eV (Case 3) is also shown in Table 4.5.1. The km obtained with the Case 3 differs considerably from those for Cases ! and 2, and also from the result of VIM.

This difference can be attributed to the following two differences between Cases 1 and 3: - The difference of the effective cross sections calculated by the

table-look-up me'hod and by the PEACO (especially for Pu-240), - The difference of scattering matrices. (especially for H in H2O).

In other words, the effects of upscattering and chemical binding. These effects were further investigated and the effect of the scattering m a l i c e s were found to be large.

Table 4.5.2 shows the contribution of each nuclide to the difference of km of Case 3 from Case 1. The contribution of Pu-239 and Pu-241 is as large as that of Pu-240. This contribution comes from the resonances of Pu-23.9 and Pu-241 around 0.3eV, through the spectrum change caused mainly by the effect of scattering matrices.

It is concluded from these results that the 1.06eV resonance of Pu-240 should be treated in the thermal energy region than in the resonance energy region. Furthermore, this resonance absorption can be calculated accurately with the group structure of lethargy width Ju=0.125 by using the shielding factor such as in the SRAC system.

- 8 3 -

JAERI-M 88・221

and there may be the efiect of upscattering and chemical binding arv~nd

leV. But. if the PEACO is used for the 1.06eV resonanc自. upscattering

and chemical binning effects ard to be neglected around 1日V. The cal-

culation with Ec=O.683eV was also made by using the PEACO routine

for the !.06eV resonance.

The resul ts c且lculated wi th the two di ff自r自nt group structure

1 i brar i es are compar日din Table 4.5.1 for tho cas日 of tho modora tort fue 1

volume ratio(Vm/Vf)=1.1. 1(1 this table. Case 1 and Case 2 show the

results obtained with the original aod new energy group structure.

The difference between Cases 1 and 2 is very small. It is also observ目d

that the resul ts for Cases 1 and 2 are in good agreement wi th :hose

calcu!ated by the continuous energy Monte Carlo code VIMれ The cross

section library of VIM has been produced61 on the basis of the JENDL-2

nuclear data. In this library， the energy pointwise cross sections of

the 1.06eV resonance of Pu-2~O were generated. The~e results show that

the absorption rate of Pu-240 1.06eV resonペnce can be obtained

accurately with the broad energy group .tructure and the table-look-up

method for self-shielding factor.

The calculated results fur the case of Ec=O.683eV (Case 3) is also

shown in Table 4.5.1. The k. obtained with the Case 3 differs consid-

erably from those for Cases and 2. and also from the result of VIM.

This difference can be attribut自d to the following two differences

between Cases 1 and 3:

-The difference of the effective cross sections calculated by the

table-look-up method and by the PEACO (especially for Pu-240)，

-The difference of scattering matrices. (especial1y for H in H20).

In other words， the effects of upscattering and chemical binding.

These effects were further investigated and the effect of the scattering

matrices were found to be large.

Table 4.5.2 shows the contribution of each nuclide to the diffe-

rence of k. of Case 3 from Case 1. The contribution of Pu-239 and Pu-.241

is as 且rge as that of Pu-240. This contribution comes from the

resonances of Pu-23~ and Pu-241 around O.3eV. through the spectrum

change caused mainly by the effect of scattering matrices.

It is concluded from these resul ts that the 1.06eV resonance of

P:.t-240 sho1l1d be treated in the thermal energy region ~han in the

resonance energy region. Furthermore. this resonance absorption can

be calculated ac.curately with the group structure of lethargy width

.du=O.125 by using the shielding factor such as in the SRAC system.

JAER1-M 88-221

R e f e r e n c e s

1) Penndorf K. e t a l . : "Some Neut ron P h y s i c a l Consequences of

M a x i m i z i n g the Convers ion R a t i o of P r e s s u r i z e d Water R e a c t o r s

Opera ted in the U r a n i u m - P l u t o n i u m C y c l e , " Nuc lea r Technology, 59,

256 ( 1 9 8 2 ) .

2) T s u c h i h a s h i K. et a l . : "Revised SRAC Code S y s t e m , " JAERI 1302

( 1 9 8 6 ) .

3) I s h i g u r o Y. e t a l . : "Proposa l of Benchmarks on Data and Methods

to C a l c u l a t e R e a c t o r C h a r a c t e r i s t i c s in High Convers ion L igh t Water

R e a c t o r s , " NEACRP-A-789 (1986 ) .

4) Taubmann C . J . "The WIMS 69-Group L i b r a r y Type 166259."

AEEW-M1324 ( 1 3 7 5 ) .

5) P r a e l R . E . et a l . "A U s e r ' s Manual for the Monte C a r l o Code

V I M . " FRA-TM-84, Argonne N a t i o n a l Labo ra to ry (1976) .

6) Mori T. e t a l . : p r i v a t e c o m m u n i c a t i o n (1987) .

T a b l e 4 . 5 . 1 C a l c u l a t e d r e a c t i o n r a t e s and c ross Tab le 4 . 5 . 2 C o n t r i b u t i o n o f s e c t i o n s f o r the c e l l w i t h V m / V / = 1 .1 each n u c l i d e t o the ( A . a b s o r p t i o n r a t e , o* i n ba rns and) d i f f e r e n c e o f A . be tween

Au deno tes l e t h a r g y w i d t h a round 1eV) Cases 1 and 3

Case 1 Case 2 Case 3 VIM

Au 0 .125 0.02 0.00125 (PEACO)

Ec(eV) 3.93 3.93 0 .683 4 .5

°Pu A 0.1518 0.1507 0.1466 0 .1555±0 .0010 aa 15.11 14.97 14.60 15.54 ± 0 . 1 0

A. 1.1269 1.1265 1.1391 1.1287±0.0024

N u c l i d e AP/P-AA/A

2 3 5 u 1.09E-4 i 3 * u 6.4 E-4 " 9 P u 3.60E-3 "°Pu 5.20E-3 m P u 1.342-3 2 4 3 P u 6.08E-5

Zr 4 .6 E-6

Ak/k 1.08E-2

P : production rate A : absorption rate

- 8 4 -

]AERI-M 88・221

References

1) Penlldorf K. et a1. "80me Neutron Physical Consequenc日s of

Maximizing th由 Conversion Ratiロ or Pressurized Wat骨r Reac tors

Operated in the Uranium-Plutonium Cycle，" Nuclear Technology， 59，

256 (1982).

2) Tsuchihashi K. et al. "Revised 8RAC Code 8ystem，" JAERI 1302

(1986).

3) I sh i guro Y. e t a 1. "Proposa 1 0 f 8enchmarks on D且ta and Me thods

to Calculate Reactor Charact自ristics in High Conversion Light Wat自r

R~actors ，" ~CRP-A-789 (1986).

4) Taubmann C. J . "The WIMS 69ベ}roup Library Type 166259，"

AEEW-M1324 (1:175).

5) Prael R.E. et al. "A User's Manual for the Montc Carlo Code

VIM，" FRA-TI，1-84， Argonne National Laboratory (1976).

6) Mori T. et al.: private communication (1987).

Table 4.5.1 Calculated reaction rates and cross sections for the cel I with Vm/Vt=1.1

Table 4.5.2 Contribution of each nucl ide to th日

difference of k. between Cases 1 and 3

(A:absorption rate， dG in barns and) ~U denotes lethargy width around 1eV)

Case 1 Case 2 Case 3 VIM Nuc 1 idc 4P/P-I1A/A

.du 0.125 0.02 0.00125 235U 1.09E-4 (PEACO) 233U 6.4 E-4

Ec( eV) 3.93 3.93 0.683 4.5 13gpu 3.60E-3 240pu 5.20E-3 241pU 1.3..Z-3

240pU A 0.1518 0.1507 0.1466 0.1555士0.0010 2・2pU 6.08E-5 110 15. 11 14.97 14.60 15.54 :!::0.10 Zr 4.6 E-6

lt. 1.1269 1.1265 1.1391 1.1287:!::0.0024 ~lt/lt 1.08E-2

P production rate A absorption rate

ー制一

J A E R I - M 8 8 - 2 2 1

4.6 Conceptual Design of SPWR (System-Integrated Pressurized Water Reactor)

K.Sako

A design study has been carried out on a new type of integrated pressurized water reactor, SPWR ( System integrated PWR ) ' . This reactor has a poison tank with valve system in the reactor vessel instead of control rod drive (CRD) system. The features of the reactor concept are as follows: (1) Simplified reactor system, (2) Reliable reactor shutdown and decay heat removal systems, (3) Flexible operation, and (4) Easy maintenance.

A 700MWe power plant of twin reactor type has been studied. Major components and systems are designed and evaluated preliminarily. As a result, this concept is expected to be feasible as the next generation power plant. SPWR Series Reactors

Figure 4.6.1 shows typical design options. SPWR is available in two basic types (Hot vessel and Cold vessel types). Furthermore, therer are many options by the location of main circulating pump (MCP). (1) Hot vessel type: an integrated reactor incorporating a primary cooling system in the reactor vessel which houses a poison (borated water) tank. (2) Cold vessel type: a reactor vessel containing borated water and housing a primary cooling system.

These reactors do not have any control rod, and their operation is controlled by oeans of negative temperature coefficient of the reactor core and by adjusting the concentration of boron in the primary water. Changes in reactivity due to fuel burn-up will be compensated by adjusting the boron concentration. In the event of an abnormality such as a drop in the delivery head of MCP, borated water is automatically injected into the reactoz core, shutting down the reactor. There is also a system to forcibly inject borated water to bring the reactor to an emergency halt during operation. Design Concept of SPWR (Reference type)

Among the wide options of SPWR , at present, H/H (Hot vessel / Hot leg) type seems to be the most realistic. A design study has been made for this type.

Table 4.6.1 shows the major design parameters. Figure 4.6.2 shows a concept of plant layout. Twin 1 IOOMWt reactors are installed in tho suppression type reactor building similar to the BWR system.

Figure 4.6.3 shows the concept of the 11OOMWt reference type SPWR reactor, SPWR-H/H-1100. Figures 4.6.4 and 4.6.5 show the cross section of the reactor.

The reactor consists of four major parts, namely shell part of reactor pressure vessel (RPV), top dome of RPV integrating SG, a poison tank which houses core structure and a single unit MCP fitted at the top of RPV.

The once-through helical coil type SG is divided into four unit headers. The poison tank surrounding the core is connected to core outlet via three I^C. .wl ic pressure valves at the upper part of the tank. These valves are operated by the pump delivery pressure. The bottom part of the poison tank is attached to the core inlet plenum through a bundle type hortaycomb structure which absorbs thermal expansion of borated water in the

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]AERI-M 88-221

4.6 Conceptual Design of SPWR (System-Integrated Pressurized Water Reactor)

K.Sako

A design study has been carried out on a new type of integrated

pressurized water reactor， SPWR ( ~yste~ integrated PWR ) '. This reactor ha~

a poison tank with valve syste圃 inthe reactor vessel instead of contr~l rod

drive (CRD) systell・ The features of the reactor concept are as follows:

(1) S i皿plified reactor syste・(2) Reliable reactor shutdown and decay heat

re ・oval syste圃s，(3) Flexible operation， and (4) Easy田aintenance.

A 700門凶e power plant of twin reactor type has been studied. 月ajor

co・ponents and syste ・s are des igned and evaluated pre t i皿inarily. As a

result， this concept is expected to be feasible as the next generation power

plant.

SPWR Series Reactors

Figure 4.6. t shows typical aesign options. SPWR is available in two

basic types (Hot vessel and Cold vessel types). Further.ore， therer are

・ ~ny options by the Iocation of main circulating pu・p (門CP).

(1) Hot vessel type: an integrated reactor incorporat ing a pri圃ary cooling

system in the reactor vessel which houses a poison (borated water) tank.

(2) Cold vessel type: a reactor vessel containing borated waler and ho~sing

a prl田arycooling syste圃.

These reactors do not have any control rod，耳nd their operation is

controlled by lIeans of negative te圃peraturecoefficient of the rcactor cor~

and by adjusting the concentration of boron in the pri圃ary wat~ ，. Changes

in reactivity due to fuel burn-up will be compensated by adjusting the boron

concentration. In the event of an abnor圃alitysuch as a drop in the delivery

head of門CP，borated water is auto・aticallyinjected into the reactol core， shutting down the reactor. There is also a syste圃 to forcibly inject

borated water to bring the ~eactor to an e圃ergencyhalt during operation.

Design Concept of SPWB (Reference type)

Among the可ideoptions of SPWR ， at present， H/H (Hot vessel / 1I0t leg)

type seems to be the most realistic. A design study has been 哩ade for this

Table 4.6.1 shows the 田ajor design para.eters. Figure 4.6.2 shows a

concept of plant layout. Twin 1100MWt reactors are Instal1ed ln thε

suppression type reactor building similar to the BWR syste圃 .

Figure 4.6.3 shows the concept of the 1100門Wt reference typc SPWR

reactor， SPWR-H/H-ll00. Figures 4.6.4 and 4.6.5 show the cross section of

the reactor.

The reactor consists of four 圃ajorparts， na圃ely shell part of reactor

pressure vessel (RPV) ， top do聞eof RPV integrating SG， a poison tank which

houses cor号 structureand a single unit 門Cpfitted at the top of RPV.

The once-through helical coil type SG is divided into four unit headers. The poison tank surrounding the core is connected to core outlet

via three :，y':.，o..lic pressure valves at the upper part of the tank. These

valves are operated by the pu・pdelivery pressure. The botto・part of the

poison tank is attached to the core ialet plenu. through a bundle type

hon~yco・ b structure which absorbs ther圃al expansion of borated water in the

J A E R I - M 88-221

poison tank. One unit of cooler is it> _o ' :-d in the upper part, of the tank to maintain the temperature of the borai^d water in the tank at I50±I0*C during the reactor operation and also to enhance the nixing by natural circulation. The natural boron concentration in the borated water is maintained higher than 4,000 ppa. The poison water is injected into the core at the reduced pump delivery pressure. Reference I) K.Sako: Conceptual Design of SPUR (Systen-integrated Pressurized Water

Reactor), ANS Topical Meeting, Seattle, Wash, May I-5, 1988.

HOT VESSEL T Y P E

Feed water

Upper interlace _ (Hydraulic

pressure valve )

Borated water (150 C)

SPWR-H/C (Sel f-pressurized)

COLD VESSEL TYPE

SPWR-C/H SPWR-C/C (Sel f-pressurized)

Main circulating pump

|_ Steam generator

Borated water .p. (Higher t

temp.) ';

Lower I, interface \ (Honeycomb)

Water level Feed

Upper ^X interface (Hydraulic

pressure valve )

Cover plate \-Water level

Fig.4.6.1 Typical SPWR options

Main steam lines -— celt."

Feed water lines 7777

5PWR-H/H 1100

Fig.4.6.2 Concept of plant layout (Twin reactor system)

•86-

]AERI-M 88・221

poison tank. 0ne unit of cooler is i~ .o '~d in the upper part of the tank

to ・aintain the te阻perature of the boraぃ d water in the tank at 150士 10・cduring the reactor oper司tion and a150 to enhance the 皿ixing by natura 1

circu1ation. The natural boron concentration in the borated water i5

lIaintained higher than 4，000 pp圃・ The poison water is injected into the

core at the reduced pu圃p delivery pressure.

Reference

1) K.Sako: Conreptual Design of SPWR (Syste圃・ integrated Pressurized Water

Reactori， ANS Topical 門eeting，Seattle， Wash，門ay1・・5，1988.

Feed water ~

water (150 C)

HOT VESSEL TYPE COLD VESSEL TYPE

SPWR・H/H SPWR・村IC SPW<<-C/H SPWR-C/C (Refer官官e) (Self;>res卸 rized) (Self;>ressuriZed)

Fig.4.6.1 Typical SPWR options

Fig.4.6.2 Concept of plant layout (Twin reactor system)

-86ー

J A E R I - M 8 8 - 2 2 1

Table 4.6.1 Major design parameters PLANT

Electric power No. of reactors

SPWR-H/H 1100 REACTOR Thermal power Coolant inlet/outlet temp. Coolant Flow rate Core outlet pressure Core pressure drop SG pressure drop CORE Equivalent core dia./height Total Uran inventory U235 enrichment Core average power density No. of fuel assemblies No. of instrumentation channel FUEL ASSEMBLY Lattice pitch (triangle) 250 mm Assembly total length 2.5 n No. of fuel rods 325 Rod dia./pitch(triangle) 9.5/13 Weight of Uran 296 kg Weight of assembly 447 leg Rod average liniar heat rate 14.1 kW/m MAIN CIRCULATING PUMP(one unit)

700 MWe 2

1 ,100 MWt 2B0/ 31 Or 2 4 , 0 3 0 t / h 13 MPa 0 . 0 3 5 MPa 0 .18 MPa

2 . 6 9 / 2.0m 35 .S ton 3 ~ 4 X 84 MWt/m' 1 20

Flow rate Delivery pressure Rotating speed Diameter of impeller NPSh (pump/available) STEAM GENERATOR Steam temperature/pressure Feed water temperature Steam flow rate No. of heat trasfer tubes Tube length, total/effective Tube inner/outer diameter Effective heat transfer area SG inner/outer diameter SG height No. of headers (in, out each) POISON INJECTION SYSTEM Natural Boron content Poison temperature Initial driving force No. of hydr. pressure valves No. of rapid opening valves

MCP seat Downcomer

26,000 t/h 0.2 3 MPa 600 rpm 1 . 5 m 55 / 420 m

285-C /5MPa 210"C 2 , 0 0 0 t o n / h 3 , 0 4 0 80 / 60 m 18 / 2 4 mm 1 3 , 7 5 0 ra' 3 . 2 / 6 . 7 . 6 m 4

>4 ,000 ppm 150± 10"C 0 .031 MPa 3 3

Coolant path ( x & )

Tank support'

Main circulating pump ( x I )

Rapid opening valve (Active shutdown system , x 3 )

- ( t o Poison tank)

^fro« SC inlet plenui)

• " l A ' Steam header

( x <t )

Pump delivery port ( x "» )

Upper Interface, Hydraulic pressure valve (Passive shutdown system, x 3 )

generator ( once-through helical coil rype )

Reactor pressure vessel ( ID = 6.6 m)

Core. Fuel assembly ( x 120 )

Lower Interface, Poison expansion absorberf x 6 )

.Poison passage ( x l )

SPWR- 1100

HT tube group ( x 8

Hydraulic pressure valve ( x 3

Poison tank outer cylinder

Poison tank inner cylinder

Thermal insulator

Core. Fuel assembly ( x 120 ) Core bare! ( Reflector )

Core bypass ( for natural circulation )

Coolant path ( Poison tank support, x 6 )

Fig.4.6.4 Cross section of SPWR (A-D) Fig.4.6.5 Cross section of SPWR (E)

- 8 7 -

JAERI-M 88‘221

"¥4i<l cil'C¥.liati<l& PUrT甲( x 1 )

Rapid QPef'ir唱 v・1ve(ACtive shlJt由 wn昨日間.x ) )

。Poasont・nkl，0・SCin i.t

p I enu・}

一"lA-St.""、header( x与}

Core. Fue1 userr，o1V ( x 120 )

t.ower interfac:e. Poi邸調、 exp・r冒ion副制民r(x 6 )

Poison p&$坦&e( x 1 )

ーーー..Irt W&円 er!iow

-ー同.Pci5rl苫、!iow

・-a---EL-a、田

，。・・Ea

、war・

・3enE

Table 4.6.1

|何回、， downct>町，er

a 。1

~I 内 POi5On町山

700 MWe 2

・，，・

954附

℃℃

04//?/

eu--nu''nugJt3EJ40

『，••

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1 .100 MWt 2BO; 310"C 24.000 t/h 13 MPa 0.035 M!?a 0.18 MPa

>4.000 ppm 150士 10'C0.031 MPa 3 3

El ・ctri.c power No~ ot: reactors

SPWR-H/H 1100

Concept of SPWR-H/H 1100

PoIs司、 tar古。uterc:yllr世erPOi5町、田川〈

irv官 rcyUnder n官円nalinsulator

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c。時国rel( Re(lect町}

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Steam t官ader/(xω}

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国 nk Il百τrunet1国ttOl1p1a時一 d四四割 (x1)

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ト皆、畢，i"8str¥JCture o( t¥JOe

虫峰田rt

Hrt¥式寓

阜市早{x 11

① Hydr1lulic P冊 ssul"evalve ( x .)

Cross section of SPWR (E) Fig.4.6.5 Cross section of SPWR (A-D)

Fig.4.6.4

J A E R I - M 8 8 - 2 2 1

A.7 Design Study of the Valves for Reactor Shutdown System of SPWR

K.Sako

A design study has been made on the valves for the reactor shutdown system of SPWR (System-integrated Pressurized Water Reactor) 1.

Three reactor shutdown systems are applied to SPWR as follows; (1) Scheduled shutdown systea

Scheduled shutdown is made by borated water injection into the primary water via a snail size piping froa outside the RPV. This systea is also used for the reactor power adjusting. (2) Emergency shutdown system (Passive Systea)

Three hydraulic pressure valves are installed in the upper plate of the poison tank as shown in Fig.A.6.3. These valves are noraally closed by supplying the priaary water froa the SG inlet plenua (puap delivery side) via strainer. In case of reducing the MCP delivery pressure, the valves are automatically opened and borated water in the poison tank is injected into the reactor core. Since th's valve is designed to have very reliable structure, no aaintenance under the reactor operation is necessary and no failure is expected. (3) Emergency shutdown system (Active system)

Three rapid opening valves are installed in the RPV top doae. As shown in Fig.A.6.3, the pipe lines of these valves are connected froa the SG inlet plenum (pump delivery side) to the poison tank upper plate. When the MCP is fully running the reactor is shut down in 3 seconds by using two valves after the signal reached to the valves. Hydraulic pressure valve

Figure A.7.1 shows the hydraulic pressure valve as an upper interface. Because this valve is a safety related component, very reliable perforaance is necessary. Engineering design points are how to Bake the higer hydraulic resistance in the annular space between the cylinder and piston, and how to aake the lower hydraulic resistance at the valve port.. Diaaeters of valve port is 200 aa, cylinder is 300 aa and the piston is 290 aa. The annular space gap is designed very big, 10 aa for diaaeter. For this gap, 50 l/s (correspond about 0.5 % of rated flow of MCP) of leakage flow is estimated. High pressure water is introduced froa SG inlet plenua via a strainer. No failure such as piston stick is expected. No disconnecting failure of the flexible joint between the valve plate and the weight is expected because this point is designed to have very high strength. Disconnecting failures of the other part introduce no safety problem. Only the valve is opened and the reactor is shut down.

Figure A.7.2 shows the characteristics of the hydraulic pressure valve at the operating temperature. The valve is designed as opened at the flow rate of higher than A0% of rated flow of MCP. The weight of aoving part is designed 170kg in the hot water. Adding to this weight, the force settled by the relation between the density difference of poison and priaary water and core pressure drop. At the full power operating condition, 1.26 ton of lifting force is generated and shut the valve tightly. No large amount of leakage is expected through the valve seat.

- 8 8 -

]AER 1圃 M 88・221

4.7 Design Study of the Valves for Reactor Shutdown System of SPWR

K.Sako

A design study has been ・ade on the valves for the reactor shutdown

syste・ofSPWR (System-integrated Pressurized Water Reactor) 1.

Three reactor shutdown syste.s are app1ied to SPWR as follows;

(1) Scheduled shutdown syste・Scheduled shutdown is 圃ade by borated water injection into the pri・ary

water via a s・a11 size piping fro. outside the RPV. This syste・is a1so

used for the reactor power adjusting.

(2) E.ergency shutdown syste ・(PassiveSyste圃)

Three hydraulic pressure valves are instal1ed in the upper p1ate of the

poison tank as shown in Fig.4.6.3. These valves are norully closed by

supplying the pri・ary water fro圃 the SG inlet plenu. (pu圃p delivery side)

via strainer. In case of reducing the門CPdelivery pressure， the va1ves are auto・atically opened and 1I0rated water in the poison tank is injected into

the reactor core. Since th;s valve is designed to have very reliable

structure， no .aintenance under the reactor operation is necessary and no

failure is expeιted.

(3) E圃ergencyshutdown syste圃 {Activesyste・}

Three rapid opening valves are installed in the RPV top do・e. As shown

in Fig.4.6.3， the pipe lines of these valves are connected fro. the SG inlet plenu圃 (pu皿pdelivery side) to the poison tank upper plate. When the門CPis

fully running the reactor is shut down in 3 seconds by using two valves

after the signal reached to the valves.

Hydraulic pressure valve

Figure 4.7.1 shows the hydraulic pressure valve as an upper interface.

Because this valve is a safety related co・ponent，very reliable perfor・anceis necessary. Engineering design points are how to ・akethe higer hydraulic

resistance in the annular space bet~een the cylinder and piston， and how to ・ake the lower hydraulic resistance at the valve port. Dia.eters of valve

port is 200 ・圃， cylinder is 300 ・・ and the piston is 290 ・・ The <lnnular

space gap is designed very big， 10 ・・ for dineter. For this gap， 50 p./s

(correspond about 0.5 4 of rated flow of MCP) of leakage flow is esti.ated.

High pressure water is introduced fro. SG inlet plenu. via a strainer. ~o

failure such as piston stick is expected. No disconnecting failure of the

flexible joint between the valv哩 plate and the weight is expected because

this point is designed to have very high strength. Disconnecting lailures

of the other part introduce no safety proble.. Only the valve is opened and

the reactor is shut down.

Figure 4.7.2 shows the characteristics of the hydraulic pressure valve

at the operat ing te・perature. The va 1 ve i s des igned as opened a t the fl ow

rate of higher than 40% of rated flow of門CP. The weight of 圃ovingpart is

designed 170kg in the hot water. Adding to this weight， the force settled

by the re

一回一

J A E R i - M 8 8 - 2 2 1

Figure 4.7.3 shows a trially fablicated 1/2 scale model and the hydraulic test at the Kishikawa Special Valve Co. Rapid opening valve

Figure 4.7.4 shows the rapid opening valve used in the active reactor shutdown system. This valve is under design by coorporation with the KishiKawa S.V.Co. The basic mechanism is similar to the conventional one. However, this valve has the following features: (1) Piping connected to the valve can be managed in RPV. (2) The valve seat can be _ „__„ ^ dismantled during the [E=OT=31 D 2 :* 30<] •» I P S J 3 ^ 1 reactor shutdown. (3) Moving test can be made without affecting the reactor performance during the reactor operation. Reference I) K.Sako: Conceptual design of SPWR (System-integrated Pressurized Water Reactor), ANS Topical Meeting, May 1-5, Seattle, Wash., 1988. ( Closed : Operating condition )

Lifting forth (ton)

Lifting forth due to differential pressure at the piston (0.2MPa at 100 % flow)

Forth settled by the relation between the density difference of poison and primary water and the pressure drop

Down forth by weight of valve movable part

( Opened : Shutdown condition )

Fig.4.7.I Hydraulic pressure valve

Pressurized air

Air cylinder J PlsTon.

20 W 60 MCP flow rate ( % )

Fig.4 .7 .2 Characteristics of the valve York,

Stroke indicator (for t e s t i s )

Valve disk Valve seat Jrt

Lift Umiter testing under reactor operation)

Grand packing

Seal flange

Skirt--V) / / lUijoiL

from SC inlet plenum

to Poison tank

Fig.4.7.3 1/2 scale model and the test GO

Fig.4.7.4 Rapid opening valve

- 8 9 -

jAERi-M 88・221

figure 4.7.3 shows a trially fablicated 1/2 scale Model and the

hydraulic test at the Kishikawa Special Valve Co.

Rapid opening valve

Figure 4.7.4 shows the rapid opening valve used in the active reactor

shutdown syste田 This valve is under design by coorporation with the

Kishi且awa S.V.Co. The basic 田echanisll is s i園ilar to the conventional one.

However， this valve has the following features:

(1) Piping connected to the valve can be 田anaged in RPV.

(2) The valve seat can be

dis.antled during the

reactor shutdown.

(3)門oving test can be

made without affecting the

reactor performance during

the reactor operation.

Referent:e

1) K.Sako: Conceptual

design of SPWR (Syste田ー

integrated P~essurized

Water Reactor)， ANS

Topical 門eetlng，門ay 1-5，

Seattle， Wash.， 1988.

内M内

内uw角uauE

・φφφ

•••••• 123

nunuRU

Liiting fon:1'l (ton)

LiCting Con:h due to dl{ferential pre:巴;vreat the piston (O.2MPa at I∞% !Iow)

( C同制 :0同市町ng∞ndition) ( 0PeneCI : Shutdownロr岨ition)

0.4 ./ Fon:h sett刷 by向印刷民tween

he de問 ItydlCCerenc:e oC PQほm町、dprimary water and首、epre酷ured問。0.2

。-0.2

Down fon:l'l by weiゆtof valve movable pan:

1∞ 40 60 80

MCP !Iow四 te( % )

Fig.4.7.2 Characteristics of the valve

Fig.4.7.3 1/2 scale ・odeland the test

Fig.4.7.1

Hydraulic pressure valve

Fig.4.7.4 Rapid openin~ valve

J A E R l - M 8 8 - 2 2 1

4.8 Preliminary Characteristic Analysis of a System-Integrated PWR in Inherently Safe Concept

T. Ise, F.Tanabe* and Oakajima*

Preliminary characteristic analysis has been performed for improvement of design of a system-integrated PWR(SPWR-H/H1100) in inherently safe concept". Figure 4.8.1 illustrates the primary system of a SPWR-H/H1100(1100 MWt.350 MWe; see Reference 1 for its design specifications) and Figure 4.8.2 displays the characteristic curve of a hydraulic pressure valve installed in place of the upper density lock in the reactor. We indicate by numerical evaluation that the reactor never fails to be passively shut down and cooled down by action of the hydraulic pressure valve at an accident .

WIMS-D is used for nuclear calculations and TRAC-PF1 for dynamic analyses of the reactor. The void and temperature coefficients which are significant as main negative feedback are shown in Fig. 4.8.3, and the modeling of the primary system for TRAC-PF1 is presented in Fig. 4.8.4.

As an example of the system behavior of severe transient, the main coolant pump trip transient is shown in the following. The pump is assumed to be tripped within 0.01 second. Since the pressure head of the pump is lost, the hydraulic pressure valve is opened by the gravity force and then the borated pool water starts flowing in the primary loop by force of the natural circulation (see Fig. 4.8.5). Consequently the reactor shuts down very quickly, as shown in Figs. 4.8. 6 and 4.8.7. As the natural circulation keeps continuing, the peak central temperature of the hottest fuel in the reactor is decreasing, as illustrated in Fig. 4.8.8.

References

1) Sako K.: "Design Study of System-Integrated Pressurized Water Reactor" Annual Rep. of Reactor Eng. Dept.,JAERI-M 87-126,ppl00-l02(1987).

2) Sako K. (Private communication).

+ Department of Reactor Safety Research; Present affiliation: Japan Institute of Nuclear Safety

* Century Research Center Corp.

- 9 0 -

]AERl・M 88・221

4.8 Preliminary Characteristic Analysis of a System-[ntegrated

PWR 1n Inherently Safe Concept

T.1se， F. T叩 abeφ 叩 dH.Nakajima.・

Preliminary characteristic analysis has加enperformed for improvement of

desIgn of a system-Integrated PWR(SPWR-H/HIIOO) in inherently safe concept lJ •

Figure 4.8.1 illustrates the primary system of a SPWR-H/HllOO(llOO附t，350附e:

see Reference 1 for its design specifications) and Figure 4.8.2 displays the

characteristic curve of a hydraulic pressure valve installed in place of the

upper density lock in the ~eactor. We indicate by numerical evaluation that the

r回 ctornever fails to be passively shut down and cooled down by action of the

hydraulic pressure valve at an accident •

WIMS-D is used for nuclear calculations and TRAC-PFl for uynamic analyses

of the reactor. The void and temperature coefficients which are significant as

ma.in negative feedback are shown in Fig. 4.8.3， and the modeling of the primary

system for TRAC-PF1 is presented in Fig. 4.8.4.

As an example of the system behavior of severe transient， the main coolant

pump trip transient is shown in the following. The pump is assumed to加 tripped

within 0.01 second. Since the pressure head of the pump is lost， the hydraulic

pressぽ evalve is opened by the gravity force and then the borated pool water

starts flowing in the prima.ry loop by fοrce of the natural circulation (see Fig.

4.8.5). Consequently the reactor shuts down very quickly， as shown in Figs. 4.8.

6 and 4.8.7. As the natural circuJation keeps continuing， the peak central tem-

perature of the hottest fuel in the reactor is decreasing， as illustrated in

Fig. 4. 8. 8.

References

1) Sako K. :“Design Study of System-1ntegrated Pressurized Water Reactor"

&U1ual Rep. of Reactor Eng. Dept..JAERI-M 87-126，pplOO-l02(1987).

2) Sako K. (Private communication).

+ Department of Reactor Safety Research: Present affiliation: Japan Institute

of Nuclear Safety

一回一

J A E R I - M 8 8 - 2 2 1

Wittr Itvtl

Feed water header (iS)

Upper intirtict. hydraulic pr iuur i valve (x3) Reactor pressure -vessel (10-S.Sm)

Poison £ tank 5 Thermal

Insulator

Connecting rib

lower interface. , honeycomb structure

Main circulating pump

Steam header (x6)

Pump delivery port (x3)

Poison cooler

. Poison downcomer

Core upper plate (x6) Core,

.fuel assembly (x12l)

, Fuel tentative storage (x30)

Tank support structure

Fig. 4.8.1 System integrated PWR11

pushing-up force by pressure difference

pushing-down force by temperature difference $. core pressure drop

..pushing-down force by weight

20 50 Coolant flow(X)

Fig. 4.8.2 Characteristic curve of hydraulic pressure valve 2'

x 1 O - 2

^ - 4 h *. -8

- l o h

Void coefficient of coolant

Void ratio(X) 5 0

Coolant temperature

c o e f f i c i e n t

4 0 0 5 0 0 6 0 0 7 0 0 Coolant tenperature(K)

Fig. 4.8.3 Main reactivity coefficients

J12 ^<k0 pressurizer \f rILL

F^R ^ initial water level

hydraulic sr?\lll = i pressure <^7fF \J]J6 <£& ^ valve { " ^ \TJ21 \ l \ 2 \

SG secondary

J22 J 23

<2>< VALVE

C^) node number " junction number

Fig. 18.4 Computational model for TRAC-PFl

- 9 1 -

Fig. 4. 8.1 System integrated PWRll

-3 x 1 0

ーl Void coefficient of coolant

唱にEIH¥』¥」

-3 l..--一一一一一一

ーーゐ一一司l..-

50 Void ratio(%)

-4 x 1 0 印。l飢 ttemperature

coefficient -2

一一hF¥S¥乏

400 500 600 700

C∞lant t'l園開rature(K)

Fig. 4.8.3 Main reactivity

coefficients

]AERI-M 88・221

2.0 push ing-up f orce by pressure difference

1、2詰 L

5 ー1.0Q，

5 I push ing-down f orce by

田邑8 0 . 5

0.0 。 20 50 100 C∞lant flow(l;)

Fig.ι8.2 Characteristic curve of

hydraulic pressure valve21

Fig. 4.8.4 Computational model for TRAC-PFl

7500 I—r-r

T—I 1—I I—I—I—I—I 1—I 1 1 1—I—I—I—I T

prinary coolant flow(total)" I

Elapsed tiae(s)

Fig. 4.8.5 HCP trip: primary coolant flow

2 0 00 (—i—i—i—I—r—i—i—i—I—i—i—i—I—I—|—I—i—i—i—i—i—i—rr

1 5 0 0

1 0 0 0

0-0 S.O 10.0 1E.0 20.0 25.0 Elapsed tiae(s)

Fig. 4.8.7 MCP trip: reactor power

-0.040-0.0 S.O M it i t_s-.0 •"• 2?. 25.0 lo.o is.o

Elapsed tiae(s) Fig. 4.8.6 HCP trip: compensating reactivities

1 0 15 Elapsed tiae(s)

Fig. 4.8.8 HCP trip: peak central temperature of hot spot fuel

10.0 15.0

llapsed ti・e(s}MCP trip: compensating reactivities

> +> o g l!?

・6j -om

:~t i apri岨 ryl c oolmt fE loI w{total)--

pri岨 ry.cool回 tf10・(noincident boron闘 ter:

I " " I 10.0 ‘・0

I!lapsed ti田 (s】

MCP trip: primary coolant flow

20.0 5.0

-2500 0.0

) s E ~ 2500

』〉開河【・

∞∞ENN-

Fig. 4.8.6

~l 000 匝 Z@

t t ド 500

Fig. 4.8.5

-会主』

OE』38a

00・0Z5 10 15

I!lapsed ti・e(s}MCP trip:開akcentral

temperature of hot spot fuel

Fig. 4. 8. 8

10.0 15.0

I!lapsed ti・e(s}MCP trip: reactor power

Fig. 4.8.7

J A E R I - M 8 8 - 2 2 1

4.9 Controllability of Load Reduction of ISER (Intrinsically Safe and Economical Reactor)

Y. Asahi

With regard to inherently safe reactors, there is a widespread concern such that they light be easily shut themselves down. This problem was scrutinized for the I S E R 1 , 2 , 3 ) . The thermal power of the ISER is 645 MWth- The steam generator is of once-through helical coil type. The other characteristics of the ISER include ; (1) the reactor pressure vessel is made of steel, and (2) the secondary system is equipped with the passive safety and shutdown system(PSSS)41 .

Load reduction can be regarded as a design basis event for a reactor control system. The 40$ load reduction was chosen as the first problem in designing the reactor control logic of the ISER. The load reduction was simulated by a sudden 40 % decrease of the turbine flow rate. The feedwater flow rate was assuied to maintain the steady state value until 100 sec and then decrease 40$ from 100 sec to 120 sec. The control logic designed for the load reduction is as follows '. (a) The reference value for the riser temperature (Triser) is 314 °C

\- I r i se r I •

(b) As soon as the signal of the load reduction is generated, the chemical volume and control system (CVCS) flow rate is made 10 tines as large as the rated value, in order to make the reactivity control by CVCS more effective. The value is always maintained during the control process, but when Tr,ser is near its reference value (±1 °C), the rated CVCS flow rate is recovered in order to avoid power oscillations.

(c) For T r i S 8 r > T r " r

r i s e r : CD The initial (steady state) boron concentration of the CVCS

feed-flow is maintained or recovered. © The main coolant pumps (MCP's) keep their present speed.

(d) For T r i 3 a r < T r e rr i s e r :

CD Boron dilution is actuated. @ In order to stop the interface flows, MCP's shall be controlled

with a time constant of 1,000 sec such that (1) when the lower interface flow is incoming, the MCP's speed

shall be twice as large as the rated value, and (2)' otherwise, they shall stop.

(e) Turbine bypass will be actuated at 75 ata, while it will be turned

JAERI-M 88・221

4.9 Controllability of Load Reduction of rSER (Intrinsically

Safe and Economical Reactor)

Y. Asahi

国ithregard to inherently safe reactor5， there i5 a widespread con-

cern such that they might be easily shut themselves do同n・Thisproblem

was scrutinized for the ISERt・2目 3) The thermal power of the ISER is 645

MWth. The steam generator is of once-through helical coil type. The

other characteristics of the ISER include (1) the reactor pressure

vessel is made of steel， and (2) the secondary system is equipped with

the passive safety and shutdown system(PSSS)4】.

Load reduction can be regarded as a design basis event for a reac-

tor control system. The 40% load reduction was chosen as the first

problem in designing the reactor control logic of the ISER. The load

reduction r..as sim1l1ated by a sudden 40 % decrease of the turbine flo叫

rate. The feedwater flo珂 ratewas assumed to maintain the steady state

value until 100 sec and then decrease 40% from 100 sec to 120 sec. The

control logic designed for the load reduction is as follows •

(a) The reference value for the ri5er temperature (Triser) is 314 oc (= Tr~'r"or)'

(b) As soon as the signal of the load reduction is generated， the chemical volume and control system (CVCS) flow rate is made 10

times as large as the rated value， in order to make the reactivity control by CVCS more effective. The value is always maintained

during the control process， but when Tr，ser is near its reference

value (土 1OC)， the rated CVCS flow rate i~ recovered in order to

avoid power oscillations.

(c) For Tr，ser > Tr~rriser

① The initial (steady state) boron concentration of the CVCS

feed -flo日 ismaintained or recovered.

② The main coolant pumps (MCP's) keep their present speed.

(d) ForTr，ser < Tre'riser' ① Boron dilution is actuated.

② In order to stop the interface floMs， MCP's shall be controlled with a time constant of 1，000 sec such that

(1) when the lower interface floM is incoming， the MCP's speed shall be twice as large as the rated valu宇， and

(2t otherwise， they shal1 stop. (e) Turbine bypass will be actuated at 75 ata， while it咽illbe turned

J A E R I - M 8 8 - 2 2 1

off at 57.5- ata. It is expected that, with the initiation of the transient, the flow

at the upper and lower hot/cold interfaces rapidly oscillates so that the borated water will enter the circulating primary system, thereby automatically lowering the reactor power. The calculation without the control logic shows (Fig.4.9.1) that for about 200 sec, thanks to the negative reactivity coefficient, the ISER can resist the automatic addition of negative reactivity resulting from entry of the borated water, but after 200 sec gradually shuts itself down. The results of the calculation with the control logic are as follows. After 160 sec, Tr,s»r (Fig .4.9.2) falls below 314 °C so that the MCP's are controlled to stop the interface flows (Fig.4.9.3), and simultaneously boron dilution is started to remove boron having entered the circulating primary system. As a result, the reactivity (Fig.4.9.4) and the interface flows (Fig.4.9.3) tend to vanish, while the reactor power (Fig.4.9.5) and T r i S e r (Fig.4.9. 2) asymptotically become constant. In these Figures, the influence of termination of the PSSS flow initiated at 600 sec can be seen. Since the control logic thus designed contains the control element with the time constant of as large as 1,000 sec, it does not conflict with the passive shutdown function of the ISER. Thus, it has been found that the ISER control system can be designed so that the reactor can follow large load variations such as a 40$ load reduction without having valves installed at the hot/cold interfaces.

This work was performed as a joint research between The University of Tokyo and JAERI.

References 1) Asahi Y., et al., "Safety of Intrinsically Safe and Economical

Reactor(ISER)", ANS Topical Meeting on Safety of Next Generation Power Reactors, May 1-5, 1988, Seattle, Washington, USA.

2) Kuwahara S., et al., "Design Study of the Primary System of the Intrinsically Safe and Economecal Reactor", Advisory Group Meeting on Planning on New Concepts of Light Water Reactors, IAEA, Vienna, Austria, Dec.9-11, 1985.

3) Oda J., et al., "A Conceptual Design of Intrinsically Safe and Econoniecal Reactor (ISER)", TCM on Advances in LWR Technology, IAEA, Washington, D.C., USA, Nov. 24-25, 1986.

4) Asahi Y., and H. Wakabayashi, "Improvement of Passive Safety of Reactors", Nucl.Sci.Eng., 96,73-84(1987).

- 9 4 -

]AER 1 -M 88-221

off at 57.5' ata.

It is eKpected that， with the initiation of the transient， the f10w at the upper and 10wer ho:/co1d interfaces rapidly osci11ates so that

the borated water will enter the circu1ating primary system， thereby a1ltomatical1y 10wering the reactor power. The ca1cu1ation without the

contro1 10gic shows (Fig.4.9.1) that for about 200 sec， thanks to the negative reactivity coefficient， the ISER can resist the automatic addi-tion of negative reactivity resulting from entry of the borated water， but after 200 sec gradually shuts itself down. The results of the calcu-

lation with the control logic are as follows. After 160 sec， Trlser (Pig

.4.9.2) falls be10w 314 oc so that the MCP's are control1ed to stop the

interface f10ws (Fig.4.9.3)， and simu1taneously boron dilution is start-

ed to remove boron having entered the circu1ating primary system. As a

result， the reactivity (Fig.4.9.4) and the interface flows (Fig ・4.9.3)

tend to vanish， whi1e the reactor power (Fig.4.9.5) and Tr.ser (Fig.4.9.

2) asymptotical1y become constant. 1n these Figures， the influence of termination of the PSSS f10w initiated at 600 sec can be seen. Since the

control 10gic thus designed contains the control element with the time

constant of as 1arge as 1，000 sec， it does not conf1ict with the passive shutdown function of the ISER. Thus， it has been found that the rSER control syst~m can be designed so that the reactor can fo11ow 1arge 10ad

variations such as a 40出10adreduction without having va1ves insta11ed

at the hot/cold interfaces.

This work was performed as a joint research between The University

of Tokyo and JAERI.

References

1) Asahi Y.， et a1.， "Safety of Intrinsica1ly Safe and Economica1 Reactor(ISER)"， ANS Topical Meeting on Safety of NeKt Generation Power Reactors， May 1・5，1988， Seatt1e， Washington， USA.

2) Ku開aharaS.， et a1.，“Design Study of the Primary System of the Intrinsically Safe and Economecal Reactor"， Advisory Group Meeting on Planning on New Concepts of Light Water Reactors， 1AEA， Vienna， Aust.ri a， Dec. 9・11，1985.

3) Oda J.， et a1.，“A Conceptual Design of Intrinsica11y Safe and

Economecal Reactor (ISER)"， TCM on Advances in LWR Technology， IAEA， Washington， D.C.， USA， Nov. 24・25，1986.

4) Asahi Y.， and H. Wakabayashi， "Improvement of Passive Safety of Reactors"， Nuc1.Sci.Eng.， 96，73・840987J.

J A E R I - M 88-221

1 — -"1 ' i

UJ S 0.8 o a.

- v\ -

UJ 0.6 > < 0.4 UJ

" 0 . 2 -

1 1 i 0 0 100 200 300

TIME (SEC)

Fig.4.9.1 Reactor Power (without control)

...ZL reference level

riser temperature

lower plenum temperature

0 400 800 1200 TIME (SEC)

4.9.2 Riser and Lower Plenum Temperatures (with control)

400 800 1200 TIME (SEC)

Fig.4.9.3 Lower Hot/Cold Interface Flow (with control)

400 800 TIME (SEC)

Fig.4.9.4 Total Reactivity (with control)

400 800 TIME (SEC)

Fig.4.9.5 Reactor Power (with control)

]AER 1・M 88・221

門川

区凶主

-F4J凶巴

300 200 ( SEC)

1∞ TIME

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J A E R I - M 8 8 - 2 2 1

4.10 A Very High Burnup Pressurized Water Reactor with Highly Enriched Plutonium Fuel Assemblies Using a Spectral Shift Concept

H.Ichlkawa, Y.Naito and H.Takano

Advanced versions of the existing commercial light water reactors (LWRs) are being developed to obtain higher level of economics and safety, and to utilize uranium resources effectively. In oder to improve the economics, it is important to study a feasibility of higher burnup and longer life fuel assemblies.

Highly enriched plutonium mixed-oxide (MOX) fuel assemblies are being considered to achieve higher fuel burnup and/or longer core life in the present pressurized water reactor (PWR). Parametric study for

1 ) nuclear characteristics was carried out for these fuel assemblies, 2) using the SRAC code system

Some burnup calculations for uniform MOX fuel rods show that an initial plutonium enrichment of more than 20wt% provides a sufficient excess reactivity to reach fuel burnup up to 100,000 MWd/tHM under the condition of almost the same moderator-to-fuel (H/F) volume ratio as the existing PWR core lattice. It is also shown that the increase of H/F volume ratio causes to increase the excess reactivity, because the optimum lattice pitch for the reactivity is considerably larger than that of the existing PWR core.

On the other hand, a large amount of the plutonium loadings makes the initial excess reactivity exceedingly large, and the moderator void reactivity coefficient becomes positive on early burnup stage.

To solve these two primary problems, the fuel assemblies which consist of highly enriched plutonium MOX fuel rods and natural/depleted U0„ rods are proposed. A typical concept of these fuel assemblies using highly enriched plutonium MOX fuel is shown in Fig. 4..10.1.

The natural/depleted U0_ rods play a role as a kind of control rods to decrease the initial excess reactivity. Besides that, withdrawing these rods from the assemblies at some burnup stages, the H/F volume ratio of the core becomes larger, and consequently, the reactivity increases, so that it is possible to compensate for reactivity decrease due to fuel burnup. That is, this methode utilizes a concept of "neutron spectral shift".

In the case of fuel assembly shown in Fig. 4..10.1, the H/F volume

]AERI-M 88・221

4.10 A Very High Burnup Pressurized Water Reactor with High1y Enriched

P1utonium Fue1 Assemb1ies Using a Spectra1 Shift Concept

H.lchikawa， Y.Naito and H.Takano

Advanced versions of 七he自x::'s七ingcommercia1 1igh七wa七日rreac七ors

(LWRs) are being developed to obtain higher level of economics and

safe七y，and七ou七ilizeuranium resources eff申ctively. In oder to

improve七heeconomics， i七 isimportan七七os七udya feasibili七Yof higher

burnup and longer life fu自1ass自mblies・

Highly enriched plu七oniummixed-oxide (MOX) fuel assemblies are

being considered七oachieve higher fuel burnup and/or longer core life

in七hepresen七pressurizedwater reac七or (PWR). ParRmetric s七udytor 1)

nuclear charac七eris七icswas carried ou七 for七hesefuel assemblies，

using the SRAC code sys七em2)

Some burnup calcula七ionsfor uniform MOX fuel rods show that an

initial plu七oniumenrichment of more七han20wt% provides a sufficien七

excess reactivity七oreach fuel burnup up to 100，000 MWd!tHM under七he

condition of almos七 thesame modera七or-七o-fuel(H!F) volume ra七ioas七he

exis七ingPWR core la七七ice. It is also shown七ha七七heincrease uf H/F

volume ra七iocauses七oincrease the excess reactivity， because七he

optimum la七七icepi七chfor 七hereac七ivityis considerably larger 七han

tha七 ofthe existing PWR core.

On七heother hand， a large amour七 of七heplu七oniumloadings makes

the ini七ialexcess reactivity exceedingly large， and七hemodera七orvoid

reactivi七ycoefficien七 becomespositive on early burnup stage.

To solve these two primary problems， the fuel assemblies which consis七 ofhighly enriched plu七oniumMOX fuel rods and na七ural/deple七ed

rods are proposed. A七ypicalconcept of these fuel assemblies using

highly enriched plutonium MOX fuel is shown in Fig. 4.10.1.

The natural!deple七edU02

rods play a role as a kind of control rods

to decrease the ini七ialexcess reac七ivi七y. Besides七ha七， wi七hdrawing

七heserods from七heassemblies a七 someburnup stages，七heH/F volume

ra七ioof the core becomes larger， and consequently，七hereactivity

increases， so七ha七 itis possible七ocompensa.te for reacti vi ty decrease

due to fuel burnup. That is，七hismethode utilizes a concept of

lIneutron spectral shif七"・

In七hecase of fuel assembly shown in Fig. 4.10.1， the H/F volume

一崎一

JAERI-M 88-221

ratio becomes twice (1.5-3.0) by withdrawing all control rods, and excess reactivity increases by about 15% k/k. Of course, the degree of reactivity recovering can be adjusted to be smaller as possible by withdrawing control rods one after another, grouping into several batches. This spectral shift concept permits very high burnup fue^ assemblies more than 100,000 MWd/tHM as shown in Fig. 4.10.2.

Another problem for the highly enriched plutonium fuel assemblies is that concerned with the moderator void reactivity. However, since the average plutonium enrichment of a fuel assembly in initial core is almost a half of M0X fuels, the moderator void reactivity coefficient for the assemblies never has positive values. Furthermore, the absolute value becomes larger and larger with proceeding of fuel burnup and/or increasing of H/F ratio by the spectral shift.

A typical core configuration constructed as the results described above is shown in Fig. 4.10.3. The neutronic characteristics of this core are as follows.

(1) A very high burnup (>100,000 MWd/tKM) of fuel assembly is obtained as shown in Fig. 4.. 1 0.2. (In this figure, burnable poison, Gd ?0,, is also considered.) (2) Excess reactivity of the core is kept low through the whole exposure time by using the spectral shift concept. (3) Moderator void reactivity coefficient is negative in any burnup stage. Neutronic studies are also carried out for the fuel assemblies

utilizing ThO in place of U0„ as control rods. The results obtained for the Th0„ rods are very similar to those for the U0 ? rods described above.

References 1) Ichikawa H. et al.: to be presented at "The 1988 International

Reactor Physics Conference", Jackson Hole, Wyoming, Sept. 1988 2) Tsuchihashi K. et al.: "Revised SRAC Code System", JAERI 1302

(1986).

jAERI-M BB・221

ra七iobecomes twice (1.5-3.0) by withdrawing all con七rolrods， and excesil reac七ivityincreases by about 15% k/k. Of course， the degree o.f

reac七ivity r申coveringcan be adjust自d to be smaller as possible by

wi七i1drawingcon七rolrods one af七日ranother， grouping into several

batches. This spec七ral shif七 concep七 permi七svery high burnup fuel

assemblies more七han100，000 MWd/tHM as shown in Fig. 4.10.2. Another problem for七hehighly enriched plu七onlumfuel assemblies

is 七hatconcerned with七hemoderator void r申3.C七ivi七y. However， since

もheaverage plu七oniumenrichmen七ofa fu自1assembly in ini七ialcore is

almos七 ahalf of MOX fuels，七hemoderator void reacもivitycoefficient

for the dssemblies never has posi七ivevalues. Fur七hermore，the absolu七e

value becomes larger and larger with proceeding of fuel burnup and/or

increasing of H/F ra七ioby七hespec七ralshift.

A typical core configura七ionconstructed as七heresul七sdescribed

above i8 shown in Fig. 4.10.3. The neu七roniccharac七eristicsof this

core are as follows.

(1) A very high burnup (>100，000 MWd/t凹 of fuel assembly is

obtained as shown in Fig. 4.10.2. (In 七hisfigure， burnable poison， Gd~O~ ， is also considered.) 2-3 (2) Excess reactivity of the core 1s kept low through the whole

exposure tirne by using the spectral shift concep七.

(3) Modera七orvoid reactivi七ycoefficient is negative in any burnup

s七age.

Neu七ronic s七udiesare also carried out for七hefuel assernblies

utilizing Th02

in place of U02

as control rods. The results ob七ained

for the Th02

rods are very similar七othose for七heU02

rods described

above.

References

1) Ichikawa H. et al.: 七obe presented at "The 1988 International

Reac七orPhysics Conference"， Jackson Hole， Wyoming， Sept. 1988

2) Tsuchihashi K. e七 al.: "Revised SRAC Code System"， JAERI 1302 (1986) •

J A E R I - M 8 8 - 2 2 1

2 4 . 24 0 .87

i * "*V

urn t : cm

Fig. 4.10.1 Typical Fuel Assembly Utilizing 20Z Plutonium Enriched MOX

1 .0 -

0.9 50 100 Burnup (GWd/t)

Fig. 4.. 10.2 Excess Reactivity vs. Burnup

Type-(A) ( 60 ass.s)

Type-(A) : such like the control rod assembly of existing PWRs; UCU rods are withdrawn/inserted by the attached driving mechanism.

Type-(B) : UO, rods are withdrawn and taken away during the reactor shutdown.

Type-(B) (133 ass.s)

Fig. 4..10.3 A Concept of Core Configuration

J AE R 1・M 88・221

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， .2

8 '.1 ..!:l:

0.9O 50 100 Burnup (GWd/t)

Fig. 4.10.2 Excess Reac七ivi七.yvs. Burnup

Type-(A) such like the control rod assembly of exist-ing PWRs; U0

2 rods are

withdrawn/inserted by the attached ariving mechanism.

Type・(B) u02

rods are withdrawn

and taken away during the reactor shutdown.

Type-(B) (133 ass.s)

Fig. 4.10.3 A Concept of Core Configuration

一随一

J A E R I - M 8 8 - 2 2 1

A-.ll Analysis of Fission Product and Minor Accinide Build-up in Very High Burnup Pressurized Water Reactor

H. Takano, H. Ichikawa, Y. Naito and M. Ido'

One of important problems to improve economics of the existing commercial light water reactors is to study a feasibility of very high burnup fuel assembly. A parametric study of nuclear characteristics for the MOX fuel assembly with high plutonium enrichment of 20 wt% showed a possibility of high burnup more than 100 GWD/t l ). The build-up of fission product (FP) and minor actinide (MA) depend strongly on burnup state. It is very important that amount of FP and MA build-up with burnup is evaluated accurately on the view point of design studies for fuel, structural materials, FP-gas plenum and nuclear fuel cycle in very high burnup pressurized water reacror (VHBPWR).

Burnup calculations for the lattice assembly described in Section 4.10 were performed with the SRAC-FPGS3 code system 2 ) developed on the basis of SRAC 3 1 and FPGS-3 4 ) codes. The calculated results are compared with those obtained for a typical pressurized water reactor (PWR) with the U-235 enrichment of 3.2 wt%.

Figs.4.11 . 1 and 2 show the decay heat change as a function of burnup and cooling times, respectively. The decay heat variation in cooling for VHBPWR is more mild than that for PWR.

Figs.4.11.3 and 4 show FP-gas build-up as a function of burnup times in VHBPWR and PWR, respectively. At high burnup time the FP-gas quantity for VHBPWR is about 3 times of those for PWR. A very long FP-gas plenum will be required for fuel design in VHBPWR, if a special device for the FP-gas release is not considered.

Figs.4.11.5 and 6 show neptunium isotopes build-up as a function of burnup time, as an example of many minor actinides. The Np-237 is a very troublesome TRU nuclide with exceptionally long half life of 2 x 106 years. Though the build-up of Np-237 for PWR is larger than that for VHBPWR in burning, the build-up quantity for VHBPWR increases faster than that for PWR with cooling time.

• I.S.L. Inc. . Tokyo

jAERI-M 88・221

今.11 Analysis of Fission Product and MinoI' Actinide Build-up

in Very High Burnup Pressurized Water Reactor

H. Takano， H. Ichika.wa， Y. ::-.faito and M. Ido'

One of important problems to improve economics of the existing

commercial light water reactors is to study a feasibility of very high

burnup fuel assembly. A parametric study of nuclear ch且racteristics

for the MOX fuel assembly wi th high plutonium enrichment of 20 wt%

showed a possibility of high burnup more than 100 GWD/tLl. The build-up

of fission prηduct (FP) alld minor actinide (MA) depend strongly on

burnup state目 It is very important that amount of FP and MA build-up

with burnup is evaluated accurately on the view point of design studies

for fue!， structurul materials， FP-gas pler".lm and nuclear fuel cycle

in very high burnup pressurized water reac!'or (VHBPWR).

Burnup c且 lculations for the lattice assembly described in Section

4.10 wer母 performed wi th the SRAC-FPGS3 code system 幻 developed on

the basis of SRAC3l and FPGS-34l codes. The calculated results are

compared wi th those obtaiロed for a typical pressurized water reactor

(PWR) wi th the U-235 enr i chment ()f 3.2 wt%.

Figs.4. 11.1 and 2 show the decay he且 tchange as a function of burnup

and cooling times， respectively. The decay heat variation in cooling

for VHBPWR is more mi ld than that for PWR.

Figs.4.11.3 and 4 show FP-gas build-up as a function of burnup

times in VHBPWR and PWR， respectively. At high burnup time the

FP-gas quant i ty for VHBPWR i s about 3 t imes of those for PWR. A very

long FP-gas plenum will be required for fuel design in VHBPWR， if a

special device for the FP-gas release is not considered.

Figs.4.11.5 and 6 show nept~~ium isotopes build-up as a function

of burnup time， as an example of m町 ly minor ac t inides. The Np-237

is a very troublesome TRU nuclide with exceptional1y long half !ife

of 2 x 106 years. Though the bui Id-up of Np-237 for PWR is larger than

that for VHBPWR in burning， the bui ld-up quant i ty for VHBPWR

increases faster than that for PWR with cooling time.

• l.S.L. Inc.， Tokyo

JAERI-M 88-221

R e f e r e n c e s

1) Ich ikawa H. , Y. Nai to and H. Takano: S e c t i o n 4.10 in t h i s r e p o r t .

2) Takano H. et a l . : to be pub l i shed

3) T s u c h i h a s h i K. et a l . : "Rev i sed SRACCode Sys tern" .JAER11302 (1987) .

4) I h a r a H. e t a l . : p r i v a t e commun ica t i on

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JAERI-M 88・221

Refprences

in this report. 1) Ichikawa H.， Y. Naito and H. Takano: Section 4.10

to be published 2) Takano H. e t a 1 . :

3) Tsuchihashi K. et al.:"RevisedSRACCode System"，JAERII302 (1987).

4) Ihara H. et al.: private communication

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F i g . 4 . 1 1 . 5 Nep tun ium i s o t o p e s b u i l d - u p as a func t ion of burnup t i m e

in VHBPWR

TIME (DAY)

4 . 1 1 . 6 Neptunium i so topes build-up as a funct ion of burnup t ime

in PWR

JAERI・M 88・221

c.，:)

10 102 10 J

=o-出量

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in VHBP¥VR

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in PWR

-101ー

J A E R I - M 8 8 - 2 2 1

4.12 Conceptual Study of Higher Actinide Burner Reactors

T.Mukaiyama, H.Takano, T.Takizuka, T.Ogawa and M.Osakabe*

The nuclear transmutation has been proposed as one of options for the high-level radioactive wastes management. The major candidate for the disposal techniques in most of nuclear countries is the permanent disposal of wastes into a certain geologic formation. The combination of the transmutation of long-lived wastes and the geologic disposal of shorter lived nuclei will technically close the fuel cycle backend and provide the visible option of the waste management to the general public.

Of the choices for the potential transmutation schemes, we have proposed an actinide burner fast reactor1'' '. Since the most of higher actinides are fissionable with fission thresholds in the several hundred keV range, they can be used as nuclear fuel in hard neutron spectrum.

Concept of Actinide Burner Fast Reactor 1) Metal-alloy fuel with Na-cooling

To achieve the harder neutron spectrum, the higher concentration of fissile materials and the lower concentration of light elements which will be the moderator of energetic neutron are required. The metal-alloy-fueled core provides considerably harder neutron spectrum than the core with oxide fuels. In our study, Np-20wt%Zr alloy and Am-Cm~5wt%Y alloy were selected as the possible actinide alloy fuels. Yttrium is mixed as a thermal diluent.

The bundle-pin type metal fueled fast reactor and the tube-in-shell type metal fueled fast reactor •' are studied as the actinide burner reactor concepts. The core consists of Am-Cm-Y alloy fueled inner core and Np-Zr alloy fueled outer core in these reactors.

2) Particle bed reactor with He-cooling The particle bed reactor concept ' was studied as an alternate

actinide burner reactor. This concept has an advantage of high efficiency in heat transfer because small particle size produces a large heat

* Present address : Ibaraki University, Faculty of Engineering

JAERI-M 88・221

4.12 Conceptual 5tudy of Higher Actinide Burner Reactors

T.Mukaiyama， H.Takano. T.Takizuka， T.Ogawa and M.Osakabe持

The nuclear transmutation has been proposed as one of options for

the high-level radioactive wastes management. The major candidate for the

disposal techniques in most of nuclear countries is the permanent disposal

of wastes into a certain geologic formation. The combination of the

transmutation of long-lived wastes and the geologic disposal of shorter

lived nuclei will technically close the fuel cycle backend and provide the

visible option of the waste management to the general public.

Of the choices for the potential transmutation schemes， we have

proposed an actinide burner fast reactor1)，2). Since the most of higher

actinides are fisstonable with fission thresh01ds in the several hundred

keV range， they can be used as nuclear fuel in hard neutron spectrum.

Concept of Actinide ~uKn~rfast Reactor

1) Metal-alloy fuel with Na-cooling

To achieve the harder neutron spectrum， the higher concentration of

fissile materials and the lower concentration of light elements which wil1

be the moderator of energetic neutron are required. The metal-alloy-fueled

core provides considerably harder neutron spectrum than the core with

oxide fue1s. 1n our study， Np-20wt%Zr alloy and Am-Cm即 Swt%Yalloy were

selected as the possib1e actinide alloy fuels. Yttrium is mixed as a

thermal diluent.

The bundle-pin type metal fueled fast reactor and the tube-in-shell

type metal fueled fast reactor3) are studied as the actinide burner

reactor concepts. The core consists of Am-Cm-Y a110y fueled inner core

and Np-Zr alloy fueled outer core in these reactors.

2) Particle bed reactor with He-co01ing

The particle bed reactor concept4) was studied as an alternate

actinide burner reactor. This concept has an advantage of high efficiency

in heat transfer because sma1l particle size produces a large heat

普 Presentaddress 1baraki University， Faculty of Engineering

JAERI-M 88-221

t r a n s f e r surface. In our s tudy , the cold fuel concept ( fuel t e m p e r a t u r e < 0.3 x

me l t i ng t e m p e r a t u r e ) i s app l i ed . A fuel p a r t i c l e c o n s i s t s of a c t i n i d e n i t r i d e k e r n e l of 1mm in d i ame te r and double t h i n s h e l l s of TiN of d i f fe ren t d e n s i t y .

3) Nuclear and thermalhydraulic calcula t ion The a c t i n i d e burner c a l c u l a t i o n code system "ABC" and JENDL-2

l i b ra ry were used for the nuclear ca lcula t ion . ABC code system cons i s t s of the code "SLAROM", the code "CITATION" and the code "0RIGEN-2".

In the bundle-pin fuel r ea r to r , the fuel configuration i s the p e l l e t diameter of 4mm and the pin pitch of 6mm. The average temperature at the fue l p e l l e t c e n t e r i s about 700 C which i s we l l below the m e l t i n g t e m p e r a t u r e . In the hot channe l , the t e m p e r a t u r e a t the fue l c e n t e r i s 820 C.

The t ube - in - she l l type fuel reactor gives very high volume f rac t ion of fuel and provides the same level of thermalhydraulic performance as the bundle-pin fuel configuration.

Resul ts The comparisons of the transmutation efficiency of various burner

reactors proposed are given in the tables 4.12.1 and 4.12.2. The combination of the larger fission to capture rate ratio and the shorter fission half-life is the index of good transmutation efficiency. In this context, clearly the actinide burner reactors proposed here represent the most attractive transmutation reactors.

References

1) T.Mukaiyama et a l . : in Proc.Intn ' l Conf.Nucl.Cross Sections for Technology(Knoxville,1979),NBS-SP-594,p552 (1980).

2) H.Murata, and T.Mukaiyama, : Atomkernenergie-Kerntech.,45,p23( 1984). 3) T. Hiraoka et a l . : Section 4.15 in t h i s r epor t . 4) F. L. Horn et a l . : BNL-38379 (1986).

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jAERI-M 88・221

transfer surface.

1n our study， the cold fuel concept (fuel temperature く 0.3 x

melting temperature) is applied. A fuel particle consists of actinide

nitride kernel of lmm in diameter and double thin shells of TiN of

different density.

3) Nuclear and thermalhydraulic calculation

The actin:ide burner calculation code system "ABC" and JENDL由 2

library were lJ se~d for the nuclear calculation. ABC code system consists

of the code "SLA;ROM・¥thecode "C1TATION" and the code "OR1GEN-2".

1n the bundle-pin fuel rea"Ctor， the fuel configuration is the pellet

diameter of 4mm and the pin pitch of 6mm. The average temperature at the

fuel pellet center is about 700 C which is well below th~ melting

temperature. 1n the hot channel， the temperature at the fuel center is

820 C.

The tube-in-shell type fuel reactor gives very high volume fraction

of fuel and provides the same level of thermalhydraulic performance as the

bundle-pユnfuel configuration.

旦些斗主主

The comparisons of the transmutation efficiency of various burner

reactors proposed are given in the tables 4.12.1 and 4.12.2. The

combination of the larger fission to capture rate ratio and the shorter

fission half-life is the index of good transmutation efficiency. 1n this

context， clearly the actinide burner reactors proposed here represent the

most attractive transmutation reactors.

References

1) T.Mukaiyama et a1. in Proc.lntn'1 Conf.Nucl.Cross Sections for

Techno1ogy(Knoxvi11e，1979)，NBS-SP-594，p552 (1980).

2) H.Murata， and T.Mukaiyama， Atomkernenergie-Kerntech.，土立，p23(l984).

3) T. Hiraoka et a1. Section 4.15 in this report.

4) F. L. Horn et a1. BNL-38379 (1986).

-103一

J A E R I - M 8 8 - 2 2 1

Table 4. 12.1 Comparison of fission to capture ratios

Nuclide LWR LMFBR BP-ABMR TS-ABMR HePR-ABNR

Np-237 - 0 0 . 1 6 0 . 9 5 1 .00 0 . 6 5 Pu-238 0.05 1 .17 3 . 1 2 3 . 1 9 2 . 5 4 Pu-239 2.4 3 . 1 0 9 . 6 1 9 . 8 7 7 .01 Am-241 - 0 0 . 1 3 0 . 6 6 0 . 7 0 0 . 4 8 Am-243 - 0 0 . 1 2 0 . 7 5 0 . 8 0 0 .51 Cm-242 0 .71 3 . 8 3 4 . 0 2 2 . 7 5

Cm-243 3.4 1 3 . 3 3 1 . 7 3 2 . 4 Cm-244 0 .02 0 . 6 0 3 . 0 2 3 . 1 6 2 . 2 5

LWR : Light Water Reactor LMFBR : Na-cooled MOX fuel fast breeder

BP-ABMR : Na-cooled bundle-pin actinide metal burner IS-ABMR : Na-cooled tube-in-shell actinide metal burner

HePR-ABNR : He cooled particle-bed actinide nitride burner

Table 4. 12. 2 Comparison of half-lives of natural decay and transmutation

( unit; year)

LWR LMFBR BP-ABMR TS-ABMR HePR-ABNR Nuclide Natural <t

Decay s - ' c m " 3 X 1 0 " 5 X 1 0 " 5 X 1 0 " 5 X 1 0 " 5 X 1 0 "

T, 2.3 2.6 2.6 2.6 "*Pu 2 . 4 X 1 0 ' T« 7.2 25 26 19

T, 1.8 2.4 2.4 2.3

T, 4X10 ' 15 7.3 7.1 7.7 " T N p 2 . 1 X 1 0 ' T, 4 2.4 6.9 7.1 5.6

T, 4 2.0 3 . 5 . 3.5 3.2

T, 3.9 3.2 3.2 3.3 " ' P u 86 T« 4.6 10 10 8.8

T, 2.1 2.4 2.4 2.4

T, 230 15 7.3 7.0 7.9 '"Am 433 T, 1 2.1 4.8 4.9 4.1

T, 0.9 1.8 2.9 2.9 2.7

T, 1X10' 20 8.8 8.4 9.6 '"Am 7 . 4 X 1 0 3 T« 9 2.4 6.6 6.8 5.4

T, 9 2.2 3.8 3.8 3.5

T, 610 11 5.9 5.7 6.1 '"Cm 18 T« 50 6.7 18 18 14

T, 13 4.2 4.4 4.4 4.4

f ; f i ss ion . c : capture . t ; fission + capture

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]AERI-M 88・221

Table 4.12. 1 Comparison of fission to capture ratios

Nuclide L¥JR u佐llR BP-A8~依 TS-A8HR HePR-A8NR

Np-237 、。 0.16 O. 95 1. 00 O. 65

Pu-238 O. 05 1. 17 3.12 3.19 2. 54

Pu-239 2. 4 3. 10 9.61 9. 87 7.01

Am-241 ー0 0.13 0.66 O. 70 O. 48

Am-243 句。 0.12 O. 75 0.80 0.51

Cm-242 0.71 3. 83 4.02 2. 75

Cm-243 3. 4 13.3 31. 7 32. 4

Cm-244 0.02 O. 60 3.02 3.16 2. 25

LWR Light Water Reactor

Lt1FBR Na-c∞led t10X fuel fast br百 der

BP-ABt1R Na-c∞led凶ndle-pinactinide r陪 talburner

TS-ABt1R Na-c∞led tu出 -in-shellactinide metal出 rner

HePR-ABNR He cooled particle-民dactinide nitride burner

Table ~. 12. 2 Comparison of half-lives of natural ~記ay and tr8I硲nutation

(凶itoyear)

LWR ll1FBR BP-ABt依百四品ぽ1iP，2R-A聞RNuc1ide Natura1 φ

[)ecay 5-'cm" I 3XI0" 5XIO'・5XIO'・5X10" 5X 10'・T， 2.3 2.6 2.6 2.6 1..Pu 2.4 X 10‘ L 7.2 25 26 19 τ. 1.8 2.4 2.4 2.3

T， 4XIO‘ 15 1.3 7. 1 7. 7 "TNp 2.1 x 10‘ Tc 4 2.4 6.9 7. 1 5. 6

r. 4 2.0 3.5 3.5 3. 2

τ， 3.9 3.2 3. 2 3.3 .. .Pu 86 T‘ 4.6 10 10 8.8

T‘ 2. 1 2. ~ 2.~ 2.4

T， 230 15 7.3 7.0 7. 9 ''''Am ~33 L l 2. 1 4.8 4.9 4. 1

T. 0.9 1.8 2. 9 2.9 2. 7

T， 1 x 10・ 20 8.8 8.4 9. 6 2・'Am 7.4XI0' τe 9 2.4 6.6 6.8 5. 4

r. 9 2.2 3.8 3.8 3.5

T， 610 11 5.9 5. 7 6.1 z・'UII 18 τe 50 6. 7 18 18 14

T. 13 4.2 4.~ ~.~ 4.4 一一一 ---ー園町圃圃圃圃圃園町聞一一一一 l 一一一一一

f fission. c; capture. t fission +岨ptUrC

J A E R I - M 8 8 - 2 2 1

4.13 Concepts of TRU Fuel and Predictions of Phase Diagrams

T. Ogawa

Burning transuranium elements (TRU) in a fission reactor is one of the future options to get rid of the long-term risk of nuclear

237 energy. Effective half life of Np, which will dominate the risk

s e of disposed waste after 10 years, could be reduced by a factor of 10 in a fast reactor. For the transmutation, hard neutron spectrum is desirable; softer spectrum reduces the fission-to-capture ratio of TRU nuclides. Successive capture reactions eventually increase the fraction of heavier nuclides. From this point of view, the most promising TRU fuel is the metallic one with high heavy-metal density and low contents of light elements. Feasibility of the all metal core consisting mostly of the waste TRU (Np, Am, Cm etc.) and probably some Pu is being investigated in JAERI. Other requirements to the TRU fuel are: (1) Accommodation to the change of TRU compositions during recycling. (2) Sufficiently high phase stability. (3) Good thermal property and achievement of high power density. (4) Ease of fabrication and reprocessing.

The last point is particularly important, but hard to answer, since it will depend on the state of the art of FBR fuel technology in some years ahead. One wants to avoid transportation of irradiated TRU fuel which emits more neutrons and alpha decay heat. And a compact fuel cycle concept based on pyrochemical processes of the metallic FBR fuel appears certainly attractive.

However, data for designing TRU metal fuel are wanted. Even in the initial conceptual phase of designing, density and melting points (solidus) have to be known as functions of TRU alloy composition; eutectics with cladding material and high-temperature thermal conductivity are the other pertinent data. TRU should be alloyed to increase melting points to achieve large linear heat rating. The alloying will also enhance burnup rate by reducing TRU contents at a given power density. It is, however, always born in mind that alloying will also soften neutron spectrum and may decrease thermal conduct iv ity.

Experimental data concerning constitution of alloys of TRU elements are too scarce. Therefore, phase diagrams of binary systems containing TRU and transition metals were predicted by a regular solution model. The results are summarized as follows:

(1) Large difference in internal pressure (ratio of atomization energy to molar volume of the element) would limit the mutual solubility of Np and Am. The system would resemble to U-RE(rare earths) systems in its phase behavior.

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JAERI-M 88・221

4.13 Concepts of TRU Fue! and Predictions of Phase Diagrams

T. Ouwa

Burning transuraniuM ele剛ents <TRU) in a f iss ion reactor is one of the future options to get rid of the long-term risk of nuclear

energy. Lk

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of dispused waste after 10 years， could be reduced by a factor of 10 in a fast reactor. F'or the trans剛utation，hard neutron sp廿ctru嗣 isdesirable; softer spectruM reduces the fission-to-capture ratio of TRU nuclides. Successive capture reactlons eventua.lly increase the f ract ion 0 f heav ler nuc J ldes. F'rom th ls pυInt of vlew， the ruost promis ing TRU fuel Is the 剛etalllcone wlth hlgh heavy-metal density and low contents of light eleMents. F'easlbillty of the all Metal core consisting Rlostly of the waste TRU (Np， Am， Cm etc.) and probably somc PU is being investigated in JAERI. Other requirements to the TRU fuel are:

(1) Acco同開。dation to the change of TRU compositions during recyc 1 ing. (2) SufficientIy high phase stability. (3) Good thermal property and achievement of high power density. (4) Ease of fabrication and reprocessing.

The last point is particularly important， but hard to answer， since it will depend on the state of the art of F'BR fuel technology in some years ahead. One wants to avoid transportation of irradiated TRU fuel which emits More neutrons and alpha decay heat. And a cOMpact fuel cycle concept based on pyrocheMlcal processes of the IIIetall ic F'BR fuel appears certa Inly attract Ive.

However， data for designlng TRU lIIetal fuel are wanted. Even In the initial conceptual phase of designing， density and 則eltlngpoints (sol idus) have to be known as functlons of TRU alloy cOlllpositlon; eutectics with claddlng 聞aterial and high-telllPerature therlllal conductivity are the other pertinent data. TRU should be alloyed to increase Melting points to achieve large linear heat rating. The alloying will also enhance burnup rate by reducing TRU contents at a given power density. It is， however， always born In lIIind that alloying will also soften neutron spectru. and剛aydecrease therlllal conductivity.

Experi踊ental data concerning constitution of alloys of TRU elelllents are too scarce. Therefore， phase diagra同sof binary syste聞S

containing TRU and transition lIIetals were predicted by a regul~r solution則。del. The results are su嗣lIIarizedas follows:

(1) Large difference in internal pressure (ratio of atolllization energy to 開。 lar volu聞eof the ele剛ent) wou ld 1 i岡 itthe lIIutual solubility of Np and A.. The systelll would rese踊bleto U-RE<rare earths) syste聞s in i ts phase behav ior.

一105-

J A E R I - M 8 8 - 2 2 1

(2) Melting point (solidus) of Am and Cm would be effectively increased by alloying with Y with a little sacrifice of neutron economy.

(3) Np would be alloyed with Zr to increase melting point: addition of 20 wtX Zr would increase it from 640'C to about 900'C (Fig. 4.13.1).

(4) Np-Zr alloys would dissolve Pu, which might be added as a driver or which is anyway generated from Np by nuclear reaction, with little effect on its solidus (Fig. 4.13.2).

The first point will have serious effects on both nuclear design and process concept. Alpha decay heat can be prohibitively large in Am-Cm alloys undiluted with Np: only a small pin-bundle without duct could be assembled at best. Hence, experiments will have to be made first on the phase diagram of Np-Am system. At present, the investigation of metallic TRU core is based on the two-alloy system, namely, Am-Cm-Y and Np-Pu-Zr.

Another problem of TRU metal fuel is rather low thermal conductivities of its main constituents, Np and Am, compared with U, although high-temperature data are lacking. If the difference in room-temperature conductivities between the uranium and TRU alloys is maintained to high temperature, the linear heat rating of the latter alloy fuel will be restricted.

As an alternative concept which potentially maximize burnup rate, particle-bed fuel was proposed which will be directly cooled by helium. If one can use TRU nitride microspheres coated with refractory material such as TiN, one may adopt cold-fuel concept. Fuel temperature can be kept lower than one third of its melting point to reduce mass transport within fuel matrix. Reduced mass transport will result in smaller swelling and internal gas pressure; thickness of the coating layer can be minimized to give large heavy metal density in the core. The coating is composed of two layers, that is, porous inner layer to absorb fission-fragment damage and outer dense layer to be a gas seal. The concept, however, may further penalize fabrication and reprocessing. In addition, nitride fuel generally

14 14 has the problems associated with N/ C. Once fabricated, however, the particle fuel may be a versatile option to give some freedom in handling and transportation.

- 1 0 6 -

]AERI-M 88・221

(2) Melting point <solidus) oi Am and C鳳 would be errectively increased by alloying with Y with a little sacrifice of neutron econo聞y.

(3) Np would be alloyed with Zr to increase melt ing po int: add it ion of 20 wtX Zr would increase it fro鯛 640・Cto about 900・C<Fig. 4.13.P.

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The first point wil 1 have serious effects on both nuclear d~sign and process concept. Alpha decay heat can be prohibitively large in Am-Cm alloys undiluted with Np: only a s圃all pin-bundle without duct cou ld be assubled at best. Hencc， exper lments w i 11 have to be made first on the phase diagram of Np-Am system. At present， the investigation of metallic TRU core is based on the two-alloy system. namely. Am-Cm-Y and Np-Pu-Zr.

Another problem of TRU metal fuel is rather low thermal conductivities of its main constituents. Np and Am， compared with U. although high-temperature data are lacking. If the difference in room-temperature conduct iv i t ies between the uran ium and TRU a 11 oys is maintained to high temperature. the linear heat rating of the latter alloy fuel will be restricted.

As an alternative concept which potentially maximize burnup rate， particle-bed fuel was proposed which will be directly cooled by helium. If one can use TRU nitride 聞icrospheres coated with refractory material such as T1N， one may adopt cold-fuel concept. Fuel temperature can be kept lower than one third of its 聞eltlngpoint to reduce mass transport withln fuel matrix. Reduced mass transport will result in s剛allerswelling and internal gas pressure; thickness of the coating layer can be 冊lnimlzed to glve large heavy metal dens i ty in the core. The coat ing is cOlllposed of two layers， that is， porous inner layer to absorb flssion-fragment damage and outer dense layer to be a gas seal. The concept. however， lIIay further penal ize fabrication and reprocessing. In addition， nitride fuel generally

1'1 ‘ has the problems associated with N/ C. Once fabricated， however， the particle fuel lIIay be a versatile option to give so聞ef reedolll in handling and transportation.

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600 10 Zr

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0 10 20 30 50 100 wt 9/.

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4‘13. 1 Fig.

J A E R I - M 88-221

4.14 Effect of Nuclear Data Uncertainty on Nuclear Characteristics in TRU-3umer Fast Reactor

H. Takano

Concept of TRU-burner fas t r e a c t o r w i t h m e t a l - a l l o y fue l s has been

s t u d i e d to a t t a i n h ighe r t r a n s m u t a t i o n e f f i c i e n c y than MOX-fueled

LMFBR as d e s c r i b e d in S e c t i o n 4 . 1 2 . The TRU-fuel c o n s i s t s of Np-237,

A m - 2 4 1 , Am-243, Cm-242 and Cm-244 which a r e produced by burnup of

33 GWd/1 in c o m m e r c i a l l i g h t w a t e r r e a c t o r s . Two kind of Np-20wtXZr

and Am-Cm-5wtXY fuel a l l o y s were c o n s i d e r e d . The fuel c o n f i g u r a t i o n

for the p e l l e t d i a m e t e r of 4 mm and the p in p i t c h of 6 mm was s e l e c t e d

from p a r a m e t e r s t u d i e s for n u c l e a r and t h e r m a l h y d r a u l ic

c h a r a c t e r i s t i c s . The b u n d l e - p i n type r e a c t o r c o n s i s t s of inner core

w i t h Am-Cm-5wtXY and o u t e r core w i t h Np-20wtXZr. The core s i z e i s

the d i a m e t e r of 120 cm and the h e i g h t of 60 cm.

N u c l e a r d a t a u n c e r t a i n t i e s for m ino r a c t i n i d e s such as Np, Am

and Cm a r e c o n s i d e r a b l y l a r g e r than those for m a j o r a c t i n i d e s such as

U-235 , U-238 and Pu-239 . As some t y p i c a l e x a m p l e s , F i g s . 4 . 1 4 . 1 , -2

and - 3 show the compar i son of Of, ain and v o b t a i n e d from th ree e v a l u a t e d

n u c l e a r d a t a f i l e s of JENDL-2. ENDF/B-IV and - V . From these f i g u r e s ,

l a rge d i s c r e p a n c i e s a r e observed among these d a t a f i l e s . These group

c o n s t a n t s were c a l c u l a t e d w i t h the TIMS-PGG d a t a p r o c e s s i n g c o d e " .

N e u t r o n i c s c a l c u l a t i o n s were pe r fo rmed on the b a s i s of

t w o - d i m e n s i o n a l d i f f u s i o n t h e o r y . The e f f e c t i v e and i n f i n i t e m u l

t i p l i c a t i o n f a c t o r s c a l c u l a t e d w i t h the t h r e e d i f f e r e n t n u c l e a r d a t a

a r e compared in Table 4 . 1 4 . 1 . The k.// o b t a i n e d w i t h the ENDF/B- V

or - I V d a t a d i f f e r c o n s i d e r a b l y by 7 and 10 X from t h a t for the JENDL-2

d a t a . F u r t h e r m o r e , i t i s observed from the c o m p a r i s o n of k - i n f i n i t y

o b t a i n e d for Am-Cm-Y and Np-Zr fuel co re s t h a t t h e r e a r e very b ig

d i f f e r e n c e s among these d a t a . Th i s l a rge d i f f e r e n c e in m u l t i p l i c a t i o n

f a c t o r s i s caused m a i n l y by the d i s c r e p a n c i e s of y - v a l u e s and f i s s i o n

c r o s s s e c t i o n s w i t h t h r e s h o l d . Hence , a c c u r a t e e v a l u a t i o n s and

e x p e r i m e n t s for n u c l e a r d a t a of minor a c t i n i d e s a r e s t r o n g l y r e q u i r e d

for more r e l i a b l e n u c l e a r d e s i g n s tudy for TRU-burner f a s t r e a c t o r .

R e f e r e n c e

1) Takano H . et a l . : " A Code Sys t em for Produc ing Group C o n s t a n t s in

F a s t N e u t r o n Energy Reg ion" , JAERI-M 82-072 ( 1 9 8 2 ) .

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JAERI-M 88・221

4.14 Effect of Nuclear Data Uncertainty on Nuclear

Characteristics in TRU-Burner Fast Reactor

H. Takano

Concept of TRU-burner fast reactor with metal-alloy fuels has been

studied to attain higher transmutation efficiency than MOX-fu日led

LMFBR as descr i b自d in Section 4.12. The TR.u-fuel consists of Np-237，

Am-241， Am-243， Cm司 242 and Cm-244 which are produced by burnup of

33 GWd/t in commercial light wat骨rreactors. Two kind of Np-20wt~Zr

and Am-Cm-5wt~Y fuel alloY5 were considered. The fuel configuration

for the pellet diameter of 4 mm and the pin pitch of 6 mm was selected

from paramet自r studies for nuclear and thermalhydraul ic

characteristics. The bundle-pin type reactor consists of inner core

wi th Am一Cm-5wt%Yand outer core wi th Np-20wt%Zr. The core s i ze i 5

the diamet自rof 120 cm and the h申ightof 60 cm.

Nuclear data un~ertaìnties for minor actinides such as Np， Am

and Cm are considerably lar・ger th且n those for major actinides such as

U-235， U-238 and Pu-239. As some typical examples， Figs.4.14.1. -，2

and -3 show the comparison of a，・ 6‘1¥ and JI obtained from three evaluated

nuclear data files of JENDL-?'， ENDF/B-IV and -V. From these figures，

large discrepancies are observed among these data files. Th自segroup

constants were calculated with th自 TIMS-PGGdata processing codell.

Neutronics calculations were performed on the basis of

two-dimensional diffusion theory. The effective and infinite mul-

tiplication factors calculated with the three different nuclear data

are compared in Tabl e 4.14.1. The k.1I obtained wi th the ENDF/B- V

or -IV data differ considerably by 7 and 10 " from that for the JENDL-2

data. Furthermore. it is observed from the comparison oC k-inCinity

obtained for Am-Cm-Y 且nd Np-Zr fue 1 cores that there are very big

differences among these data. This large difference in multiplication

Cactors is caused mainly by the discrepancies of JI-values and Cission

cross sect:ons wi th threshold. Hence. accurate eva.luations and

experiments for nuclear data of minor actinides are strongly required

for more reliable nuclear design study for TRU-burner fast reactor.

Reference

1) Takano H. et al.:"A Code System Cor Producing Group Constants in

Fast Neutron Energy Region". JAERI-M 82-072 (1982).

JAERl-M 88-221

Energy (eV)

Fig .4 .14 .1 Comparison if f i s s i o n i-rn s sect ions of Am-241

F ig .4 .11 .2 Comparison of inelast ic scattering cross sections of

of Np-237

-109 -

jAERI-M 88・221

: 0.11...

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Fig...14.Z Comparison of inclastic scattering cross sections or

of Kp-237

-1凶一

Fnerqy

J A E R I - M 88-221

3.50 * - • - » • - f - * - « • • - • • • • - • - • ^

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l i l t m i l

EN0FB5 JENDL2

t i i i t i i i ii i i i i t i t i 4-4-4-4-H-H- 4' 1

• » " " 1 ,..,..; i 10° 10s 1 0 2 1 03 IO 4

Energy (eV)

F i g . 4 . 1 4 . 3 Comparison of v - / a t u e s for Cm-244

10* 10'

Tab le 4 . 1 4 . 1 Compar i son of m u l t i p l i c a t i o n f a c t o r s

n u c l e a r k -e f f k - i n f i n i t y

da t a Am-Cm-5wtXY Np-20wtXZr

JENDL-2 1.069 1.465

ENDF/B-V 0.998 1.319

ENDF/B-IV 0.971 1.274

-110 -

jAERI-M 88・221

5.00 [...........;..

一印DFB5一+ー JENDL2

μ・・・・・・・4・・・・4・H 叫“"““件"““却炉.......H・・・

4ー+・噌ー+ー+・ー

I~ I~ I~ I~ I~

3.00 10'1

4. 。=一司、，l

Fig.4.14.3 Comparison of 11皿 laluesfor Cm-244

Table 4.14.1 Comparison of multiplication factors

k-inf in i ty k-eff nuclear

Np-20wt"Zr Am一Cm-5wt"Ydata

1.155 1.465 1.069 JENDL-2

1. 207

1. 237

1. 319

1. 274

ENDF/B-IV 0.971

ENDF/B-V

J A E R 1 - M 8 8 - 2 2 1

A.15 High Breeding / High Burnup Fast Reactor — Proposal of FP Gas Purge / Tube-in-Shell Metallic Fuel Fast Reactor

T. Hiraoka, K. Sako, H. Takano and M. Osakabe

Most essential feature of fast reactors is plutonium production especially in the country with poor energy resources as Japan. However, after transfer to mixed oxide fuel from metallic, fast reactor sacrificed always breeding capability through its development history. The construction cost of FBR is nearly two times of LWR's and breeding ratio is around 1.2. The doubling time is 30-40 years which is nearly equivalent to a reactor life. Therefore, produced F'.utonium must be quite expensive.

Recently developed metallic fuel, which was once abandoned, recalls possibility to generate hard neutron spectrum which gives high breeding ratio. A new concept of the FP gas purge/tube-in-shell type metallic fuel assembly is proposed here to achieve high plutonium breeding capability.

The concenpt is schematically illustrated in Fig. 4.15.1 and the cross section of the fuel assembly at its center is shown in Fig. 4.15.2.

Each of cooling tube penetrates centrally-holed hexagonal metallic fuel pellets triangularly packed in a closed hexagonal shell. The shell is filled with stagnant sodium to fill gaps between adjacent fuel pellets and cooling tubes which allow expansion of fuel due to swelling. Filling sodium makes thermal conduction in the shell better and traps the most of the fission products other than inert gas which is released into primary sodium coolant through the gas purge section at the top of the assembly or lead to the external FP gas treatment facility by a guide tube.

Axial thermal expansion of cooling tubes is absorbed by bellows at bottom end of each cooling tube or the sliding tube structure at the top. Fuel is DU-Pu-10%Zr alloy as developed by ANL. Structural material is SUS-316(in future, probably HT-9).

Major dimensions and compositions of the core are summarized in Table 4.15.1. The volume fraction of fuel in the core region is beyond 50 % and the total sodium fraction can be suppressed to below 40 %.

The FP gas purge/tube-in-shell type metallic fuel assembly has advantages as follows: 1. High fuel fraction and low sodium fraction in the core provides

very hard neutron spectrum and therefore, high breeding ratio. 2. Initial excess reactivity is unnecessary. 3. FP gas purging shortens the length of the fuel assembly by

eliminating the long FP gas plenum as in the conventional pin type assembly.

4. Burnup of fuel is not limited by high FP gas pressure.

]AERI-M 88・221

4.15 High Breeding I High Burnup Fast Reac.tor

一-Proposal of FP Gas Purge / Tube-in-Shell Metallic Fuel Fast Reactor

T. H1raoka， K. Sako， H. Takano and H. Osakabe

Host essential feature of fast reactors 1s pluton1um product1on espec1ally 1n the country w1th poor energy resources as Japan. However， after transfer to m1xed ox1de fuel from metallic， fast reactor sacr1f1ced always breed1ng capab111ty through 1ts development h1story. The construct1on cost of FBR is noarly two times of LWR's and breeding ratio is around 1.2. The doubllng time is 30-40 years wh1ch 1s nearly equ1valent to a reacto!' life. Therefore， produced ~'.utonium must be quite expens1ve.

Recently developed metallic fuel， which was once abandoned， recalls possibi1ity to generate hard neutron spectrum which gives high breed1ng rat10. A new concept of the FP gas purge/tube-in-shell type metal11c fuel assembly is proposed here to achieve high plutonium breed1ng capab11ity.

The concenpt 1s schematically illustrated in Fig. 4.二5.1 and the cross section of the fuel assembly at 1ts center is shown in Fig. 4.15.2.

Each of cooling tube penetrates centrally-holed hexagonal metallic fuel pellets triangularly packed in a closed hexagonal shell. The shell is filled with stagnant sodium to fill gaps between adjacent fuel pellets and cooling tubes wh1ch allow expans10n of fuel due to swel11ng. Fi1ling sodium makes thermal conduction 1n the shell better and traps the most of the f 1ss10n products other than 1nert gas wh1ch 1s released 1nto pr1mary sod1um coolant through the gas purge sect10n at the top of the assembly or lead to the external FP gas treatment faci11ty by a gu1de tube.

Ax1al thermal expans10n of cooling tubes 1s absorbed by bellows at bottom end of each coo11ng tube or the sliding tube structure at the top. Fuel 1s DU-Pu-10%Zr alloy as developed by ANL. Structural mater1al 1s SUS-316(in future， probably HT-9).

Hajor d1mensions and compos1tions of the core are summarized 1n Table 4.15.1. The volume fraction of fuel in the core region is beyond 50 % and the total sodium fraction can be f;11ppressed to below 40も.

The FP gas purge/tube-in-shell type metallic fuel assembly has advantages as follows: 1. High fuel fraction and low sodium fraction 1n the core provides

very hard neutron spectrum and therefore， high breeding ratio. 2. Initial excess reactivity is unnecessary. 3. FP gas purg1ng shortens the length of the fuel assembly by

eliminating the long FP gas plenum as in the conventional pin type assembly.

4. Burnup of fuel 1s not l1mited by high FP gas pressure.

-111ー

JAERI-M 88-221

5. The closed shell structure of the fuel assembly makes it easy to deal with venting of FP gas.

6. Good neutron economy can be obtained because of relatively thin cooling tube. Safety aspect of FP gas purging should be considered. Strictly

speaking, the FP gas purging breaks the innermost barrier of FP containment. However, cooling tubes are not high pressure boundaries as in the case of conventional pin-bundle type assembly. Issue of creeping due to inner high pressure disappears by FP gas purging.

A standard 700 MWe tank-type plutonium fuel fast breeder is assessed as summarized in Table 4.15.2.

By a two-dimensional diffusion code CITATION with 70 tjroup JFS-3-J2 cross section set, criticality was analized about the homogenous core of 1.3 m in radius and 1.0 m in height. Blanket thickness is 50 cm. The required plutonium enrichment is 8.7 % (7.8 % in metallic fuel). The neutron spectrum is very hard and the initial breeding ratio is 1.90.

Burning characteristics was analyzed by code PHENIX. The amount of 400 kg of fissionable plutonium, Pu-239 and Pu-241, can be produced by 300 day continuous burning a year. The doubling time is 6.1 years.

The neutron spectrum is so hard that breeding of plutonium is possible within realistic time period, even with the initial core of low enrichment uranium of 11 %. Even after loading of generatd plutonium to the initial uranium core to compensate uranium burning, accumulation of extra plutonium will reach the necessary amount of a fresh plutonium core within about ten years. If fresh low enrichment uranium is continuously fed to the uranium core, the generaton time for another plutonium core is only 4.4 years. Thus, plutonium breeding can be initiated from a low enrichment uranium core.

Table 4.15.1 Dimension of Component of Core(mm)

Unit Cell Lateral Distance 11.34 Inner Lateral Distance Inner Radius of Coolint Tube 2.75 of Shell 151.69 Coolinc Tube Thickness 0.25 Thickness of Shell 3.00 Opening Radius of Fuel Pellet 3.40 Thickness of Pud Part 5.00 Lateral Distance of Fuell Pellet 11.00 Outer Lateral Distance

of Fuel Cell Assembly 150.63

JAERI-M 88・221

5. The closed shell structure of the fuel assembly makes it easy to deal with venting of FP gas.

6. Good neutron economy can be obtained because of relatively thin cooling tube. Safety aspect of FP gas purging should be considered. Strictly

speaking， the FP gas purging breaks the innermost barrier of FP containment. However， cooling tubes are not high pressure boundaries as 1n the case of conventional pin-bundle type assembly. Issue of creeping due to inner high pressure disappears by FP gas purging.

A standard 700 MWe tank-type pluton1um fuel fast breeder 1s assessed as summarized 1n Table 4内 15.2.

By a two-dlmens1onal d1ffuslon code CITATION wlth 70 匂roupJFS-3-J2 cross section set， criticallty was analized about the homogenous core of 1.3 m in radius and 1.0 m in height. Blanket th1ckness is 50 cm. The required plutonium enrichment is 8.7 亀

(7.8 亀 1nmetal11c fuel). The neutron spectrum 1s very hard and the initial breeding rat10 is 1.90.

Burning characterist1cs was analyzed by code PHENIX. The amount of 400 kg of fissionable plutonium， Pu-239 and Pu-・241， can be produced by 300 day continuous burning a year. The doubling time i5 6.1 year5.

The neutron 5pectrum i5 50 hard that breeding of plutonium i5 possible within rea11stic time period， even with the initial core of low enrichment uranium of 11 も Even after loading of generatd plutonium to the initial uranium core to compensate uranium burning， accumulation of extra plutonium will reach the necessary amount of a fresh plutonium core w1thin about ten years. If fresh low enrichment uran1um 1s continuously fed to the uranium core， the generaton time for another plutonium core 1s only 4.4 years. Thus， plutonium breeding can be initiated from a low enrichment uranium core.

Table 4.15.1

Dimension of Component of Core(mm)

Unit Cell L~t.ral Dist~nee 11.3~ Inner R~dius of Coolin， rube 2.75 Coolin， Tube Thickness 0.~5 目penin，R~dius of Futl Pellet 3..0 L~ter~l Distance of Fuell P，ll，t 11.00

Inner L~ter&l Distlnc・of Sh.ll Tbickness of Shell Thickness of Pud P~rt Outer Lateral Distance of Fuel Cell Asse.bly

151.59 3.00 5.0。

150.69

J A E R I - M 8 8 - 2 2 1

Table 4.15.2 Standard Plutonium Reactor

Core Radius 1.3 D Core Height 1.0 a Blanket Thickness 50 co Fuel Composition, X

Pu 7.8 011(0.250 82.2 Zr 10.0

Plutoniua Enrichment 8.7 Z No. of Fuel Unit Cel 1 in a assembly 169 Ha. of Assemblies 234 Fuel Loading, kg

Fuel Ketal 43,328 Pu, DU 39,028 Pu 3,416 Pu-239.241 2,455

1.032 1.90 40S

keff Initial Breeding Ratio Production of Pu-239,-241 per year(300 day burning), kg Doubling Time, year (300 day burning per year) Paver

Average Linear Power Density 470 W/ea Average Power Density 350 KW/JZ Thermal Power 1,880.X«to Electric Pover 725 HWe

fru surfaea of primary eoolin? sodium

uppar axial blaokaO

louar axial blaaxat

antranca nozzla

F? gas purja nachaalin

„~ pad

«cP sas planum

coolln? tuba faul pallac

. f i l l i n g aodlua

fual/blaok.t alamant

• coalla? tuba

.coollr.9 sodium

pad Fig. 4.15.2 Cross section of fuel assembly

Fig. 4.15.1 Concept of FP gas purge/ tube-in-shell type metallic fuel assembly

Cor哩 RadiusCore Hei，ht Blanket !hickness

]AERI-1¥I 88-221

Table 4.15.2

S七audardPlutonium Reactor

1.3固

1. 0 m SO Cl:l

keff Initial 8reedinc Ratio Production af Pu・239，-241

1.032 1. 90

Fuel Composition， % Pu 7.8

per Year(300 day burninc)， kc Doublinc !illle， year

DU(0.2%) 82.2 Zr lO.O

PlutロniulllEnrich白色nt 8.7 % lIo. 日fFuel Unit Cell in a a.sse:ubly No.ロfAssublies Fuel Loadinc， k， Fuel l!eta.l Pu， DU PU Pu-239，241

tro ・ su::!'~・c・ of-

Fi' 9'，1，3

pr1m~rY <:Qoling sodlum

169 234

<¥3，328 39，028 3，416 2，455

(300 day burnin， per year) Pa官官r

Aver畠，eLinear Paller Dens ity Avera，e POller Density !bernl PQII哩rElectric POller

n su P<l.rS・"・c:h...a..Lsna

iP ，，，:1 pl・nwa

f1111n!l $od 1弘司

‘'1'¥1・z‘><1‘lbl&D.k.t

.h・11~Ube

:u・l/l>laa.l<・z・1・m・n.c:ort'

coal1.dfJ eube

巴。oIL"':9' :lc41wn

low.r ax1~1 blank・z

470 'rI/c:! 350 KY.{ Q

1，860 !!'Ilt!I 725 l!'i/e

p~4 Fig. 4.15.2 Cross section of fuel assem;，ly

en巳ranc:.no:zl.

Fig. 4.15.1 Concept of FP gas purge/ tube-in-shell type metallic fuel assell、bly

JAERI-M 88-221

4.16 Study of Nuclear Characteristics of Nitride- and Carbide-Fueled LMFBRs S. Iijima, H. Yoshida and H. Takano

Nuclear characteristics of nitride- and carbide-fueled liquid metal fast breeder reactors have been investigated as part of the study for an innovative fast breeder reactor. The potential of these advanced fuels lies mainly in the higher fissile atom density and better thermal conductivity than the oxide fuel. These properties allow for a higher linear power, higher breeding ratio and a lower fuel inventory, thus reducing doubling time for fast reactor cores and improving the utilization of fuel resources. The material properties for nitride, carbide and oxide fuels are compared in Table A.16.1.

Following reactor parameters were calculated in r z core models of nitride- and carbide- fueled LMFBRs: (1) a breeding ratio, (2) a burn-up reactivity loss and (3) reactivity coefficients for a sodium void worth and a Doppler reactivity worth. Selection of calculation models was made based on two criteria: (1) k «-f=1.025 at the end of an equilibrium cycle(550 EFPD; equivalent full power d;iy) and (2) maximum powers of the inner and the outer core are kept at the same level by adjusting fissile enrichments at the beginning ol life. The core volumes are the same in both nitride- and carbide-fueled cores. The rz model is shown in Fig. A.16.1.

The core calculations were made by the CITATION-FBR code and the burn-up calculations by the PHENIX code. The results for breeding ratios and burn-up reactivity loss are given in Table 4.16.2. High conversion ratios in the core regions of these advanced fuel LMFBRs allow for the design of a long life core. The reactivity coefficients are given in Table 4.16.3. The sodium void worths and the Doppler worths are about the same in the nitride- and the carbide- fueled LMFBRs.

The reactor physics parameters of the nitride- and carbide-fueled LMFBRs were investigated. In these calculations it was made clear that nitride- and carbide-fueled LMFBRs showed high breeding performance and had attractive potential in development of advanced fast breeder reactors.

JAERI-M 88・221

4.16 Study of Nuclear Characteristics of Nitride-and

Carbide-Fueled LMFBRs

S. Iijima. H. Yoshida and H. Takano

Nuclear characteristics of nitride- and corbide-fueled liquid

metal fast breeder reactors have been investigated as part of lhe

study for an innovative fast breeder reactor. The potential of

these advanced fuels lies mainly in the higher fissile atom density

and better thermal conductivity than the oxide fuel. Thesl'

properties allow for a higher linear power. higher breeding ralio

and a lower fuel invenlory. lhllS reducing dOllbling lime for fast

reactor cores and improving the utilizatlon of fuel resources. The

material properties for nitride. carbide and oxide fuels are 仁om-

pared in Table 4.16.1.

Following reactor parameters were calculated in rz core models

of nitride- and carbide- fueled LMFBRs: (1) a breeding ralio. (2) a

burn-up reactivity loss and (3) reactivity coefficients for a sodium

void worth and a Doppler reactivity worth. Selection of calculalion

lTJodels was made based on two criteria: (1) keff=1.02S at the end of

an equilibrium cycleCSSO EFPD; equivalenl full power dllY) atld

(2) maximum powers of lhe inner and the ouler core are kepl al Ihe

same level by adjusting fissile enrichmenls al the beginning 01

life. The core volumes are the same jn bolh nitrjdp- and carbide-

fueled cores. The rz model is shown in Fig. 4.16.1.

The core calculalions were made by the CITATION目 FBR code and

the burn-up calculations by the PHENIX code. The results for breed-

ing ratios and burn-up reactivity loss are given in Table 4.16.2.

High conversion ratios in the core regions of thpse advanιpd f ue 1

LMFBRs allow for the design of a long 1 ife core. The rea ιt i v i l Y

coeffici巴nts are given in Table 4.16.3. The sodium void worths and

the Doppler worths are about the same in the nitride- and thc

carbide- fueled LMFBRs.

The reactor physics parameters of the nitride- and carbide-

fueled LMFBRs were investigated. 1n these ca]culations it was made

clear that nitride- and carbide-fueled LMFBRs showed high breeding

pprformance and had attractive potential in development of advanced

fast breeder reactors.

JAERI-M 88-221

Table 4.16.1 Material property of advanced fuels

Nitride Carbide Oxide

Melting point(°C) ~2800 Density (g/cm 3) 14.30 Fissile material density (g/cm ) 13.52 Thermal conductivity (W/m.K) (at 773K) 12 (at 1773K) 16

.~2300 13.60

2 7 50 11.08

4.0 2.3

Table 4.16.2 Result for principal physics parameters

Nitride Carbide

k e f f : B0L E0EC

Burn-up reactivity loss(%Ak/k) Breeding ratio Total Core Blanket

1.0432 1.0377 1.0250 1.0250 1.77 1.22

1.261 1.360 0.871 0.941 0.390 0.419

B0L : a beginning of life, E0EC : an end of equilibrium cycle

Table 4.16.3 Results for reactivity coefficients

Nitride Carbide

Sodium void worth (IAk/k) BOL 1.587 1.864 EOEC 2.352 2.632

Doppler reactivity worth (%Ak/k) ( a )

BOL -1.066 -1.108 EOEC -0.987 -1.019

(a) 300 K to 900 K

JAER[-M 88・221

Table 4.16.1 Material property of advanced fuels

Nitride Ca了bide Oxide

『ーーーー『ーーーーーーー--ーーーーーーーーーーーーーーーーーーーーーーーーーーーーーーーーーーーーーーー

Melting point(OC) ，"，，2800 白.....2300 2750

Density (g/cm3) 14.30 13.60 11.08

Fissile material

density (g/cm3) 13.52 12.96 9.75

Thermal conductivity (W/m.K)

(at 773K) 12 16 4.0

(at 1773K) 16 1 7 2.3

Table 4.16.2 Result for principal physics parameters

Nitride Carbide

eff: BOL 骨

1.0432 1.0377

EOEC 1.0250 1.0250

Burn-up reactivity loss(%A k/k) 1. 77 1. 22

Breeding ratio

Total 1 .261 1.360

Core 0.871 0.941

Blanket 0.390 0.419

発 BOL a beginning of life， EOEC an end of equilibrium cycle

Table 4.16.3 Results for reactivity coefficients

Nitride Carbide

Sodium void worth (% .a..k/k)

BOL 1.587 1.864

EOEC 2.352 2.632

Doppler reactivity

worth (%Ak/k)(a)

BOL -1.066 -1.108

EOEC ー0.987 -1.019

(a) 300 K to 900 K

AXIAL REFLECTOR

AXIAL BLANKET

50 I INNER CORE

Clad dia. : 9.40mm Clad thick: 0,38mm P/D : 1.178 Smear Den.: 80%TD

OUTER CORE

AXIAL REFLECTOR

RADIAL BLANKET

Fuel : 43.6 v/0 Na : 38.5 v/0 SUS : 17.9 v/o

Fuel Na SUS

55.4v/o 28.3v/o 16.3v/o

RADIAL REFLECTOR >

Clad dia:12.0mm P/D :1.07 Smear D :90%TD

- R 106 138 130

Fig.4.16.1 R-Z reactor model (unit

210 cm)

」〉何見【

∞∞'NN-

AXIAL AXIAL REFLECTOR REFLECTOR

AXIAL BLANKET RADIAL RADIAL

BLANKET REFLECTOR

INNER CORE OUTER Fuel 55.4v/o

CORE Na 28.3v/o SUS 16.3v/o

Clad dia. 9.40mm Fuel 43.6 v/o

Clad thick: 0.38mm Na 38.5 v/o Clad di口:12. Olffil P/D 1.178 sus 17.9 v/o P/D : 1.07 Smear Den.: 80%TD Smear D :90%TD

50 l-】由

130 210 (unit cm)

138 106 。。

R-Z reactor model Fig.4.16.1

J A E R I - M 8 8 - 2 2 1

5. Fusion Neutronics

The research activities covered under this title during the term are categorized in three subjects: integral experiments with analyses, cross section measurements and developmental plans for the future.

In the first subject, a steady progress has been made on the blanket engineering benchmark experiments of JAERI-USDOE collaboration: a) the phase 2A experiments, the first series in the large-scale cesc blanket system of a closed geometry, were analysed by a code, MORSE-DD. b) a new series of experiments, Phase 2B, was conducted with a Be liner inside the system. A combination of the results obtained in these two series provides valuable data on the performance of Be neutron multiplier to the tritium production in lithium oxide solid blanket concepts. In addition, the effect of Be was examined in a benchmark experiment of simple geometry where a Be layer was sandwiched between two Li 20 regions. Nuclear heating is an important neutronics parameter in the integral experiment. A series of experiment on gamma-ray heating was performed using thermo-luminescence dosimeters. The comparison with the calculations- revealed deficiencies in the currently available secondary gamma-ray production libraries.

As for the second subject, systematics for activation cross section of various reaction types were formulated based on the 140 reactions measured so far at FNS in a unified fashion, providing a better quality systematics compared with the precedent ones. The cross sections of Al, Nb and Zr around 14 MeV were determined anew in absolute basis. The results acertained the reported values of the activation cross sections, mentioned above, which had been determined taking the cross section of 2 7Al(n,a) 2 1 ,Na reaction in the ENDF/B-V dosimetry file as the standard. An experimental equipment was installed for a new series of cross section measurement on secondary gamma-ray production. A research work started to determine the neutron source characteristics of the fusion devices, taking the system of the JAERI-USDOE experiment as an example.

On the third subject, a plan was proposed to undertake integral neutronic experiments for and of the blanket test module of next fusion device. An investigation was also made on the feasibility of the concept and its usage of very high intensity 14MeV neutron source of 101!*n/s for the use of fusion nulear testing. (Tomoo Nakamura)

]AERI-M 88-221

The research activities covered under this title during the term

are categorized in three subjects: integral experiments with analyses，

cross section measurements and developmental plans for the future.

In the first subject， a steady progress has been made on the

blanket eng1neering benchmark exper1ments of JAERI欄 USDOEcollabora-

tion: a) the phase 2A exper1ments， the f1rst ser1es 1n the large-scale

cest blanket system of a closed geometry， were analysed by a code，

MORSE-DD. b) a new ser1es of exper1ments， Phase 2B， was conducted w1th

a Be liner 1ns1de the system. A combinat10n of the results obta1ned 1n

these two ser1es prov1des valuable data on the performance of Be neut-

ron mult1plier to the tritium production in lithium oxide solid blan-

ket concepts. In addition， the effect of Be was examined 1n a bench-

mark experiment of simple geometry where a Be layer was sandwiched

between two Li20 regions. Nuclear heating is an important neutronics

parameter in thE! integral experiment. A series of experiment on gamma-

ray heating was performed using thermo-luminescence dosimeters. The

comparison with the calculations revealed deficiencies in the

currently availa，ble secondary gamma-ray production libraries.

As for the second subject， systematics for activation cross

section of variclus reaction types were formulated based on the 140

reactions measuI'ed so far at FNS in a unified fashion， providing a

better quality systematics compared with the precedent ones. The cross

sections of Al， Nb and Zr around 14 MeV were determined anew in

absolute basis. The results acertained the reported values of the

activation cross sections， mentioned above， which had been determined

taking the cross section of 2?Al(n，α)2 句Nareaction in the ENDFjB-V

dosimetry file as the standard. An experimental equipment was

installed for a new series of cross section measurement on secondary

gamma-ray production. A research work started to determine the neutron

source characteristics of the fusion devices， taking the system of the

JAERI-USDOE expeI'iment as an example.

On the third subject， a plan was proposed to undertake integral

neutronic experiments for and of the blanket test血oduleof next

fusion device. An investigation was also made on the feasibility of

the concepL and its usage of very high intensity 14MeV neutron source

of 101'・njsfor the use of fusion nulear testing. (Tomoo Nakamura)

JAERI-M 88-221

5.1 Phase-IIB Experiment of JAERI/USDOE Collaborative Program on Fusion Blanket Neutronics

Yukio Oyaraa, Seiya Yamaguchi, Koichi Tsuda, Yujiro Ikeda, Chikara Konno, Hiroshi Maekawa, Tomoo Nakamura, Karl G. Porges ,

* * Edgar F. Bennett and Richard F. Mattas

A neutronics test for beryllium neutron multiplier and lithiuir-oxide breeder blanket model has been carried out at the Fusion Neutronics Source (FNS) facility in the Japan Atomic Energy Research Institute, as the JAERI/USDOE collaborative program on fusion blanket neutronics.1 The program aims at providing experimental data on neutronics parameters, e.g., tritium breeding and at examining those prediction accuracies. The present experiment, Phase-IIB, is one of the Phase-II experimental series in " closed geometry," i.e., the DT neutron source and the test blanket are enclosed by lithium carbonate ( Li 2C0 3 ) and polyethylene ( PE ) layers. The Li 2C0 3 region works as a reflector to simulate a full-covered blanket and the PE as the neutronic isolation zone against room-returned neutrons. The Phase-IIB arrangement simulates a full-covered blanket with front beryllium zone configuration to be compared with the Phase-IIA arrangement with the partial-coverage by beryllium only on the test blanket.

Figure 1 shows the experimental arrangement where the rotating neutron target is contained in a cavity. A 5 mm-thick stainless steel, simulating the first wall, and 50 mm-thick beryllium layer were newly added to the Phase-IIA system. The measurements were performed in configurations with and without a front beryllium region of 50 mm-thick, and with and without a first wall steel of 5 nun-thick for the test blanket surface. The measured parameters were:l) tritium production rate (TPR) distributions by the on-line techniques of Li-glass and NE213 scintillators, liquid scintillation counting method with Li-metal foils, Li 20 pellets and Li 20 blocks, 2) in-assembly neutron spectra by an NE213 detector and a. recoil-proton gas proportional counter, and 3) reaction rates by several activation foils. These distributions were measured along the central axis using the experimental drawer installed in the test blanket. The neutron

* Argonne National Laboratory

JAERI-M 88・221

5.1 Phase-IIB Experiment of JAERI/USDOE C011aborative Program

on Fusion Blanket Neutronics

Yukio Oyama， Seiya Yamaguchi， Koichi Tsuda， Yujiro Ikeda，会

Chikara Konno， Hiroshi Maekawa， Tomoo Nakamura， Karl G. Porges ， ~ *

Edgar F. Bennett and Richard F. Mattas

A neutronics test for beryllium neutron multiplier and lithiurr-

oxide breeder blanket model has been carried out aC che Fusion

Neutronics Source (FNS) faci1ity in the Japan Atomic Energy Research

Institute， as the JAERI/USDOE collaborative program on fusion blanket

neucronics.1 The program aims at providing experimencal data on

neutronics parameters， e.g.， tritium breeding and at examining those

prediction accuracies. The present experiment， Phase申 EB，1s one of

the Phase-II experimental series i.n " c10sed geometry，" i.e.， the DT

neutron source and the Cest b1anket are enc10sed by 1ithium carbonate

( Li2C03 ) and po1yethy1ene ( PE ) layers. The Li2C03 region works as

a reflector to s1mulate a full-covered blanket and the PE as the

neutronic isolation zone against room-returned neutrons. The Phase-.IIB

arrangement s1mulates a full-covered blankec w1th front beryllium zone

configurat10n to be compared w1th the Phase-IIA arrangement wich che

partial-coverage by bery11ium on1y on the test b1anket.

F1gure 1 shows the experlmenta1 arrangement where the rotatlng

neutron target 1s conta1ned 1n a cavity. A 5 mm-thick stainless steel，

simu1ating the f1rst wa11， and 50 mm-thick bery11ium layer were new1y

added to the Phase-IIA system. The measurements were performed in

conf1gurations with and w1thout a front beryllium region of 50

mm-thick， and w1th and w1thout a first wa11 stee1 of 5 m血-th1ckfor

the test blanket surface. The measured parameters were:l) tritium

productlon rate (TPR) d1str1but10ns by the on-11ne techniques of

Li-g1as5 and NE213 scint土llators，liquid scin'c111ation counting method

with L1唱 metalfOi1s， Li20 pe11ets and Li20 blocks， 2) in-assembly

neutron spectra by an NE213 detector and a recoil-proton gas

proportiona1 counter， and 3) react:ion rates by several act:ivatian

foil~. These distributions were measured along the central axis using

the experimental drawer 1nstalled 1n the test blanket. The neutron

* Argonne National Laboratory

一118-

J A E R I - M 8 8 - 2 2 1

flux distribution on the inner surface of the source cavity was also measured by 9 3Nb(n,2n) 9 2 Nb and l 9 7Au(n,y) 1 9'Au reactions.

The measured results of Phase-IIB experiment provide the effect of neutron reflection at the beryllium layer on the neutronic parameters in the test blanket zone by comparison with the Phase-IIA results.2 Figure 2 shows the ratios of the tritium production rates of both experiments on the central axis of the test blanket zone. The effect of the inner beryllium layer increased the TPR of 7Li (T 7) by LO % over the whole test blanket and the TPR of 6Li (T 6) by a factor more than 2 in the front region of the test zone. The first wall effect of 5 mm-thick decreased the T 7 by 10 % over the whole region and the T s by 5 % at the rear region in the test blanket. From these results, the total tritium breeding ratio (TBR) is discussed for a full-covered blanket configurations in comparison with that of a partial-covered blanket.

The VS activities were supported by the United States Department of Energy, Office of Fusion Energy.

References

1) Nakamura T. and Abdou M. A.:"Summary of Recent Results from the JAERI/US Fusion Neutronics Phase I experiments," Proc. Int. Conf. 7th Topical Meeting on the Technology of Fusion Energy, Fusion Technol., 10., NO.3, Part 2A, 541 (1986).

2) Oyama Y., et al.:" Phase II Experimental Results of JAERI/USDOE Collaborative Program on Fusion Blanket Neutronics Experiment," submitted to the Proceedings of Int. Symp. on Fusion Nuclear Technology, Tokyo, April 11-15 (1988).

]AERI-M 88・221

f1ux distr1but10n on the 1nner surface of the source cav1ty was a1so

measured by 93Nb(n ， 2n)92~b and 19'Au(n，y)19SAu react10ns.

The measured results of Phase-IIB exper1ment prov1de the effect

of neutron reflection at the beryl11um layer on the neutronic

parameters in the test b1anket zone by compar1son with the Phase-IIA

results.2 F1gure 2 shows the ratios of the tr1t1um product10n rates of

both exper1ments on the central ax1s of the test blanket zone. The

effect of the 1nner beryl11um layer increased the TPR of 'Li (T7) by

10 % over the whole test blanket and the TPR of 6L1 (T6) by a factor

more than 2 1n the front reg10n of the test zone. The f1rst wall

effect of 5 mm-th1ck decreased the T， by 10 ~ over the whole reg10n

and the T6 by 5 % at the rear region 1n the test blanket. From these

resu1ts， the tota1 tr1t1um breeding rat10 (TBR) 1s d1scussed for a

ful1-covered blanket configurat10ns 1n compar1son with that of a

partial-covered blanket.

The US activities were supported by the United States Departmant

of Energy， Office of Fusion Energy.

References

1) Nakamura T. and Abdou M. A.: "Summary of Recent Results from the

JAERI/US Fus10n Neutronics Phase 1 experiments，" Proc. Int. Conf.

7th Topical Meeting on the Technology of Fusion Energy， Fusion

Technol.， 10， NO.3， Part 2A， 541 (1986).

2) Oyama Y.， et a1.: 11 Phase II Experimental Results of JAERI/USDOE

Collaborative Program on Fusion Blanket Neutronics Experiment，"

submitted to the Proceed1ngs of Int. Symp. on Fus10n Nuc1ear

Technology， Tokyo， April 11-15 (1988).

PHASE-IIB EXPERIMENT

\ \ \ \ \ \ \ \ \ \ \

o O r-< a:

-*- BEF WITH FW/PIBA-BEF --«>- BEF W/0 FW/PIGA-8EF - » - FIEF W/0 FW/HBA-BEF

- i — i i i i i i

•20 0 20 40 60 80 AXIAL DISTANCE (CM)

Pig. 5.1.1 Phase-IIB experimental arrangement. Pig. 5.1.2 Comparisons of tritium production rates of 7l.i between Phase-IIA and -11B result Squares shows inner beryllium effect.

』〉開

∞∞'NN-

PHASE-IIB EXPER1MENT

「・一--I!)--

使配

内ぽ

刷引の悼の肱

m刷川

CEunun

CEr『

HHnunu

-E''rf

uHH刊

庇庇佐

I~二己???:;

庄止トト

0一ト《庄

80 o 20 40 60

AXIAL 015TんへJCE(口11

5.1.2 Comparisons of trilium production rates I~ i且・Fig・5.1.1 Phase-IIB experi町~ntal arrangement・

uet¥.le邑n Phase-IlA and -IIs results.

Squares shoys inner beryllium effect.

of 71.i

J A E R I - M 8 8 - 2 2 1

5.2 Analysis of Neutronics Parameters Measured in Phase II Experiments of the JAERI/USDOE Collaborative Program

M. Nakagawa and T. Mori

Fusion blanket neutronics parameters measured in the phase 11 experiments have been analysed. Since the phase II experimental system simulates a fusion environment better than the system in phase I as described in Refs. 2) and 3), the analysis is useful to evaluate the present status of prediction uncertainty of tritium breeding ratio and other neutronics parameters in fusion blankets. The main items of measurements were the source characteristics, reaction rate distribution in the Li 90 test zone, tritium production rate and neutron spectrum.

In the analysis, the calculations have been performed by using a Monte Carlo code MORSE-DD, a discrete ordinate code DOT-DD and the group cross section set DDL/J3P1 which has been produced from JENDL3/PR1. In the calculation of activation rates, ENDF/B-IV and the FNS file which includes the activation cross sections measured at the FNS by Ikeda et a l . 4 >

For the source characterization, the activation rates measured at the inner surface of L i 2

C 0 3 container and at the front surface of the Li,0 test zone have been calculated. The result shows that the calculated values with ENDF/B-IV agree well with the measured values for Au(n,2n) and Al(n.a) but overestimate those by 10 - 20% for Ni(n.p) and Nb(n,2n), and underestimate those by about 10% for Ni(n,2n), while the FNS file can predict well the measured values for most reactions except for Ni(n,p) which is overestimated by several percent. An example of the map of C/E values is shown in Fig.5.2.1.

The C/E values for the reaction rates distributions in the test zone are summarized in Table 5.2.1. In the table are shown the C/E values for the systems with and without a beryllium multiplier. A similar trend is observed on the space dependence of the C/E values for threshold reactions. It shows that the C/E values decrease with increasing distance from the front surface of the Li,0 test zone. In addition, the C/E values in the system with beryllium are generally smaller than those in the reference system without beryllium.

With respect to the tritium production rate(TPR), the differences

JAERI-M 88-221

5.2 Analysis of Neutronics Parameters ~Ieasured in Phase II

Experよmentsof the JAERI/USDOE Collaborative program

M. Nakagawa and T. Mori

Fusion blanket neutronics parameters measured in the phase I1

experiments have been analysed. Since the phase [1 experimental

system simulates a fusion environment better than the system in phase 1 )

1.' as described in Refs. 2) and 3)， the analysis is useful to

evaluate the present status of prediction uncertainty of tritium

breeding ratio and other neutronics parameters in fusion blankets.

The main items of measurements were the source characteristics，

reaclion rate dislribution in the Li?O test zone， tritium production

rate and neutron spectrum.

In the analysis， the calculations have been performed by uSing a

河onteCarlo code MORSE-DD， a discrete ordinate code DOT-DD and the

group cross section set DDL/J3Pl which has been produced from

JENDL3IPRl. In the calculation of activation rates， ENDF/B-IV and the

FNS file which includes the activation cross sections measured at the 4)

FNS by Ikeda et al.

For the source characterization， the activation rates measured

at the inner surface of LiCO container and at the front surface of 2--3

the Li?O test zone have been calculated. The result shows that the

calculated values with ENDF/B-IV agree well with the measured values

for Au<n，2n) and Al(n，α) but overestimate those by 10 - 20~ for

Ni(n，p) and Nb(n，2n)， and underestimate those by about 10~ for

Ni<n，2n)， while the FNS file can predict well the measured values for

most reactions except for Ni(n，p) which is overestimated by several

percent. An example of the map of C/E values is shown in Fig.5.2.1.

The C/E values for the reaction rates distributions in the test

zone are summarized in Table 5.2.1. In the table are shown the C/E

values for the systems with and without a beryllium multiplier. A

similar trend is observed on the space dependence of the C/E values

for threshold reactions. It shows that the C/E values decrease with

increasing distance from the front surface of the Li?O test zone. In

addition， the C/E values in the system with beryllium are generally

smaller than those in the reference system without beryllium.

Wi th respect to the tri tium production rate(TPR)， the differences 主ー

一121-

J A E R i - M 8 8 - 2 2 1

With respect to the tritium production rate(TPR), the differences of several percent can be seen among the measuring methods. Figure 5.2.2 shows the C/E values of TPR where the TPR was measured by the zonal method. The agreement between the measured and calculated values is satisfactory in the range z= 0 - 30 cm. The integrated TPR provides useful information for an estimation of tritium breeding ratio in a fusion blanket. The C/E values for three measuring methods are shown in Table 5.2.2 for the systems with and without a multiplier. In the case of zonal method, those values of integrated TPRs and a multiplication factor are close to unity but the other two methods show the deviations of several percent.

References 1) Nakamura T. and Abdou M.A.: Fusion Technol. 10.541(1986). 2) Nakamura T. and Abdou M.A.: Int. Symp. of Fusion Nuclear Technol.

Tokyo(1988). 3) Nakagawa M. et al.: Int. symp. on Fusion Nuclear Technol.

Tokyo (1988), 4) Ikeda Y. et al.: JAERI 1312(1988).

Table 5.2.1 Range of C/E values for reaction rates along the central channel of the test zone

Reaction System C / E range 5 aNi(n,2n) S 7Ni Ref. 1.03 - 0.77*

Be-Sand. 1.08 - 0.80* 93Nb(n.2n)92Nb Ref. 1.05 - 0.91* 1 9 7Au(n.2n) 1 9 6Au Ref. 1.02 ~ 0.91 2 7 Al(n,d) M Na Ref. 1.01 - 0.87

Be-Sand. 0.99 - 0.74 5 8Ni(n,p) 5 8Co Ref. 1.08 ~ 0.90*

Be-Sand. 1.06 - 0.87* n 5 I n ( n , n ' ) u 5 I n Ref. 1.23 ~ 0.97

Be-Sand. 1.13 ~ 0.96 1 9 7 Au(n,n l 9 'Au Ref. 0.98 ~ 0.78

Be-Sand. 1.07 ~ 0.81**

1 c of calculated value < 5 X. * Activation cross section by Ikeda et al. ** Except the region around the beryllium.

]AERI-M 88圃 221

With respecl to lhe tritium production rate(TPR)， the differences

of several percent can be seen among the measuring methods. Figure

5.2.2 shows the C/E values of TPR where the TPR was measured by lhe

zonal melhod. The agreement between the measured and calculaled

values is satisfactory in the range z= 0 -30 cm. The integrated TPR

provides useful information for an estimation of tritium breeding

ratio in a fusion blanket. The C/E values for three measuring methods

are shown in Table 5.2.2 for the systems with and without a multiplier，

In the case of zonal method， those values of integrated TPRs and a

multiplication factor are close to uni ty but the other two melhods

show the deviations of several percent.

References

1) Nakamura T. and Abdou M.A.: Fusion Technol. よ旦，541 ( 1986) •

2) Nakamura T. and Abdou M.A.: Int. Symp. of Fusion Nuclear Technol.

Tokyo (1988).

3) Nakagawa M. et al.: Int. symp. on Fusion Nuclear Technol.

Tokyo (1988)

4) Ikeda Y. et al.: JAERl 1312(1988).

Table 5.2.1 Range of C/E values for reaction rates along

the cenlral channel of the test zone

Reaction System C I E range

S8Ni (n， 2n) 57Ni Ref. 1. 03 '" 0.77・Be-Sand. 1.08 '" 0.80・

93Nb (n， 2n) 92Nb Ref. 1.05"， 0.91・197Au(n，2n)196Au Ref. 1.02 '" 0.91

27Al (n， d)2~Na Ref. 1. 01 句 0.87Be-Sand. 0.99 '" 0.74

SSNi (n， p) sSCo Ref. 1.08 ... 0.90・Be-Sand. 1.06 ... 0.87・

1l5In(n，n，)115In Ref. 1. 23句 0.97

197Au(n， y) 19SAu Be-Sand. 1.13 ... 0.96 Ref. 0.98 .... 0.78 Be司 Sand. 1.07 ... 0.81帥

1σof caLcuLated vaLue < 5 %. 事 Activationcross section by Ikeda et al. ** Except the region around the beryllium.

J A E R I - M 8 8 - 2 2 1

Table 5.2.2 C/E values of integrated tritium production rate of natural lithium

Method Reference With beryllium Multiplication'

Zonal 1.03 1.00 On-1ine 1.07 0.97 Fo i l 1 .00 0.92

0.98 0.91 0.92

ratio of with beryllium to reference i a-3-5 %

8~20% (in LigCOal

Fig.5.2.1 C/E map for Ni(n,p) Co activation rates at the inner surface of the container. The upper and lower values are calculated with ENDF/B-IV and FNS files, respectively

( »lO' )

20.0 30.0 *0.0 50.0 Distance from front surface < em )

Fig.5.2.2 C/E values of zonal tritium production rate in the reference system

J 八 ERI-~ 88・2Z1

Table 5.2.2 C/E values of integrated tritium productlon rat号口f

natural lithium

."Ie thod Reference With beryJ J lum 刈uJtipJ lcationa

On-I ine

Fo i 1

-・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・・-----------------

a ratio o( wi th beryll ium to reference

58... . .58 Fig.5.2.1 C/E map for UUNi(n.p)UUCo activation rates at the inner

surface of the container. The upper and lower values are

calculated wi th ENDF/B-IV and FNS fi les. respectively

( _，0・}1. 2Q .... 守司~

;:::r竺TJJJJJJJ!?::::::::::;:::::::::::;:::

0.40 L...E三h..e2L..___...J

a.却0.0 10.0 20.0 却 .0 柑 .0

D1・包副lcefro. fron‘ ・urf・ce( c・3

50.0 50.0

Fig.5.2.2 C/E values of zonal tritium produclion rate

in the reference system

J A E R l - M 8 8 - 2 2 1

5.3 Blanket Benchmark Experiment on a Beryllium-Sandwich Lithium-Oxide Cylindrical Assembly

H. Maekawa, S. Yamaguchi, Y. Ikeda, Y. Oyama and C. Konno

As the fourth clean benchmark experiment using the FNS facility, an integral experiment on a lithium-oxide cylindrical assembly with beryllium neutron multiplier has been carried out. This assembly, "Be-sandwich Li 20" was selected from following reasons : i) Beryllium is the most promising neutron multiplier regarding the

fusion blanket materials, ii) A large discrepancy was observed between calculated and measured

tritium production rates (TPR) in the engineering benchmark experiments on Li 20 assemblies with Be-region. A sectional view of the experimental arrangement is shown in Fig.

5.3.1. The cylindrical assembly had a 630 mm area equivalent diameter. A 50.8 mm-thick Be-region was sandwiched by a 50.6 and a 506 mm-thick Li 20 regions. Lithium-oxide and beryllium blocks were stacked to simulate a cylinder just in the same manner as those in the previous

.,, 2,3,4) three assemblies An experimental channel a set of sheath and drawer made of type

304 stainless steel was placed at the center of the assembly. Special-sized blocks, some of which had an experimental hole, were loaded in the drawer in order to set detectors and foil samples. The d-T neutron target was located at 200 mm from the front surface of the assembly on the central axis.

Measured items and their methods are summarized as follows : i) Fission-rate distributions

• micro-fission chambers of U-235, U-238, Np-237 and Th-232. ii) Tritium production-rate distributions

• liquid scintillation method with Li 20 pellets. • a pair of 6Li and 7Li glass scintillators (for T 6 ) . • indirect method with NE213 spectrometer (for T 7 ) . • self-irradiation method with LiF TLDs.

iii) Reaction-rate distributions • foil activation method using Al.Au, Co, Fe, In, Mn, Nb, Ni, Ti, Zn and Zr.

]AERl・M 88・:?:!l

5.3 B1anket Benchmark Experiment on a Beryl1ium-Sandwich

Lithium-Oxide Cylindrica1 Assemb1y

H. Maekawa， S. Yamaguchi， Y. Ikeda， Y. Oyama and C. Konno

As the fourth c1ean benchmark experiment using the FNS facility，

an lntegral experiment on a lithlum田 oxidecylindrical assembly with

beryllium neutron multiplier has been carried out. This assembly，

"Be-sandwich Li20" was selected from following reasons

i) Beryl11um 1s the most promising neutron multip1ier regard1ng the

fusion blanket materials.

11) A large discrepancy was observed between ca1culated and measured

tritium production rates (TPR) in the engineering benchmark

exper1ments on Li20 assemblies with Be-region. 1)

A sectlonal view of the experimenta1 arrangeUlent is shown in Fig.

5.3.1. The cy1indrical assemb1y had a 630 mm area equiva1ent diameter.

A 50.8 mm-thick Be-region was sandwiched by a 50.6 and a 506 mm-thick

Li20 regions. Lithium-oxide and bery1lium blocks were stacked to

simulate a cy1inder just 1n the same manner as those in the previous 2，3，4)

three assemblies-'-'

An experimental channe1 ___ a set of sheath and drawer made of type

304 stainless steel ___ was placed at the center of the assembly.

Special-sized blocks， some of which had an experimenta1 hole， were

loaded in the drawer in order to set detectors and foi1 samples. The d-T

neutron target was located at 200 mm from the front surface of the

assembly on the centra1 axis.

Measured items and their methods are summarized as fo110ws

i) Fission-rate distributions

• micro-fission chambers of U-235， U-238， Np-237 and Th-232.

ii) Tritium production-rate distribut10ns

• 1iquid scintil1ation method with Li20 pel1ets.

• a pair of 6Li and 7Li glass scinti1lators (for T6).

• indirect method with NE213 spectrometer (for T7).

• se1f-1rradiation method with L1F TLDs.

i11) Reaction-rate distributions

• f011 activation method us1ng A1，Au， Co， Fe， In， Mn， Nb， Ni， Ti，

Zn and Zr.

一124-

J A E R I - M 8 8 - 2 2 1

iv) In-system neutron spectra • small sphere NE213 spectrometer. • multi-foil activation method.

Figure 5.3.2 shows tritium production-rate distributions measured by the four methods. As the TPR distributions of TLD were not measured absolutely at the present, the data of T 6 (TPR of 6Li) at 224 mm in the figure are normalized to 1.80 x 10~ 2 8 and the other TLD data are shifted relatively. Although those data were measured independently by quite different methods, a good agreement was observed within their experimental errors. The C/E value (calculated to experimental values ratio) of 27Al(n,a)2<*Na is shown in Fig. 5.3.3. The calculation was executed by DOT3.5 with the 125-group cross section set FSX125/J3T based on JENDL-3T.

The following facts are pointed o it from the present study : i) Measured data are consistent among each other and have high

reliability, ii) Useful benchmark data have been provided in addition to the

previous benchmark data of three assemblies of Li 20, C and Li20-C. iii) The accuracy of Be nuclear data in JENDL-3T is insufficient for

applying to fusion neutronics.

References 1) Youssef M. Z., et al.: "Analysis for Phase II Experiments of the

JAERI/US Collaborative Program on Fusion Blanket Neutronics, Part II: Tritium Production and In-System Spectrum," Int'l Symp. on Fusion Nuclear Technology, MO-02, Apr. 10-15, 1988, Toky, Japan.

2) Maekawa H., et al.: "Fusion Blanket Benchmark Experiments on a 60 cm -thick Lithium-Oxide Cylindrical Assembly," JAERI-M 86-182 (1986).

3) Maekawa H., et al.: "Benchmark Experiments on a 60 cm-Thick Graphite Cylindrical Assembly," JAERI-M 88-034 (1988).

4) Maekawa H., et al.: "Fusion Blanket Benchmark Experiment on a Lithium-Oxide Cylindrical assembly with Graphite Reflector," To be published in JAERI-M report.

5) Rhoades W. A., Mynatt F. R.: "The D0T-III Two Dimensional Discrete Ordinates Transport Code," ORNL/TM-4280 (1979).

JAE Rト M 88・221

1v) 1n-system neutron spectra

• s思畠11sphere NE213 spectrometer.

• mu1t1-f011 act1vation method.

F1gure 5.3.2 shows tr1t1um product10n-rate distributions measured

by the four methods. As the TPR d1str1but10ns of TLD were not meJsured

abs01ute1y at the present， the data of T6 (TPR of 6Lt) at 224 mm 1n the

f1gure are norma11zed to 1.80 x 10-28 and the other TLU data are sh1fted

relat1ve1y. A1though those data were measured 1ndependent1y by qu1te

different methods， a good agreement was observed w1th1n the1r experi-

menta1 errors. The C/E va1ue (ca1cu1ated to exper1menta1 va1ues rat10)

of 27A1(n，α)24Na 1s shown 1n F1g. 5.3.3. The ca1cu1at1on was executed by 5)

DOT3.5-' with the 125-group cross sectioll set FSX125/J3T based on

JENDL-3T.

The f011ow1ng facts are p01nted Olt from the present study

i) Measured data are cons1stent among each other and have high

reliability.

11) Usefu1 benchmark data have been prov1ded i.n add1t10n to the

prev10us benchmark data of three assemb11es of L120， C and L120-C.

111) The accuracy of Be nuc1ear data 1n JENDL-3T 1s insuff1c1ent for

app1y1ng to fus10n neutron1cs.

References

1) Youssef M. Z.， et a1.: "Ana1ysis for Phase II Exper1ments of the

JAERI/US C011aborat1ve Program on Fus10n B1anket Neutron1cs， Part 11:

Tritium Production and 1n-System Spectrum，" Int・1Symp. on Fus10n

Nuc1ear Techn010gy， ~IO-02 ， Apr. 10-15， 1988， Toky， Japan.

2) Maekawa H.， et a1.: "Fupion B1anket Benchmark Experiments on a 60 cm

-th1ck Lith1um-Oxide Cy11ndr1ca1 Assemb1y，" JAERI-M 86-182 (1986).

3) Maekawa H.， et a1.: "Benchmark Experiments on a 60 cm-Th1ck Graphite

Cy1indrica1 Assemb1y，" JAERI-M 88-034 (1988).

4) Maekawa H.， et al.: "Fusion B1anket Benchmark Exper1ment on a

L1thium-Oxide Cy1indrica1 assemb1y with GraphiこeRef1ector，" To be

pub11shed 1n JAERI-M report.

5) Rhoades W. A.， t1ynatt F. R.: "The DOT-II1 Two J)1mensiona1 D1screte

Ord1nates Transport Code，" ORNL/TM-4280 (1979).

-125--

-200—f50.6+50.8+- - 5 0 6 -

T fcirgei

> d _ -

Unil ram

L i 2 0 Be Li,0

Fig. 5.3.1 Experimental arrangement.

Be-sandwich L i 2 0 assembly

i . i -

0.7. 20 40 60 80 D i s t a n c e f r o i t h e t a r g e t ( c i J

Fig. 5.3.3 Comparison of C/E values for 2 7 Al(n,a) 2 l *Na in Be-sandwich L i 2 0 assembly.

Be Sandw ich L i 2 0 Assembly

10 -27

10 -28 ;

10 -29

4 Li-Gloss * Li20 Pellet * TLD * NE213

4 * £

14 * T

N K t ®

A 1 IE <t (

10 -30

T 7 f*

200 4-00 600 800 D i s t a n c e f r o m t h e t a r g e t (mm)

Fig. 5.3.2 Tritium production rates in Be-sanwich Li 20 assembly.

トー2∞ーーー+切.6+50.8+一一一-506一一一一ー→

しi20Assembly Sandwich Be ヲ''

向。「tnu

-EEE.，

lizO 8e

三互コ-~- -~--

Uml 何曹司

』〉切符日・

∞∞'NN-

Experimental arrang剖 lent.Fig. 5.3.1

Be-sandwich

a 1-li -01 ass 「

4ゐ[!.l. CT li 20 Pe I I e t

dD 中lLD 「

即 4ll +旺213

生中生 ~ 4D T 6 一

!1ti人李圃

4D ベ.

Uf:ITf¥l e

; ~<L IJ-

巴 4J q:

しi20 assembly 1.3 -

MMal 27Al(n，α)24No 1.2

'-' 0.9

600 4-00

1 0 -30

Distance

target

Tritium production rates in Be-sanwich Li20 ass佃 bly.

the f r om

Fig. 5.3.2

[ c ・]

Comparison of C/E values for 27Al(n，α)2匂Nain Be-sandwich L hO ass釧 bly.

hr 9 e t t h e

0.7_1-: 寸 u 40

Distance fro・Fig. 5.3.3

J A E R 1 - M 8 8 - 2 2 1

5.4 Measurements and Analyses of Gamma-Ray Heating in Lithium-Oxide, Graphite and Iron Slab Assemblies Irradiated with D-T Neutrons

3. Yamaguchi, H. Maekawa, K. Kosako and T. Nakamura

The measurements and analyses of y-ray heating distributions in simulated fusion blanket assemblies irradiated with D-T neutrons were made by the interpolation method using thermoluminescence dosimeters

•> 3 ) (TLD) and by using a 2-D transport code, respectrively. ' The assemblies were lithium-oxide (Li,0), graphite (63 c r x 61 cm ) and iron (100 cm* x 95 cm C) slabs (see Fig. 5.4.1).

The measurements of y-ray dose were made with four kinds of TLDs of 7LiF(Mg), Mg 25i0 u(Tb), Sr25i0u(Tb) and Ba,Si0u(Tb). Four TLD ampules for each kind were enclosed in a polyethelene bag of 0.05 mm chick and sec along che central axis of each assembly. The effective atomic number, the annealing condition and the heating condition are shown in Table 5.4.1. Neutron fluence was determined so that no correction was necessary for supralineality. The absolute neutron yield was determined by the associated j-particle method. The neutron contribution to total dose was estimated by the calculated neutron

4) response of each TLD and calculated neutron rlux obtained by transport calculation. The measured y-ray heating rates were shown in Figs. 5.4 2. x 4. The total experimental errors were ±20 a, 30 % (See Table 5.4.2). The data at the front of Li 20 and graphite assemblies were excluded because of bad signal-to-noise ratios.

Calculations were made with the 2-D transport code D0T3.5 and two coupled neutron and y-ray multi-group cross section sets GICX40 ( 42-n, 21-y energy group) and GICXFNS(135-n, 21-y energy group). As the neutron and Y - r a y source for calculations, the target neutron spectrum calculated by a Monte Carlo method and the target y-ray spectrum measured with a 5.04 cm x 5.04 cm NE213 liquid scintillator were used, respectively. The results of calculations are shown in Figs. 5.4.2 "v 4. Calculations are shown for source of neutron only, and that of both neutron and y-ray. The contribution of target y-ray is large for Li 20, graphite and snail for iron assemblies. The discrepancy of the y-ray heating calculated using both cross section sets is attributed to the difference of the secondary y-ray production

JAERI-M 88・221

5.4 :-leasurements and Analyses of Gamma-Ray Heating in

Lith1um-Oxide， Graphite and lron Slab Assemblies

Irradiated with D-T Neutrons

S. Yaruaguchi，日. ~aekawa ， K. Kosako and T. Nakamura

The measurements and analyses of y-ray heating distributions in

simulated fusion blanket a昌品母国bliesirradiated with D-T neucrons were 1 )

made by the interpolatエonmethod" using thermoluminescence dosimeters ヲ， 3)

(TLD) and by using 品 2-Dtransport code， respec:r工vely.-，-， !he

assemblies were lithium-ox1de (L120I， graph1te (63 cmや x o L cm t) and

T¥ C 1ron (100 亡温ν X 95 c:n-) slabs (see Fig. 5.4.1).

The measurements of y-ray dose were made with four kinds of TLDs

o! 7LiF(Mg)， MgzSiO 、(Tb)，SrzSiO. (Tb) and BazSiO. (Tb). Four TLD

amoules !or each kind wer色色nclosed in a polyethelene bag o! 0.05 m皿

chic瓦丑nd sec along che central axエsof each asse皿bly. The effective

acomic number， the annealing condit1on and the heat工ng condition are

snowp.エn:able 5.4.1. ~eutron flu巴ncewas determined so that no

こorrec:10nwas necessarァ forsupralineality. The absolute neutron

:r工eldwas determ1ned ~y the assoc:ated J-parti己1emethod. でhe neutron

contごibution to total dose was est1mated by the calculaced neutron 4)

response of each 7LD" and calculated neutron flux obcained by

tr丑nsport calculation. The :neasured y-ray heat1ng rates were shown 1n

Figs. 5.4 2. ~ 4. The total experimental errors were 土20~ 30 ~ (See

Table 5.4.2). The daca at the front of L1z0 and graphite assemblies

were excluded because of bad signal-to-口01se ratios.

Calculations were made w工ththe 2-D transport code DOTJ.5 and cwo

coupled neucron and y-ray mu1ci-grollp cross section sets GICX4Q ( . 42-n， 21-y energy group) and GICXFNS(135-n， 21-y energy group).

As the neutron and y-ray source for calculations， the target neutron

spectrUIll ca1culaced by a i10nte Carlo mechod and the carget y-ray

spectrum measured w1ch a 5.04 cm x 5.04 cm NE213 liquid scintillator

were used， respectively. The results of calculacions are shown in

Figs. 5.4.2 ~ 4. Calcu1ations are shown for source of neutron on1y，

and that of both neutron and y-ray. The contribution of target y-ray

1s large for Li20， graphite and s~all for 1ron assemb11es. The d1s-

crepancァof the y-ray heating calculaced using boch cross section sets

1s atcribuced to the d1fference of che secondary y-ray production

J A E R I - M 8 8 - 2 2 1

cross seccion libraries (ENDF/B-IV, P0P0P4). The calculated results are not in good agreemenc wich measured ones in boch cases wich and wichouc cargec y, excepc for che Li 20 wich GICXFNS.

References

1) Tanaka S. and Sasamoto N. : J. Nucl. Sci. Technol. 2_2 (L985) 109. 2) Yamaguchi 5., ec al. : "Gamma-ray Heating Measurement wich Thermo-

luminescence Dosimencers in a Simulated Fusion Blanket Assembly," JAERI-M 87-126 (1987) 107.

3) Yamaguchi S., ec al.: "Measurements of Gamma-ray Heating in Lithium-Oxide, Graphite and Iron Slab Assemblies Bombarded by D-T Neutrons," Proc. Inc. Symp. on Fusion Nucl. Technol., Tokyo, April 10-15 (1988) ; to be published in J. Fusion Eng. Design (1988).

4) Hashikura H., ec al, : J. Faculty of Eng., Univ of Tokyo 3_9_ (1987) 7.

Table 5.4.1 Specification of TLDs

7LiF IlgiSiO.. SrjSiOu Ba2Si0u amoule size 2.0 • diam. x 12 am long effective atomic number 3.2 U.l 32.5 49.9

annealing condition 400 "C, 30 min. • 80 «C. 10 h 5 0 ° ° C' 3° "*'

Time 60 s 60 s 60 s 60 s hearing condition Temp. (ini.) 60 °C 60 °C 60 "C 60 "C

Temp.(final) 370 "C 350 "C 420 °C 430 °C

Table 5.4.2 Experimental Errors

item error Measurement of TL-cocal dispersion

calibration 2 •». 15 : 2 ->. 10 X

estimation of TL-neutrou neutron response io -v. 20 ; extrapolation or interpolation < 20 ; neutron yield 1.5 X total 20 •v 30 S

]AERI-M 88・221

cross seccion libraries (ENDF/B-IV， POPOP4). The calculated results

are not in good agreement with measured ones in both cases with and

without target y， except Eor the LizO with GICXFNS.

References

1) Tanaka S. and Sasamoto N. J. Nucl. Sci. Technol. 22 (1985) 109.

2) Yamaguchi S.， et al. "Gamma-ray Heating :1easurem~nt with Thermo-

luminescence Oosimenters in a Simulated Fusion Blanket Assembly，"

JAERI-M 87-126 (1987) 107.

ユ) Yamaguchi S.， et al.: "!-1easurements of Gamma-ray Heating in

Lithium-Oxide， Graphite and Iron Slab Assemblies Bombarded by O-T

Neutrons，" Proc. Int. Symp. on Fusion Nucl. Technol.， Tokyo， April

10-15 (1988) tO be published in J. Fusion Eng. Oesign (1988).

4) Hashikura H.， et al， J. Faculty of Eng.， Univ of Tokyo 39

(1987) 7.

Table 5.4.1 Specification of TLOs

'LiF !".g2SiO、 Sr!SiO、amoule size 2.0 IIIID. dia皿. l{ 12 :cm lon耳

eifectユveatolllJ.C num国 r 8.2 11.1 32.5

400・C，30 min. annea斗 ngcondition

令 80.C， 10 h 500・C，30 lIIin.

Ti皿塵 60 5 60 5 60 s

h臼 tingcondition !emp.(ini.) 60 .C 60 .C 60 .C

Temp. Cfinal) 370 .C 350・c 420 .C

Table 5.4.2 Experimental Errors

ltem enor

出 asurementoi TL-total

estヰZ温 Eェ。noi TL-neutron

E:ctrapolat:1.on or inter凹 lacヰOQ

B.I!!utron yield

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calibration

neutrロロ response

一1却一

2 "1. 15 ::

2、10::

10 "1. 20 :

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L.5 ::

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Ba ~SiO"

430・c

JAERI-M 88-221

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Fig.5.4.J

J A E R I - M 8 8 - 2 2 1

5.5 Absolute Cross Section Measurements for 2 7Al(n,a) : i ,Na, 9 0Zr(n,2n) 8 9Zr and 9 3Nb(n,2n) 9 2 T nNb Activation cross Section at Neutron Energy Range from 13.3 to 15.0 MeV by Means of Associated a-Particle Counting Method

Y. Ikeda, C. Konno, K. Oishi*. K. Toraioka*. K. Kosako and Y. Oyama

Since the activation cross section is usually obtained relatively to the standard reaction cross section, high accuracy in the standard data is required. As reported previously , the activation cross sections were measured at FNS in relative to the 2 7Al(n,a) 2 1 ,Na cross section taken from ENDF/B-V dosimetry file. In order to validate the cross section data as the absolute values, the absolute cross section measurement were carried out for the reaction of 2 7Al(n,a) 2 HNa as well as 9 0Zr(n,2n) 8 9Zr and 9 3Nb(n,2n) 9 2 mNb at neutron energy range from 13.3 to 15.0 MeV. The associated a-particle

2) counting method was employed to determine the absolute neutron flux.

The foils of Al(10 mm in diameter and 0.1 mm in thickness), Nb and Zr(10 mm X 10 mm X 0.1 mm) were placed at a distance of 10 cm from the D-T target center by using a polystylene foam supporter. The error for the sample positioning was estimated to be less than 1 %. The neutron flux at the sample was derived from the neutron yield and the distance from the target with help of the neutron spectrum calculated by the Monte Carlo code M0RSE-DD with a precise model of the target assembly. The calculated spectrum was verified by the data measured with the time of flight technique. The overall error of the neutron flux was estimated to be less than 3.5 %. After irradiation by D-T neutrons for 4 hours, the reaction rate for each reaction was obtained from the gamma-ray counting using a Ge detector. The errors of the reaction rate determination were less than 2 %.

The cross sections measured are shown in Figs. 5.5.1 to 5.5.3 together with data previously measured and evaluated results. 27Al(n.g)2'*Na

The present data were in good agreement with the data in ENDF/B-V within the experimental errors as shown in Fig. 5.5.1. This indicates that all activation cross sections measured at FNS are ensured to be the absolute value because the data were obtained relative to the data of the ENDF/B-V.

* Shimizu Corporation

]AERI・M 88・221

5.5 Absolute Cross Section Measurements for z7A1(n.α)"4Na， 9OZr(n.2n)89Zr

and 93Nb(n.2n)9zmNb Activation cross Section at Neutron Energy Range

from 13.3 to 15.0 MeV by Means of Associated (l-Particle Counting Method

党台

Y. Ikeda， C. Konno， K. 01shi ~， K. Tom1oka". K. Kosako and Y. Oyama

Since the activation cross section is usually obtained relat1vely to

the standard react10n cross section. h1gh accuracy 1n the standard data 1s 1 )

required. As reported previously-'， the activation cross sections were

measured at FNS 1n relative to the 2'Al(n，α) 2句Nacross sect10n taken from

ENDF/B-V dosimetry file. In order to va11date the cross section data as the

absolute values， the absolute cross section measurement were carried out for

the reaction of 27A1(n，α)24Na as wel1 as 90Zr(n，2n)89Zr and 93Nb(n.2n)92~b

at neutron energy range from 13.3 to 15.0 MeV. The assoc1atedα-partic1e 2)

counting method-' was employed to determine the abso1ute neutron flux.

The fo11s of A1(10 mm 1n diameter and 0.1 mm 1n th1ckness). Nb and

Zr(10 mm X 10 mm X 0.1 m皿) were placed at a di~t~nce of 10 cm fru田 thcD-T

target center by uS1ng a polystylene foam supporter. The error for the

sample pos1tion1ng was estimated to be less than 1 %. The neutron f1ux at

the sample was der1ved from the neutron y1eld and the d1stance from the

target w1th help of the neutron spectrum calculated by the Monte Carlo code 3)

MORSE-DD-' w1th a prec1se model of the target assembly. The calcu1ated

spectrum was verif1ed by the data measured with the time of flight

techn1que. The overa11 error of the neutron flux was estimated to be less

than 3.5 %. After 1rradiat1on by D-T neutrons for 4 hours， the react10n

rate for each reaction was obtained from the gamma-ray counting us1ng a Ge

detector. The errors of the reaction rate determination were less than 2 %.

The cross sections measured are shown in Figs. 5.5.1 to 5.5.3 together w1th

data previously measured and eva1uated resu1ts.

27Al(n.α) 2'+Na

The present data were in good agreement with the data 1n ENDF/B司 Vwith-

1n the experimental errors as shown 1n Fig. 5.5.1. This indicates that a11 1)

activation cross sections measured at FNS-' are ensured to be the absolute

value because the data were obtained re1ative to the data of the ENDF/B-V.

* Shimizu Corporation

-]30ー

J A E R I - M 88-221

9 Z r ( n . 2 n ) a 9 Z r

Since this reaction has a high threshold energy at 12 MeV, it is a good index for the direct component of D-T neutron. The present data as shown in Fig. 5.5.2 are in excellent agreements with the data measured at FNS relative to 2 7Al(n,a) 2 1 ,Na of ENDF/B-V.

? 3Nb(n,2n) 9 2 5Nb Recently, the 9 3Nb(n,2n) 9 2 cJb has been recognized as a substandard

reaction for the D-T neutron flux monitor. As shown in Fig. 5.5.3, the present data are in good agreement within experimental errors with the data measured previously at FNS relative to 27Al(n,a)2,*Na. However, the present data give systematically lower values by 1 - 3 % than the previous ones. It indicates that the cross section value of 455 mb around 14 MeV is preferable to that of 464 mb which has been used as the standard value.

References 1) Ikeda Y.,"Activation Cross Section Measurements for Fusion Reactor

Structural Materials at Neutron Energy from 13.3 to 15.0 MeV Using FNS Facility," JAERI 1312 (1988).

2) Maekawa H., et al.;"Neutron Yield Monitors for the Fusion Neutronics Source (FNS)," JAERI-M 83-219 (1983).

3) Nakagawa M. and Mori T.;"MORSE-DD; A Monte Carlo Code Using Multi-Group Double Differntial Form Cross Section," JAERI-M 84-126 (1984).

}AERI・M 88・221

9Zr(n，2n)89Zr

Since this reaction has a high threshold energy at 12 MeV， it is a good

index for the direct component of D-T neutron. The present data as shown in

Fig. 5‘5.2 are in excellent agreements with the data measured at FNS rela-

tive to 27Al(n ， α)2~Na of ENDF/B-V.

F3Nb(n ， 2n)92~b

Recently， the 93Nb(n ， 2n)92~b has been recognized as a substandard

reaction for the O-T neutron flux mon1tor. As shown 1n F1g. 5.5.3， the

present data are 1n good agreement w1th1n exper1mental errors w1th the data

measured previously at FNS relative to 27Al(n，α)2‘Na. However， the present

data give systematically lower values by 1 -3 % than the previous ones. It

indicates that the cross sect10n value of 455 mb around 14トIeVis preferable

to that of 464皿bwh1ch has been used as the standard value.

References

1) Ikeda Y.，"Activation Cross Section Measurements for Fus10n Reactor

Structural Mater.ials at Neutron Energy from 13.3 tu 15.0 MeV Using FNS

Facility，'t JAERI 1312 (1988).

2) Maekawa H.. et al. ;"Neutron Yield Monitors for the Fus10n Neutron1cs

Source (FNS)，" JAERI-M 83-219 (1983).

3) Nakagawa M. and Mori T.; "MORSE-DD; A Monte Carlo Code Using Multi-Group

Double Differntial Form Cross Sectエon，"JAERI-M 84-126 (1984).

J A E R l - M 8 8 - 2 2 1

Neutron Energy [MeVJ

Fig. 5.5.1 Cross sections for the 27Al(n,cr)21*Na reaction.

9CC 800

700 600 500 400 i-300 200 100

/ • P"esent a a t i

C N D F / 8 - / O r. I V . d i •87 C S .L .So th r i s T B A S.M. Q I I I I I "71 V . . ^ r i n i n D n c t •73 O O . S . N e l h e . e r •72 Z Y . < t n o i - 7 2 W A.Abboud •69

* . . : ) , < ! •67 5 3 . . . P - ! 3 t < o o a • 6 ;

Neutron Energy (MeV]

• Present gate O r. i l i a . •87 O J .Leurec • 8 ! A 7 . 9 . 3 r » n •80 V D . S . N e 1 h t « « j •;B O E.rlol ub •76 Z M. Sorminn •70 M A.Piu1 sen " 0

13 Neutron

Energy (MeV)

Fig. 5.5.2 Cross sections for the

90Zr(n,2n)',9Zr reaction.

Fig. 5.;.3 Cross sections for the

9 3Nb(n,2n) 9 2 mNb reaction.

]AERI-M 88-221

7y t'~ • P'e"nl d ・1・一一一一 JENOI・2

JENDL -3 T ー←叫ー ENDfl自-vo ~， Kudo '84 D凡 1，51111nhoe . 83 ムJ.C!II k， I 'S.! マLu ~i・ n "ll n • ，~9 01 日.Ayve' ・・8X 11. Von・ch '.'0 刷例.日o，m・nn '65 被 A.PauIH:n '65 壬J，P，Bull.. '63

ドー

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En e r 9 y Neutron

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Fig. 5.ニ.3Cross sections for the

93Nb(n，2n)92mwb reaction.

Fig. 5.5.2 Cross sections for the

90Zr(n，2n)89Zr reaction.

J A E R I - M 8 8 - 2 2 1

5.6 The Formulation of Systematics for (n,p), (n,np), (n,a) and (n,2n) Reaction Cross Sections Based on the Data Measured at FNS

C.Konno and Y.Ikeda

A hundred and forty neutron activation cross sections at energy range from 13.3 to 15.0 MeV have been measured'-) on the unified experimental condition at FNS facility for various structural materials of fusion reactor. Based on these data, we determined the coefficients of formulas for systimatics of (n,p), (n.np), (n,a), and (n,2n) reaction cross sections. (n,p) reaction

In Fig. 5.6.1 che present data at 14.9 MeV are plotted along with the previously reported data as a function of (N-Z)/A, where N, Z, and A were neutron, proton, and mass numbers of the target nuclei, respectively. It is evident that the cross section decreases as the (N-Z)/A increases. Using present daca, the systematics was expressed by a simple formula based on the statistical model, given by,

a, . ' 36.18-CA1/3 + l)2-exp(-31.468-S), (1) S = (N-Z)/A.

It is significant to note that our data lie in a narrow band along the line derived from eq. (1), whereas data previously reported in literatures show a rather large divergency. (n.np) reaction

According to the increase of energy from 13 to 15 MeV, che cross sectio values of the (n.np) reaction increase rapidly. In Fig. 5.6.2 the cross sections measured are plotted as a function of (N-Z)/A. The upper and lower curves were drawn to fit the data at 14.9 and 13.3 MeV, respectively. From this result, it was suggested that a great care must be taken into the formulation of the systematic trend for (n,np) reaction because of strong energy dependence of the cross section. The systematics for this reaction cross section at 14.9 MeV was expressed by,

Of*, ™ \ " 6 8 0 . 0 - C A 1 / 3 • l ) 2 - e x p ( X ) . ( 2 ) ( n . n p )

X - - 1 0 5 . 0 5 - S + 3 7 0 . 5 9 - S 2 - 4 7 1 . 4 7 - S 3 .

( n , a ) r e a c t i o n

The systematic trend for (n,a) reaction is very similar to that for (n,p) reaction. The systematics is given by,

]AERI-M 88圃 221

5.6 The Forrnulation of Systematics for (n，p)， (n，np)， (n，α) and (n，2n)

Reaction Cross Sections Based on the Data Neasured at FNS

C.Konno and Y.lkeda

A hundred and forty neutron activation cross sections ar energy

range from 13.3 to 15.0 MeV have been measured1) on the unif1ed

experimental condition at FNS facility for various structural

田昌terialsof fusion reactor. Based on these data， we determlned the

coefficients of formulas for systimatics of (n，p)， (n，np)， (n，α)， and

(n，2n) reaction cross sections.

(n，p) reaction

In Fig. 5.6.1 the present data at 14.9 MeV are plotted along with

the previously reported data as a function of (N-Z)/A. where N. Z， and

A were neutron， proton， and mass numbers of the target nuclei，

respectively. It is evident that the cross section decreases as the

(N-Z)/A increases. Using present data， the systematics was expressed

by a simple formula based on the statistical model， given by，

'" 36.18・(A1/3+ 1)2・exp(-31.468・S)， (1)

(n.p) S (N-Z)/A.

It is significant to note that our data lie in a narrow band along the

line derived from eq. (1)， whereas data previously reported in

literatures show a rather large divergency.

(n.np) reaction

According to the increase of energy from 13 to 15 MeV， the cross

sectio values of the (n，np) reaction increase rapidly. In Fig. 5.6.2

the cross sections measured are plotted as a function of (N-Z)/A. The

upper and lower curves were drawn to fit the data at 14.9 and 13.3

MeV， respectively. From this result， it was suggested that a great

care must be taken into the formulation of the systematic trend for

(n，np) reaction because of strong energy dependence of the cross

section. The systematlcs for this react10n cross sect10n at 14.9 MeV

was expressed by，

(n，np) s 680.0 ・(A1/3+ 1)2・exp( X )， (2)

X .. -105.05・S+ 370.59・S2 -471.47・S3.

(n，(I) reaction

The systematic trend for (n.α) react10n 1s very similar to that

for (n，p) react10n. The systemat1cs 1s g1ven by，

JAERI-M 88-221

a, > - 2 4 » 2 0 - ( A 1 / 3 + l ) 2 - e x p ( - 3 4 . 5 4 6 - S ) . (3) (n,2n) reaction

The systematics for the (n,2n) reaction cross section Is shown in Fig. 5.6.3 where the data are plotted as a function of (N-Z)/A. In contrast to the (n.charged particle) reactions, the (n,2n) reaction gives a profile where the cross section for lighter mass region (Z < 30) increases rapidly and it increases as the mass number increases. For the data of heavy mass nuclei (A > 100), it becomes almost constant. This figure indicates that almost all data are encompassable within a narrow band along a curve which was formulated as

lnCT, , , - 7.402-U - 1.434-exp(-26.566-S)). (4) (n,^n;

Consequently, it is demonstrated that cross section measurement at FNS makes the analysis of systematic trend very precise.

References

1) Ikeda Y. et.al.: "Activation Cross Section Measurements for Fusion Reactor Structural Materials at Neutron Energy from 13.3 to 15.0 MeV Using FNS Facility," JAERI 1312 (1988)

io •

Z I 0 2

«C«\

Re a c t i o n ( n . p ) En = 1 1 . 9 MeV

•a o" « SIM, 51V , ^^o* .117

SSCg ». OWH, a * S

• ? r t i * n e Data

O Oeh«r £xD«r la«f iea l Dae a

0.0 0.20 0.05 0.10 0.15 Asymmetry Psrametsr (N-Zl/A

Fig. 5.6.1 Syscimat.2 cs of (n,p) reaction cross sections at 14.9 MeV

JAERI-M 88・221

• 24、20・(A1/3 ~ 1)2.exp(ー34.546・5)• (n，α}

(n，2n) reac1:ion

The systematics for the (n，2n) reaction cross sect10n is shown in

5.6.3 where the data are plotted as a function of (N-Z)/A.

In Fig.

contrast to the (n，charged particle) reactions， the (n，2n) reaction

gives a profile where the cross sect10n for 11ghter mass region (Z <

30) increases rapidly and it increases as the mass number increases.

it becomes a1most For the data of heavy mass nuc1ei (A > 100)，

constant. This f1gure ind1cates that a1most a11 data are encompassable

within a narrow band along a curve which was formulated as

lnσ ・7.402・(l - 1.434'exp(ー26.566・5)) • (n，2n) (4)

Consequently， it is demonstrated that cross section measurement at

FNS makes the analysis of systematic trend very prec1se.

References

1) Ikeda Y. et.al.: "Activation Cross Section Measurements for Fus10n

Reactor Structural Materials at Neutron Energy from 13.3 to 15.0

MeV Using FNS Fac1lity，" JAERI 1312 (1988)

円eaction (n.p) 3 E刀=14..9門eV i

¥kぞ i31NI ，'-. "TI

<<c.¥' <11呪・』ーで町.~ !21"~可 4

ヮム J .'。¥¥，;'l句0"

¥aK1saai

吋0.20

Pr...nc D・c• • L

O. 15 IN-ZJ IA

Oc拘・~ !zD.r1..ftc~1 o.C&

0.05 O. 10 Asymmetry P!ramet~r

" :; 10 I

-司一-句"の

Fig. 5.6.1 Systimatょ:sof (n，p) reaction cross seccions at 14.9 MeV

J A E R I - M 88-221

l 0V w

E \ i *MI

'• \1 r.e 3 c : i a n I n . n a • Ei = U . 9 MeV o £.1 = 13 .2 MeV

I n . d )

" io 3 L

10 0.0

JTi 117, %l. i S5.VJI," \ ? • ' « « •

0.20 0.05 0.10 0. IS •ijymmetry ?»rame:ir (N-Il/A

Fig. 5.5.2 Sysciaacics of (n.ap) raac- ion cross sec t ions ac 13.3 and U . 9 MeV

;- he sc i i on In, ^n , '• En = U . 9 MaV

1 0 3 _ E t

1 £.0*

10" 0.0

"J3n Sc I00H-

I'RV «Zr-

« C 4

0.05 0 .10 0.15 Asymmetry Parameter (N-Zl/A

g. 5.6.3 Syscemacics of (n,2n) reaction cross sections ac 14.9 MeV

]AERI-M 88・221

…リベ勺いの、

J-hm喝ハ爪JJWUU

眠、¥¥‘一一

-¥EF--「-il--Etlac-一=:---・Il--E-afi--10

。=e--uuの円R0・ハ}

ー...-::

0.10 I-"!rame~:r

。.05.l~ Y mme t r y

sec:ニロns5.6.2 Sys:~~ac~cs oe (n ，~?) "呈ac:~on c"oss

ac 13.3 and 14.9 ~eV

11払 ι27J・・・一一・三・・ーーャーUZr- -- I

SSTfU12-・両町・.7R~ → 町 ... ;~~ペロir 叫・岨C• • 1.5<:よ35r'IUa・コ

/ ~ 町 1・.l'F i ~e~i w :~Nt

~ _J - : ，

I i 0.20

Re!!ct i口口 [n.2n En = i<i..9 ，"'1:V

(N-ZJ/A O.OS 0.10

Asymmetry Psrsrneter

=こ，

ICLI--Eヒ」卜・LEE-[|「FELl「1LD

Eg--ouma軒"。、ω

言ig. 5.6.3 Sys~ematics or (n，2n) reac:ion cross sec:主ons a t 14 . 9 Me V

JAERI-M 88-221

5.7 Measuring System for Secondary Gamma-ray Production Cross Section at FNS

Y. Ikeda and C. Konno

The production cross section of gamma-rays emitted via interactions with 14 MeV neutrons is important in the nuclear heating calculation for the fusion reactor materials. Experiments have been carried out in many

1-3) facilities and provided a large number of data at neutron energy range from 1 MeV to 20 MeV for the major structural materials. However, it has been pointed out that there still exist considerably large discrepancies among data reported in the various authors. In particular, the lack in the reliable -data at 14 MeV was quoted often from a view point of data evaluations.

In order to provide systematic experimental data for the secondary gamma-ray production cross section at 14 MeV for the new nuclear data evaluation, an experimental progarm has been planned utilizing the intense pulsed D-T neutron at FNS. For this purpose, the measuring system has been installed.

A germanium detector(Ge) is the main detector of the system which was specially designed to be inserted into a large detector shielding housing. The detector system configuration is shown in Fig. 5.7.1. In order to reduce the background due to the Compton scattering in the detector, a compton suppression system was employed by using a BGO scintillator having special shape, in which the Ge was inserted. In order to increase the suppression ratio for the high energy gamma-ray to be detected, the Ge detector was inserted from the upper side to be located in the middle of the BGO, as shown in Fig. 5.7.2. This configuration increases the solid angle for the Compton sacttering components to be counted.

The soild angle subtened by the main Ge detector with respect to the irradiation sample should be determined accurately to obtain accurate cross section data. The collimator 50 cm in length (40 cm; iron and 10 cm; Pb) and 3 cm in diameter was placed between the Ge and the sample. The collimation effect was measured by scanning the 6 0Co gamma-ray source across the cross sectional plane at the sample position. The profile of the gamma-ray intensity is shown in Fig. 5.7.3. It is clear that the the sample size of 3 cm in diameter is within the flat portion of the profile. It indicates that if the sample of less than 3 cm in diameter is used, any corrections for the

]AERI-M 88・221

5.7 Measuring System for Secondary Gamma-ray Production

Cross Section at FNS

The production cross sect10n of ga皿ma-raysem1tted v1a 1nteract1ons

w1th 14 MeV neutrons 1s 1mportant 1n the nuclear heat1ng calculat10n for the

fus10n reactor matar1als. Exper1ments have been carr1ed out 1n many 1-3)

fac111t1es. -， and prov1ded a large number of data at neutron energy range

from 1 MeV to 20 MeV for the major structural materials. However， it has

been po1nted out that there st111 exist cons1derab1y 1arge d1screpanc1es

among data reported in the var10us authors. In particular， the lack in the

re11able .data at 14 MeV was quoted often fron a v1ew po1nt of data

eva1uations.

In order to prov1de systemat1c exper1menta1 data for the secondary

gamma-ray product10n cross sect10n at 14 MeV for the new nuc1ear data

eva1uat10n， an experimental progarm has been planned ut111z1ng the intense

pulsed D-T neutron at FNS. For th1s purpose， the measur1ng system has been

installed.

A german1um detector(Ge) is the main detector of the system which was

specially designed to be 1nserted 1nto a large detector shielding housing.

The detector system conf1gurat1on is shown in F1g. 5.7.1. In order to

reduce the background due to the Compton scattering 1n the detector， a

compton suppress10n system was employed by using a BGO scinti11ator having

special shape， in wh1ch the Ge was inserted. In order to increase the

suppress10n rat10 for the h1gh energy gamma-ray to be detected， the Ge

detector was 1nserted from the upper s1de to be located in the middle of the

BGO， as shown 1n F1g. 5.7.2. Th1s configuration increases the solid angle

for the Compton sacttering components to be counted.

The soild angle subtened by the main Ge detector with respect to the

irradiation sample should be determined accurately to obtain accurate cross

section data. The collimator 50 cm in length (40 cm; iron and 10 cm: Pb) and

3 cm in diameter was placed between the Ge and the sample. The collimation

effect was measured by scann1ng the 60Co gamma-ray source across the cross

sectional plane at the sample position. The profile of the gamma-ray

1ntens1ty 1s shown in F1g. 5.7.3. It is clear that the the sample size of 3

cm in diameter 1s with1n the flat port1on of the prof11e. It 1nd1cates that

if the sample of less than 3 cm 1n diameter is used， any correct1ons for the

J A E R I - M 8 8 - 2 2 1

soild angle is not necessary; only the efficiency at the position is needed for the determination of the cross section.

As a preliminary test for the system, the first experiment with use of D-T neutron pulse has been carried out for the iron, copper, carbon, aluminum and lithium samples. Figure 5.7.4 shows the measured gamma-ray spectrum for the iron sample at the angle of 125° with respect to the incident d + beam direction. A prominent gamma-ray line at 847 keV from 5SFe(n,n'Y) reaction was observed in the spectrum obtained by a measurement for less -nan two hours. From the consideration that it took more than several days to get enough statistics in the same type of experimental configuration at other facilities, it is well recognized that the present experimental configuration and the measuring system are significantly efficient. It is expected that the present system will provide systematic experimental data to reinforce the nuclear data evaluation in the future.

References 1) Lachkar J., et al.; Nucl. Sci. and Eng., 55, 168(1984). 2) Bell 2. W., et al.;"Neutron Flux Measurements at 22 ra Sattion of the Oak

Ridge Linear Accelerator Flight Path No. 8," ORNL/TM-8514 (1983). 3) Dickends, et al.;"Cross Sections for Gamma-ray Production by Fast

Neutrons for 22 Elements Between Z-3 and Z»82," Nucl. Sci. and Eng., 62,

Fig. 5.7.1 Experimental system congiguration for the secondary gamma-ray production cross section measurement.

jAERI-M 88・221

soi1d ang1e is not necessarYi on1y the efficiency at the position is needed

for the determination of the cross section.

As a pre1iminary test for the system， the first experiment with use of

D司 Tneutron pu1se has been carried out for the iron， copper， carbon，

a1uminum and 1ithium samp1es. Figure 5.7.4 shows the measured gamma-ray

spectru皿 forthe iron samp1e at the ang1e of 1250 with respect to the

incident d+ beam direction. A prominent gamma-ray 1ine at 847 keV from

S6Fe(n，n'y) reaction was observed in the spectrum obtained by a measurement

for 1ess .nan two hours. From the consideration that it took more than

severa1 days to get enough statistics in the same type of experimenta1

configuration at other faci1ities， it is we11 recognized that the pres邑nt

experimental configuration and the measuring system are significantly

efficient. It is expected that the present system wi1l provide systematic

experimenta1 data to reinforce the nuclear data eva1uation in the future.

References

1) Lachkar J.， et al.; Nucl. Sci. and Eng.， 55，168(1984).

2) Be11 Z. W.， et a1. ;"Neutron F1ux l1easurements at 22 m Sattion of the Oak

Ridge Linear Acce1erator Flight Path No. 8，" ORNLITM-8514 (1983).

3) Dickends， et al.; "Cross Sections for Gamma-ray Production by Fast

Neutrons for 22 E1ements Between Z-3 and Z"82，" Nuc1. Sct. and Eng.， 62，

515(1977).

Fig. 5.7.1

production cross seccion measurement.

J A E R I - M 88-221

Fig. 5.7.2 The cross s e c t i o n a l view

of the gamm-ray d e t e c t o r .

•-•••-•

center axis

60, Co at ! . 5 m

-6 -t -I 0 2 1 6

Radial Distance (ca)

Fig. 5.7.3 The colimetor c h a r a c t e r i s t i c s measured by 6 0 Co gamma-ray source.

U 1UJ) 8

i63 ( 7 6 G « )

-J^»-wij

200 400 600

Channel Number

d 800 1000

Fig. 5.7.4 The secondary gamma-ray spectrum due to Fe(n.nxg) react ion.

- 1 3 8 -

jAERI-M 88・221

• .-・.-.•

60 Co at :'5 m

center &11.' S

• •

‘' •

auζ33dhaw』句量・aV9-Zu--ω昌

"句'1・9弓l

‘J吻 l'刊 1

s • 6 -2 -4 -6

Radial D1stance (cm)

The colimetor characteris:ics

measured by 60Co gamma-ray

source.

Fig.5.7.3 The cross sectional view

of the gamm-ray detector.

Fig. 5.7.2

(l<. 1(3) B

‘u s

向HW3

。。

y‘ S96 ('"Ge)

戸川

WCロ司王U¥帥HE20u

1000 800 600 400 200

。。Channel Number

The secondary ga阻皿a-rayspectrua due to Fe(n，nxg) react1on.

-1詔一

F1g. 5.7.4

J A E R I - M 8 8 - 2 2 1

5.8 Determination of D-T Neutron Source Strenghth and Position by Foil Activation Technique

The neutron source strength and location are key terms in planning and in interpreting results of both fundamental neutronics experiments as well as engineering tests in next D-T burning fusion devices such as FER, NET, and ITER. Thus, an R&D how to measure the source terms is required in advance of the design of the testing devices, since severe environmental restrictions are expected in addition to the fact that the geometry is tight and complicated and the source distributes voluminously in plasma.

A feasibility test of the experimental approach for the determination of D-T neutron source strength and position was carried out in the course of Phase-II experiment of JAERI/USDOE collaborative program on fusion blanket neutronics. The neutron flux distribution inside the Phase-II system which consists of a Li 2C0 3 container and Li 20 testing material, forming the wall of a cavity in which the D-T source is positioned, were measured by the foil activation technique. The source terms was obtained from the mapped activation foils.

The cross sectional view of the Phase-II closed system is shown in Fig. 5.8.1. The system consists of a 20 cm thick Li 2C0 3 container, a 5 cm thick polyethylene reflector and a 60 cm thick Li 20 testing zone. The D-T rotating target is located within the cavity on the symmetrical axis. The D-T neutrons were generated by bombarding a tritium target with a d beam. This closed geometry system has great advantages for a better simulation of fusion environment, in comparison with Phase-I system, i.e., open geometry with external source.

Fifty two sets of activation foils, Al, Ni, Nb, and Au, were emplaced inside of the cavity to obtain energy dependent flux distributions from reaction rates of 2 7Al(n,a) 2 1 ,Na, S 8Ni(n,2n) S 7Ni, 5 8Ni(n,p) S 8Co, 9 3Nb-(n,2n) 9 2 mNb, x 9 7Au(n,2n) i 9 6Au and » 9 7Au(n, Y) 1 9 3Au. In particular, a large number o f foils were placed on the front surface of the Li 20 test region to provide precise knowledge of flux profile. After heavy irradiation by D-T neutrons, each reaction rate was deduced from y-ray counts measured by a Ge detector. The neutron yield was monitored by a 2 3 2 T h fission counter calibrated by the associated a-particle counting method. All reaction rate data were obtained with experimental error within ± 5%.

jAERI-M 88・221

5.8 Determination of D-T Neutron Source Strenghth and

Position by Foil Activation Technique

The neutron source strength and location are key terms in planning

and in interpreting results of both fundamencal neutronics experiments as

well as engineering tests ~n next O-T burning fusion devices such as FER，

~ET ， and ITER. Thus， an R&D how co measure the source terms is required in

advance of the design of the tp.sting devices， since severe environrnental

restrictions are expected in addition to the fact that the geometry is

tight and complicated and the source distribut邑svoluminously in plasma.

A feasibility test of the experimental approach for the determin-

at工onof O-T neutron source strength and position was carried out in the

course of Phase-II experiment of JAERI/USDOE collaborative program on 1 )

fusion blanket neutronics. -， The neutron flux distribution inside the

Phase-II system which consists of a Li2C03 container and Li20 testing

material， forming the wall of a cavity in which the O-T source is

posit土oned，were measured by the foil activation technique. The source

terms was obtained from the mapped activation foils.

The cross sectional view of the Phase-II closed system is shown in

Fig. 5.8.1. The system consists of a 20 cm thick Li2C03 concai九er，a 5 cm

thick polyethylene ref1ector and a 60 cm thick Li20 testing zone. The O-T

rotating target is 10cated within the cavity on the sym皿etrica1axis. The

D-T neutrons were generat ed by bombarding a tritium target W1Eh a d +beam •

This c10sed geometry syste皿 hasgreat advantages for a better simulation

of fusion environment， in ccmparison with Phase-I system， i.e.， open

geometry with externa1 source.

Fifty two sets of activation fOi1s， A1， Ni. Nb， and Au， were emp1aced

inside of the cavity to obtain energy dependent f1ux distributions from

reaction rates of 2?Al(n，α)2処Na，s8Ni(n，2n)S?Ni， s8Ni(n，p)S8Co， 93Nb-

(n ， 2n)92~b ， 19?Au(n，2n)196Au and 19?Au(n，y)199Au. In particu1ar， a large

number oc foi1s were placed on the front surface of the LizO test region

to provide precise know1edge of flux profile. After heavy irradlation by

O-T neutrons， each reaction rate was deduced from y-ray counts measured by

a Ge detector. The neutron yie1d was monltored by a 232Th flsslon counter

ca1ibrated by the associated a-particle counting method. All reaction

rate data were obtained with experimenta1 error within t 5~.

J A E R I - M 8 8 - 2 2 1

The source position and strength in Phase-II system were determined from 9 3Nb(n,2n) 9 2 wb reaction races distribution inside the cavicy. The least square method was adopted in the determination of Che position. The formulation is given as follows:

N N S * I I (ri 2Ri-rj 2Rj) 2. ( to be minimum)

i-1 J-l

r i2-(x 0-x i) 2 + (y0-yi)2*-(zo-Zi)2 ,

where N : number of measured data, R- : the reaction-rate at measuring point No.i, (x0> ^0' Z(P : c' l e coordinate of the neutron source position to

be estimated, (xi> ^i> zi) : c n e coordinate of measuring position No.i.

Then, the neutron strength, Y n, is determined by

N I Airr^Ri

* n " ^ ' NCT

where a- reaction cross section Results are shown in Table 5.8.1 along with the actual position and

the strength as derived from the neutron Eonitor. The case using all reaction data (Run#l) underestimated the neutron strength by 6%, while the case using only four data in the forward direction (Run#2) overestimated it by 7%. The underestimation is attributed to the decrease in reaction rate due to a flux depression around the D-T target, especially in the direction of 90°. The case excluding those data, (Run#3), gave good agreement of the strength with reference data.

A good agreement in the source position and strength was observed between the estimated values utilizing the mapped reaction rate data and actual values. It suggested that the foil activation technique is a promising technique to measure the D-T source position and strength, i. e., the fundamental items for plasma diagnostics directly related to D-T burning rate.

References 1) Nakamura T. and Abdou M. A.;"Overview of JAERI/USDOE Collaborative

]AERI-M 88・221

The source posit1on and strength in Phase-II system were determined

from 93Nb(n ， 2n)92~b reaction rates distribution inside the cavity. The

least square method was adopted in the determination of the posi:ion. The

formulation is given as follows:

N N S. L L (r12R1-rj2Rj)2， ( to be [町m山土

1圏 1j畠 l

r12・(XO-X1)2+(YO-Y1)2+(ZO・Zi)2

~ number of measured data，

Ri: the reaction-rate at measuring point No.1，

(Xo， yo， zo) the coordinate of the neutron source position to

be estimated，

(Xi' Yi' zi) the coord1nate of measuring position No.i.

Then， the neutron strength， Yn， is determined by

内''--

-可--

tu『唱

W川円、lf』

where σreaction cross section

Results are s~own 1n Table 5.8.1 along with the actual position and

the strength as derived from the neutron mcnitor. The case using all

reaction data (RunQl) underestimated the neutron strength by 6%， while the

case using only four data in the forward direction (RunI2) overestimated

it by 7%. The underestimation 1s attributed to the decrease in reaction

rate due to a flux depression around the O-T target， especially in the

direction of 90 0• The case excluding those data， (RunI3)， gave good

agreement of the strength with reference data.

A good agreement in the source posiこionand strength was observed

between the estimated values utilizing th吟 mappedreaction rate data and

actual values. It suggested that th~ fo11 act1vat1on technエque1s a

promising technique to measure the O-T source position and strength， i.

e.， the fundamental items for plasma diagnostics d1rectly related to O-T

burning rate.

References

1) Nakamura T. and Abdou M. A.; "0verv1ew of JAERI/USDOE Collaborati'.re

J A E R I - M 8 8 - 2 2 1

Program on Fusion Blanket Neutronics Experiments," Int. Symp. on Fusion Nucl. Technol., Apr. Tokyo, Japan (1988)

Table 5.8.1 Estimated source terms by Nb foil mapping data.

Items Case Reference Run*l** Run#2*** Runrf>3****

Neutron Yield 2.847X1016 2.681X1016 3.059X1016 2.954X1016

(n/4TT at RNT) ( + 1.7 %)

F o s i t i o n X 0 .0 - 0 . 7 -L .7 - 1 . 0

( cm ) Y 0 .0 0.02 1.5 0 .21

Z 77 .1 75.7 79.1 78.0

* Obtained by the neutron monitor. ** Using all of Nb data *** Using only 4 data at forward direction. **** Using selected 21 data. The error sources in the estimation of neutron yield are in the reaction rate ( ± 3 ^ 4 % ) , the cross section of 9 3Nb(n,2n) 9 3

(± 4 %) and the positions of foils (± 1.4 % 2.5 % ) .

Polyethylene Refleclor

OY////Y/////A "///

/ - L , 2 C 0 3 Conioiner'

Fig. 5.8.1 Cross sec t iona l view of the Phase-II experimental assembly.

- 1 4 1 -

JAERI-M 88・221

program on Fusion Blanket Neutron1cs Experiments，" Int. Symp. on Fusion

Nucl. Technol.， Apr. Tokyo， Japan (1988)

Table 5.8.1 Estimated source terms by Nb foil mapping data.

Items Case Reference Runi1.. RunI2... RunI3.......

* Neutron Yield 2.847X1016 2.681X1016 3.059X1016 2.954XI016

(n/4πat 悶 T) ( + 1.7 %)

Fosition X 0.0 -0.7 -1.7 -1.0

( cm ) Y 0.0 0.02 1.5 0.21

Z 77 .1 7:'.7 79.1 78.0

p」-l n

月炉

、ム

.、目・

*** Using only 4 data at forward direction.

**** Using selected 21 data.

The error sources in the estimation of neutron yield are in the

reaction rate (~ 3 ~ 4 %)， the cross section of 93Nb(n，2n)93m

(~ 4 %) and the positions of foils (~ 1.4 ~ 2.5 %).

Polyelhylene Rellecfor

ー，

し5.5 IZilI. 民Z ~) [cml

Fig. 5.8.1 Cross sectional view of the Phase-II experimental assembly.

J A E R I - M 88-221

5.9 A Plan of Neutronics Experiments for Next Fusion Engineering Facilities Using FNS

H. MAEKAWA

Next fusion engineering facilities, FER, NET, TIBER, OTR and new project ITER, have been proposed. In the case of FER , a reactor engineering test is planned using blanket test modules (BTM). The mission of this engineering test is to confirm the tritium breeding ability, to demonstrate the tritium continuously recovering system, to achieve actual temperature for electrical power generation and to obtain the basic test data of reactor outside in the environment under typical fusion conditions. Almost the same engineering tests are proposed for the other facilities. Neutron parameters, such as neutron fluence, neutron spectrum, tritium production rate and so on, are essential for the assessment of these test results. There are, however, two weak points in the engineering experiments using FER. First, the spectrum in BTM is not the same as that in the case of full-coverage breeding blanket. Second, as the operating condition and machine time will be restricted, e.g., the neutron flux at the first wall is about 4.4 x 10 1 3

/s/cm2 and the burning time is up to several hundreds sec, it is not so easy to measure the neutronic parameters. On the other hand, the FNS facility at JAERI has a large adaptability, i.e., the neutron yield rate can be varied widely in consideration of detector sensitivity. Moreover, it is easy to simulate the neutronic field in the cases of both partial and full-coverage blankets with assistance of enclosures (See Fig. 5.9.1). In order to support the engineering tests and to make amends for the test data, a plan of neutronics experiments using FNS is proposed here based on the status of techniques for measurement of neutron parameters and survey calculations. Figure 5.9.2 shows the tritium production-rate distributions in the BTM torus model.

The present plan of bar-chart for neutronics experiments relevant to the engineering tests of FER is shown in Fig. 5.9.3 along with the FER master plan. The plan has two phases. The objective of the first phase is to obtain the data required for detailed design of the blanket

This paper was presented at Int'l Symp. on Fusion Nuclear Technology.

]AERI-M 88・221

5.9 A P1an of Neutronics Experiments for Next Fusion

Engineering Faci1ities Using FNS

日. MAEKAWA

Next fus10n engineering fac111t1es， FER， NET， TIBER， OTR and new 1 )

project 1TER， have been proposed. In the case of FER-'， a reactor

engineer1ng test 1s p1anned using b1anket test modu1es (5TM). The

miss10n of th1s engineer1ng test is to conf1rm the trit1um breeding

ab11ity， to demonstrate the tritium cont1nuous1y recover1ng system， to

ach1eve actua1 temperature for electr1ca1 power generation and to obtain

the bas1c test data of reactor outside in the environment under typical

fusion conditions. A1most the same engineering tests are proposed for

the other faci1ities. Neutron parameters， such as neutron f1uence，

neutron spectrum， tritium production rate and 50 on， are essent1a1 for

the assessment of these test resu1ts. There are， however， two weak

p01nts in the engineer1ng experiments us1ng FER. F1rst， the spectrum 1n

BTH 1s not the same as that 1n the case of fu11-coverage breed1ng

b1anket. Second， as the operating cond1t10n and machine t1me wi11 be

restricted， e.g.， the neutron flux at the f1rst wa11 1s about 4.4 x 1013

/s/cm2 and the burning t1me is up to severa1 hundreds sec， it is not so

easy to measure the neutronic parameters. On the other hand， the FNS

faci1ity at JAER1 has a 1arge adaptab111ty， i.e.， the neutron yie1d rate

can be varied wide1y 1n consideration of detector sensitiv1ty. Moreover，

1t 1s easy to simu1ate the neutronic f1e1d 1n the cases of both part1al

and fu11-coverage blankets with ass1stance of enc10sures (See F1g.

5.9.1). 1n order to support the eng1neer1ng tests and to make amends for

the test data， a plan of neutronics exper1ments using FNS 15 proposed

here based on the status of techniques for measurement of neutron

parameters and survey calculations. Figure 5.9.2 shows the tritium

production-rate distributions in the 8TH torus model.

The present p1an of bar-chart for neutronics exper1ments relevant

to the engineering tests of FER is shown 1n Fig. 5.ヲ.3along with the

FER master p1an. The plan has twc phases. The objective of the first

phase is to obtain the data required for detai1ed design of the blanket

2) Th1s paper was presented at 1nt'l Symp. on Fusion Nuc1ear Technology.

-142ー

J A E R I - M 8 8 - 2 2 1

test modules using both partial and full-coverage simulated blankets. The JAERI/USDOE Collaborative Program on Fusion Blanket Neutronics can play a role in this first phase.

The object of the second phase is to obtain the complementary data for the engineering tests. In the case of BTM mockup experiments, it is advisable to arrange the composition of enclosure and simulated blanket assemblies so as to produce the neutron spectrum as similar as possible to those in the BTM placed in FER. Therefore further survey calculations and mockup experiments could be necessary.

In advance of tne engineering tests on site of FER, the neutronics pre-test should be carried out, i.e., the original blanket test module will be placed in the test bench (enclosure) at FNS. From these experimental data, we might obtain a proper method to convert the local tritium production rates measured to the total tritium produced. Development and confirmation of neutron diagnostics are another important task in the first phase. Some of them will be carried over to the second phase. The same type plan might be applied and contribute to the engineering tests of the other next fusion facilities proposed. This plan can also contribute to the ITER project. For this neutronics experiments, the FNS facility is the best field in the world at present.

Following items should be considered in the design of BTM: (1) Positions of samples; (2) Methods for setting and removing the samples; (3) Induced activities; (4) Monitoring method of basic neutron parameters (fluence and

spectrum). We can examine synthetically the engineering test <-*ata including

the basic neutron data by the use of neutronic data of the first and second phases, and by assistance of calculatlonal results.

References i) "Conceptual Design Study of Quasi-Steady State Fusion Experimental

Reactor (FER-Q) Part 2," JAERI-M 85-175 (1985) (In Japanese). 2) Maekawa H.: "A Plan of Neutronics Experiments for Next Fusion

Engineering Facilities Using FNS," Int'l Symp. on Fusion Nuclear Technology, M2-03, Apr. 10-15, 1988, Tokyo, Japan.

3) Nakamura T., Abdou M. A.: "Overview of JAERI/USDOE Collaborative

J A E R 1 ~ M 88・221

test modules using both partial and full-coverage simulated blankets. 3)

The JAER1/USDOE Collaborative program on Fusion 81anket Neutronics-' can

play a role 1n this first phase.

The object of the second phase is to obtain the complementary data

for the engineering tests. ln the case of 8TH mockup exper1ments， 1t 1s

advlsable to arrange the composition of enclosure and s1mulated blanket

assemblies so as to produce the neutron spectrum as similar as poss1ble

to those 1n the 8TH placed 1n FER. Therefore further survey calculat10ns

and mockup exper1ments ~ould be necessary.

lnョdvanceof tne engineering tests on s1te of FER， the neutron1cs

pre-test should be carried out， i.e.， the original blanket test module

will be placed in the test bench (enclosure) at FNS. From these

experimental data， we might obtain a proper method to convert the local

tritium production rates measured to the total tritium produced.

Development and confirmation of neutron diagnostics are another

important task in the first phase. Some of them will be carried over to

the second phase. The same type plan might be applied and contribute

to the engineering tests of the other next fus10n facl1itles proposed.

This plan can also contribute to the ITER project. For thls neutron1cs

experiments， the FNS faci11ty 1s the best f1eld 1n the world at present.

Followlng ltems should be cons1dered !n the des1gn of 8TH:

(1) Posltions of samples;

(2) Hethods for setting and removing t~e samples;

(3) 1nduced activities;

(4) Honitoring method of basic neutron parameters (fluence and

spectrum) •

We can examine synthetically the engineering test〆ataincluding

the basic neutron data by the use of neutronic data of the first and

second phases， and by assistance of calculational resu!ts.

References

1 )・ConceptualDesign Study of Quasi-Steady State Fusion Experimental

Reactor (FER-Q) Part 2，" JAER1-M 85-175 (1985) (1n Japanese).

2) Haekawa H.: "A Plan of Neutronics Experiments for Next Fusion

Engineering Facilities Using FNS，" Int・1Symp. on Fusion Nuclear

Technology， M2・03，Apr. 10-15， 1988， Tokyo， Japan.

3) Nakamura T.， Abdou M. A.: "Overview of JAERI/USDOE Collaborat1ve

J A E R I - M 8 8 - 2 2 1

enclosure ( L i 2 C 0 3 , S 5 5 l 6 , etc.!

/ / / < / I

Rotating " \ Taraet

Simulated Blanket

•*y| Assembly & L i , 0 , etc.) §_ 10"

/ / / / / / /

I I T" - F u l l Coverage Blanket --Blanket Test Model

or x > Blanket Test § Module

Fig. 5.9.1 A sketch of experimental arrangement for mockup experiment and pre-test using FNS.

10"' 660 670 680 690 700 710

Distance from the Torus Axis (cm! Fig. 5.9.2 Tritium production-rate

distributions in the BTM torus model.

Yenr 1990 1 9 9 5 2000

C o n c e p t u a l Deal in C o n s t r u c t i o n J S o l f - l t n i t Ion

i \j^ E x p e r i a e n t F E R

: D e t a i l e d O e s i en E n « i n e e r i nt T e s t

; / k : k

B l a n k e t Test Nodu le

Oeta 1 1 e d D e s 1 « n

a c t u r 1 n f

B l a n k e t Test Nodu le

Nanu a c t u r 1 n f

Eiuer i aen is tt FNS

Hem ron ics

N e u t r o n 0 i ac nos t ics

First P h a s e 1

S e c o n d : P h a s e Eiuer i aen is tt FNS

Hem ron ics

i P a r t i a r d o c k -

UP N o c k - u p

~i ' : P r e - T e s t J

Hem ron ics

' t B T H' D e v e l o p m e n t t C o n f i r m a t i o n ^

N o c k - u p J\ >

Hem ron ics

' t B T H' D e v e l o p m e n t t C o n f i r m a t i o n ^ ' :

Hem ron ics

N e u t r o n 0 i ac nos t ics — p i

FNS r ac i1i ty

A c c e l e r a t o r N o d i M c a P 1 a n n ij|c ;

t ip 11

8u iId i nt .J. N o d i f ica t i 0 1 |

Fig. 5.9.3 A plan of neutronics experiments for FER using FNS.

]AERI-M 88・221

点、一ーFulI Coveroge Blonkel ¥ ---810耐 ITesl Model

?¥l ¥ ¥T6

一ーよニニー『司・・ー・一一-ー，ーー--旦ユジ

」----z --=

旬旬=孟• • 圃

--‘--

• • 句

• 、、‘

.t 、‘、‘

‘.、、、、--

10・4

<lJ 仁J"-コo u、、、、ε ・5O 10" 1:: ー炉・・Cコ、、

<lJ ← Eコ0::

Enclosure (Li 2C03 ， SS~16 ， etc

Tes↑5 ← L>

=コてコCコ』田-

Eコ二一」↑

A sketch of experimental arrang伺 entfor mockup experiment and pre-test using FNS.

Fig. 5.9.1

710 700 690 680 670

[cmJ Torus Axis

Tritium production-rate distributions in the BT阿torus mode 1 .

the from Dislance

Fig. 5.9.2

110;l:BUshj九;F 一一一l'一一→

l !?"!!?山:l -ー『一一ーーで:.，

一 --_P山止叩・岬 L! ↓ T B T M" I Mo c k‘UP 、4

l叫に

2000 Ve.r

01，・ nketTest Nodul.

以[xper i・enLs

II FN5 Ncutronics

Neutron 日』・IAoslics

.¥cceler.tor

FN5 r・cI 1 I 11

8uildin，

A plan of neutronics experiments for FER using FNS.

Fig. 5.9.3

J A E R I - M 8 8 - 2 2 1

5.10 An Intense DT Neutron Source and Test Conditions for Fusion Nuclear Technology Research

Yukio OYAMA, Seiya YAMAGUCHI and Chikara K0NN0

Fusion neutron irradiation tests for reactor components are required to clarify the critical issues under radiation environment and to estimate design margins for fusion nuclear technology (FNT) development. This is significant to reduce a construction and maintenance costs of reactor.

A lot of high intense neutron irradiation test facilities, e.g., FMIT , INS , EURAC , etc., have been proposed for fusion material development. However, they have not been realized yet because of high construcion and operation cost, and the other defects such as small irradiation volume and low neutron efficiency. A plasma confinement neutron source like a next fusion test device needs much technological development efforts. On the other hand, an accelerator-based DT neutron source has a lot of advantages for the irradiation experiments: it has high reliability and controllability of the operation; its operating and construction costs are lower than the others because of low tritium inventry and a small number of components; and also it is easy to place the measuring devices and the equipments for simulation of fusion environments, because of simple target structure.

In the present work, we investigate the possibility of nuclear tests for the fusion reactor components using an accelerator-based DT neutron source, separately from material irradiation test..We provide here a concept of feasible DT neutron source which can be realized as an extension of the existing technology, and examine the conditions for such a nuclear test concept.

The most critical problems to be considered for a DT neutron generator are in the ion sources and target systems. At present, large current ion sources have been intensively developed for neutral beam injection to raise a plasma temperature. The beam current at the target is limited by heat removal capability of the available cooling system. Table 5.10.1 shows the target performance for various DT neutron sources including ones at the conceptual design stage. The conventional metal target system using Ti or Sc has a limited heat removal capability below 400 MW/m2. Since a certain amount of irradiation volume and high total neutron yield are required for the component test, a large area source with a broad beam is effective

JAERI-M 88・221

5.10 An Intense DT Neutron Source and Test Conditions for 1)

Fusion Nuclear Technology Research

Yukio OYAMA， Seiya YAMAGUCHI and Ch1kara KONNO

Fusion neutron irrad1at10n tests for reactor components are requ1red

to clar1fy the cr1t1cal 1ssues under rad1at10n env1ronment and to estimate

des1gn marg1ns for fus10n nuclear technology (FNT) development. Th1s 1s

significant to reduce a construction and maintenance costs of reactor.

A lot of high intense neutron 1rrad1at1on test facilit1es， e.g.， 2) ...~3) _..~ .~4)

FMIT-'， INS-'， EURAC~' ， etc.， have been proposed for fusion material

development. However， they have not been rea11zed yet because of high

construcion and operation cost， and the other defects such as small

irradiation volume and low neutron efficiency. A plasma confinement neutron

source like a next fusion test device needs much technological development

efforts. On the other hand， an accelerator-based DT neutron source has a

lot of advantages for the irradiation exper1ments: 1t has h1gh re11ab11ity

and control1ab11ity of でheoperationo 1ts operat1ng and construct10n costs

are lower than the others because of low tritium inventry and a small

number of components; and also it 1s easy to place the measuring devices

and the equ1pments for simulation of fusion environments， because of s1mple

target structure.

工nthe present work， we investigate the possibility of nuclear tests

for the fus10n reactor components using an accelerator-based DT neutron

source， separately from material irradiation test. .We provide here a

concept of feas1ble DT neutron source which can be realized as an extens10n

of the existing technology， and examine the conditions for such a nuclear

test concept.

The most critical problems to be considered for a DT neutron generator

are 1n the 10n sources and target systems. At present， large current 10n

sources have been 1ntensively developed for neutral beam 1njection to raise

a plasma temperature. The beam current at the target is 11mited by heat

removal capability of the available cooling system. Table 5.10.1 shows the

target performance for various DT neutron sources includ1ng ones at the

conceptual design stage. The conventional metal target system us1ng T1 or

Sc has a limited heat removal capab11ity below 400 MW/m2 • Since a certa1n

amount of irradiation volume and high total neutron yield are required for 5)

the component test， a large area source-' with a broad beam 1s effect1ve

J A E R I - M 8 8 - 2 2 1

for use of che conventional metal cargec. Thus, a large area source of % 0.1 m in radius is proposed as a feasible and low cose DT neucron source. This source can provide a neutron emission density of 1 x 10 l ? n/sec,'m2 and a cocal neucron yield of 3 x 10 1 5 n/sec, with a deuceron beam above 30 A-200 keV and with 200 MW7m2 heat density lower chan chat of RTNS-I1.0

Nuclear tests using the DT neucron source will be performed with some equipments for simulating fusion environments, i.e., magnet, chermo-concrolled insulator, cricium purge system and so on. Figure 5.10.1 presents the flux map of the large area target in comparison with a point source. The large area source reduces che maximum flux compared with a point source, but scill keep an irradiation volume of -UD.007 m 3 above 1 x 10 1 S n/m2/sec which corresponds to neutron wall load of 0.02 MW/m2 as well as the point source of the same neutron yield. The maximum flux is 5 x 10 1 6 n/m2/sec and che fluence for a 100 hr-operation is 2 x 102* n/m2. This level is adequate to perform the irradiation cesc for integrated effect. The calculated result shows that this tesc sample can produce a tritium of 1 Bq (27 ijCi) in a second. The increase race of average tritium gas concentration in a breeding material becomes to 3.7 MBq/mJ/sec (I x 10"" IjCi/cm-Vsec). Since a detectable limit of tritium gas concentration is x 37 kBq/m3 (1 x 10" s pCi/cm 3), this tritium production rate is acce> :able to perform che recovery tesc. Heac deposicion race produced by chis arrangement is 100 kW/m3 at the front of the first wall. This heat provides a temperature increase rate of 0.02 °C/sec for the first wall under the thermal insulation.

References

1) Oyama Y., et al.:" An Intense DT Neutron Source and Test Condicions for Fusion Nuclear Technology Research," Proc. Inc. Symp. on Fusion Nuclear Technology, Tokyo, April (1988)

2) Trego A.L., et al. : Nucl. Technol/Fusion, 4_, 695 (1983). 3) Armstrong D.D., ec al.: Nucl. Instrum. Meth., 145, 127 (1977). 4) Kley W. and Bishop G.R.:" The JRC-Ispra Fusion Reactor Materials

Test and Development Facility," Nuclear Science and Technology, November 1984, EUR 9753EN (1984).

5) Crawford J. C. and Bauer W.:" Large Area Solid Target Neutron Source," Proc. Int. Conf. on Radiation Test Facilities for the CTR Surface and Materials Program, Argonne, 1975, ANL/CTR-75-4, Argonne National Lab.

JAER(-νI 1¥8・221

for use of the conventional metal target. Thus， a large area ョourceοf ~_

0.1 m 1n rad1us 1s proposed as a feasib1e and low cost DT neutron 5ource.

This source can provide a neutron emission density of x 101勾円/sec:m2 and

a tota1 neutron yie1d of 3 x 1015 n/sec，日1tha deuteron ?eam above 30 時)

A-200 keV and with 200河W/m2 heat density lower than that or RT~S-ll.

Nuc1ear tests using the OT neutron source w111 be performed with some

equipments for simulating fusion environmencs， i.e.， magnec， thermo-

contro11告d insulator， trit1um purge 5ystem and 50 an. Figure 5.10.1

presents the f1ux map of the 1arge area target in comparison w1th a point

source. The 1arge area source reduces the maximum f1ux compared with a

po1nt source， but st111 keep an 1rradiation volume of ~0.007 mJ above 1

X 1016 n/m2/sec which corresponds tO neutron wa11 load of 0.02 ~/m2 as

well as che point source of the same neutron yield. The maximum flux is 5 x

1016 n/m2jsec and che f1uence for a 100 hr-oper昌tion1s 2 X 10 2 句 n/m 2 • This

1eve1 1s adequace to perform the irradiation test for integrated effect.

The calculaced resulc shows thac chis cest samp1e can produce a tritium of

1 Bq (27 UCi) in a second. The increase race of average tr1tium gas

concentration in a breeding material becomes tO 3. 7 ~q/mJ!sec (1 x 10-~

μCi/cmJ/sec). Since a detectab1e 1i阻止tof tritium ~as concentration lS ~ 37

kBqjmJ (1 x 10-6 UCi/cmJ)， this tritium production rate 15 acce[ :able to

perform the recovery test. Heat deposition rate produced by this arrange-

ment is 100 kW/mJ at the front of the first wa11. This heat provides a

temperature increase rate of 0.02 oC/sec for the first wall under the

thermal insulatj.on.

References

1) Oyama Y.， et a1.:" An Intense DT Neutron Source and Test Condi tions

for Fusion Nuclear Technology Research，" Proc. Int. Symp. on Fusion

Nuclear Techno1ogy， Tokyo， Apri1 (1988)

2) Trego A.L.， et a1.: Nucl. TechnoljFusion， ~， 695 (1983).

3) Armstrong D.D.， et a1.: Nucl. Instrum. Meth.， 1丘三， 127 (1977).

4) Kley W. and B1shop G. R. :" The JRC-Ispra Fusion Reactor Mater1als

Test and Deve10pment Faci1ity，" Nuclear Sc1ence and Technology，

November 1984， EUR 9753EN (1984).

5) Crawford J. C. and Bauer W.:" Large Area So11d Target Neutron Source，'・

Proc. 1nt. Conf. on Radiation Test Fac111ties for the CTR Surface and

Mater1als Program， Argonne， 1975， ANLjCTR-75-4， Argonne Nat10na1 Lab.

J A E R I - M 8 8 - 2 2 1

227 (1975). 6) Logan C. M. and Heikkinen D. W.: Nucl. Instrum. Meth., 200, 105

(1982).

Table 5.10.1 Target performances for DT neutron source facilities.

Facility Target Current Voltage Heat density

FNS 800rpm Ti/H20

RTNS-II 5000rpm Ti/H20

Sandia Fixed Sc/H20

RTNS-II, p-grade

3000rpra,4m cladded Ti/

INS D-gas jet

23 mA 350 kV

NBINS liquid-Li drive-in 50 A or Li-D

8kW/cmJ

125 mA 375 kV 40kW/cm2

200 mA 200 kV AOkW/cm'1

1.5 A 200 kV 200kW/cm'

200 kV 200kW/cm2

160 kV 80-320kW/cm2

Conceptual design study

•v. 20

S - 3 . 1 4 1 1 0 <Vs

R T» 10cm

Neutron forger / -J2JJ2 ' •, * ., »;, foraer 2xi0 \<f SxlO'1 2*10 iO' n/s

0 10 20 30 40 50 60

R [ c m ]

Fig. 5.10.1 Comparison of neutron flux discribucions for a large area

target with a point source.

)AERI-M 88・221

227 (1975).

6) Logan C. 礼 an.dHe1kk1nen O. W.: Nucl. Instrum. Mp. t~.. 辺旦. 105

(1982).

Table 5.10.1 Target performances for OT neutron source fac111t1es.

Fac1lity Target

FNS 800rpm T1/H20

RT:-lS-II 5000rpm

Ti/H20

Sandia Fixed Sc/HzO

RTNS-II"， 3000rpm・4m中up-grade cladded Ti/H20

* INS D-gas jet

* NBINS liquid-Li dr1ve-in or Li-D

'" Conceptual des1gn study

50トId'-

O 10 20

Current Voltage

23 mA 350 kV

125 mA 375 kV

200 mA 200 kV

1. 5 A 200 kV

1.1 A 200 kV

50 A 160 kV

-" -s ・3.¥41¥0'-"/s

ー-Rr-¥Ocm

---Rr-¥ cm

21¥d' ?

10・， n/s

30 40 50

[ cm J

Heat dens1ty

8kW/cm2

40kW/cm2

200kW/cru2

200kW/cm2

80-320kW/cm2

F1g. 5.10.1 Comparison of neucron flux distr1but1ons for a large area

target with a po1nt source.

J A E R 1 - M 8 8 - 2 2 1

6. Radiation Shielding

First, there are reported two experiments using d-T neutron source of the FNS. One uses the 0* beam line. In order the objective of the experiment is to investigate the behavior of 14 MeV neutrons incident to a large cavity simulating the Neutral Beam Injector room. The measured neutron spectra were analyzed with various codes and data, and comparison between them is discussed. The another one utilized the 80* beam line to the first target room. Many kinds of reaction rates in a large iron cylindrical assembly were measured by activation foil method. Analysis shows that benchmark experiments for deep penetration are important as well as proper estimation of cross sections.

Numerical evaluation was performed about shielding design for three advanced marine reactors under operational condition and hypothetical accident. Among these three types, the integral-type PWR was most appropriate from the viewpoint of shielding design.

Next, two streaming studies, experimental and/or analytical, are reported. One is concerning the additional shield design for gamma rays to compensate the shield irregurarities of reprocessing plant. The additional shield was successfully embedded in the concrete shield without increasing the shield thickness. The another one is the design study of beam holes for thermal neutrons installed in the upgraded JRR-3- The analysis shows that the tangential beam hole is best to obtain a high quality beam, that is, a beam free from high energy neutrons and gamma rays.

In addition, a set of data has been obtained by transport calculation of exposure buildup factors for slant penetration through slab shields from point isotropic gamma-ray source, to be used in the point kernel codes such as QAD or G33- The accuracy was confirmed in comparison with measured data.

All these enhancement of data and methods as well as aquisit-ion of benchmark data by experiments appeared in this chapter will be fully presented at the Seventh International Conference on Radiation Shielding in September 1988 at Bournemouth.

(Tomoo Suzuki)

lAERI-M 88・221

6. Radiation Shielding

First， there are reported two experiments using d-T neutron

source of the FNS. One uses the 0・beamline. In order the object-

ive of the experiment is to investigate the behavior of 14 MeV neu-

trons incident to a large cavity simulating the Neutral Beam Injec-

tor room. The measured neutron spectra were analyzed with various

codes and data， and comparison between them is discussed.

The another one utilized the 80・beamline to the first target

room. Many kinds of reaction rates in a large iron cylindrical ass-

embly were measured by activation foil method. Analysis shows that

benchmark experiments for deep penetration are important as well as

proper estimation of cross sections.

Numerical evaluation was performed about shielding design for

three advanced marine reactors under operational condition and hy-

pothetical accident. Among these three types， the integral-type PWR

was most appropriate from the viewpoint of shielding design.

Next， two streaming studies， experimental and/or analytical，

are reported. One is concerning the additional shield design for

gamma rays to compensate the shield irregurarities of reprocessing

plant. The additional shield was successfully embedded in the con-

crete shield without increasing the shield thickness. The another

one is the design study of beam holes for thermal neutrons install-

ed in the upgraded JRR-3. The analysis shows that the tangential

beam hole is best to obtain a high quality beam， that is， a beam

free from high energy neutrons and gamma rays.

In addition， a set of data has been obtained by transport cal-

cu1ation of exposure buildup factors for s1ant penetration through

s1ab shie1ds from point isotropic gamma-ray source， to be used in

the point kernel codes such as QAD or G33. The accuracy was confi-

rmed in comparison with measured data.

411 these enhancement of data and methods as we11 as aquisit-

ion of benchmark data by experiments appeared in this chapter will

be fu11y presented at the Seventh International Cor.ference on Rad-

iation Shie1ding in September 1988 at Bournemouth.

(Tomoo Suzuki)

J A E R 1 - M 8 8 - 2 2 1

6.1 Experiment and Analysis on the Behavior of 14 MeV Neutrons Incident to a Large Cavity

H. Nakashima, S. Tanaka, A. Hasegawa, H. Kotegawa and T. Suzuki

It is important to estimate the behavior of the d-T neutrons incident to a large cavity such as the Neutral Beam Injector (NBI) box in Tokamak fusion reactor. An experiment has been carried out with an experimental assembly having a large cavity, using the d-T neutron source of FNS.

The experimental assembly simulating the NBI box is made of the mortar of 2.04 m height, 1.44 m width and 1.54 m length, covered with 0.02 m thick type-304 stainless steel and has the cavity of 1.56 m height, and 0.96 m width and length. The neutrons generated at the rotating target in the second target room travel through a cylindrical duct of 40 cm diameter and next a rectangular throat of stainless steel whose cross section is 40 cm X 40 cm, and arrive at the inlet of the cavity. The distance from the target to the inlet of the cavity is 5.93 m. Figure 6.1.1 illustrates a schematic drawing of the experimental setup.

The fast neutron and gamma ray spectra in the cavity were obtained using a 5.06 cm-long and 5.06 cm-diameter cylindrical NE213 liquid scintillation counter and an unfolding code F0RIST. The measured spectra for the neutron and gamma-ray are shown, with the measured points, in Figs. 6.1.2 and 6.1.3, respectively. The neutron reaction rate and gamma-ray dose rate distributions were also measured using a 14 mm diameter spherical NE213 liquid scintillation counter and 2 3 2 T h and 2 3 5 U fission chambers. These data were normalized to the absolute values by using the a-particle monitoring system. These data will be available for evaluating for the data and methods for the NBI shielding design.

The measured neutron spectra were analyzed with the Monte Carlo code MCNP, using either of the MCNP cross section libraries based on ENDF/B-IV or JENDL. The comparison between the measurement and the calculations is shown, with the measured point, in Fig. 6.1.4. The calculation using ENDF/B-IV overestimates spectra in the energy region between 6.0 and 11.0 MeV, while the calculation using JENDL is in good agreement with the measurement within experiment and calculation

JAERI-M 88・221

6.1 Experiment and Analysis on the Behavior of 14 MeV

Neutrons Incident to a Large Cavity

H. Nakashima， S. Tanaka， A. Hasegawa， H. Kotegawa and T. Suzuk1

It is 1mportant to e9t1mate the b申hav10rof the d-T neutrons

1nc1dent to a large cavity such as the Neutra1 Beam Injector (NBI)

box 1n Tokamak fus10n reactor. An exper1ment has been carr1ed out

with an experimental assembl.y hav1ng a large cavity， using the d-T

neutron source of FNS.

The exper1mental assembly s1mulat1ng the NBI box 1s made of the

mortar of 2.04 m he1ght， 1.44 m w1dth and 1.54 m length， covered w1th

0.02 m thick type-304 stainless steel and has the cavity of 1.56 m

he1ght， and 0.96 m width and length. The neutrons generated at the

rotating target 1n the second target room travel through a cy11ndr1cal

duct of 40 cm diameter and next a rectangular throat of stainless

steel whose cross sect10n 19 40 cm X 40 cm， and arr1ve at the 1nlet of

the cav1ty. The distance from the target to the 1nlet of the cav1ty

19 5.93 m. Figure 6.1.1 111ustrates a schemat1c draw1ng of the

exper1mental setup.

The fast neutron and gamma ray spectra 1n the ca¥"ity were

obtained us1ng a 5.06 cm司 longand 5.06 cm-d1ameter cy11ndr1cal NE213

11quid scint111ation counter and an unfo1d1ng code FORIST. The

measured spectra for the neutron and gamma-ray are shcwn， w1th the

measured points， in F1gs. 6.1.2 and 6.1.3， respect1vely. The neutron

react10n rate and gamma-ray dose rate d1str1but10ns were a1so measured

us1ng a 14 mm diameter spherical NE213 1iquid sc1ntillat10n counter

and 232Th and 235U fission chambers. These data were normalized to

the absolute values by using theα-particle monitoring system. These

data will be available for evaluat1ng for the data and methods for the

NBI shielding design.

The measured neutron spectra were analyzed w1th the Monte Carlo

code MCNP， using either of the MCNP cross section libraries based on

ENDF/B-IV or JENDL. The comparison between the measurement and the

calculat10ns is shown， with the ~easured point， in Fig. 6.1.4. The

calculation using ENDF/B-IV overest1mates spectra 1n the energy region

between 6.0 and 11.0 MeV， while the calculation us1ng JENDL 1s 1n good

agreement with the measurement with1n exper1ment and calculat10n

-149一

J A E R I - M 8 8 - 2 2 1

errors. It is confirmed that the MCNP code is a effective tool for the NBI design calculation.

In order to examine the applicability of two-dimensional discrete ordinate codes and the group constants to this experimental arrangement, the experimental data are also analyzed using the DOT3.5 and BERMUDA-2DN codes with various group constants based on ENDF/B-IV and JENDL. The results are shown in Fig. 6.1.5 with the experimental one. The BERMUDA-2DN calculation represents measurement in shape, but both calculations underestimate. The difference between the measurement and the DOT calculation is caused by the two-dimensional geometry approximation and the approximation for the angular distribution of neutron scattering with the finite terms in the Legendre expansion. While, the discrepancy between the measurement and the BERMUDA-2DN calculation is caused by the former one because this code adopes P,,, in the Legendre expansion. It is thought that the BERMUDA-2DN code is a better tool for this type of shielding problem, in which the back-scattered neutrons are dominant.

Fig. 6.1.1 Schematic drawing of the experimental setup.

JAERI-M 88・221

errors. It 1s conf'irmed that the MCNP code 1s a effect1ve to01 for 、

the NBI des1gn calculat1on.

In order to exam1ne the app11cabi11ty of two-d1mens10na1 d1screte

ord1nate codes and the group constants to th1s exper1menta1 ar-

rangement， the exper1mental data are a1so analyzed us1ng the OOT3.5

and BERMUDA-2DN codes w1th var10us group constants based on ENDF/B-IV

and JENDL. The resu1ts are shown 1n F1g. 6.1.5 w1th the exper1menta1

one. The BERMUDA-20N calcu1at1on represents measurement 1n shape， but

both calculat10ns underest1mate. The d1fference between the

measurement and the OOT calculat10n 1s caused by the two-d1mens1ona1

geometry approximation and the approx1mat1on for the angular

d1str1but1on of neutron scatter1ng w1th the finite terms 1n the

Legendre expansion. Wh11e， the d1screpancy between the 皿easurement

and the BERMUDA-20N ca1culat1on 1s caused by the former one because

this code adopes ~∞ in the Legendre expans1on. It 1s thought that the

BERMUDA-2DN code 1s a better tool for th1s type of sh1e1d1ng problem，

1n wh1ch the back-scattered neutrons are dom1nant.

1st Target RoolII

Acc.Roo.

F1g. 6.1.1 Schemat1c draw1ng of the experimental setup.

-1回一

J A E R I - M 8 8 - 2 2 1

0 . 0 5 , 0 1 0 . 0 1 5 . 0 NEUTRON ENERGY IMeV)

Fig. 6.1.2 Measured spectra for fast neutron in the cavity.

0.0 5 . 0 GAMMA RAY ENERGY IMeVI

1 0 . 0

Fig. 6 .1 .3 Measured spectra for gamma-ray in the cavity.

0 . 0 5 . 0 1 0 . 0 1 5 . 0 NEUTRON ENERGY (MeVI

Fig. 6 .1.1 Comparison among the measured and calculated spectra using HCNP code with ENDF/B-IV and JENDL.

0 . 0 5 . 0 1 0 . 0 1 5 . 0 NEUTRON ENERGY IMeV)

Fig. 6.1.5 Comparison among the measured and calculated spectra using the DOT3.5 and BERMUDA-2DN code with on JENDL.

- 1 5 1 -

lAERI-M 88・221

• P... XIY1Z60 • P". X3Y1260 • P... xsγ1260

建1口・10〉匂'

ぞ炉、4・e色、、

5 t 〉

E E 10・12

+ P". XIYIZ60 φP... lClYIZ60 + P...渇Y1Z6D

10'・

週 10・10>

さ10・11

i 10・12

10 ・130.0 10.0 5.0

GAけ仏 RAYEN回GYlMeVI

Fig. 6.1.3 Measured spectra for gamma-ray 1n the cavity.

Fig. 6.1.2 Measured spectra for fast neutron in the cavity.

5.0 10.0 ぽIJTF削 D巴司GYlMeVJ

P・..X3Y1Z60 ・E..pt.

ー一白1，.1民子札l)A-2聞+JElO..I鑓

ーー臼1，.I∞[3. S+JElO..I P7-S16

10-・r; p.，. X3YIZ6目。E..pt.

ートc.ie. IIOF+JElO..I →-白1，.IIOF.EtUI8-IVl

‘ ' . . . . . ・. . . ・.

:; 10・・2

さω u

宅10-10 〉

* ・6E

三 10叶 1

~ ..J u...

i 1012

:; 1 0-'

ぞ 10・10

ぞe、4・e・6E U

三 10・11

~ ..J u...

喜一言

10・日

1 0 -13

0.0 10・13

0.0 15.0

Fig. 6.1.5 Comparison among the measured and calculated spectra using the DOT3.5 and BER阿UDA-2DNcode with on JENDL.

5.0 10.0 ぽUT町制 EIDむY(他VI

5.0 10.0 ぽUTR創 rn子むYI地VI

Fig. 6.1.4 Comparison among the measured and calculated spectra using MCNP code with ENDF/B-IV and JENDL.

J A E R I - M 8 8 - 2 2 1

6.2 Measurement and Analysis on Reaction-Rate Distribution in a Large Iron Cylindrical Assembly

K. Oishi , Y. Ikeda, C. Konno, K. Tomioka and T. Nakamura

Iron is one of the most Important materials in fusion facilities, not only as a structural material, but also for shielding. However, at present, there exist many differences among nuclear data libraries for estimating the neutron spectrum in iron assemblies'. Disagreements between experimental and calculated results was also reported by Hertel et al. 2 In this study, measurements of reaction rate distribution were performed using FNS to verify currently available nuclear data libraries for iron.

A large cylindrical iron assembly, 1m in diameter and 0.95m in thickness, was set at 0.2m from the D-T target along the incident d beam and was irradiated by 14 MeV neutrons for 10 hours. The total neutron yield at the target was 7.31xl015. The ten kinds of reaction rates such as 5 6Fe(n, p), 2 7Al(n, a), and 5 8Ni(n, 2n) were measured using foil activation method. The detector was placed at various points ranging from the front surface to the 0.95m depth along the central axis in the assembly. The obtained reaction rate distribution is shown in Fig.6.2.1. For high threshold reactions, reaction rates decreased exponentially to the fourth powers, while the 1 9 7Au(n, y)

reaction decreased to the second powers. Calculations were carried out using the two-dimensional transport calculation code DOT 3.5 with a 42 group structure library based on ENDF/B-III (GICX40) and 125 group structure ones based on ENDF/B-IV, JENDL-3PR1 and JENDL-3T.

The reaction rates were compared between the experiment and the calculations. Calculated to experimental (C/E) values are shown in Fig.6.2.2(aH(d) for each libraly. In the case of GICX40 (42 group), calcuraced rates conformed closely to measured ones for 2 7Al(n,a), 9 3Nb(n, 2n), and S 6Fe(n, p) reactions, which had large sensitivities around 14 MeV neutrons. However, for the other reactions C/E values varied from 0.8 to 1.2 even at the front surface where the 14 MeV neutron flux dominates the field. There existed discrepancies for the prediction of neutron spectrum usin GICX40, because consistencies of several reactions, whose sensitive energy regions were different, could not be obtained. The results of calculation for the 125 group

JAERI-M 88・221

6.2 Measurement and Ana1ysis on Reactlon-Rate Distribution

in a Large 1ron Cy1indrica1 Assemb1y

* * K. Oishi ， Y. Ikeda， C. Konno， K. To阻止oka and T. Nakamura

Iron is one of the most 1mportant matel1:'1a1s 1n fus10n fac111t1es，

not on1y as a structura1 mater1a1， but a1so for sh1e1d1ng. However， at

present， there ex1st many d1fferences among nur.1ear data 11brar1es for

est1mat1ng the neutron spectrum 1n 1ron assemb11es1. D1sagreements

between experimenta1 and calculated resu1ts was a1so reported by

Hertel et a1.2 1n this study， measurements of reaction rate distr1bu-

tion were performed using FNS to verify currently ava11ab1e nuc1ear

ciata 11braries for 1ron.

A 1arge cylindrical 1ron assembly， 1m in diameter and 0.95m in

th1ckness， was set at O.2m from the D-T target a10ng the inc1dent d+

beam and was irrad1ated by 14 MeV neutrons for 10 hours. The tota1

neutron y1e1d at the target was 7.31x1015. The ten k1nds of react10n

rates such as 56Fe(n， p)， 27Al(n，α)， and seNi(n， 2n) were measured

us1ng fo11 act1vat10n method. The detector was placed at var10us

points rang1ng from the front surface to the 0.95m depth along the

central axis 1n the assembly. The obtained reaction rate dlstribution

1s shown 1n F1g.6.2.1. For high threshold reactions， react10n rates

decreased exponentially to the fourth powers， while the 197Au(n， y)

react10n decreased to the second powers. Calculations were carried out

using the two-dlmensional transport calculat10n code DOT 3.5 with a 42

group structure 11brary based on ENDF/B-111 (GICX40) and 125 group

structure ones based on ENDF/B-1V， JENDL-3PRl and JENDL-3T.

The reaction rates were compared between the experiment and the

calculations. Calculated to experimental (C/E) values are shown 1n

Fig.6.2.2(a)~(d) for each libraly. 1n the case of GICX40 (42 group)，

calcurated rates conformed closely to measured ones for 27Al(n，α) ，

93Nb(n， 2n)， and s6Fe(n， p) reactions， which had large sensitiv1ties

around 14 MeV neutrons. However， for the other react10ns C/E values

varied from 0.8 to 1..'2 even at the front surface where the 14 MeV

neutron flux dominates the field. There existed discrepancies for the

prediction of neutron spectrum us1n GICX40， because cons1stencies of

several reactions， whose sensiti~e energy regions wexe different，

could not be obtained. The results of calculation for the 125 group

JAERI-M 88-221

structure libraries using ENDF/B-IV and JENDL-3T agreed well each other within 10%, but the differences between the measured and the calculated values increased as the detector position became deeper in the assembly. Large underestimations of calculated reaction rates by about 0.5 using JENDL-3PR1 were observed at the detector position 2-0.5m for the high threshold reactions. In general, for all libraries, underestimations of calculated result to experimental one increased as the depth of detector position increased. The reason was probably due to uncertainties of Fe cross sections in the libraries. It is expected that a small uncertainty in the cross section, ever though it is less than ±5%, generates a large difference in the transport calculation according to the increase in the thickness of the assembly. The effect of multiple scattering in the assembly was very large, because the mean free path of 14 MeV neutrons in the iron is about 5cm.

As a result, we can conculude that not only the proper estimation of cross sections, but also benchmark experiments at the deep position to obtain a practical factor for the adjustment of the calculated result to experimental ones are very essential for the shielding of fusion reactors.

[ This work was conducted under the Cooperatice Program between JAERI and Shimizu Corporation.]

Reference 1) Mori S. et al., : JAERI-M 87-083 2) Hertel N.E. et al., : " Transmission of Fast Neutrons Through an

Iron Sphere," Fusion Technol. vol.9. 345, Mar. 1986 10'

'"""---, J

Fig. 6.2.1 Reaction rate distribution based on experiments

}AERI-M 88・221

structure 1ibraries using ENDF/B-IV and JENDL-3T agreed we11 each

other with1n 10%， but the differences between the measured and the

ca1cu1ated va1ues 1ncreased as the detector pos1t1on became deeper in

the assemb1y. Large underest1mat1ons of ca1cu1ated react10n rates by

about 0.5 using JENDL-3PRl were observed at the detector position

Z-O.5m for the high thresho1d reactions. In genera1， for a11

1ibrar1es， underest1mat1ons of ca1cu1ated result to experimental one

increased as the depth of detector position 1ncreased. The reason was

probably due to uncertaint1es of Fe cross sect10ns 1n the 11brar1es.

It is expected that a sma11 uncerta1nty in the cross sect1on， ever

though it 1s 1ess than ~5% ， generates a 1arge difference in the

transport ca1cu1at1on according to the increase in the thickness of

the assembly. The effect of multip1e scatter1ng in the assemb1y was

very 1arge， because the mean free path of 14 MeV neutrons in the iron

is about 5cm.

As a resu1t， we can concu1ude that not only the proper est1mation

of cross sections， but a1so benchmark experiments at the deep position

to obtain a practica1 factor for che adjustment of the ca1culated

resu1t to experimenta1 ones are very essentia1 for the shie1d1ng of

fusion reactors.

[ This work was conducted under the Cooperat1ce Program between

JAERI and Shimizu Corporation.]

Reference

1) Mori S. et a1.， JAERI-M 87-083

2) Herte1 N.E. et a1.， " Transm1ssion or Fast Neutrons Through an

Iron Sphere，" Fusion Techno1._vo1.9，345， Mar. 1986

:、い¥三部 ji r2ti 、氏、.三三;3151]

Reaction rち ted1str・ibutioobased 00 exper-i田eats

Fig. 6.2. 1 li.32; 、-j

oj人 1-153-

J A E R I - M 8 8 - 2 2 1

2.0 2.0

(a) (b)

! , In In.n*) I ' l l In.C\ "W In.In) !!'*uln.T) "Zr ln.2nl " » l n . J n l

7- ' - - . "Tl ln.pl

— » — "Zi fa .p j . . - . . - . •'iflln.nl

- 2 0 0 200 +00 Z ( mm J

600 800 1000 - 2 0 0 1000

Fig. 6.2.2 C/E values of reaction rate. Calculation was perforned using (a) ENDF/B-III (42group), (b) ENDF/B-IV, (c) JENDL-3PR1, and (d) JENDL-3T

JAERI-M 88・221

2.0 ， 2.0 一一-Thin-91---0・ー ii:.I~ (fI':.~ '1 -・ー・ー-;í~! !n.~) ーー『四ーー，.叫，hiln.. znl -ーー'b---，. ln.TI

1.5ト;;~ !o ， ~o!

曹¥ 町司甲ー田ーー，ーーーー-ー-叫 T司，』hsn.'H 4n，イ 1.5トー-ー・ 4・抽10.Z.) -田ー・・‘Ti(o.p) -一・一 ;;!f， !.，p!

払EJーー・~司ー， --NI (品.p)

'r: 弘J

¥1.0 ι2

ロ v、、0.5

0.0 -200 D 200 4.00 600 800 1000 -200 。200 4.00 600 800 1000

Z ( mm ) Z ( IM¥ )

(a) (b)

2.0 2.0

一円台町In.，・・・・・・4 s1e1g -ZElm.9l

一-0・・;;:!~ (.._~'I -4--2F ln ln.i1

-ー・ー・AII...1 ーーーー---S EAl in .. a l

一一 I，UJRJ 1，phJAt-U In.28 1 ーーー・ー ii~~ (n~T! ーーーー-ー ;å~!~(向， TI 一........・;~~".~，，~ ー--ーーー.;;~! ~n.~" ~

1.5ト‘.出'..Zol 1.5

ーー唱ー..柚 (n.2，， 1

-争噌ー・・‘Ti '..，1 -一--;Jih ?-1...，.."zh tn.，i ;;~ !"'! -・宇_-:J、、一-・・一 !II'o.p 1 ーーー咽.-ー.......Ni 1t. .~J .、、な守、、--' 目可Z、‘

:10卜 aa型四，豆・、4・~:、、' k - 、彦、、‘

0.5トまえぎ

、¥J 0.5

.. . . +

内Hu・

• AH 内

nH】v• 肉

uAH】w•

。。内

。ーωm

U司，-nu

0.0 ・200

'・nHV

内向M

角川u

nHM・

勾，‘内Hu'

Fig. 6.2.2 C/E vヨluesof 問 actionI"3te. calculation岨 S 陣 rfor'll凶 using

(a) ENDF/B-III (42group)， (b) ENDF/B-IV， (c) JENDL-3PR1， and

(d) JENDl-3T

J A E R I - M 8 8 - 2 2 1

6.3 A Comparative Study of Shielding Design Between Advanced Marine Reactors in Concept

T. Ise and Y. I ton*

We have performed a comparative study of shielding design for three advanced marine reactors in concept(semi-integral PWR, integral PWR and self-pressurized integral PWR) at operational condition and hypothetical accident through numerical evaluation. Shielding design of the semi-integral PWR(see Fig. 6.3.1 (a))" resembles that of the looped PWR(the existing PWR) and shielding design of the integral PWR(see Fig. 6.3.1 (b)) u looks like that of self-pressurized integral PWR(refer to Reference 1 about detailed shielding designs such as thicknesses and materials of shield). So here are shown only the results of comparison between the semi-integral PWR and the integral PWR.

Common calculational procedure and shielding material ingredient between the reactors have been adopted for fair evaluation(see Fig. 6.3.2 about the calculation flow). Main results of numerical comparison for operational condition and hypothetical accident are listed in Table 6.3.1 and gamma-ray dose distributions for (r, z) geometry are illustrated in Fig. 6.3.3. The results show that the integral PWR is more appropriate than the semi-integral PWR in view of shielding design, although not here discussed about duct streaming through main coolant pipe.

Reference

1) Ise T., Narita H. and Itoh Y.

Table 6.3.1 Numerical comparison of shielding design

(A) at operational conditiondnrem/h) Semi-integral PVR Integral PWR

Primary shield

side 3 lxlO z

6x10 side n 3x10"

Secondary 7 2x10" shield MCPC'N) 7 2x10'

3x10" 2x10-

top n 2x10"' 9X10"3

7 8xl0- 2 3xl0" s

(8) at hypothetical accident(rem) Secondary 7 6 4 shield Note: not contain streaming front main coolant pipe

(by D0T3.5 with P5.S100).

Evaluation of Trial Design Studies for an Advanced Marine Reactor (7) - Shielding (I): Evaluation of Shielding Design-," JAERI-M 88-055(1988)(in Japanese).

+ Office of Nuclear Ship Research and Development

jAERI-M 88・221

6.3 A Comparative Study of Shielding Design Between Advanced

トlarineReactors in Concept

T.Ise釦 dY.ltoh'

We have perforrned a cornparative study of shielding design for three advanced

rnarine reactors in concept(semi-integral PWR， integral P聞 andself-pressurized

integral PWR) at ope~ational condition and hypothetical accident through numeri-

cal evaluation. Shielding design of the semi-integral PWR(see Fig. 6.3.1 (a))I)

resembles that of the looped PWR(the existing PWR) and shielding design of the

integral PWR(see Fig. 6.3.1 (b)) Il looks like that of self-pressurized integral

PWR(refer to Reference 1 about detailed shielding designs such as thicknesses

回 drnaterials of shield). So here are shown only the results of comparison be-

t嶋田 thesemi-integral PWR and the integral P問.

Cω~on calculational procedure and shielding rnaterial ingredient between the

r回 ctorshave been adopted for fair evaluation(see Fig.6.3.2 about the calcula-

tion flow). Main results of numerical comparison for operational condition釦 d

hypothetical accident are list-

ed in Table 6.3.1加 d胆 runa-ray

dose distributions for (r，z)

geometry are il1ustrated in Fig.

6.3.3. The results show that the

int唱ralPWR is more appropriate

than the semi-integral PWR in

Table 6.3.1 Numerical comparison

of shielding design

(A) at ooeratiQnaI condition(mrem/h)

Primary

shield

Semi-integraI附RIntegraI附R3 5xlO' lxlO' 6xlO 3xlO-‘ 3xlO-' 2xlO-I 2xlO・2

side n

Secondary view of shielding design， aト shield

though not here discussed about

duct streaming through rnain coo-

lant pipe.

Reference

担丘三旦」top n

nv-nu内

--a---EA'EA

van-vAVA

-q'uoo

9xlO・3

3xlO・2

{B) at hYPOthetical accident(re~)

se∞ndary 6 4

shield Note: not contain streaAing [roll main c∞lant pipe

(by DOT3.5削 thP..5100).

1) 1 se T.， Nar i ta H.叩 d1 toh Y.:“Evaluation of Trial Design Studies for an Ad-唱 ncedMarine Reactor (7)ー Shielding(1): Evaluation of Shielding Design-，"

JA回トM88-055(1988)(in Japanese).

+ぽficeof Nuclear Ship Research and Development

J A E R I - M 8 8 - 2 2 1

ship bottom shield

side shield

bottom shield

(a) Semi-integral PWR

secondary shield

containment

•iser

blowoff tank pressure balance

valve nary shield

(b) Integral PWR

Fig. 6.3.1 Shielding design of marine reactors

V I T A M I N - C

A M P E X - I I

crosssection library for AHPX-!U7ln.36r)

processing for ANISN calculation

calculation of neutron and gamma-ray spectra (171n,367;P s.S,i),and production of group constants calculation of neutron and gamma-ray fluxes, and dose rates (8n.4T:Ps,S48/P5.S100)

Q A D - C C

gamma-ray source in primary coolant

calculation of gamma-ray dose rates for three-dimensional geometry

OR 1 G E N - 2

(A) at operational condition

calculation of gama-ray

source according to core burn-up

release into containment

Q A D - C G

Semi-integral PWR:Integral PWR noble gas:100X.halogen:50t others:1J

Self-pressurised PWR noble fas:0.1X.halogen:0.05X others: 0.001 J

calculation of dose rates and integrated dose for S days

(B) at hypothetical accident

Fig. 6.3.2 Shielding calculation flow

JAERI-M 88-221

(a) Semi-integral PWR (b) Integra1開R

Fig. 6.3.1 Shielding design of marine reactors

crosssect ion li brary

for Iu'IPX-I(l'7ln.36，)

S出国叫aryshield

process ing for

AIIISN calculation ? gamma-ray source 1n

pri旭町 c∞lant

calculation of neutron and g.掴祖-rays関ctra

(l71n. 367: P.. S，，).ωd

production of group

constants 国 lculationof neutron

副首ig副_-rayfluxes.

副首id国 erates

(8n. 47 :P.. 548/P.. 5100)

l QAD-CG I

(A) at operational condition

calculation of ga闘lIl-ray

source according to core

calculation of gallllll8.-

ray dose rat~s for

three-dimensional

geo鵬 try

加rn-up印刷・intes同 l附R:lntesral附Rnぬleps:l00%.hal何回:50%

rel担調印toC開 tau鴨川 0凶町$:1%

SelC-pressurised阿R

noble ps:O.l%.haloll凹 :0.05%

。凶ers:札ω11

calculation of dose rates and

integrated dose for 5 days

(B) at hypothetical accident

Fig. 6.3.2 Shielding calculation f10w

一156-

*—•

1 p fD

3 *.—

c o -1 •o 3 -j

distance from

core c

enter(cm)

1130.∞ 1112.∞

IOJ6.ω

"1.20 950.ω 9l8.∞

』〉切河

-ap目白

.. 9 . d . 、

8ilil.'

L、+ト斗回市淀刊りげ1hSH

∞∞

ω∞∞個初∞伺却∞曲曲

m∞曲羽田知朗純白田崎柿崎崎川同

ωω刊柑

剛一町制問削

mmmm一町一印刷一制抑制哨一町一川仰山一山山叩制

m一期制

mMm知加抑制

(g)包言8285占

832士宮H4

-】切吋

EEZ ~~。

「¥“

日凶!川lf.l肌f.I山~ïl._一一一-よ一」山………L山…'-L…Lいι…ιいム…

21・.226.・

.... ".‘ 0.0

200.90 。l2.・O12・.ω125.∞ 山 .ω81.ω 81.40 7~.∞

Integal PWR

Gamma-ray dose rate distributions for upper-half pan of the reactor

、‘，JZ

D ，，.、

Fig. 6.3.3

JAERI-M 88-221

6.4 Design Method of Compensational Shield for Shield Irregularities of Reprocessing Plant

A. Yantaji* and T. Suzuki

Design method has been studied for Iron condensational shield for slant duct and slit, straight slit, and offset slit In concrete shield wall against gamma radiation to compensate lowering of shielding efficiency caused by these shield Irregularities, with applying similar design concept used in the previous works of the compensational shield for straight duct 1', double bent duct a >, partially penetrating duct a > and the German Nuclear Standard for the design of double bent duct 4 1. The characteristics of the present design method are to compensate lowering of the shielding efficiency without increasing the concrete wall thickness and to determine the dimensions of the compensational shield using the densities of the concrete and iron, the duct diameter, the slit width and the concrete wall thickness independent of the incident gnmma-ray energy. The calculations of the dimensions of the compensatlonal shield are described In reference 5). Experimental verification of the present design method has been carried out using the research reactor JRR4. Figure 6.4.1 shows the experimental geometry where the condensational shield is inserted in the irregularities of the shield wall and acts as a duct wall for straight duct, and a slit wall for slant and straight slits. The amount of radiation behind the shield wall were reduced effectively by the compensational shield, and the dose rates higher than that of the bulk shield wall were restricted only in the area near the slit exits, as shown in Figs.6.4.2-5 B ). The experiments were analyzed using a multigroup single scattering code G33-GP S > in order to assess the applicability of the code to these geometries. The calcu-lational results are also shown in Figs.G.4.2-5. The agreement between the calculations and the experiments is good. The G33-GP code was then applied to calculations for reprocessing plant configurations with various incident gamma-ray energies, Incident angles, concrete wall thicknesses and shield irregularities. The calculations showed that the shielding performance of the shield irregurarities with compensational shield is approximately equal with that of a bulk shield wall s ).

» Ship Research Institute

JAERI・M 88司 221

6.4 Design Method of Compensational Shield for Shield

Irregularities of Reprocessing Plant

A. Yamnji-and T. Suzukl

Design mctho(! has been studled for iron compcnsatlonlll shicld for

slant duct and sllt， stralght sllt， and offsct sllt ln concrctc shlcld

wllll agalnst gllmma rllcllatlon to compcnsatc .lowerlng of shlcldlng effl-

ciency caused by these shleld lrregularltles. wlth applylng slmllar dc-

slgn concept used ln the prevlous works of the compensational shlcld for

straight duct 11 • doub.le bent duct 21 • partiallY penetrating duct31 and

the German Nuclear Standard for the design of double bent duct 41 • Thc

characteristics of thc present design method are to compensate lowcrfng

of the shielding efficiency without increasing the concrete wal1 thick-

ncss and to determinc the dimensions of the compensatlonal shield using

thc densitles of the concrete and lron. the duct diameter. the slft

width and the concrete wall thickness indcpendent of the incldcnt gnmma-

ray cnergy. The calculatlons of the dlmenslons of the compensatlonal

shield are descrlhed ln reference 5). Experlmental verlflcatlon of thc

present design mcthod has been carrled out uslng the rescarch rcactor

JRR4. Figure 6.4.1 shows the experllcntal geomctry where the compcnsa-

tional shield 1s inscrted in the irregularities of the shield wall and

acts as a duct wall for straight duct. and a slit wall for slant and

straight slits. The amount of radlation behlnd the shleld wall were rc-

duced effectlvely hy the compensat1onal ship.ld. and the dose rates h1gh-

er than that of the bulk sh1eld wall were restrlcted only in the area

ncar the slit ex1ts， as shown 1n Figs.6.4.2-551 • The exper1ments ~cre

analyzed using a田ultigroupslngle scattering code G33-GpSI in ordcr

to assess the apP1icabillty of the code to these geometrics. The calcu-

lational results are a1so shown 1n Figs.6.4.2-5. The agreement bctween

the calculations and the experi.ents 1s good. The G33・GPcode was then

applied to calculations for reprocessing plant conflguratlons with

various 1ncident gam聞a-rayenerg1es， incldent angles. concrete wall

thicknesses and shield irregularlties. The calculations showed that the

shielding perfor聞anceof the shleld il・regurarltieswith compensational

shield is approximately equal with that of a bulk shield wal15).

* Ship Research Institute

-1同一

JAERI-M 88-221

References 1) Yaraaji A., Nannta S. and Salto T.: J. At. Energy Soc. Jpn.dn

Japanese),29[6], 555(1987). 2) Ficbig R. and Yamajl A.: Atomkerncncrgic,22[2], 105(1973). 3) Fiebig R. and Yamnj I A.: Atorakerncncrglc.22[2|, 113(1973). 4) DIN Dcutschcs Instltut fuer Nornung c.V.: DIN25427,(1977), Bcuth

Verlag. 5) Yamaji A.:"Deslgn Method of Compcnsatlonnl Shield for Shield

Irregularities of Reprocessing Plant," to be submitted to 7th International Conference on Radiation Shielding (1988).

6) G33-GP: RSIC Computer Code Collection, CCC-494. (1986).

JRR4 Core (405Lx34.4wx60' / Sfs

ms ^s E I -fcr -ft

No. I PQPi

j , 150 T (Vtofer)^ 7 7^

•JJL. / ////XTA f//f//ft

••01 g Compensational §i,f /Shield

Shield Wall

Core Tank

W;/VA (Air) i ////////y/ 150

(Concrete )

Experimental Hole

Compensat ional S h i e l d

Dimensions in cm , , , „„ (3 ) Offset Slit

F i g . 6 . 4 . 1 Exper imenta l geometry

- 1 5 9 -

jAERI-M 88・221

Rcfcrcnccs

1) YamaJl A.. XJ悶ntaS. and Salto T.: J. At. Encrgy Soc. .lpn. (ln

Japancsc).辺[6).555(1987).

2) Flcblg R. and YamaJl A.: Atomkerncncrgle.22(2). 105(1073).

3) Flcblg R. and Yamnjl A.: Atomkerncncrglc.22[21. 113(1973).

4) DI~ Dcutschcs Instltut fucr Nor聞ungc.V.: DIS25427. (1977). scuth

Vcrlag.

5) YamaJl A. :"Dcslgn河cthodof Co~pcnsatlonnl Shlcld fつrShlcld

Irregularltics of Reprocesslng Plant." to be sub冊ittedto 7th

International Conference on Radlatlon Shlcldlng (1988).

6) G33-GP: RSIC Cornputer Code Collectlon， CCC-494. (1986).

Core Tank

、;マZ401

(Concrefe ) ( 1) Slonf Duct & Slit¥

Expenmentol

Dimenslons in cm

Fig. 6.4.1 Experimental geometry

-159一

Meosurlng Line

2) Stroighf Slit /

J A E R I - M 8 8 - 2 2 1

c? 1.5

i 1 r D = 3.5cm

• Experiment o G 33-GP

0 = 6.2 cm

-20 0 20 40 60 80 100 X (cm)

Fig. 6.A.2 Ratio of dose rates behind the wall with and without the slant duct. The slant angle 9»30°, and the duct diameter D-3.5cm and b.2crr.. o

.2 0.8 -

• Experiment o G 3 3 - G P

=3=^®-

o 1.5 I

08 fl=30°

-g-Q-»-Q p- -o ij y a — _ ] I I I I I I ' ' ' ' '

0 20 40 60 80 100 X (cm)

Fig. 6.A.3 Ratio of dose rates behind the wall with and without the slant slit with 11 mm in width.

_ ^ 0 = 2 ° ' r

• Experiment OG 33-GP

: I i I 1 1 1 : \ 8 • °°

1 H //

• 61 -

! )l\p /A

it if 1

' * Experiment y\y .OG 33-GP eg

-30-15 0 15 30 X (cm)

6.A.A Ratio of dose rates behind wall with and without the straight with 9mm in width.

0.01 : E i i i i i i _ ... -40 -20 0 20 40

X (cm) Fig. 6.A.5 Ratio of dose rates behind the wall with and without the offset slit with 27mm in width.

只 8= 15。

1 ト・~叩→c.-r-宇-=eo 0.8 c 〆守、、、....

0:: 1.5汗 J8 = 3o.v

1十-Q=Q畠・cQ:-:Q::----:Q::----:9;:----Q-~ιー@→0.81 川

0. 20. 40. 60. 80. 10.0.

X (cm) F~g. 6.4.3 Ratio of dose rates behind the wa11 with and w1thout the slanc s1it with 11 mw 1n width.

目 30 -15 0 15 30

X (cm)

Fig. 6.4.4 Ratio of dose rates behind the wa11 with and without the straight 51it with 9mm in width.

星1.5

]AERI-M 88・221

1 5 jr--.---，---r----r--，--，---，----，

1~宇~弘一 0= 3.5 cm _= a:Al I ト←----・，、m ---..Qー_-0-一+寺'・ 4

0.7トー~.2 芋・ Experiment ~ & 15 ~ 0 G 33 -GP

I ' 0 = 6.2 cm

-20 0 20 40 60 80 1 00

X (cm) Fig. 6.4.2 Ratio of dose rates behind the wa11 with and without the slant duct. The s1ant ang1e e橿 30.，and the duct diameter D・3.5cmand b.2cr.:. 2

1.5 • Experiment o G33 -GP

• Experiment

o G 33-GP

。--田町民

-40.・20 0 20 40.

X (cm I

Fig. 6.4.5 Ratio of dose rates behind the wa1l with aod without the offset s1it with 27mm io width.

JAERI-M 88-221

6.5 Physics and Shielding Design of High Quality Neutron Beam Hole in a Research Reactor — Comparison Between Tangential and Radial Beams —

T. Ise, T.Maruo* and H.Kawasaki*

It is said that the tangential beam hole, which does not look at, reactor core, can produce high quality neutron beam, that is, high thermal neutron flux with low background noise. Here are shown some advantages of the tangential beam over the conventional radial beam through numerical comparison.

The calculational procedure11 combining the discrete ordinates method(D0T 3.5) and the Monte Carlo method(MORSE-CG) has been adopted considering physics( neutron flux) and shielding(dose rate) calculations of duct streaming in complicated geometry, as shown in Fig. 6.5.1. Calculational geometry for a beam hole is illustrated in Fig. 6.5.2. It is assumed that both a tangential beam hole and a radial beam hole are installed in the upgraded JRR-3 research reactor(HTR-type fueled, light water cooled and heavy water reflected reactor), as depicted in Fig. 6.5.3. The notations A and B shown in the figure are the locations for evaluating thermal flux and dose rate, respectively.

Figure 6.5.4 shows that the thermal neutron fluxes are almost equivalent between the tangential and radial beams but the fast neutron fluxes are visibly different, in particular, the fastest neutron fluxes are. Figure 6.5.5 displays that gamma-ray spectra are extensively different between the tangential and radial beams. Table 6.5.1 summarizes the comparisons of thermal neutron flux and dose rate. The results shows that the tangential beam hole can produce high thermal neutron flux with minimal fast neutron and gamma-ray contamination, as expected.

Reference

1) Ise T., Maruo T. et al. :"Shielding Analyses for Design of the Upgraded JRR-3 Research Reactor • 2 — Shielding of Neutron Beam Holes — , " JAERI-M 85-105 (1985)(in Japanese).

+ Department of Research Reactor Operation * Century Research Center Corp.

JAERI-M 88・221

6.5 Physics and Shielding Design of High Quality Neutron Beam

Hole in a Research Reactor

ーー ComparisonBetween Tangential and Radial Beams --

T. Ise， T.Maruo. and H.Kawasaki"

It is said that the tangential国amhole， which does not look at reactor

core， can produce high quality neutron beam， that is， high thermal neutron flux

with low background noise. Here are shown some advantages of the tangential凶訓

over the conventional radial国訓 throughnumerical comparison.

The calculational procedureLI combining the discrete ordinates method(DOT

3.5) and the Monte Carlo method(MORSE-CG) has been adopted considering physics(

neutron flux) and shielaing(dose rate) calculations of duct streaming in comp-

1 icat剖 geometry，as shown in Fig. 6.5.1. Calculational geometry for a be訓 hole

is illustrated in Fig. 6.5.2. It is assumed that both a tangential be訓 holeand

a radial beam hole are installed in the up~radeà JRR-3 research reactor(MTR-type

fueled， light water cooled and heavy water reflected reactor)， as depicted in

Fig. 6.5.3. The notations A and B shown in the fi別 reare the locations for eva-

luating thermal flux and dose rate， respectively.

Figure 6.5.4 shows that the thermal neutron fluxes are almost equivalent

between the tangential and radial beams but the fast neutron fluxes are visibly

different， in particular， the fastest neutron fluxes are. Figure 6.5.5 disp!ays

that gamma-ray spectra are extensively different between the tangential and ra-

dial回ams.Table 6.5.1 summarizes the comparisons of thermal neutron flux and

dose rate. The results shows that the tangential beam hole can produce high

thermal neutron flux with minimal fast neutron and gamma-ray contamination. as

ex開cted.

Reference

1) Ise T.， Maruo T. et al. :“Shielding Analyses for Design of the Upgraded JRR-3

Research Reactor . 2一-Shielding of Neutron Beam Holes ---，" JAERI羽田島105

(985) (in Japanese).

+ De回rtmentof Research Reactor 0問 ration

J A E R I - M 8 8 - 2 2 1

/oic-«tc\

AMPX-I

OOT 3.5

Nuclear Cross Section Library ( I 7 1 n , 3 6 r l

Processing Nudeor Cress Sections

Producing Group Conitonls ( S n , 7 r ]

Radiation Distribution Around the Reactor Core

j j ™ , . . . Rodiation Flux through w * o t " U l Reoclor Pool with Beam Hole

OOT 3.5 Rodioticn Distribution through Reactor Wall with Beam Ho!e

Auxiliary Shield

Fig. 6.5.1 Calculational flow

Fig. 6.5.2 Calculational geometry (e.g.: tangential beam)

87.78 109.54

i ' 2 -J 1—5.5 ^~Aluminum Ailoy

Heavy Woter J \ _ z R M e t of p 0 0 |

(a) Tongentiol Hole

75 109.54

Aluminum Alloy

Reactor Pool

lb) Radial Hole

Dimensions in cm.

Fig. 6.5.3 G-eonetrical configuration of tangential and radial bea» holes (location A: to compare fluxes; location B: to compare dose rates)

]AERl‘M 88・221

州£梅宮 C同盟SICt刷 Ubt司円(171 n ， 361 I

Pr田担問陥岨回rC咽 S皿1-

内包也凪吋 Gro.咽 C朗11叩 11(a n • 711

R凶岨f刷、。同Irl凶 lionAroo.凪d1111 R伺C附 Cor・Rω同liOn円ux 1附加gh雨明町内d酬 IhB凹m 陶1e

Rodi副田nOisuibu1ion lhr明削除岡町 WolIwilh Beom Ho:"

OOT 3.5

Fig. 6.5.2 Calculational geometry

(e. g.: tangential beam) Fig. 6.5.1 Calculational flow

109.54

「-一Aluminum Ailoy

Reoclor Pool

Aluminum Alloy

30 101 Tonqenliol Hole

Oimensions in cm.

白巴

4If--・2.

Fig. 6.5.3 Geometrical configuration of tangential and radial随掴 holes

(location A: to comparc fluxes; location B: to co.pare dose rates)

109.54

一162-

J A E R I - M 8 8 - 2 2 1

10" • Tongentiol

Radiol i

- i o ' ° ' •. E

1 s,o- r 1 2

• _ _ _ ;

<" 1 0 *

*) Z I 0 '

• _ _ _ ;

<" 1 0 *

*) Z I 0 ' •

10' — i

i n » • . > <

8 7 6 5 4 3 2 1

Energy Group Number

7 6 5 4 3 2

Energy Group Number

6.5.4 Comparison of neutron spectra Fig. 6.5.5 Comparison of gamma-ray spectra (group no. 8: thermal flux) / gamma-ray energy degradation

\ with increase of group no. /

Table 6.5.1 Comparison of thermal neutron flux and dose rate between tangential and radial beam holes

Type of beam

Thermal flux (n/cm2 • s)

Dose rate(mrem/h) Type of beam

Thermal flux (n/cm2 • s) Neutron Gamma-ray

Tangential 8.2xl0 1 0 1.6 4.3x10

Radial 8.7x10'° 4.6x10 5.3x102

Tangential Radial

]AERI-M 88・221

10"トー一-TOIlgenl岨1

・ーー-Rod咽l 一一 T町制iol

τ10'0

E 、、S 10'

310・s

星101

IO~ 8765432

Energy Gr<凹 pNumber

ー・『

-ーー-Rodiol

;叩L-UE 3 10同

E ，- 10'

一3 10 ・

7 6 5 4 3 2

Energy Group ~umber

Fig. 6.5.4 Comparison of neutron spectra Fig. 6.5.5 Com開 risonof g創runa-rayspectra

(group no. 8: thermal flux) I gamma-ray energy àegradatio~

¥ with increase of group no. }

Table 6.5.1 Comparison of thermal neutron flux卸 ddose

rate between tangential and radial同訓 holes

Type of Thermal fl ux Dose rate(mrem/h)

出創E (n/cm2 • s) Neutron G剖¥s!a-ray

Tangential 8.2x1010 1.6 4.3xlO

Radial 8.7xlOIO 4.6x10 5.3xl02

Tangential 1 l 一一一

Radial 1.1 29 12

JAERI-M 88-221

6.6 Exposure Buildup Factors for Slant Penetration through Slab Shields from Point Isotopic Gamma-Ray Source

Y. Kanai", T. Suzuki, .V. Xarlyama" and K. Takeuchi-

For practical calculation in gamma-ray shield design, the point-kernel codes, such as QAD and G-33, always use buildup factors for a point isotropic source in an infinite medium. In recent years the standard reference data of the exposure buildup factors have been provided by American Nuclear Society standards committee working group ANS-S.4.3.

The point-kernel calculations using the buildup factors for a point source in an Infinite medium provide considerable underestimation with greater obliquity for gamma-ray slant penetration through a slab shield. There have been, however, no buildup factors usable in practical calculations for slant penetration through a slab fro* a point isotropic source.

To find a solution to this problem, we applied a two-dimensional discrete ordinates direct integration code, PALLAS-2DCY-FX1>, to calculations of gamma-ray penetration through slabs from point isotropic sources. The PALLAS calculations were made in two-dimensional cylindrical model with an adequately large radius and with the same thickness as that of the slab. A penetration angle is defined as the angle between the line connecting the source point and detector point and the normal line to the slab. For PALLAS calculations the source point was fixed at the origin and detector points were selected as the radius mesh points corresponding to the specified penetration angles.

Before applying to systematic calculations of the buildup factors, we verified the accuracy of PALLAS calculations by comparing with measurements carried out by Kanemori2'. The experiments were made using several slab shields for a Co-60 point source, and exposure buildup factors have been given as a function of penetration angle. To confirm the applicapable limitation of the buildup factor for a point isotropic source in an infinite medium, QAD-CGGP3' and G-33GP"*' calculations were also performed for an iron slab of 11.64 cm.

• Ship Reseach Institute

jAERI-M 88・221

6.6 Exposure Buildup Factors for Slant Penetration through

Slab Shields Erom Point Isotopic Gamma-Ray Source

Y. Kanal-. T. Suzukl. N. Nariyama-and K. Takeuchl-

For pract1cal calculat10n 1n gamma-ray shleld deslgn. t1，e polnt-

kerncl codcs. such as QAD and G・33. always use bu!ldup factors for a

polnt lsotroplc source 1n an 1nflnlte Icdlum. In reccnt years the

standard reference data of the exposure bUl1dup factors have bcen

provlded by Amerlcan Nuclear Soclety standards commlttee worklng group

ANS・6.4.3.

The point-kernel calculatlons uSlng the bUlldup factors for a polnt

source in an lnflnlte mediuI provlde conslderable underestiaatlon wlth

greater obliquity for gamma-ray slant penetratlon through a slab sh1cld.

There have becn. however. no buildup factors usable in practlcal calcu-

lations for slant penetration through a slab frol a polnt Isotropic

source.

To flnd a solution to th1s proble踊. we app11ed a two-dimcnslonal

dlscretc ordlnates dlrect lntegratlon code. PALLAS-2DCY-FX1). to

calculatlons of gRmMa-ray penetratlon through slabs frol polnt Isotropic

sources. The PALLAS calculatlons 1Vere ・ade ln two-di.enslonal

cYllndrlcal .odel wlth an adequately large radlus and wlth the sa踊e

thickness as that of the slab. A penetration angle 1s def1ned as the

angle between the llne connectlng the source point and detector point

and the norlal 11ne to the slab. For PALLAS calculatlons the source

pOint was fixed at the origln and detector po!nts were selected as the

rad1us lesh points correspondlng to the specified penetratlon a~gles.

Before applylng to syste・atlccalculatlons of the buildup factors.

we verified the accuracy of PALLAS calculatlons by cOlparing with

leasurements carried out by Kanelori2). The eKperllents were lade using

several slab shlelds for a Co-60 polnt source. and exposure buildup

factors have been glven as a function of penetratlon angle. To conflr祖

the applicapable li.itatlon of the buildup factor for a point 1sotropic

source in an lnflnlte lediuI. QAD-CGGP31 and G-33GP41 calculatIons were

also perfor・edfor an iron slab of 11.64 c..

• Ship Reseach Instltute

J A E R I - M 8 8 - 2 2 1

Figure 6.6.1 shows a comparison of calculations by the PALLAS, QAD and G-33 with the experimental values of the exposure buildup factors for 11.64 cm thick iron slab as a function of penetraion angle. For further verification of the accuracy of PALLAS, the other comparisons of the PALLAS calculations with experiments in the case of 22.8-, 9.6- and 4.8-cm-thick aluminum slabs are presented in Fig.6.6.2. Although PALLAS calculations depict the ripples of the ray effects In thinner slabs, fairly good agreement is obtained between PALLAS calculations and experiments except beyond 70 degree of 4.8 cm slab, where the PALLAS values is 23 * higher.

For a practical use, PALLAS calculated exposure buildup factors are provided for ordinary concrete slabs of thicknesses up to 10 mfp, for penetration angles in cosine from 1.0 to 0.25, and for source energies between 0.661 and 6.13 MeV. The exposure buildup factors are partially displayed only for lower two source energies in Table 6.6.1. In order to check the PALLAS calculational accuracy of the exposure buildup factors for ordinary concrete, the continuous energy Monte Carlo, MCNP 8 )

calculations were also made for 1 and 2 mfp thick slabs. The PALLAS results are In reasonable agreement.

The new data of the exposure buildup factors should be used to avoid extremely large underestimations encounterd often in <7ose-rate estimations outside slab shields calculated with the point-kernel code such as QAD using usual point source buildup factors in infinite medium.

References 1) Takeuchi K.. Sasamoto N. and Kanai Y.: "PALLAS-2DCY-FX: A Code for

Direct Integration of Transport Equation In Two-Dimensional (R,Z) Geometry," JAERI-M 84-244 (1985).

2) Kanemori Y.:"Measurements of Dose Buildup Factors in Slab Materials for Co-60 Isotropic Source," J. Atom. Eng. Soc. Japan, 8, 304 (1966).

3) QAD-CGGP: RSIC Computer Code Collectin CCC-493 (1986). 4) G33-GP: RSIC Computer Code Collection CCC-494 (1986). 5) "MCNP - A General Monte Carlo Code for Neutron and Photon

Transport," LANL (1981).

JAERI-M 88・221

Fl凹 re6.6.1 shows a co皿parlsonof calculatlons by the PALLAS. QAD

and G・33with the experi.ental values of the exposure bulldup factors

for 11.64 c~ thlck lron slab as a functlon of penetralon angle. For

furthtr verlflcatlon of the accuracy of PALLAS. the other comparlsons of

the PALLAS calculatlons with experlments ln the case of 22.8・， 9.6- and

4.8・cm-thlckaluJlll・IUIIslabs are presented ln Flg. 6.6.2. Al though PALLAS

calculatlons deplct the rj~ples of the ray effects ln thlnner slabs.

falrly good agreeJlent Is obtalned betwecn PALLAS calculatlons and

experlments except beyond 70 degree of 4.8 CJl s1ab. where the PALLAS

values 1s 23 % hlgher.

For a practical use， PALLAS calculated exposure bul1dup factors are

provided for ordinary concrete slabs of thicknesses up to 10 mfp， for

penetration angles in cosine fro. 1.0 to 0.25， and for source energles

between 0.661 and 6.13阿eV. The exposure blllldup factors are partially

displayed only for lower two source energles in Table 6.6.1. In order

to check the PALLAS calculatlonal accuracy of the exposure bul1dup

factors for ordinary concrete， the continuous energy Monte Carlo，問CNpS)

calculatlons were also圃adefor 1 and 2 mfp thlck slabs. The PALLAS

results are in reasonable agree.ent.

The new data of the exposure buildup factors should be used to

avoid extremely large underestl・atlonsencounterd often In ~Qse-rate

estimations outside slab shlelds calculated with the polnt-kernel code

such as QAD using IIsual polnt source buildup factors ln lnflnlte .ediu..

References

1) Takeuchl K.， Sasa・otoN. and Kanal Y.: "PALLAS-2DCY-FX: A Code for

Dlrect Integration of Transport Equation ln Two-Dl.ensional (R，Z)

GeoJletry，" JAERI-M 84-244 (1985).

2) Kane.orl Y.:"Measure.ents of Dose Bul1dup Factors ln Slab Materlals

for Co-60 Isotroplc Source，" J. Ato.. Eng. Soc. Japan， ~， 304

(1966) .

3) QAD-CGGP: RSIC Co・puter Code Col1ectln CCC・493(1986).

4) G33-GP: RSIC Co.puter Code Collection CCC-494 (1986).

5) "MCNP A General Monte Carlo Code for Neutron and Photon

Transport，" LANL (1981).

J A E R I - M 8 8 - 2 2 1

Table 6.6.1 Exposure buildup factors for concrete slabs 0.661MeV

mfp COS0 O «1 0 . 7 5 0 . 5 0 0 . 2 5

1 . 0 1 . 8 3 2 . 2 4 2 . 7 4 5 . 1 1 1 . 8 5 + 0 . 0 1 * 2 . 0 6 + 0 . 0 1 2 . 5 5 J 0 . 0 2 4 . 5 8 + 0 . 0 8

2 . 0 2 . 9 5 3 . 9 9 6 . 4 9 2 7 . 0 2 . 9 5 + 0 . 0 2 3 . 6 8 ± 0 . 0 3 5 . 5 0 1 0 . 0 6 3 4 . 0 + 1 . 2

4 . 0 5 . 8 9 9 . 3 8 2 2 . 3 2 . 8 9 ( + 3 ) * *

6 . 0 9 . 4 2 1 7 . 2 6 2 . 3 8 . 1 9 ( + 5 )

8 . 0 1 3 . 4 2 7 . 2 1 . 7 5 ( + 2 ) 2 . 7 9 ( + 8 )

1 0 . 0 1 7 . 9 3 8 . 4 5 . 7 2 { + 2 ) 9 . 4 8 ( + 1 0 )

1.25MeV 1 . 0 1 . 7 2 2 . 0 4 2 . 4 8 4 . 2 7

1 . 7 0 ± 0 . 01 1 . 9 1 ± 0 . 01 2 . 3 2 + 0 . 0 1 3 . 9 7 + 0 . 0 7

2 . 0 2 . 6 1 3 . 3 6 5 . 0 5 1 5 . 9 2 . 6 0 ± 0 . 02 3 . 1 8 + 0 . 03 4 . 5 2 + 0 . 0 5 1 9 . 6 + 0 . 7

4 . 0 4 . 6 7 6 . 5 9 1 2 . 9 9 . 0 0 ( + 2 )

6 . 0 6 . 5 6 1 0 . 1 2 6 . 4 1 . 5 3 ( + 5 )

8 . 0 8 . 7 1 1 3 . 9 5 0 . 9 3 . 0 6 { + 7 )

1 0 . 0 1 0 . 7 1 7 . 2 1 . 0 K + 2 ) 1 . 8 0 ( + 1 0 )

* MCNP data with the statistical error ** Read as 2.89xl0+3

u O u. Q. T3

o Kanemori's experiment

— PALLAS-2D r — G-33GP

QAD-CGGP 102

Iron slab deleclor

o o u. Q. =>

3 I/) O a. x tu

Kanemori's experiment

20 40 60 80 Angle :9(deg.)

Fig.6.6.1 Exposure buildup factors for Iron slab

101 _L 20 40 60 80

Angle(deg.) Fig.6.6.2 Exposure buildup factors

for alualnua slabs

JAERI・M 88・221

Table 6.6.1 Exposure bul1dup factors for concrete slabs

一一F

コマ，-nu 0.25

2.24 2.06士0.01

3.99 3.68士0.03

5.11 4.58tO.08

27.0 34.0土1.2

2.89(+3)術禽

8.19(+5)

2.79(+8)

9.48(+10)

山一山一

込山一υ

cosSo '"' 1

品開

'A-句

nu一nv-

・・

35-55一9

88-99一B

・・岨・・・・

•. 3ム

、，ZR

l.75(+2)

5.72(+2)

4.27 3.97!0.07

15.9 19.主旦三9.00(+2)

1.53(+5)

3.06(+7)

l.80(+10)

.1. 25MeV

2.48 2.32土0.01

5.05 4.52:!:O.OS

17.9 10.0

1.01(+2)

MCNP data with the statistica1 error Read as 2.89xlO+3

17.2 10.7 10.0

Kanemori's experiment

一一 PALLAS-2D

102ι一一 G-33GP長一一.QAO-CGGP

。Kanemori's experirnent

- PALLAS-2D

AI slab

.... 0 ‘ ... υ U I.L

a コ司コ

。』コωoaH凶

‘-0 ・d

0... コ刀

コ 1ω10'/・，

二一-ー・...-...・:ょω』コmoak凶

20 40 60 A n 9 r e ( d eg. )

Flg.6.6.2 Exposure buildup factors

for aluainu・slabs

80 。

20 40 60 80 Angle: e (deg.)

Fig.6.6.1 Exposure buildup factors

for lron slab

nυ nu nu

J A E R I - M 8 8 - 2 2 1

7. Reactor and Nuclear Instrumentation

In the field of Nuclear Instrumentation Systei (N1S) for the reactor control and protection, a Wide Range Monitoring System (IRMS) was trially fabricated for the valuation test of the NIS of the HTTR under planning. The system was tested in the research reactor JRR-4 and a high-temperature fission counter-chamber developed in the laboratory was used in the test as the neutron sensor of the WRMS. Various operating performances such as operating characteristics and stability of the system up to 800"C and monitoring linearity over 9 decades were measured in the test, and it was confirmed that the system meets all operating requirements for the NIS of the HTTR.

In the area of In-core Instrumentation. N-type thermocouples were tested in combination with various kinds of sheath metals to investigate possibility of utilisation of such kind of thermocouples for the in-core temperature monitoring in the HTTR. The drift test of electromotive force(emf) of the thermocouples has been carried out at 1200'C and the test hours reached 3000h at the end of March 1988. As far as the test result is overviewed. the N-type thermocouples with the Nicrosil or Inconel sheath seem to have low drift performance of the emf.

In the development of the Fuel Failure Detection(FFD) system for the HTGR fuels(cpf). the performance of an F.P. precipitating detector was improved by optimising the speed and timing of the precipitating wire movement in addition to the selective detection of the short life nuclides- As a result of such improvements, the system detected fuel failure of the cpf of the order of 10"5.

Extensive studies were also carried out in the field of Non-Destructive Measurement of TRU in waste drums, and an epoch-making result has been obtained; It is the success of non-destructive/quantitative measurement of TRU in the waste drums without any information on the containing waste matrix materials. This new measurement technique must be used widely in the field of TRU waste management in the future-

The assessment was also made on the result of development of an In-Situ Non-Dest.-uctive Measurement System for evaluation of radioactive inventory of the decommissioned power plants, and it was concluded that the developed system makes the inventrory evaluation easy and has sufficient accuracy for such purpose-

In addition to these results, various kinds of studies are in progress such as investigation for formation of magnetic gratings by electron/laser beam and the improvement of gamma-ray spectrum analysis.

(Naoaki fakayama)

JAERI -M 88・221

7. Reactor and Nuclear Instrumentation

[n the field of Nuclear [nstru田entationSyste・(N[S)for the reactor cont-

rol and protection. a Wide Rnnge Monitoring System (WRKS)曹astrially fabricat-

e1 for the vali~ation test of the NIS of the HTTR under planning. The syste田

was tested in the resenrch reactor JRR-4 and a high-te皿peraturefission count-

er-chn田berdeveloped in the laboratory was used in the test as the neutron

sensor of the WRMS. Various operating performances such as operating charact-

eristics and stability of the system up to 800・Cand皿onitoringlinearity over

9 decades胃eremeasured in the test. and it was confir皿edthat the syste皿

meets al1 operating requirements for the NIS of the HTTR.

In the area of In-core Instru皿entation. N-type ther田ocouples胃eretested

in combination胃ithvarious kinds of sheath皿etalsto investigate possibility

of utilisation of such kind of thermocouples for the in-core temperature moni-

toring in the HTTR. The drift test of electromotive force(emf) of the thermo司

couples has been carried out at 1200'C and the test hours reached 3000h at the

end of ~arch 1988. As far as the test result is overviewed. the N-type ther血0-

couples胃iththe Nicrosil or Inconel sheath see皿 tohave low drift perfor皿ance

of the emf.

[n the develop田entof the Fuel Failure Detection(FFD) syste・forthe HTGR

fuels( cpf). the performance of an F. p. precipi tating detector胃as 皿provedby

opti皿isingthe speed and timing of the precipitating宵ire皿ove田entin addition

to the selective detection of the short life nuclides. As a result of such

improve田ents. the system detected fuel failure of the cpf of the order of 10・5

Extensive studies were also carried out in the field of Non-Destructive

lIeasure圃entof TRU io waste dru・s. and an epoch-皿akingresult has been obtain-

ed: It is the success of non-destructive/quantitative田easurementof TRU in

the胃astedru圃swithout any infor皿ationon the containing waste matrix materi-

als. This new measure血enttechnique皿ustbe used胃idelyin the field of TRU

胃aste皿anagementin the future.

The assessment胃asalso皿adeon the result of develop血entof an In-Situ

Non-Dest~uctive leasure圃entSystea for evaluation of radioactive inventory of

the deco皿皿issionedpower plants. and it was concluded that the developed

syst告E 田akesthe inventrory evaluation easy and has sufficient accuracy for

such purpose.

In additioo to these results. various kinds of studies are

(Naoaki Jakaya皿a)

J A E R I - M 8 8 - 2 2 1

7.1 Development of Nuclear Instrumentation System for HTTR

N.Wakayama, H.Yamagishi, H.Itoh and K.Sakasai

A Wide Range Reactor Monitoring System (WRMS) which can measure boch neutron flux density ranging about 10 decades and reactor period has been trially fabricated for the validation test of the Nuclear Instrumentation System (NIS) for the High-temperature Engineering Test Reactor (HTTR). The system includes a High-Temperature Fission Counter-Chamber (HTFC) which had been developed to be used as the sensor of the NIS for the HTTR. D" 4)

The system was tested in the Research Reactor JRR-4 from March 1987 to May 1987 to confirm the overall operating performance of the System. The HTFC was inserted in a capsule with a electric furnace which can heat the HTFC up to the temperature of 850°C in order to simulate the operational environment at the HTTR. The capsule was installed at the side of the reactor core of JKR-4. At first the output linearity of the system was measuied at ordinary temperature at the various neutron flux density obtained by changing the reactor power. It was confirmed the system has an expected performance at ordinary temperature. Next, the temperature was raised up to 600°C and the linearity of the system was measured again in the same way. Some examples of the test results are shown in Fig. 7.1.1 and Fig. 7.1.2. Fig. 7.1.1 shows the plateau curve of the HTFC at 600°C under the thermal neutron flux density of 2.1x10* N t n/m .s. The plateau characteristics shows the WRMS can operate at 150-200V under such high thermal neutron flux density. Fig. 7.1.2 shows an output linearity as a function of thermal neutron flux density at 600°C. With the results of this test, it was confirmed the WRMS has a good linearity in measuring range over 9 decades of thermal neutron flux density. It means the WRMS can monitor overall range of the reactor power from the neutron source level to the level of full power with a good linearity. Next the temperature was furthermore raised up to 800 °C in order to inspect the operability of the system as a Post Accident Monitor. Fig. 7.1.3 and 7.1.4 shows the plateau curve of the HTFC under the neutron flux density of 2.8xl0*4 N^/m'.s and an output linearity at 800 C, respectively. There was no notable change between the result at 800 *C and 600°C. It was. shown the system can b°

]AERI-M 88・221

7.1 Development of Nuclear Instrumentation System for HTTR

N.Wakayama， H.Yamagishi， H.ltoh and K.Sakasai

A Wide Range Reactor Monitoring System (WRMS) which can measure

both neutron flux densicy ranging about 10 decades and reactor period

has been tria11y fabricated for the va1idation test of the Nuc1ear

Instrumentation System (NIS) for the High-temperature Engineering Test

Reactor (HTTR). The system includes a High-Temperature Fission

Counter-Chamber (HTFC) which had been developed to be used as the

sensor of the NIS for the HTTR. 1)-4)

The system was tested in the Research Reactor JRR-4 from March

1987 to May 1987 to confi=m the overall operating performance of the

System. The HTFC was inserted in a capsule with a e1ectric furnace

which can heat the HTFC up to the temperature of 8500C in order to

simulate the operational envir~nment at the HTTR. The capsule was

installed at the side of the reactor core of J~j・ 4. At first the

output linearity of the system was measu~ed at ordinary temperature at

the various neutron flux density obtained by changing the reactor

power. It was confirmed the system has an expected performance at

ordinary temperature. Next， the temperature was raised up to 6000C

and the linearity of the system was measured again in the same way.

Some examp1es of the test results are shown in Fig. 7.1.1 and Fig.

7.1.2. Fig. 7.1.1 shows the p1ateau curve of the HTFC at 6000C under

the thermal neutron flux density of 2.1x1014 Nth/m2.s. The

p1ateau characteristics sho~s the WRMS can operate at 150・200Vunder

such high thermal neutron flux density. Fig. 7.1.2 shows an output

linearity as a function of thermal neutron flux density at 600oC.

With the resu1ts of this test， it was confirmed the WRMS has a good

linearity in measuring range over 9 decades of thermal neutron flux

density. It means the WRMS can monitor overall range of the reactor

power from the neutron source level to the 1evel of full power with a

good linearity. Next the temperature was furthermore raised up to 800

oc in order to inspect the operability of the system as a Post Accident

Monitor. Fig. 7.1.3 and 7.1.4 shows the p1ateau curve of the HTFC

under the neutron f1ux density of 2.8x1014 Nth/m2.s and an

output linear~ty at 800 C， respectively. There was no notahle chan~包

between the result at 8000c and 600DC. It wa~ shown the system can b宅

-}回ー

JAERI-M 88-221

used is a Post Accident Monitor. From the results of these tests, it was concluded the reactor power can be measured satisfactorily wich the WRhS at the temperature range from ordinary temperature up to 800°C.

As the next step of the work, it is being discussed to test the WRMS in a actual environment of a High-Temperature Gas Cooled Reactor, including effects of the electro-magnetic interference, as a Joint Experiment under the agreement of cooperation between KFA and JAERI.

References 1) Wakayama N. et al. : "Development of Fission Chambers for

High-Temperature Reactors," Symposium proceedings on Nuclear Power Plant Control and Instrumentation, Vol.11 IAEA-SN-226/32, 243 IAEA (1978).

2) Wakayama N. et al. : "Reactor Eng. Div. Annual Report" JAERI-M 3393, 77 (1979).

3) Wakayama N. et al. : "Reactor Eng. Div. Annual Report" JAERI-M 9032, 98 (1980).

4) Wakayama N. ec al. : "Reactor Eng. Div. Annual Report" JAERI-M 9672, 90 (1981).

JAERI・M 88・221

used ~s a Post Accidenc Monicor. From the results of chese cests， ic

was concluded che reaccor power can be measured satisfactorily wich

the '.~Rl'，5 at the temperature range from ordinary temperature up to

8000c. As che next scep of the work， it is being discussed to tesc che

日時15 in a actual environment of a High ・TemperatureGas Cooled Reactor， including effects of the electro-magnetic interference， as B Joinc

Experimenc und自rthe agreement of cooperacion between KFA Bnd JAERI.

References

1) ¥.Jakayama N. et al. "Development of Fission Chambers for

日igh・TemperatureReactors，" 5ymposium proceedings on Nuclear

Power Plant Control and Instrumentation， Vol.II

I岨 A・5N・226/32，243 IAEA (1978).

2) ¥.Jakayama N. et a1. "Reactor Eng. Div. Annual Report川

JAERI-M 8393， 77 (1979).

3) Wakayama N. et al. 川ReactorEng. Div. Annual Report川

JAERI-M 9032， 98 (1980).

4)ωakayama N. et a1. "Reaccor Eng. Div. Annual Repor't"

JAERI.M 9672， 90 (1981).

一169-

S OH UJ

U J a:

UJ 13 (X or UJ o

PLATEAU CURVE

DETECTOR TEMP.

NEUTRON FLUX

600 °C

2 . 1 x l 0 1 4 N t h / m 2 . s

Z3 Q -

50 100 150 200

HIGH VOLTAGE (V)

F i g . 7 . 1 . 1 P la teau c u r v e a t 600°C

L I N E A R I T Y OF WIDE RANGE MONITORING SYSTEM

HIGH VOLTAGE

DETECTOR TEMP.

_ / " s s s

600 °C /

io 5 io 6 io' io a io 9 io 1 0 io" io" io" i o M i o , s

NEUTRON FLUX (NtK/m2.s)

Fig.7.1.2 Linearity of Wide Range Monitoring System tested in JRR-4 at 600°C

』〉拘

m-E玄∞∞・MM-

ノノ

・・向u

HIGH VOLTAGE 10'1

6000C DETECTOR TEMP. 10.2

Z一口広UV品目

LU」FUZ}

区ロ」こ

ZEZU口Z江区凶ロ

6000C 14.... ， 2

2.1xlO"Nth/m~.s

PLATEAU CURVE

DETECTOR TEMP.

NEUTRON FLUX

{uz-DgUM-Mh凶」FUZ}

区巴」F240ZMUZ江区凶ロ

-ELロド

コ仏」F

コロ

10・10

トコaLFコロ

10') O

.0 ，I¥ ._12 . _'3 ...‘ .5 10" 10" 10'~ 10" 10" 10 10

5 200 150 100 50 10

8 10' 10&

HIGH VDLT円GE (V) NEUTRDN FLUX (Nth/m

Fig.7.1.2 Linearity of Wide Range Monitorin包

Syste冊 tested in JRR.4 at 6000C

Fig ・7.1.I Plateau curve at 6000c

z a ac u

§ 2 O

UJ C3 z a: a: UJ a

PLATEAU CURVE

DETECTOR TEMP. NEUTRON FLUX

800 °C 2 . 8 x l 0 1 4 N t h / m 2 . s

10 SO 100 150 200

HIGH VOLTAGE (V)

F i g . 7 . 1 . 3 Plateau curve at 800°C

LINEARITY OF WIDE RANGE MONITORING SYSTEM

HIGH VOLTAGE ; ISO V DETECTOR TEMP. ; 800°C

10S 108 107 10' 10S 10 1 0 10" 10 B 10 , S 10 M 10 , S

NEUTRON FLUX (Nth/o a .sJ

F ig .7 .1 .4 L i n e a r i t y of Wide Range Moni tor ing System tested in JRR-4 at 800*0

LINEARITY OF WIDE RANGE

MONITORING SYSTEM

HIGH VOLTAGE

DETECTOR TEMP. 10.1

』〉開河

玄∞∞-NN-

-uz-GqU包

MhuaFUE】

go--zロE凶UZEMhuロ目

BLEトコ色」

Fコロ

10-1也

{ugω区町民凶』「

ωz-区白」「

-zozuロzgguロ-BL巴」

Fコ色」

Fコ白

800CC 14 ，.~~ ，_ 2

2.8xlO~~Nth/m~.s

PLATEAU CURVE

DETECTOR TEMP.

NEUTRON FLUX

10，1

!日4Ml

10'5 0

10' 200 150 100 50 10

1・101510

HJGH VIlLTAGE (VI NEUTRCN FlUX (N.h/回2.$J

Fi邑・ 7.1.4Linearity of Wide Range Monitoring

System lested in JRR-4 at 800~C Fig.7.1.3 Platcau curve at 800・c

J A E R I - M 8 8 - 2 2 1

7.2 High-Temperature Out-Pile Tests of N-Type Thermocouples

M. Yamada, K. Ara and N. Wakayama

A new type of thermocouples, so-called Nicrosi1/Nisi 1 thermocouple (N-type TO, seemed to be the most suitable sensors for in-core gas-temperature measurement in the high-temperature engineering test reactor (HTTR). The actual experience on the use of N-type TCs is, however, very little so that a long-term out-pile test of them having defferent types of sheaths was started at a high-temperature He-gas environment.

The chemical composition of N-type TC is shown in Table 7.2.1 as compared with that of K-type. One see the differences in content of Cr and Si: these differences make N-type more resistant to high-temperature than K-type. The electromotive force (emf) characteristic of N-type is almost the same as that of K-type, as shown in Fig. 7.2.1.

The N-type TCs under the test have sheaths with four different materials; i.e., inconel 600, incoloy 800, hasteloy X and Nicrosil. The insulater is MgO in common. The tests are being performed at the temperature of 800 °C, 1000 °C and 1200 °C. K-type TCs and R-type ones are also tested for comparison and as references. All testing TCs were calibrated prior to the start of the tests; and their emf's are compared with that of reference R-type TCs. The results of 1200 °C test are shown in Fig. 7.2.2: N-type TC having Nicrosil sheath seems to have the least amount of emf drifts, so far forth. The test will be continued for the total test hours of more than 10000.

J:¥ERI・M 88-221

7.2 High-Temperature Out-Pile Tests of N-Type Thermocouples

M. Yamada， K. Ara and N. Wakayama

A new type of thermocoup1es， so-ca11ed Nicrosi1/Nisil therrπm問l悶OCOl巾1e

(N-type rc)， seemed to be the most suitab1e sensors for ;n-core gas拘 tempera-

ture measurement i n the h i gh-tempera ture eng i neeri n9 tes t reactor (HTTR).

The actua1 exper;ence on the use of N-type TCs is， however. very 1itt1e so

that a 1ong-term out-pile test of them having defferent types of sheaths was

started at a h;gh-temperature He-gas environment.

The chemica1 compos;t;on of N-type TC is shown in Tab1e 7.2.1 as

compared with that of K-type. One see the differences ;n content of Cr and

Si: the~e differences make N-type more resistant to high・temperaturethan

K-type. The e1ectromotive force (emf) characteristic of N-type is a1most

the same as that of K-type. as shown in Fig. 7.2.1.

The N-type TCs under the test have sheaths with four different

mater‘ia1s; i.e.， inconel 600， inco1oy 800， hasteloy X and Nicrosil. The

insu1ater is MgO ;n common. The tests are being performed at the

temperature of 800 oC， 1000 oc and 1200 oC. K-type TCs and R-type ones are

also tested for comparison and as references. All testing TCs were

ca 1 i brated !=r i or to the start of the tests; and thei r emf ・sare compared

with that of reference R-type TCs. The resu1ts of 1200 oc test are shown in

Fig・7.2.2: N-type TC having Nicrosil sheath seems to have the 1ea~t amount

of emf dr; fts， so f ar forth. The test wi 11 be cont i nued for the tota 1 test

hours of more than 10000.

一172-

J A E R I - M 8 8 - 2 2 1

Table 7.2.1 Chemical composition of N-type and K-type thermocouples

Cr Si Fe Mg N1 Al Mn

Nicrosi1 13.85 -14.61

1.43 -1.47

0.024 -0.1 4.0.081 bal . - -

Nis i l <0 .008 4.23 -4.88

0.054 -0.41

0.07 -0.18 bal . - -

Chromel 9.46 0.4 0.2 - ba l . - -

Alumel - 1.2 0.1 - bal . 2.0 1.75

= 0 •

40 -»*-> > 3. K / /

« / N / c 30 . X

10 / /

- 2 - i

• M 0 2 4 5 8 10 12

TSMP (XIOO'C)

- - 1 0

Fig. 7.2.1 Thermoelectromotive forces of N-type and K-type thermocouples

]AERI-M 88・221

Chemica1 composition of N-type and K-type thermocoup1es

Tab1e 7.2.1

Cr Si Fe Mg Ni A1 川n

Nicrosi 1 511431o O24K0 081 ba 1. . . -14.61 ・1.47 ・0.1

Ni s i 1 4000814231o山 1007lbal . . -4.88 ・0.41 ・0.18

Chrome1 9.46 . .

Alumel上二 11.2

"- ，

ba1. 2.0

-SIri--1111111「

R岡山

1ヲ・圃

TEMP (XIOO'C)

10 s ロ4 ヲー

11111111Lilli--'ーー」

C-x)弘

Thermoe1ectromotive forces of N-type and K-type thermocoup1es

7.2.1 Fig.

&S-0.5

. 1 .0

t / i -1 .5

- 2 . 0

- 2 . 5

Ts T r

T e m p e r a t u r e i n d i c a t i o n of s a m p l e TC T e m p e r a t u r e i n d i c a t i o n of r e f e r e n c e TC

I I I I L

(TC lype/Shealh) — N lype/lnconel 600

-a- - N t y p e / N i c r o s i l

•fa N - t y p e / l n c o n e l 600

ft,--N t y p e / l n c o l o y 800

ft N - t y p e / l n c o l o y 800

K - l y p e / l n c o n e l 600

1000 2000 EXPOSURE HOURS

Tig. 7.2.2 EMF drifts of N-type thermocouples due to exposure to temperature of 1200 °C, in comparison with that of K-type

-----* ----占一ー___*~ (TC -Iype/Shealh)

~割t JK一一-N-Iype/lnconel 600

入\--r← -L人---.[J~---.ーベl--N Iype/Nicrosil

h哩ー0.5I

JMOFJX

』〉問

m-・玄∞∞sMM-

合¥¥一円、、士----ー-~

ご¥:~~吋--一.¥¥可一一食¥¥向 -N-Iypc川 o凶 6∞

合ーーー一ー-lを-- 、、、食___________N--type/lncoloy 8∞ -句食------N-type/lncoloy 800

lH叶品

一-K-Iypc/lnconel 6∞ Ternpcralure indicillion of sarnple TC

Ternperalure indicalion of reference TC

2000 EXPOSLJRE 1I0URS

EMF drifts of N-type thermocoup1es due to exposure to temperature of 1200 oC， in comparison with that of K-type

rig. 7.2.2

J A E R I - M 8 8 - 2 2 1

7.3 Development of Fuel Failure Detection System for High Temperature Gas Cooled Reactors S. As ami, H. Terada, N. Wakayama, M. Obar.a , Y. Yoshida, N. Kobayashi and K. Fukuda

Development of a high sensitivity fuel failure detection (FFD) system is in progress for a high temperature engineering testing reactor (HTTR) under designing in the JAERI. The performance of the FFD system for the coated particle fuel (cpf) was improved by mean of the selective detection of short-life noble-gas fission products (fp). The selective detection was achieved by reducing soak chamber volume (1000ml to 100ml) and shortening of soak time (lOmin. to lOsec.) of the wire precipitator of the system as shown in Fig. 7.3.1 and Fig. 7.3.2. The system could detect cpf failure of around 0.04%.^'^

As the result of the shortening of soak time, average wire speed increased and one rounding time of precipitation wire became not long enough for decreasing radio-activities collected on the wire sufficiently. In order to overcome this problem, the precipitation wire movement was controlled by a photo wire position sensor and stepping motor driving so that the same part of the wire was not used for collecting fp again within six rounds of wire. This control is possible because the precipitation point (a) in Fig. 7.3.2 collects fp 10 seconds in the chamber and then moves to point (b) in the scintillator within very short time, and a certain length of wire without any deposited fp exists between point (a) and (b) •

Figure 7.3.3 shows an experimental result of effect of background reduction on the counting rate of the precipitator using this driving control.

Figure 7.3.4 shows a difference of the responses between precipitators using the improved wire driving method and the conventional driving method in the fp monitoring from cpf under irradiation during the reactor start-up of the JMTR. The figure shows that the improved driving precipitator could detect the increasing new born fp depending on the reactor power but the conventional driving precipitator could not because of so high background counting level.

+ Irradiation Division II, Department of JMTR project. ++ Fuel Irradiation and Analysis Laboratory, Department of Fuels and Materials Research

JAER j-M 88・221

7.3 Development of Fuel Failure Detection System for High

Temperature Gas Cooled Reactors

S. Asami， H. Terada， N. Wakayama， M. Obata+， Y. Yoshida，

N. Kobayashi++ and K・Fukuda

Development of a high sensitivity fuel failure detection (FFD)

system is in progress for a high temperature engineering testing reactor

(HTTR) under des igning in the JAERI. The performanc己 of the FFD system

for the coated particle fuel (cpf) was improved by mean of the selective

detection of short-life nob1e-gas fission products (fp). The se1ective

d巴tectionwas achieved by reducing soak chamber volume (lOOOml to 100m1)

and shortening of soak time (lOmin. to lOsec.) of the wire precipitator

uf the system as shown in Fig. 7.3.1 and Fig. 7.3.2. The system could

1)， 2) detect cpf failure of around 0.04%.

As the reslJ1 t of the shortening of soak time， average wire speed increased and one rounding time of precipitation wire became not 10ng

enough for decreasing radio・activities co11ected on the wire

sufficiently. 1n order to overcome this problem， the precipitation wire movement was contr011ed by a photo wire position sensor and stepping

motor driving so that the same part of the wire was not used for c01-

1ecting fp again within six rounds of wire. This control is possible be-

cause the precムpitation point ② in Fig. 7.3.2 collects fp 10 seconds in

the chamber and then moves to point⑤ in the scintillator within very

short time， and a certain length of wire without any deposited fp exists

between point①and⑤.

Figure 7.3.3 shows an experimental result of effect of backgrou~ci

reduction on the counting rate of the precipitator using this driving

control.

Figure 7.3.4 shows a difference of the responses b~tween

precipitators using the improved wire driving m~thod and the conven-

tional driving method in the fp monitoring from cpf under iτradiation

during the reactor start-up of the JMTR. The figure shv~s that the im-

proved driving precipitator could detect the increasing new born fp

depending on the reactor power but the conventional driving precipitator

could not because of so high background counting level.

+ Irradiation Division II， Department of JMTR project. ++ Fuel Irradia-

tion and Analysis Laboratory， Department of Fuels and Materials Research

J A E R I - M 8 8 - 2 2 1

As a result of the evaluation of these experimental data, it was concluded that the improved FFD system detect fuel failure of the order of 0.0012 of coated particle fuels.

Reference 1) Terada H. et al,: " Development of Fuel Failure Detection System

for a High Temperature Gas Cooled Reactor," IEEE Trans. NS-35C3) (1988)

2) Terada H. et al.: " Reactor Eng. Dev. Annual Report (Apr. 1986 -Mar. 1987)," JAERI-M 87-126 190(1987)

Irradiation rig

•— J fuel

sample (normal)

Signal EleC, ic Instruments 8, Cu?+' is

-( Intact)

XL 9 J Wire control

precipitator -•exhaust

JMTR Fig. 7.3.1 Scheme of the Experimental FFD System

Scintillator Photo Multiplier Soak Chamber

(1000cm3 — 100cm3) Wire Winding Drum

Fig. 7.3.2 Schematic Structure of the Precipitator

JAERI-M 88・221

As a result 白fthe eva lua t ion 0 f these expe rimenta 1 da ta， i t was

conc1uded that the improved FFD system detect fue1 failure of the order

of 0.001% of coated particle fue1s.

Reference

1) Terada H. et a1，:川 Developmentof Fuel Fai1ure Deteccion System

for a High Tempeuture Gas Cooled Reactor，" IEEE Trans. NS-35(3)

(1988 )

2) Terada H. et a1.: " Reactor Eng. Dev. Annua1 Report (Apr. 1986 ・

Mar. 198n，" JAERI・M87・126190(1987)

Irradlatlon rlg

」Eler二 'IcInstruments 3. C";，，t， ~ls

門口一門U

Wlre control

exnous'!"

Fig. 7.3.1 Scheme of the Experimenta1 FFD System

品切-

hu-m-c一一

∞一m

S一山ハun

Photo Mul tipller

10min.

10sec.

Stepping Motor

Precipitotion Wire Wlre Posltlon Sensor

Fig. 7.3.2 Schemacic Structuτe or the Precipicacor

J A E R I - M 8 8 - 2 2 1

1.6 1.5 1.4 1.3 1.2' 1.1-H

....+•••

Wire Rounding Time * 20min.

+ : When same collecting part of wire was used. A pile-up effect of back ground was observed.

D : When a new wire driving sequence was used.

—-B- -a B ED

Number of Wire Rounding .3 Effect cf Background Reduction on the Counting Rate

of the precipitator i g -

irrodiatlon sample : 85F-5A3

"c zs o U L —

Conventional Wire Driving

10 20 30 40

Reactor Power (MWt)

Fig. 7.3.A Difference of the Responses of Precipitators

]AERI・M 88・221

1.4 Wire Roundin~ Time • ?Omin.

+ When same collecting part of wire was used. A pile-up effect of back ground was observed. 1.2

ロWhena new wire driving sequence was used. →ニ

{同】

ωNZOE」。Z)⑪』。巴mMC一一巳コ。

-EE--1

白自s-.ー『ーー【一一一一e-ー一ー一一ーー] ご

Number 01 Wire Rounding

cf Beckgroune Reouction on :he Counting Rate preclplt8tOr

4 3 内，』

J.';..) =.::ec: 0: : ne .

Irrcaictlon somple : 85i=-5A3

ー同園-・・

Conventional Wire Drivlng

Jilo--:J，1111140

-D 40 30 10

(ωAυ}

。-uzoc--Z20U」OFD=a

一υ@」仏

Recctor Power (MWt)

Fig. 7.3.4 Differen~e of rhe Responses of Precipitators

J A E R I - M 8 8 - 2 2 1

7.4 Development of Nondestructive Measuring Techniques of Transuranic Elements in Waste Drums

H. Gotoh, N. Wakayama, M. Haruyama, H. Yamagishi, M. Suzuki, Y. Yoshiiura and Y. Takeuchi

In a programme to develop a nondestructive detection system of transuranic elements in waste drums, extensive experimental studies were carried out on the passive neutron technique, the differential die-away technique and an operating test of the neutron detecting system under high gamma-ray dose environment. (1) Passive neutron technique

Experiments were conducted by using an assay chamber with graphite walls of 30 cm thickness. An encapsulated sample (Pu activity = 48.8mCi, alpha activity = 98.7mCi) containing about lg of Plutonium produced from BWR spent fuel was placed at the center or at a position of about half radius of 2001 drums. Various kinds of neutron counting characteristics were taken.

Table 7.4.1 shows the measured values of total Plutonium activity and total alpha activity calculated from the coincidence counting rate and the gross neutron counting rate taken for sever°l kinds of matrices. The analysis was carried out without any prior knowledge of the matrix. It is expected that the reliability of the measurement could be improved with refining the analysis method. (2) Differential die-away technique

Experiments were conducted by using an assay chamber with graphite walls of 20 cm thickness. The outline of the system has been described previously.') Encapsulated Plutonium samples with net weight covering 0.001 through lg were distributed in 2001 drums containing various kinds of matrices and responses to pulsed D-T neutrons were taken. Examples of the responses were reported previously.*•'

Table 7.4.2 shows the preliminary results obtained by analyzing those responses for several kinds of drums. The detection sensitivities are values for samples placed at positions about 5 cm deep from the surface of drums. (3) Influence of gamma-rays on neutron detection system

At how high exposure rate, neutron detection systems can be operated? We made special neutron counting systems in order to detect neutrons with He-3 counters in high gamma-ray environment. A neutron measuring test of the system was carried out in the Cobalt-60 Irradiation Facility in JAERI in

]AERI-M 88・221

7.4 Development of Nondestructive Measuring Techniques of

Transuranic Elements in Waste Drums

H. Gotoh. N. Wakaya・a.M. Haruyua. H. Va・agishi.M. Suzuki，

V. Voshi・uraand V. Takeuchi

In a progralle to develop a nondestructlve detection system of trans-

uranic e!e・ents in waste dru・s.extensive experi・enta!studies were carried

out on the passive neutron technique. the dlfferential die-away technique

and an operating test of the neutron detecting systel under high ga園田a-ray

dose environment.

(1) fassive neutron techniQue

Experi ・entswere conducted by using an assay cha・berwith graphite

walls of 30 c圃 thickness.An encapsulated sa・ple(Pu activity = 48.8皿Ci， a 1-

pha activity = 98.7圃Ci)containing about 19 of plutoniu圃 produced fro問 Bl¥'R

spent fuel was placed at the center or at a position of about half radius of

2001 dru圃s. Various kinds of neutron counting characteristics were taken.

Table 7.4. I shows the圃easuredvalues 01 total plutoniu皿 activityand

total alpha activity calculated fro. the coincidence counting rate and the

gross neutron counting rate taken for severol kinds of 圃atrices. The

analysis was carried out without any prior knowledge of the皿atrix. It is

expected that the reliabilitr of the圃easurelentcould be 聞provedwith

refining the analysis lethod.

(2) D ifferential die-aWhY techniQue

Experilents were condu~ted by using an assay cha圃berwith graphite

walls of 20 c・thickness.The outline of the syste・hasbeen described pre-

Viously.J) Encapsulated plutoniul sa・pleswith net weight covering 0.001

through Ig were distributed in 2001 dru.s containing various kinds of

・atricesand responses to pulsed D-T neutrons were taken. Exa圃plesof the

responses were reported p町eviously.2)

Table 7.4.2 shows the preli・inaryresults Jbtained by analyzing those

responses for several kinds of dru・s.The detection sensitivities are values

for sa圃plesplaced at positions about 5 c・deepfrol the surface of dru圃s.

(3) 1 nfluenc~ßf ga..a-rays on neutron detection syste圃

At how high exposure rate. neutron detection syste ・scan be operated?

'e lade special neutron counting systels in order to detect neutrons with

He-3 counters in high ga..a-ray environlent. A neutron・easuringtest of the

syste.曹ascarried out iu the Cobalt四 60Irl'adiation Faci 1 ity in JAERI in

一178-

J A E R I - M 88-221

comparison with a conventional proportional counting systei. A high sensitive He-3 counter having 150 cps/nv sensitivity was used in both system as the neutron detector.

Figure 7.4.1 shows the integral bias curves of the He-3 counter with our systea at 0, 100, 200, 500 and 1000 R/h, with and without a neutron source. It can be understood that the systei operates well at least at the level of several hundred R/h. Figure 7.4.2 shows those of the sane counter with the conventional systei at 0, 1 and 5 R/h, with or without the neutron source. This systei did not operate well at the level of 20 R/h.

References 1) H. Gotoh et al. : JAERI-M 86-125, P.187 (1986) 2) H. Gotoh et al. : JAERI-M 87-126, P.193 (1987)

Table 7.4.1 Measured values of TRU activities in 2001 drums

Kinds of latrix position of

Pu saiple

weight (kg)

est iiated activity Kinds of latrix position of

Pu saiple

weight (kg) total Pu

(•CI) total a (•CI)

empty drui center 29.8 48 128 cloth outer ring

center 71 66

61 127 103

river sand outer ring center

341 97 112

128 111

concrete outer ring center

332 81 163

111 64

borosi1icate glass outer ring center

368 66 46

105 158

Ta^le 7.4.2 Detection sensitivity of Pu-239 in 2001 druis Kinds of aatrix weight

(kg) sensitivity

(nCi/g)

eapty drua 29.8 11.4 cloth 71 4.37 river sand 341 2.14 concrete(hull and end-pieces contained) 551 28.9 concrete 332 88.3

]AERI-M 88・221

cOlparison with a conventional proportional counting systel. A high sensi-

tive He-3 counter having 150 cps/nv sensitivity was used in both syste田 as

the neutron detector.

Figure 7.4.1 shows the integr‘al bias curves of the He-3 counter wi th

our systell at O. 100.200.500 and 1000 R/h. wlth and without a neutron

source. 1 t can be underst.ood that the systel operates we 11 at. least at the

level of several hundred R/h. Figure 7.4.2 shows those of the sa皿ecounter

with the conventional systel at O. 1 and 5 R/h. with or without the neutron

source. This syste圃 didnot operate well at the levei of 20 R/h.

References

1) H. Gotoh et al. JAERI-M 86-125. P.187 (1986)

2) H. Gotoh et al. JAERI-M 87-126. P.193 (1987)

Table 7.4.1 Measured values of TRU activities in 2001 dru聞S

Kinds of utrix position of weight esti圃atedactivity

Pu sa圃ple (kg) total Pu tot81α (・Ci) {圃Cj)

empty druI center 29.8 48 128

c10th outer ring 71 66 127 center 61 103

river sand outer ring 341 97 128 center 112 111

concrele ouler ring 332 81 111 center 163 64

borosilicate glass outer ring 368 66 105 center 46 158

Ta~le 7.4.2 Detection sensitivity of Pu-239 in 2001 dru圃s

Kinds of ・atrix 宵eight sens i ti v i ty (kg) (nCi/g)

e・ptydrUl 29.8 11. 4

cloth 71 4.37

river sand 341 2.14

concrete(hull and end-pieceεcontained) 551 28.9

concrete 332 88.3

J A E R I - M 8 8 - 2 2 1

V* ^ "A

"•rutron S'Mir

> * * * . * *' £-..-

• ,->.-!.-> . |.

Fig. 7.4.1 Integral bias curves of the newly designed neutron detection systei

:"iejt:

i-jithou* neutron Eource

- ^ -ii -T -

r :n 2 L'

I I * - 1

101 V 10 «. Li

_i i ' J 50 100 ise

channel 2C0

Fig. 7.4.2 Integral bias curves of the conventional neutron detection systei

]AERI-M 88・221

]一ヨ守百三唱『一次

RFい回一

-p-一回

川川

U司』『パ」

『川'『・1

、』町パ・h

e--』戸・・(」

へ子uti・I:.n三‘1:.:..;r.:。テ

・・ v ・・ 0 ・i

1r:"1...ぃ'，，'1. =， Ip"..~ I '"ー

是ヨ守ヨ-=-- F門 . -芳苛豆←←-ふ‘:.ii‘

コi‘. '.咽

744?1 .J、

:. -.....s=-".b，，~ ーー揃岬亀岨・• ...、‘→ 。ー・ 6 、、ト必ーー・曹司....

ゼメ胃込.~.--... :J見、芯--

a ヨ.

白るい

¥1 I ". 1 1

ぃ".，品、。・...-..・.

-円。.:， 1 .. :

司ご

・‘.幅

、円記

"..，・A

. 巴也

l 『・

一一

~ ーー

。I， -'

ー，戸内』込1";"1

戸山1・-、!、

、ー、.

Intesral bias curves of the newly desisned neutron

detectio'l syste・F'ig. 1.4.1

i=『コヨヨ13-Jヨ「二三二一言ヨ1雪司1司ヨヨ]日C

c戸 1，

:"j寸'.J了:‘ ':';"1 ':'. '，，1. テ

.. riがitr心n =':，Ijj・ I:.~

.I~I: '=，ぞ・L士 0"-，，" ; .~....ル・へ hム・二日Ci;.'， 1"刊 i 川ご・

f、Fう町、.::.I::'t..l

1ち巴

」でEト-

事7-¥.jj寸ーーーーーー旬、

場4 九

fJ一一手=--=--::..-::s:，

d ー・ 4!'f .

品目 1

ii--i ~

~ F~ o E，0 100

乙hanne

+ ・ー

... f" ・1守.;.i・，1，~ ;-r(IU~

'11" .'"

!1I2 lii

Integral bias curves of the conventional r削 tron

Jetection syste・

-1別一

Fi!. 7.4.2

J A E R I - M 8 8 - 2 2 1

7.5 Development of an In-situ Nondestructive Measurement System for Measuring Radioactivities Inside Contaminated Pipe

M. Katagiri, N. Wakayama and H. 11 oh

An in-situ nondestructive and quantitative measurement techniques were developed for measuring radioactivities inside pipes of primary cooling system to estimate the residual radioactive inventory before decommissioning of nuclearpower plants. The prototype in-situ measuring system has been fabricated and an in-situ measurement was carried out in JPDRCBWR) to confirm the performance and usefulness of the system.

A contaminated pipe of a forced circulation system in JPOR was chosen for the performance test or' the system. The outer diameter of the pipe is 32131m and the thickness 19mm and the pipe was full of old cooling water.

The outline of the measuring system is illustrated conceptually in Fig.7.5.1. A "transportable equipment" consisting of a scanning device, N-; gas cooled Ge gamma-ray detector with collimator and shielding, N? gas cylinder, pulse amplifier, A-0 converter and t e l e m e t e r i n g system was provided in the JPOR reactor enclosure and the scanning device with the gamma-ray detector was set up on the forced circulation pipe. The "data analysis system" for analyzing the measured data and displaying the results ,vas provided in a laboratory near the JPDR building. AC-line modems and telemeter devices were used for the mutual communication of control and data signals between the in-situ instrument and data a n a l y s i s system. The radioactivities were measured at the eight points near the bent section of the forced circulation pipe in JPDR. In order to make quasi-toaographic measurement and analysis for estimating radioactivities on the inner surface of the pipe and in the water separately, scanning measurement was done for radial direction in each point to be measured. The measuring time of 1800 to 3000 seconds per 1 radia! point was required to determine the radioactivities because of low level radioactivities.

As the result of the measurement and analysis, it was found that the almost ali radioactivities were deposited on the inside surface of the pipe and there was no detectable radioactivity in the water in the pipe. The detected plate-out radioactivity was only Co-60. Figure 7.5.2 shows measurement result with the position of measuring points. The radioactivity concentration is in the range between 24 and 42 nCi/cm^. The radioactivity of the measuring point 4 (the corner of the piping) was the highest.

- 1 8 1 -

]AERI・M 88・221

7.5 Dcvelopment of an In-situ Nondestructive Measurement System for

Measuring Radioactivities Inside Contaminated Pipe

~. Katagiri. N. Wakaya田a and H. Itoh

An in-situ nondestructive and quantitative 皿easure田ent techniques 市邑re

developed for 四easuring radioactivities inside pipes of priraary cooling SYS-

tem to estimate the residual radioactive inventory before decom皿issioning ot'

nuclearpower ptants. The prototype in-situ measuring system has been fabri-

cated and an in-situ 田easurementwas carried out in JPDR(BWR) to confir田 the

performance and usefulness of the syste田.

A conta田inated pipe of a forcec circutation syste田 in JPDR was chosen

for the perfor皿ance test or the system. ihe outer d ia圃eter of the pIpe is

32~~m and the thickness 19mm and the pipe was futt of otd cooI ing water.

The outline of the 圃easuring system is i t tustrated conceptuat Iy in

Fig. i.5.l. A "transportable equipment" consisting of a scanning device. N?

gas cooled Ge gamma-ray detector wlth col t i四ator and shield!ng.'12 gas

cylinder・puIse amplifier. A-D converter and telemetering s，/ste:u was

p~ovided in the JPDR reactor enclosure and the scanning device with the

g:lmma-rlY detector 'Nas set UP on the forced circulation pipe. The "data

analysis 5ystem" for analyzing the measured data and displaying the resutts

，~as provided in a laboratory near the JPDR building. AC-line mode:ns a 白d

telemeter devices were used for the mutual co圃圃unicationof control and data

signals between the in-situ instrument and data anatysis syste皿. The

radioactivities w号re 阻easuredat the eight points near the bent section of

the forced circuiation pipe in JPDR. [n order to 圃ake quasi-to圃ographic

~easurement and analysis for estimating radioactivities on the inner surface

of the pipe aod in the water separatelY. scanning 圃easure圃ent was done for

radial direction in each point to be田easured. The 圃easuring ti皿eof 1800 to

3000 seconds per 1 radial point was required to determine the radioac-

tivities because of low levei iadioactivities.

As the result of the measure圃ent and analys is. i t was found that the

~Imost al i radioactivities 胃eredeposited on the inside surface of the pipe

and there was na detectahle radioactivity in the water in the pipe. The

detected plate-out radioactivity was only Co-60. Figure 7.5.2 shows ~easure

lIent result with the position of 置easuring points. The radioactivity con-

centration is in the range between 24 and 42 nCL'cm2. The radioactivity of

the孟easuringpoint 4 (the corner of the pjping) 胃as the h ighest.

-181一

JAERI-M 88-221

It is confirmed that the radioactivities of the order of nCi/cm' can be measured using the prototype in-situ measur. iut system. It was also confirmed that the system can measure the radioactivity on the inside surface of the pipe and in the water in the pipe separately by a nondestructive measure. It is shown through these evaluation tests that the prototype in-situ measuring system developed has relatively high sensitivity and very useful for nondestructive measurement for evaluating radioactive inventory in the decommissioning power plants.

Fig. 7. 5. 1 The outline of the measuring system

- 1 8 2 -

JAERI-M 8[1-221

lt is confir圃edthat the radioactivitiρS of the order of nCi!cm2 can be

圃easuredusing the prototype in-s:tu measur、"，"Csystem. lt was also con-

fir皿ed that the syste皿 can田easucethe radioactivity on the inside surface

of the pipe and in the water in the pipe separately by a nondestructive

皿easure. lt is shown through these evaluation tests that the PI'ototype in-

5 i tu圃easul'ingSyste田 developedhas relatively high sensitivity and very

useful for nondestructive圃easurementfor evaluating radioactive inventory

in the deco聞皿issioningpower plants.

iL ~ b 0 r • t 0 r r/

fJPOFt-conL.rol rool11J lJ?口R 1¥....¥0< Enclo竃 u< <1

Fig. :. 5. 1 The outline of the measuring system

JAERI-M 88-221

N u c l i d e : C o - 6 0

- 3?0mm 320mm

Forced C i r c u l a t i o n Pipe Outer Diameter:320mm Thickness :19mm

Plnlcoiil act iv i ty onI Activity In the (he Ins die surface

nCi / c m 2

2 6 . 4*o, 1

2 3 . 3*o- *

3 0 . 9*o- B

4 2. 4 * o • r»

2 7 . 1 * 0 - 4

ii C I / g

* N o N i;

2 5 . 4 * 0 - i

2 4, 9.to. n

2 3 .0±«- a

* :Less t h a n 0 .5nCi /g

Fig. 7. 5. 2 The results of in-s i tu measurements at the bent section of a forced circulation pipe in JPDR

- 1 8 3 -

Forced circulatiα1 pipe αlter oiellAlter:320mm '1llickness :19mm

Fig. 7.5.2

JAERI-M 88・221

1511111ggOO川

。.，_.

Nuclide :Co-60

dlylly In Ihe Ihe Insl.le 51/1 face Iνa ler

今卜一一一一|一一一一nCi/cm2 IICI/g

ホ26. '1士 0.111 NoNIl

2 23.3士(). 11 I 11

3 3 [J. 9士O田 61 グ

JJ---一 [一一-j〉イ

ー一一一一一一一一ー一一一一'12.1¥士O.r; 11

F J 27.'1圭 O.iI 11

G 25. '1士 O.11 H

7 ?'1.9tn.円 "

。 2 3. fl士O.:1 11

一一一一一一一十一牢:LessUtan O.5nCiノq

The results of in-situ rneasurernents at the bent section of a forced circulation pipe in JPDR

J A E R I - M 8 8 - 2 2 1

7.6 Formation of Magnetic Gratings by Electron/Laser Beam Irradiation

K. Ara and H. Yagi

On the basis of changing the aagnetic characteristics by electron/laser beaa irradiation to steels, formation of nagnetic gratings has been studied experimentally. The results obtained froa the experiment shows that available aagnetic gratings are formatted by beaa irradiation and characteristics of the nagnetic gratings depend on soae kinds of steels. The characteristics of the gratings were confiraed froa analyses of aetallurgical aicroscope image, measurements of hardness by Micro-Vickers aethod and measurements of residual magnetization intensity. The aagnetic gratings are applicable to a magnetic encoder and a scale beaa for high temperature use, because they have the heat-resisting characteristics. Heat-resistant sensors constructed by these magnetic gratings can be applied to robotics and instrumentations in the high temperature fields; e.g. on the nuclear power plants, the iron manufacture plants and the ceramics plants etc.. Formation of magnetic gratings by beaa irradiation and sensing methods

The experimental apparatus of the beaa irradiation is shown in Fig. 7.6.1. A magnetic grating is formatted on test pieces by electron or laser beam irradiation. There is no difference on the characteristics of the both magnetic gratings which are formed by the electron beam and the laser beam. An optimum condition of beaa irradiation is determined by an examination of the residual magnetization patterns. The conditions of beam to make one line of the magnetic gratings with 2~3mm width are of the output power of lkW and of the scanning speed of 15mm/s with just focusing on the surface. Miniaturization of the magnetic gratings is future problem but it is limited by sensitivity of the magnetization detection. Figure 7.6.2 shows the concep-tional drawing of the magnetic scale consisting of the magnetic gratings and sensor head. Characteristics of the magnetic gratings

Several kinds of steels were used to examine the characteristics of aagnetic gratings. The sample steels are listed on Table 7.6.1 with the specifications. (A) A case of ferroaagnetism

The listed sample steels were irradiated by electron and laser beaa for making test pieces. Hardness distributions were measured on a section of irradiation spots, because of a good correlation between the magnetic hardness

- 1 8 4 -

jAERI-M 88・221

7.6 Formation of Magnetic Gratings by Electron/Laser Beam Irradiation

K. Ara and H. Yagi

On the basis of changing the皿agneticcharacteristics by electron/laser

bea圃 irradlatlonto steels， forll8tlon of圃agnetlcgratlngs has been studled

experimentally. The results obtained from the experi・entshows that ava i 1-

able magnetic gratings are for圃attedby beal irradiation and characteristics

of the皿agneticgratings depend on so圃eklnds ot steels. The characteristlcs

of the grat ings were contirlled fro皿 analysesof・etallurgical lIicroscope

i田age. 回easurelentsof hardness by Micro-Vickers lIethod and leasure圃entsof

residual magnetization intensity. The magnetic gratings are applicable to a

皿agneticencoder and a scale bea圃 forhigh te圃peratureuse， because they

have the heat-resisting characteristics. Heat-resistant sensors constructed

by these田agneticgratings can be applied to robotics and instru圃entations

in the h igh te圃peraturefields: e.g. on the nuclear power piants， the iro九

manufacture plants and the cera圃icsplants etc..

Formation of ma~netic ~ratings by bea聞 irradiationand sensing methods

The experimentaI apparatus of the beu irradiation is shown in Fig.

7.6.1. A magnetic grating is formatted on test pieces by electron or iaser

bea田 irradiation. There is no difference on the characteristics of the both

magnetic gratings which ~re for皿edby the electron bea圃 andthe laser beam.

An optillu園 conditionof beal irradiation is deterained by an exa.ination of

the residual lagnetization patterns. The conditions of bea圃 to皿akeone line

of the lIagnetic gratings with 2--3圃圃 width are of the output power of 1kW

and of the scanning speed of 15.圃/swith just focusing on the surface. Mini-

aturization of the回agneticgratings is future proble・but it is li・ited by

sensitivity of the 圃agnetizationdetection. Figure 7.6.2 shows the concep-

tional drawing of the aagnetic scale consisting of the aagnetic gratings and

sensor head.

Characteristics of the圃agneti c gra ti n買S

Several kinds of steels were used to exaaine the characteristics of

magnetic gratings. The s8lple steels are listed on Table 7.6.1 with the

specifications.

(A) A case of fel・I・0圃agneti5・The listed salple steels were irradiated by electron and laser bea皿 for

laking test pieces. Hardness distributions were leasured on a section 01 ir-

radiation spots， because of a good correlation between the ・agnetichardness

J A E R I - M 88-221

and the mechanical hardness. A temperature dependency was examined to direct-current magnetization characteristics. The results obtained from the examinations of these test pieces are as follows; 1) On the ferromagnetic steels, formation of magnetic gratings is confirmed. 2) On the hardened sample by beam irradiation, a residual magnetization intensity >s proportional to degree of hardening. 3) A slope of the B-H curves is modified to high pitch but a maximum magnetic flux density decrease by heating. 4) A change of magnetic characteristics is little up to 300'C, since the Curie temperature of the ferromagnetic steels is high enough. (B) A case of non-magnetism

In the same way as the case of ferromagnetic materials, test pieces of austenitic stainless steel were made by beam irradiation. A different texture from the austenitic base was observed by metal lographic tests. This experimental fact shows that a ferromagnetic ferritic texture was precipitated due to rapid cooling of the melting spot. The Ni-equivalence value and the Cr-equivalence value give us several percent of ferritic precipitation grains on the Schaeffler's phase diagram. The results obtained from the examinations and the measurements of the permeability are as follows; 1) The ferromagnetic texture(ferritic grains) was precipitated. 2) An amount of the ferritic precipitation grains was measured to 1.5—^2.0%. 3) The hardness of melting part is equal to based part.

The characteristics of the thermal demagnetization were tested on the magnetic gratings of this case. The residual magnetization has not changed by long term heating without decrease on the first stage. The age-processing (eq. suitable heat processing) based on this temperature behavior have an effect on making stable heat-resistivity. 4) The precipitated ferritic grains have the high Curie temperature.

Examples of magnetization pattern are shown in Figs. 7.6.3 and 7.6.4 on the non-magnetism and the ferromagnetism. In these feasibility studies, the authors have good anticipation to make a practical model of the heat-resistant encoder and scale. This method of producing the heat-resistant magnetic gratings was applied for Japanese patent.

- 1 8 5 -

]AERI-M 88・221

and the 田echanical hardness. A temperature dependenc~ was exa皿ined to

direct ・current・agnetizationcharacteristics. The results obtained fro皿 the

exaainations 01 these test pieces are as follows;

1) On the ferro圃agneticsteels. for田ationof圃agneticgratings is confir田ed.

2) On the hardened sa圃pleb~ bea田 irradiation. a residual 田agnetization in-

tensity ;s proportional to degree of hardening.

3) A slope of the B-H curves is modified to high pitch but a maximu田 mag-

netic flux density decrease by heating.

4) A change of ugnetic characteristics is little UP to 300・C. slnce the

Curie te皿peratureof the ferromagnetic steels is high enough.

(B) A case of non-magnetlsm

ln the sa圃eway as the case of ferromagnetic uterials. test pieces of

austenitic stainless steel were made by bea周 irradiation. A different tex-

ture fro田 theaustenitic base WaS observed b~ 皿etallographic tests. This ex-

perimental fact shows that a ferromagnetic ferritic texture was precipitated

due to rapid cooling of the melting spot. The Ni-equivalence value and the

Cr-equivalence value give us several percent of ferritic precipitation

grains on the Schaeffler's phase diagram. The results obtained fro圃 theex-

a田inations and the皿easure圃entsof the per圃eability are as follows:

1) The ferro皿agnetictexture(ferritic grains) was precipitated.

2) An amount of the ferritic precipitation grains was圃easuredto 1.5~2.0%.

3) The hardness of melting part is equal to based part.

The characteristics of the ther踊alde圃agnetizatic.n胃eretested on the

magnetic gratings of this case. The residual圃agnetizationhas not changed

by long term heating without decrease on the first stage. The age-processing

(eq. suitable heat processing) based on this te圃peraturebehavior have an

effect on皿akingstable heat-resistivity.

4) The precipitated fer・riticgrains have the high Curie te・perature.Exa皿plesof aagnetization pattern are shown in Figs. 7.6.3 and 7.6.4 on

the non-田agnetis圃 andthe ferro・agnetis圃. In these feasibi 1 i ty studies. the

authors have good anticipation to 圃akea practical ・odel of the heat-

resistant encoder and scale. This aethod of producing the heat-resistant

aagnetic gratings was applied for Japanese patent.

-185一

J A E R I - M 88-221

Table 7.6.1 The designation and sDecifications of the sample steels Designation f Specifications Ferromagnet ism

Carbon steel for machine structural use; S35C Carbon steel for machine structural use; equivalent to A IS 11035

SPCC Cold rolled carbon steel sheet for general use; equivalent to that defined by ASTM A 109-72

SUS410 Martensit'ic stainless steel; equivalent to AIS1410 SUS430 Ferritic stainless steel; equivalent to A IS [430 Invar Low thermal expansion

Non-magnet ism SUS304 Austenitic stainless steel; equivalent to A IS 1304

Electron/laser beam

Melting spot

Formatted gratings <£/ Movement of specimen

Fig. 7.6.1 Experimental apparatus of the beam irradiation.

Vertical component, magnetized with electromagnet

Fig. 7.6.3 Magnetization pattern of the SUS304.

Heat-resistant sensor head (pick up coi 1)

Scale beam

///////W//// <=

Magnetic gratings Fig. 7.6.2 A conceptional drawing

of the magnetic scale.

(a) Vertical Component

^-\y-^7 ^ Arrows %r\ow electron-beam

irradiated positions.

(b) Horizontal Component

V\ V\ T^ y„

v v \y Fig. 7.6.4 Magnetization pattern of

the S35C.

- 1 8 6 -

]AERI-M 88・221

Table 7.6.1 The designation and soecifications of the sa田olesteels

Designation Specifications Ferromagnetis!!l

S35C Carbon steel for machine structural use; equivalent to AISII035

SPCC Cold rolled carbon steel sheet for general use; equivalent to that defined by ASTM A 109-72

SUS410 Martensitic stainless steel: equivalent to AISl410 SUS430 Ferritic stainless steel; equivalent to AISI430 1 nve.r Low thermal expansion

ーーーーーーーーーー--ーーーー -・・・・・・・・・圃・・回・・・・司・・・・..._--唱----ー-_.・・ 4・・調----"・・・----"・・a・・・・・・圃._----圃圃圃・・・・・・圃・・・・・・圃---圃・・・・・・・・--司---.. 聞

Non-magnet i sm SUS304 Austenitic stainless steel; equivalent to AISI304

Electron/laser bea田

Formatted gratings Move田entof specimen

A conceptional drawing of the magnetic scale.

Fig. 7.6.2

11 1 L ___.Lーペブーにブ'...J

Experi皿ental apparatus of the beam irradiation.

7.6.1 Fig.

A'〒。w' SI'噌w elec:lron-be・m.rr・dja'aCl oos・l.ons.

(bJ Hori:tontal C。開。。円enl

的凶凶明=〈ωz-LF同目的Z凶ex=-』

u--凶=。，、z

-aw nH

aしWnLn

nv・'A-L

m幽wwau

nv，nu

SM翁ιw

nuφιF

.，Lauφι

φLnHnv

Fae。。しw

awau，aa

凶凶凶同

」F一一目

Z凶ロ

H=」h-u一-F凶

Z口42

Magnetization pattern of the S35C.

Fig. 7.6.4

ーーーーー一一」

Magnetization pattern of the SUS304.

Position

Fig. 7.6.3

J A E R I - M 8 8 - 2 2 1

K. Teranishi* and H. Gotoh

In the previous report *', the authors described a calculation method of gamma-ray peak area. This method can be applied for the peaks having slopes in the lower and upper background regions.

In this report, we systematize the measuring results using this calculation method. When we deal with a peak, it is important to take into account next two conditions: (A) life can use either the linear expression of counts in the lower and upper background regions or the rational one; the symbol which we use for each is L or R. respectively. (B) ffe can neglect the slopes of the lower and upper background regions or

not; it is expressed as flat (F) or inclined (I). Combination of these conditions makes four basic kinds of methods,

namely L.F, L.I. R.F and R.I methods for the peak area calculation as shown in Fig. 7.7.1, In the mathematical expression, R.I method is the most complex and L.F method is the most conventional. R.I method without slopes in the background regions corresponds to R.F method and the former with fixed peak position becomes L.I method. L.F method corresponds to L.I method without slopes in the background regions and at the same time it means R.F method with fixed peak position. The peak area is determined by the sum of the surplus counts over the baseline from the start channel p2 to the end channel p3 in the peak region. An algorithm resulted from these methods was coded in the BASIC language for a personal computer PC-9801 to determine the peak position, the peak area and their standard error.

Results for the nine peaks of Eu-152 gamma-rays using these four methods are summarized as follows.' (1) For the area determination of a peak having statistically poor distribution such as the 586 keV peak, all the methods are applicable and the differences among the methods are not detectable. (2) In case of a peak with statistically good distribution such as the 122 keV peak, there are markedly different slopes in the lower and upper background regions respectively, therefore both L.l and R.1 methods are available.

* Hitachi Ibaraki Technical College, Hitachi Ltd.

- 1 8 7 -

JAERI-M 88・221

K. Teranishix and H. Gotoh

In the previous report 1) the authors described a calculation 回ethod

of ga田圃a-raypeak area. This 皿ethodcan be applied for the peaks having

slopes in the lower and upper background regions.

In this report， we systematize the measuring results using this cal-

culation 田ethod. lYhen we deal with a peak， it is 圃portantto take into ac-

count next two conditions:

(A) We can use either the linear expression of counts in the lower and upper

background regions or the rational one; the subol which we use for each is

L or R. respectively.

(B) We can neglect the slopes of the lower and upper background regions or

not; it is expressed as flat (F) or inclined (I).

Combination of these conditiuns makes four basic kinds of methods，

namely L.F. L. [， R.F and R.I 皿ethodsfor the peak area calculation as shown

in Fig. 7.7.1. In the mathe圃aticalexpression， R.I 圃ethod is the most com-

plex and L.F method is the most conventional. R.I 闇et h od w i t h 0 u t s 10 pe s i n

the background regions corresponds to R.F皿ethodand the former with fixed

peak posi tion becolles L. I田ethod. L.F method corresponds to L. f method

without slopes in the background regions and at the saJIe time it means R.F

田ethodwith fixed peak position. The peak area is deter田ined by the su目。f

the surplus counts over the baseline from the start channel p2 to the end

channel p3 in the peak region. An algorith皿 resultedfro圃 these皿ethodswas

coded in the BASIC language for a personal co圃puterPC-9801 to determine the

peak position， the peak area and tr.eir standard error.

Results for the nine peaks of Eu-152 gamla-rays using these four

阻ethodsare sU.larized as follows:

(1) For the area deterlination of a peak having statistically poor distribu-

tion such as the 586 keV peak， all the methods are applicable and the dif-

ferences among the圃ethodsare not detectable.

(2) fn case of a peak with statistically good distribution such as the 122

keV peak. there are田arkedlydifferent slopes in the lower and upper back-

ground regions respectively， therefore both L. 1 and R.I methods are avai 1-

車 Hitachi lbaraki Technical College， Hitachi Ltd.

J A E R I - M 8 8 - 2 2 1

(3) When we deal with a peak having the different slopes in the lower and upper background regions and a few channels in the peak region (when using bunched data), R.I method is more suitable than L. I method because the peak position of the former is automatically deterained in accordance with the shape of the peak. (4) In contrast to (3), for a peak having no slope in the background regions and a few channels in the peak region (when using bunched data), all the methods are applicable and the differences among the methods are not detectable. (5) We examined the fluctuation in the magnitude of the peak areas by changing the start channel of the narrow background of 10 channels in width in the range of the wide background of 50 channels respectively in both background regions. In case of the 122 keV peak by using R.I method, the fluctuations of the peak area determination were within the statistical errors of the peak areas by using this method.

Reference 1) Gotoh H. and Teranishi K. : A Rational Estimator for Peak Area Calculation, JAERI-M 87-126,pp.205-207(1987).

- 1 8 8 -

]AERI-M 88・221

(3) When we deal with a peak having the different slopes in the lower and

upper background regions and a few channels in the peak region Cwhen using

bunched data)， R.I method is lUore suitable than L. I圃ethodbecause the peak

position of the former is auto皿aticallydetermined in accordance with the

shape of the peak.

(4) In contrast to (3)， for a peak having no slope in the background regions

and a few channels in the peak region Cwhen using bunched dat!1)， al I the

methods are applicable and the differences a皿ongthe methods are not detec-

table.

(5) We examined the fluctuation in the magnitude of the peak areas by

changing the start channel of the narrow background of 10 channels in width

in the range of the wide background of 50 channels respectively in both

background regions. In case of the 122 keV peak by using R.I J日 thod， the

fluctuations of the peak area determination Were within the statistical er-

rors of the peak areas by using this method.

Reference

1) Gotoh H. and Teranishi K. A Rational Estimator for Peak Area Calcula-

tion， JAERI-M 87-126，pp.205-207(1987).

J A E R I - M 8 8 - 2 2 1

I P , ? », m i\

Pulic Hciqhl t channclt I

( a ) L.F method

I P ] t> Pj m (i

Pulic Height I channels )

t I P, P V ~ P, m n

Pulic Htiqni (channel! I

(b) R.F method

l e c i -qrountf

ftoe uqion "UPK»

9fcun4 riqMn

j r- ; 1

j ^ = ^ «/*

1 »ii-1* I I

h, i..

1 li 1 1 I ',

Pulie Helqhi (channcii)

(c) L.1 method (d) R.[ method

k.l=start and stop channel number of the lower background region, p2,p3=start and stop channel number of the peak region, p=channel number which gives maximum counts Cp in the peak region, p=peak position which is defined as the channel position with a rational number which is such that half of the sum of the counts above the baseline in the peak region lies to its left and half to its right, m,n=start and stop channel number of the upper background region , hl,h4=height at p2 and p3 respectively, h2,h3=height of both sides of the step at p , sl,s4=height of the step of the multi-stage line in the lower and upper sides of a peak

Fig. 7.7.1 Four basic kinds of methods for peak area calculation

]AE R l-M 88圃 221

v.阿，

....・.. ..... It脚・

Pu， I・"・・・、......，・-..・"・."聞

百gc-au

、嶋君、au

」引ーーー

9・・・ "・幡・・lowlI

*.. ~.曲帽

F岬刷

、-z・3AWM

(b) R.F method

内Jlu tlclqhl { ehonntl‘l

(a) L.F method

U"..-， "CI~ ~ ...... r叫嗣

内岨 ..~・"ーーーーーI.ltWtr IOU-

w・H噌『叫剛

-ug-Z叫、=gdou

u，... M凶 -

.. .. 肺d

..~酬

-m-aaeau

、Aw--Euu

P，II¥ pl.官

PUIU Helqhl (chonrtcU I

P， I cnannfls I

(c) L. I method

Ifclqhl 向』叫

(d) R. I method

!ower background region， k.l=start and stop channe! number of the

p2.p3=start a日dstoレchanne! number of the peak region，

in the peak region，

the channel position with

p=channel number which gives maximu悶 countsCp

ii=peak position which is defined as a rational

the base 1 i ne

left and half to its right，

m.n=start and stop channel number of the upper background region J

h1.h4=height at p2 and p3 respective!y，

the counts above that half of the sum of

in the peak region lies to its

i s such which number

and upper the lo¥ver In 1 i ne

h2.h3=height of both sides of the step at P J

the ;nulti-stage the step of sl.s4=height of

sides of a peak

Four basic kinds of 田ethods for peak area calculation

一189-

Fig.7.7.1

J A E R I - M 8 8 - 2 2 1

8. Reactor Control, Diagnosis and Robotics

As for the reactor control, a new code has been developed to analyze the dynamic behavior of the nuclear propulsion ship "Mutsu". Typical dynamics simulations show a good agreement with "RETRAN" calculations. Continuous study has been made on a CAD system for control system analysis with respect to simulating nonlinear system. A knowledge-based closed-loop control method was developed and applied to optimal control problems involving reactor poisoning in a high flux thermal reactor. With a view to enhance operational safety of High Temperature Test Reactor (HTTR), a computerized control and instrumentation system is studied with special attention on further automation of control, improvement of man-machine interface and so on.

As for the diagnostics study, continuous effort has been made on the expansion of a computer code STAR-II as the second version of the original code STAR. In order to perform on-line data acquisition and recording of reactor noise, the development of a data recording system is in progress. The goodness of system identification using Multivariate Autoregressive modeling method has been studied by computer simulation based on a 2-dimensional feedback system using a hybrid computer. A method of nonstationary noise analysis was tested and demonstrated using frequency-time spectra and was applied to the analysis of Borssele reactor noise data. The experimental data of in-core sodium boiling obtained at the Russian fast reactor BOR-60 have been analyzed.

As for robotics research and development, theoretical study on manipulator inverse kinematics has produced three new approaches: a polynomial expression of kinematic relation, a minimization technique of mathematical programming and a direct derivation of individual joint solutions. As an alternative to the matrix operation for spatial position control of a robot manipulator, an expression of recursive relations suitable for computer aided generation of the manipulator kinematics is proposed. In the development of remote handling technology, the amphibious robot manipulator JARM-25 has been modified and improved based on the findings from its mockup test. Concerning the robotic remote handling technology for fusion reactor, a preliminary study was also initiated to develop new design concept for invessel robot manipulators.

(Yoshikuni Shinohara)

JAERI-M BB-221

8. Reactor Control. Diagnosis and Robotics

As for the reactor control， a new code has been developed to

analyze the dynamic behavior of the nuclear propulsion ship "Mutsu".

Typical dynamics simulations show a good agreement with "RETRAN" cal-

culations. Continuous study has been made on a CAD system for control

system analysis wi th respect to simulating nonlinear system. A

knowledge-桐 basedclosed-loop control method was developed and applied to

optimal control problems involving reactor poisoning in a high f1ux

thermal reactor. With a view to enhance operational safety of High Tem-

perature Test Reactor (HTTR) ， a computerized control and instrumentation

system is studied with special attention on further automation of con-

trol， improvement of man-machine interface and so on.

As for the diagnostics study， continuous effort has been made on

the expansion of a computer code STAR-II as the second version of the

original code STAR. In order to perform on-line data acquisition and

recording of reactor noise， lhe development of a data recording system

is in ~rogress. The goodness of system identification using Multivariate

Autoregressive mode1ing method has been studied by computer simulation

based on a 2-dimensional feedback system using a hybrid computer. A

method of nonstationary noise analysis was tested and demonstrated using

frequency-time spectra and was applied to the analysis of Borssele

reactor noise data. The experimental data of in-core sodium boiling ob-

tainea at the Russian fast reactor BOR-60 have been analyzed.

As for robotics research and development， theoretical study on

manipulator inverse kinematics has produced three new approaches: a

polynomial expression of kinematic relation， a minimization technique of

mathematical pl'ogramming and a direct derivation of individual joint

solutions. As an alternative to the matrix operation for spatial posi-

tion control of a rohot manipulator， an expression of recursive rela-

tions sui table for computer aided generation of the manipulator

kinematics is propos~d. In the development of remote handling technol-

ogy， the amphibious robot manipulator JARM-25 has been modified and im-

proved based on the findings from its mockup test. Concerning the

robotic remote handling technology for fusion reactor， a preliminary

study was a1so initiated to deve10p new design concept for invessel

robot manipulators.

(Yoshikuni Shinohara)

-1伺ー

JAERI-M 88-221

8.1 Dynamics Analysis Code for Nuclear Ship Propulsion Reactor

K. Nabeshima, J. Shimazaki, T. Kusunoki*, T. Yokomura+ and Y. Shinohara

Most of the dynamics analyses for the nuclear ship "Mutsu" have been carried out by using "RETRAN" code 1 ' , originally developed for accident analysis of nuclear power plants. This code requires much computation time and lacks models of some components which are important for dynamics simulation. A new analysis code has been developed in order to analyze dynamic behavior of nuclear ship induced by large disturbances by modifying an already developed PWR plant dynamics analysis code "PWRDYN '.

Main modification was made in the secondary system model which plays an important role in nuclear ship dynamics, including some special components such as the propulsion system and astern turbine. A simplified dynamic model of the nuclear ship "Mutsu" is shown in Fig. 8.1.1. The secondary system is composed of a turbine, a propulsion system, heat exchangers and control valves, and is coupled to the primary system through a steam generator.

Typical dynamic simulations for cases of rapid power increase/ decrease, crash astern/ahead, etc. were performed for the nuclear ship Mutsu model. Fig. 8.1.2 shows the responses of the plant variables in case of rapid power increase. The power increased from 27% to 100% in 30 sec, where 27% power is base load of Mutsu. The results of simulations are reasonable and in good agreement with RETRAN calculations with a maximum error of about 10%. The CPU time for a time step of 50 ms is less than 1/5 of real time and about 20 times shorter than that required in RETRAN simulation. It can be concluded that this code is effective for dynamics analysis of nuclear ship plants and can be utilized in a nuclear ship simulator.

At present this code is going to be modified for dynamics analysis of the integrated pressurized water reactor for nuclear ship, one of the trial design by JAERI. This reactor plant with a once-through steam generator has more sensitive response than the Mutsu plant with a U-tube steam generator.

+ Office of Nuclear Ship Research and Development.

- 1 9 1 -

JAERI-M 88・221

8.1 Dynamics Analysis Code for Nuclear Ship Propulsion Reactor

K. Nabeshima， J. Shimazaki， T. Iくusunoki+，T. Yokomura+ and

Y. Shinohara

Most of the dynamics analyses for the nl.iclear ship "Mutsu" have been

carried out by using !'RETRAN" codeJ)， originally developed for accident

analysis of nuclear power plants. This code requires much computation time

and lacks models of some components which are important for dynam ics

simulation. A new analysis code has been developed in order to analyze

dynamic behavior of r.uclear ship induced by large disturbances by modify-

ing an already developed PWR plant dynamics analysis code "PWRDYN，，2).

Main modification was made in the secondary system model which

plays an important role in nuclear ship dynamics， including some special

components such as the propulsion system and astern turbine. A simplified

dynamic model of the nUclear ship "Mutsu" is shown in Fig. 8.1.1. The

secondary system is composed of a turbine， a propulsion system， heat ex-

changers and control valves， and is coupled to the primary system through

a steam generator.

Typical dynamic simulations for cas~s of rapid power increase/

decrease， crash astern/ahead， etc. were performed for the nuclear :，!1ip

Mutsu model. Fig. 8.1.2 shows the responses of the plant variables in case

of rapid power increase. The power increased from 27% to 100% in 30

sec， where 27% power is base load of Mutsu. The results of simulations

are reasonable and in good agreement with RETRAN calculations with a

maximum error of about 10%. The CPU time for a time step of 50 ms is

less than 1/5 of real time and about 20 times shorter than that required

in RETRAN simulation. It can be concluded that this code is effective for

dynamics analysis of nuclear ship plants and can be utilized in a nuclear

ship simulator.

At present this code is going to be modified for dynamics analysis of

the integrated pressurized water reactor for nuclear ship， one of the trial

design by JAERI. This reactor plant with a once-through steam generator

has more sensitive response than the Mutsu plant with a U-tube steam

generator.

+ Office of Nuclear Ship Research and Development.

-19] -

JAERI-M 88-221

Reference 1) EPRI NP-1850:"RETRAN-02, A Program for Transient Thermal-Hydraulic

Analysis of Complex Fluid Flow Systems" (1981) 2) Yokobayashi,M.:"PWRDYN: A Computer Code for PWR Plant Dynamics

Analysis," JAERI-M 82-064 (1982) (in Japanese)

Low Pressure Steai

Fig .8 .1 .1 Plant Model of Nuclear Ship "Mutsu"

-T 1 1 1 1 1 r-

* z Normalized Steam Flow -

T 1 1 1 1 1 1 1 1 r--t 1 P-

« J - i 1 1 1 1 1 1 r-—i 1 r 1 1 1 1 r-

Normalized Reactor Power

- i — i — i — i — I

Ship Speed (knot)

- • . . ' . - ' - — ' — ' — ' — • — ' — i — ' — • — ! — i 1 — ' — 0.00 10.03 ItO.OO 240.00 J20.00 400.00 410.00 JiO.OO.

T l f l E I S c C )

Fig .8 .1 .2 Plant Response (rapid power increase)

- 1 9 2 -

JAERI-M 88・221

Reference

1) EF'Rl NP-1850:"RETRAN-02， A Program for Transient Thermal-Hydraulic

Analysis of Complex fluid flow Systems" (1981)

2) Yokobayashi，M.:"PWRDYN: A Computer Code for PWR Plant Dynamics

Analysis，lI jAERI-M 82・064(1982) (in j apanese)

Fig. 8 .1.1 Plant Model of Nuclear Ship IIMutsu"

言ーーーo

Normalized Steam flow

Norma¥ized Reactor Power

-0・岨

唱。ω島国

a-S回

Ship Speed (knot)

g a ・-0.00 10.00 1・0.00 240.0・'20.00 ‘00.0・4・0.00 ・0.00

T1MEI SEC)

Fig. 8.1. 2 Plant Response (rapid power increase)

J A E R I - M 8 8 - 2 2 1

8.2 Nonlinear Simulation Method in the CAD System for Control System Analysis of Nuclear Plant

J. Shimazaki and K. Nabeshima

A CAD(Computer Aided Design) system for control system analysis (Version 1) has been developed^ and modified ' in its dynamic simulation method for more wide use. This system has good characteristics in the methods of describing dynamic model, dynamic simulation, interactive control design system and data management. Time interval used in dynamic simulation can be chosen so as to evaluate the transient within 2000 time steps, because the discrete model of nuclear plant with a relatively large time interval is made by the Pade approximation method newly developed. The CAD system is designed in the form of state-space representation for effective simulation of a large scale system. Therefore transfer function model useful to simulate control systems and simple dynamic components is reformed into a state-space representation (observer canonical form) and then the discrete model is autohdtically generated in the revised CAD system.

In the dynamic simulation there are three methods for simulation of nonlinear system as follows : (1) Nonlinear function method corresponding to components with nonlinear static characteristics. (2) Artificial control variable method extended variables introduced for nonlinear terms in the form of state-space representation. (3) Linearization method the linearization for each time step. The above three methods hold their steady state consistent with exact solution. However, dynamic responses depend on the method used and time interval. The third method can be generally used for treatment of nonlinear dynamic simulation, but it takes much time to make a discrete model even if the second order Pade approximation is applied. Figure 8.2.1 shows the difference among those methods and exact solution in case of a dynamic simulation of reactor neutron kinetics with temperature reactivity feedback. The exact solution is calculated by the Runge-Kutta method with a very small time interval chosen (ten times smaller than by the proposed methods). The second method gives the almost same response to the result made by the linearized method or the exact solution. This method has advantages in computation time and simple

]AERI-M 88・221

8.2 Nonlinear Simulation Method in the CAD System for Control

System Analysis o[ Nuclear Plant

J. Shimazaki and K. Nabeshima

A CAD(Computer Aided Design) system for control system ana1ysi5

(Version 1) has been developed1) and modified2) in its dynamic simula-

tion method for more wide use. This system h&s good characterist:Lcs in

the methods of describing dynamic model， dynamic simulation， interactive

control design system and data management. Time interval used in dynamic

simulation can be chosen so as to evaluate the transient within 2000

time steps， because the discrete model of nuclear plant wi th 3 rela-

tively large time interval is made by the Pade approximation method

ne~ly developed. The CAD system is designed in the form of state-space

representation for effective simulation of a large scale system. There-

fore transfer function model usef ul to simulate control systems and

simple dynamic components is reformed into a state-space representation

(observer canonica1 form) and then the discrete model 1s autol'utically

generated 1n the revised CAD system.

In the dynamic simulation there are three methods for simulation of

nonlinear system aS follows

(1) Nonlinear function method ---corresponding to components with non-

linear static characteristics.

(2) Artificial control variable method ---extended variables introduced

for nonlinear terms in the form of state-space representation.

(3) Linearizat工onmethod ---the linear工zationfor each time step.

The above three methods hold their steady state consistent with exact

solution. However， dynamic responses depend on the method used and time

interval. The third method can be generally used for treatment of non-

li:J.ear dynamic simulation， but it takes much time to make a discrete

model even if the second order Pade approximation is applied. Figure

8.2.1 shows the difference among those methods and exact solution in

case of a dynamic simulation of reactor neutron kinetics with tempera置

ture reactivity feedback. The exact solution is calculated by the Runge-

Kutta method with a very small time interva1 chosen (ten times smaller

than by the proposed methods). The second method gives the almost same

response to the result made by the linearized method or the exact solu-

tion. This method has advantages in computation time and simple

J A E R I - M 8 8 - 2 2 1

modification in the CAD system.

References

1) Shimazaki J., Shinohara Y.: "A CAD System for Control System Design and Evaluation (Version 1)," JAERI-M 82-180 (1982) (in Japanese).

2) Shimazaki J., Usui H., Shinohara Y.: JAERI-M 83-129,182-183 (1983).

Reactor Power(variation from normalized power)

Disturbance : 0.2$ step

Method (2) (almost same as Method (3) and Exact Solution)

Linearized Model

•fi B-

n . n I. i i i i i i i t i I i .i..t i i i i i i I i i i i i i i i P I 0-0 0.500 1.00 1-50

TIME(sec) Fig.8.2.1 Dynamic Simulation by the Proposed Method

JAERI-M 88・221

modification in the CAD system.

References

1) Shimazaki J.， Shinohara Y.: "A CAD System for Control System Design

and Evaluation (Version 1)，" JAERIート182-180 (1982) (in Japonese).

2) Shimazaki J.， Vsui H.， Shinohara Y.: JAE!日町ト183-129，182-183 (1983).

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0.1 Linearized Model

0.0 0.0 1.50

Fig.8.2.1 Dynamic Simulation by the Proposed Method

J A E R I - M 8 8 - 2 2 1

8.3 Study of Optimal Reactor Control Using an AI Method

Y. Shinohara

A knowledge-based closed-loop control method has been applied to optimal control problems involving reactor poisoning in a high flux thermal reactor.

Although optimal reactor control problems have been solved analytically for the models which are much simplified for ease of mathematical treatment, rigorously optimal control laws can seldom be obtained analytically for more complex processes in an actual reactor. Therefore, practical methods for near-optimal or practically optimal closed-loop control of reactor processes may be obtained with the help of heuristics.

In the method presented here, a knowledge-based technique like an expert system is used and qualitative and quantitative control rules are constructed based on the knowledge of the structure of the optimal control laws and/or control program patterns obtained for a simplified process model. The basic assumption adopted is that the essential features of the optimal control for a simplified model will be maintained also for more complex process models if the simplified model is a first order approximation to the real process. Such an assumption is considered reasonable based on the results of various optimization studies made so far. Thus, the knowledge of optimal control laws obtained in the phase space theoretically for a two-dimensional process model as well as the knowledge of optimal control program patterns obtained in the time domain by numerical methods such as optimum seeking methods is used as a basis of developing near-optimal control rules for more realistic process models.

The knowledge base consists of basic qualitative and quantitative control rules, adjustment rules, process dynamics and constraints on control and state variables. At each control time step, the system state is estimated from directly measurable process variables using a state variable estimator. Then the control signals generated based on the estimated current system state and the corresponding control rules are checked, prior to their execution, for their validity and safety with respect to the constraints by predictive evaluation of the system state

]AERI-M 88・221

8.3 Study of Optimal Reactor Control Using an AI Nethod

Y. Shinohara

A knowled.ge-凶 sed closed-loop control method has been applied to

optimal control problems involving reactor ~コisoning in a high flux

thermal reactor.

:U though optimal reactor control proble悶 have been solved

analytically for the models which are much simplified for ease of

mathematical treatment， rigorously opヒimalcontrol laws can seldom be

obtained analytically for more complex processes in an actual reactor.

Therefore， practical methods for near-optimal or practically optimal

closed-loop control of reactor processes may be obtained with the help

of heuristics.

In the method presented here， a knowled.ge-based technique like an

e~rt system is used and qualitative and quantitative control rules are

constructed based on the knowledge of ~he structure of the optimal

control laws and/or control program patterns obtained for a simp1ified

process model. The basic assumption adopted is that the essential

features of the optimal control for a simplified model will be maintained

a1so for more complex process models if the simplified model is a first

order approximation to the rea1 process. Such an assumption is considered

reasonable based on the results of various optimization studies made so

far. Thus， the knowled.ge of optimal control laws obtained in the phase

space theoretically for a two-dimensional process model as well as the

knowled.ge of optimal control program patterns obtained in the time domain

by numerical methods such as optimum seeking methods is used as a basis

of developing near-optimal control rules for more realistic process

models.

The knowled.ge base consists of basic qualitative and quantitative

control rules， adjustment rules， process dynamics and constraints on

control and state variables. At each control time step， the system state

is estirr泡 ted from directly measurable process variables using a state

variable estimator. Then tlle contrつ1signals generated based on the

estimated current system state and the corres~コnding control rules are

checked， prior to their execution， for their validity and safety with

respect to the constraints by predictive evaluation of the system state

-195一

J A E R I - M 8 8 - 2 2 1

using the process dynamics model of the system. If the control signals are not feasible they are modified according to the adjustment rules until they become feasible.

The present method was implemented on a microcomputer and was applied to obtain near-optimal solutions to the optimal reactor shutdown control problems involving reactor poisoning as concrete examples. Simulation study was performed for the following two problems using point reactor models.

The first one is the time-optimal reactor shutdown control problem involving only xenon poisoning. The basic control rules are constructed from the knowledge of the optimal control laws and program patterns which have been obtained theoretically for a simplified model in which the state variables are the xenon and iodine concentrations and the control variable is the neutron flux. In order to demonstrate the validity of the present method, it was first applied to the simplified model. The results obtained by this method are shown to be close enough to those obtained theoretically. The more complex process model simulated includes the neutron flux as a state variable and other reactivity effects.

The second one is the problem of optimal reactor shutdown control for minimizing the reactor poisoning after shutdown. In this problem both xenon and samarium poisonings are taken into consideration. This problem has not been solved theoretically even for a simplified process model. Therefore, the basic control rules are constructed based on the knowledge of the features of optimal control program patterns which has been obtained numerically using an optimum seeking method with random search.

The results of computer simulation show that the solutions obtained by the present knowledge-based control method are sufficiently close in most cases tested to the optimal solutions obtained theoretically or numerically and, therefore, this method gives near-optimal solutions to the optimal reactor control problems studied. The near-optimality of the solutions obtained is also checked by surveying different results obtained by modifying parametrically the timings and magnitudes of the control actions.

jAERI-M 88・221

using the process dyn創nicsmodel of the system. If the control signals

are not feasible they are modified according to the adjustment rules

until they become feasible.

The present method was implemented on a microcomputer and was

applied to obtain near-optimal solutions to the optin凶 1reactor shutdown

control problems involving reactor poisoning as concrete examples.

Simulation study was performed for the following two problems using point

reactor models.

The first one is the time-optimal reactor shutdown control problem

involving only xenon poisoning. The basic control rules are constructed

from the knowledge of the optimal control laws and program patterns which

have been obtained theoretically for a simplified model in which the

state variables are the xenon and iodine concentrations and the control

variable is the neutron flux. 1n order to demonstrate the validity of the

present method. it was first applied to the simplified model. The rEゴsults

obtained by this method are shmm to be close enough to those obtained

theoretically. The more comp~ex process model simulated includes the

neutron flux dS a state variable and other reactivity effects.

The second one is the problem of optimal reactor shutdown control

for minimizing the reactor poisoning after shutdown. 1n this problem both

xenon and samarium poisonings are taken into consideration. This problem

has not been solved theoretically even for a simplified process model.

Therefore， the凶 siccontrol rules are constructed based on the knowledge

of the features of optimal control program patterns which has been

obtained numerically using an optimum seeking method with random search.

官leresults of computer simulation show that the sulutions obtained

by the present knowledge-based control method are sufficiently close in

most cases tested to the optimal solutions obtained theoretically or

numerically and， therefore， this method gives near-op七imalsolutions to

the optimal reactor control problems studied.叩 enear-optimality of the

solutions obtained is also checked by surveying different results

obtained by modifying parametrically the timings and magnitudes of the

control actions.

一196-

JAERI-M 88-221

8.4 A Design Study on Computer Control of HTTR Plant

J.Shimazaki and K.Suzuki

Fundamental design study of HTTR has completed and safety review by the government is going on FY-1988. In the design study of control and instrumentation system of HTTR, emphasis is placed on the development of advanced control and diagnostic systems for enhancing operational safety of HTTR.

An integrated supervisory control and instrumentation system is planned to be installed in HTTR so as to further advance automation of plant control and to meet various kinds of operation modes of HTTR. This system consists of micro-computer based systems for the individual control systems of HTTR such as reactor inlet/outlet helium-gas temperature controls, and of a supervisory control system to integrate them for start-up and shut-down operations, as shown in Fig.8.4.1.

The operational sequence control panel is specially designed in the main console to do the sequence control easily by simple push-bottom operation. Using this panel and CRT displays where operational sequences and important safety related parameters should be given, one man control of HTTR plant operation is possible. In order to supervise the state of plant operation and record the plant data, a process computer is applied for the display of plant safety summary and of the state of plant control for the operators but not for direct control of plant. The computerized control and instrumentation system is designed with the special attention from the reliability and safety points of view to the functions of self-checking, backup, fail-safe, redundancy and man-machine communication.

JAERI・M 88・221

8.4 A Design Study on Computer Control of HTTR Plant

J.Shimazaki and K.Suzuki

Fundamental design study of HTTR has completed and safety review by

the government is going on FY-1988. 1n the design study of control and

instrumentation system of HTTR， emphasis is placed on the development of

advanced control and diagnostic systems for enhancing operational safety

of HTTR.

An integrated supervisory control and instrumentation system is

planned to be installed in HTTR so as to further advance automation of

plant control and to meet various kinds of operation modes of HTTR. This

system consists of micro-computer based systems for the individual con-

trol systems of HTTR such as reactor inlet/outlet helium-gas temperature

controls， and of a supervisory control system to integrate them for

start-up and shut-down operations， as shown in Fig.8.4.1.

The operational sequence control panel is specially designed in the

main console to do the sequence control easily by simple push-bottom

operation. Using this panel and CRT displays where operational sequences

and important safety related parameters should be given， one man control

of HTTR plant operation is possible. 1n order to supervise the state of

plant operation and record the plant data， a process computer is applied

for the display of plant safety summary and of the state of plant con-

trol for the operators but not for direct control of plant. The com-

puterized control and instrumentation system is designed with the spe-

cial attention from the reliability and safety points of view to the

functions of self-checking， backup， fai1-safe， redundancy and man-

machine communication.

processor

F i g . 8 . 4 . 1 HTTR i n t e g r a t e d supe rv i so ry c o n t r o l systen

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system control superv1sory integrated HTTR Fig.8.4.1

J A E R I - M 8 8 - 2 2 1

8.5 Development of Computer Code STAR-II for Dynamics Analysis and Diagnosis of Nuclear Reactor System

K. Hayashi and K. Watanabe

The STAR-II, a computer code for dynamics analysis and diagnosis of nuclear reactor systems based on reactor noise analysis, has been developed as the second version of the original code STAR K It is programmed for a hybrid computer EAI HYSHARE-600 system in order to facilitate effective handling of the noise data in analog form. As the digital part of the system Gould 32/77 is used for which the operating system is MPX-32 VI.4 and the programming language is FORTRAN77+.

The STAR-II includes all programs necessary for the analysis work from data acquisition to display of graphs. The functions of these programs are listed in Table 8.5.1.

Two types of the data acquisition programs which use the CPU memory only for short time-range data and the magnetic tape device for long time-range data are provided. The control programs for the amplifiers and A/D converters and the utility programs for sampled data file management are also provided.

The analysis programs are classified into four groups as follows: (1) spectral analyses using Blackman-Tukey (BT) method, (2) univariate autoregressive (UAR) model fitting and its application, (3) multivariate AR model fitting and its application, (4) higher order spectral analysis.

The group (1) is applied to basic signal analysis and dynamics analysis of single-input/single-output systems and also used for checking the validity of a fitted AR model.

The group (2) is used for non-stationary signal analysis. The group (3) is used for signal analysis, system dynamics analysis

of a feedback system in both time and frequency domain and signal prediction analysis.

The group (4) are used for non-linear signal and system analysis. Finally, various diagnostic studies are also possible by combined

use of these programs.

Reference 1) Hayashi K. : "Computer Code STAR," JAERI-M 9761 (1981).

JAERI-M 88・221

8.5 Deve10pment of Computer Code STAR-II for Dynamics

Analysis and Diagnosis of Nuclear Reactor System

K. Hayashi and K. Watanabe

The STAR-II， a computer code for dynamics analysis and diagnosis of

nuclear reactor systems based on reactor noise analysis， has been

developed as the second version of the original code STAR1). It is

programmed for a hybrid computer EAI HYSHARE-600 system in order to

facilitate effective handling of the noise data in analog form. As the

digital part of the system Gould 32/77 is used for which the operating

system is MPX-32 Vl.4 and the programming language is FORTRAN77+.

The STAR-II includes al1 programs necessary for the analysis work

from data acquisition to display of graphs. The functions of these

programs are listed in Table 8.5.1.

Two types of the data acquisition programs which use the CPU memory

only for short time-range data and the magnetic tape device for long

time-range data are provided. The control programs for the amplifiers

and A/D converters and the utility programs for sampled data file

management are also provided.

The analysis programs are classified into four groups as follows:

(1) spectral analyses using Blackman-Tukey (BT) method，

(2) univariate autoregressive (UAR) model fitting and its application，

(3) multivariate AR model fitting and its application，

(4) higher order spectral analysis.

百le group (1) is applied to basic signal analysis and dynamics

analysis of single-input/single-output systems and also used for check-

ing the va1idity of a fitted AR model.

The group (2) is used for non-stationary signal analysis.

The group (3) is used for signal analysis， system dynamics analysis

of a feedback system in both time and frequency domain and signal pre-

diction analysis.

百legroup (4) are used for non-linear signal and system analysis.

Fina11y. various diagnostic studies are a1so posニib1e by combined

use of these programs.

Reference

1) Hayashi K. "Computer Code STAR，" JAERI-M 9761 (1981).

Table 8.5.1 List of programs of STAR-II and their functions

Programs Function

STR.POTS Set up gains of analog amplifiers in front of A/D converter STR.SMP1 Data acquisition in single shot sampling mode (save to disc) STR.FMGR Data file utilities for small size file on disc or tape

- data file directory - information block editor - data file converter for STAR-I and ECN type data files - data file copy (disc ond tape)

STR.F0U1 Display and print out of small size file data S.STAHDZ Standardization of data variances for small size file data

STR.SMP2 Data acquisition in continuous sampling mode (save to tape) STR.TMGR Data file utilities for large size file on tape

- data file directory - information block editor - print out of file data - output of file data to 0/A converters

STR.ONST Estimation of instantaneous basic statistics series using large size data

S.HUIXOK Estimation of auto and cross correlation functions using small size data

STR.ONCR Estimation of auto and cross correlation functions using large size data

S.PLTCOR Display and print out of correlation functions S.HULSPE Estimation of power and cross spectra (DT-method) S.PLTSPE Display and print out of spectra S.SCLFRE Estimation of frequency response and coherence functions for

single-input/single-output systems S.PLSFRE Display and print out of frequency response and coherence

functions

Programs Function

S.DISPEC Estimation of auto and cross bispectra and bicoherence (DT) S.FHISPE Estimation of auto and cross bispectra and bicohcrence (FF7) S.PLTBIS Display and print out of bispectra and bicoherence

STR.UAR2 Fitting of normal and instantaneous univariate AR model (Yule-WolkerpBurg.Kitagava-Akaike and Harple method)

STR.P3PN Application analysis using normal and instantaneous UAR models - Display and print out of model parameters - Estimation and display of auto correlation - Estimation and display of (frequency-time) power spectrum

S.HULVA2 Fitting of multivariate AR model S.PLTMAR Display and print out model parameters STR.PRDC One step prediction based on HAR model S7R.XPRD Display of predicted time series data

S.STAR44 Time domain analysis using HAR model - Estimation and display of impulse and step response

functions of overall open loop system and sub-system of a feedback system

- Estimation of time constant of transfer function

Frequency domain analysis using HAR model S.ARSPEC Fourier transform of HAR coefficients and Estimation of power

and cross spectra S.NOSPSD Estimation and display of noise source spectrum S.HARNPC Estimation and display of noise power contributed spectrum and

noise power contribution ratio S.HARFRE Estimation ond display of frequency response functions of

overall open loop system or sub-systems of a feedback system Fitting and display of transfer function

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Table 8.5.1 List of programs of STAR-II and their functions

Progfa・s Funct且副首 Programs Function

STR.POTS Set up g81ns of 8na10g a~p11fiers 1n front of A/D converter S.DrSPEC Estimation of outo and cross bispectra and bicoherence (DT)

STR.SHPl Data acqu1s1t1an 1n slngle shot sampling mQde (save to disc) S.FsISPE Estimotion of auto and cross bispcctra and bicohcrence (FFT)

STR.FHGR Data flle util1ties for s..811 size file on disc or tape S.PLTsIS Display and print Out of blspectra and bicoherence

-datn fl1e dlrectory

-1nforoation block ed1tor STR.UAR2 Fitting of normal and lnstantaneous unlvariate AR model

-data fl1e converter for STAR-I and ECN type data fl1es (Yule-Wnlker，surg，litagava-Akaike and Marple創ethod)

-d8ta file copy (disc and tape) STR.PSPN Application aoalysis using norlO31 and instantaneous UAR

STR.t'OUl Displ町 8ndprlnt out of smal1 size f11e dat8 models

S.STANDZ St8ndardizatlon of data voriances for small slze file data ーDlsplayand print out o{酎 delparameters

ーEst1mat10nnnd display of auto correlat10n

STR.SHP2 Data acquisition 1n cont1nuous sampling mode (sove to tape) -~stimation and display of (frequency-tiee) þo~er spectru掴

STR.TMGR Data file util1ti回 forlarge sizc file on tape

-data file direc:tory S.HULVA2 Fitting of mu1c1variate AR lIodel

-1nfor・ationblock ed1tor S.PLl刊AR Display ond print out model parameters

-print out of file dota STR.PR以 Oncstep prediction based on HAR aodel

-output of f11e data to D/A converters STR.XPRD Display of predicted ti.e series data

STR.ONST Estiaation o[ instantaneous bas1c statistics series using S.STAR44 Time domain analysis using HAR model

large s!zc data -Estimot10n and display of i.pulse and step response

S.HULCO宜Estiaationof 8UtO 8nd cross correlation functions using functions of overall open loop syste・andsub-syste・ofa

511a11 si:te da ta feedback system

STR.ONCR Estlaotion of auto and cross correl・t10nfunct10ns using ーEstimot10nof t1・~ constant of trensf~r function

lorge sizc data

S.PLTCOR Display and pr1nt out of correlat10n functions Frequency domain analysis using HAR ~odel

S.附 LSPE Esti副 tionof power and cross spectra (sT唱ethod) S.ARSPEC Fourier tronsform of HAR coeff1cients and Esti岡 tionof po~er

S.PLTSPE Display and print out of spectra and cross spectra

S.SGLFRE Estlmat10n of frequency response and cohcrence functions for S.NOSPSD Estimation and display of noise source spectru園

single-input!single-output systemS S.HARNPC Estimot10n ond disp10y of noise powcr c:ontrlbutcd spectrum and

S.PLSFRE D1sploy and pr1nt out of frequency response and coherence noise power contribution ratio

functions S.HARFRe Est1mot10n ond display of frequency response functions of

overall open loop system or sub-sJste.s of 8 feedback system

Fitting and display of transfer function

JAERI-M 88-221

8.6 Development of Nonstationary Reactor Noise Signal Recording System

K. Hayashi

A reactor noise signal recording system is being developed for studying early detection and diagnosis of anomalies in a nuclear power plant based on reactor noise analysis. This new recording system is capable of recording not only stationary but also transient signals observed during reactor power changes. It performs automatically a series of functions as follows: (1) separation of noise signals from base sensor output signals, (2) elimination of frequency components outside a desired frequency

band, (3) amplification of the noise signals up to an input voltage range of

A/D-converters, (4) digitization of noise signals, (5) transfer to a digital recording device in a host computer system.

In this fiscal year, the signal conditioner has been developed corresponding to the above mentioned items (2) and (3).

The signal conditioner is a device which eliminates useless frequency components outside the desired frequency passband and amplifies signals up to a voltage level high enough for A/D conversion or analog tape recording. In noise signal measurement, it is necessary that the device be adjusted rapidly to time-varying properties of input signals. However, conventional programmable conditioners are not capable of adjusting the amplifier gain because it is difficult to find an optimal gain value without monitoring the variation of signal properties.

In order to solve the above mentioned problem, a new technique of setting the gain of DC amplifier is used in this signal conditioner. It consists of a signal processing part and a control part. The former consists of 16 boards of channel unit each containing a high-pass filter, a low-pass filter, a DC amplifier, an attenuator and a gain controller. The latter consists of an internal microprocessor and a communication interface to an external controller. The block diagram of the system is shown in Fig.8.6.1. The specifications for each part are shown in Table 8.6.1.

In each channel unit, the filters and attenuator elements are programmable and directly controlled by manual switch operation or ex-

JAERI-M 88・221

8.6 Development of Nonstationary Reactor Noise Signal Recording System

K. Hayashi

A reactor noise signal recording system is being developed for

studying early detection and diagnosis of anoma1ies in a nuclear power

plant based on reactor noise ana1ysis. This new recording system is

capab1e of recording not on1y stationary but a1so transient signa1s ob-

served during reactor power changes. It performs automatica11y a series

of functions as fol1ows:

(1) separation of noise signals from base sensor output signa1s， (2) elimiuation of frequency components outside a desired frequency

band，

(3) amplification of the noise signa1s up to an input voltage range of

A/D-converters，

(4) digitization of noise signals， (5) transfer to a digital recording device in a host computer system.

In this fiscal year， the signal conditioner has been developeo cor-

responding to the above mentioned items (2) and (3).

The signal conditioner is a device which eliminates useless

frequency components outside the desired frequency passband and

amplifu::i sigコalsup to a voltage 1evel high enough for A/D conversion

or ana10g tape recording. In noise signal measurement， it is necessary

that the dev工cebe adjusted rapid1y to time-varying properties of input

signa1s. However， conventional programmable conditioners are not capable

of adjusting the amplifier gain because it is difficult to find an op-

timal gain value without monitoring the variation of signal properties.

In order to solve the above mentioned problem， a new technique of

setting the gain of DC amplifier is used in this signal conditioner. It

consists of a signal processing part and a control part. The former con-

sists of 16 boards of channel unit each containing a high-pass filter， a

low-pass filter， a DC amplifier， an attenuator and a gain controller.

The latter consists of an internal microprocessor and a communication

interface to an external controller. The block diagram of the system is

shown in Fig.8.6.1. The specifications for each part are shown in Table

8.6.1.

In each channel unit， the filters and attenuator elements are

programmable and directly control1ed by manual switch operation or ex-

J A E R I - M 8 8 - 2 2 1

ternal computer commands. The DC amplifier element is controlled by the gain controller. The gain controller has two functions as follows: 1) Search for an optimal gain value based on the input signal level. It

- detects current signal peak levels with a time constant 1, 10 or 100 sec,

- determines optimal gain ( voltage range ), - generates a control signal.

2) Setup of a gain value of the DC amplifier - to an optimal gain value by external computer command or manually, - to a gain value specified by manual switch or external computer.

An internal microprocessor and an IEEE-488 interface are prepared for the communication between this equipment and the external computer as a controller.

Table 8.C.] Specifications of signal conditioner

Channel Unit Number of channels Input and output voltages High-pass filter

Cut-off frequency Roll-off Type

Low-pass filter Cut-off frequency Roll-off Type

DC Amplifier Gain

Attenuator Coefficient

Gain controller Time constant

16 Maximum +/-10 Volts

Off,0.01,0.1 Hz -24 dB/oct Butterworth

Off,10,20,50,100,200,500,Ik,1.6k Hz 24 dB/oct

Butterworth

1,2,5,10,20,50,100

1,10,100 sec

Control Part CPU Interface

6809 IEEE-488

JAERI・M 88・221

ternal computer commands. The DC amplifier element is controlled by the

gain controller. The gain controller has two functions as follows:

1) Search for an optimal gain value based on the input signal level. It

-detects current signal peak levels with a time constant 1， 100r

100 sec，

-determines optimal gain ( vo1tage range )，

-generates a control signal.

2) Setup of a gain value of the DC amp1ifier

-to an optimal gain value by external computer command or manua11y， -to a gain value specified by manual switch or external computer.

An lnternal microprocessor and an IEEE-488 interface are prepared for

the communication between this equipment and the external computer as a

control1er.

Table 8.G.1 Specifications of signal conditioner

Channel Unit

Number of channels 16

Input and output voltages Maximum +/ー10Volts

High-pass filter

Cut-off frequency Of f ，0.01，0. 1 Hz

Roll-off -24 dB/oct

Type Butterworth

Low-pass filter

Cut-off frequency Off，10，20，50，100，200，500，lk，1.6k Hz Roll-off 24 dB/oct

Type Butterworth

DC Amplifier

Gain 1，2，5，10，20，50，100

Attenuator

Coefficient 1，0.2

Gain controller

Time constant 1，10，100 sec

Control Part

CPU 6809

Interface IEEE-488

J A E R I - M 8 8 - 2 2 1

Signal In ©-

3 n a l

| Signal

HP i i LP i ' ' r " I T i TT____ i f - i i—, l i J Level I i Gain , u .. -rU

nn.ir, <-i, I C n n t r n n p r ' Mom to * -rfscn 1 P PANEL SW tart »C- 1

PANEL SW I I I Condition

SW i i

T IGO t | M

HP [LI*-A*G Y dp I I I I I I I I I I UJ ' ' '

u i y v^ -JJu

State Monitor

-Cf Scale -O* Overload

\ ^ Remote/ Local-SW raasK—TOW

—CLP -0*G -O*G0 -OC kTR/L

JJJJ. H I M IIUl. JcL Internal SW

PIO HP,LP, R/L GO, A,G,C M

*~3ecorde *

1 Ch. Address SW

CH Channel Common Bus >

Channel Unit

Margin-Adj.

Wo o 0 s e c C o n t r o 1 S l ' g n a l out ^ o G-1,2,3 C-2(G-4)

Vrl ; 5.0 V Vr2 ; 2.0 V Vr3 ; 1.0 V Vr4 ; 0.5 V Vr5 ; 0.2 V Vr6 ; 0.1 V

m G0-l,2,3

Data Out

G-1,2,3 C-1,2 Control Signal In

Gain Controller

Fig .8 .6 .1 Block diagram of s igna l condit ioner

jAERI・M 88・221

! I ー一一_..Ji Ir一一一・一--.-11

Channel Common Bus

Signal In

問argin-Adj.

Vrl 5.0 V

Vr2 2.0 V

Vr3 1.0 V

"r4 0.5 V

Vr5 0.2 V

Vr6 0.1 V

Channel Unit

Gain Controller

Fig.8.6.1 Block diagram of signal conditioner

Scale iOverload

" ，，'

J A E R t - M 8 8 - 2 2 1

8.7 Study on the Goodness of System Identification Using Multivariate AR Modeling

K. Hayashi, Y. Shinohara and K. Nabeshima

In order to evaluate the goodness of system identification using a Multivariate Autoregressive (MAR) modeling method, a simulation study was perlormed using a hybrid computer.

The simulated model was based on a 2-dimensional feedback system and was added with the coloured process which is characteristics of an AR model. The transfer functions Gil, G12, G21 and G22 of each sub-system were set up as a low-order time-lag system, and as the noise source Nl and N2 were used independent Gaussian white noises. The test of system identification using MAR modeling method was performed using several data sets sampled from two observation signals XI and X2 in this simulation model.

The following results were obtained through the fitting test. - The power spectra of the signal XI and X2 can be estimated correcr.ly. - The residual covariances of the fitted MAR model are smaller than those of Nl and N2 and the gain values of the obtained frequency response functions of Gl and G2 compensate these errors.

- Estimation errors for Gl and G2 are large in high frequency range. - Estimations of the system noise source spectra and frequency response functions of XI and X2 are also bad in the same frequency range.

- The estimates of frequency response function of the closed-loop system and noise power contribution ratio are good for overall characteristics although the estimates have some bias errors in low frequency range and large errors in high frequency range.

- The estimation of time constants of the sub-systems or the closed-loop system give relatively correct values if the estimated system is of low-order and if the sampling interval and the maximum calculation point in time axis are properly selected.

Reference 1) Hayashi K., Shinohara Y. and Nabeshima K. : Paper presented at

SMORN-V (1987).

]AER[-M 88・221

8.7 Study on the Goodness of System Identification Using

Multivariate AR Modeling

K. Hayashi， y， Shinohara and K. Nabeshima

In order to evaluate the goodness of system identification using a

Multivariate Autoregressive (MAR) modeling method， a simulation study

was perlormed using a hybrid computer.

The simulated model was based on a 2-dimensional feedback system and

was added with the c010ured process which 1s characteristics of an AR

model. The transfer functions Gll， G12， G21 and G22 of each sub-system

were set up as a 10w-order t1me-lag system， and as the noise source Nl

and N2 were used independent Gaussian white noises. The test of system

identification using MAR modeling method was performed using severa1

data sets sampled from two observation signals Xl and X2 in this simula-

tion mode1.

The fo110wing results were obtained througil the fitting test.

-The power spectra of the signa1 Xl and X2 can be estimated correc~ly.

-The residual covariances of the fitted MAR mode1 are sma11er than

those of Nl and N2 and the gain values of the obtained frequency

response functions of Gl and G2 compensate these errors.

-Estimation errors for Gl and G2 are 1arge in high frequency range.

-Estimations of the system noise source spectra and frequency response

functions of Xl and X2 are a1so bad in the same frequency range.

ー Theestimates of frequency response function of the closed-100p system

and noise power contribution ratio are good for overall

characteristics a1though the estimates have some bias errors in 10w

frequency range and 1arge errors in high frequency range.

-The estimation of time constants of the sub-systems or the c1osed-loop

system give re1atively correct va1ues if the estimated system is of

low-order and if the samp1ing interval and the maximum calculation

point in time axis are properly selected.

Reference

1) Hayashi K.， Shinohara Y. and Nabeshima K. Paper presented at

SMORN-V (1987).

J A E R I - M 8 8 - 2 2 1

8.8 A Method of Nonstationary Noise Analysis Using Instantaneous AR Spectrum and Its Application to Borssele Reactor Noise Analysis

K. Hayashi and Y. Shinohara

Nonstationary phenomena in reactor noise observed during transient operation such as reactor start-up and shut-down contain very important information sources which are very useful for diagnosis of nuclear power plant. Therefore test and demonstration of a method of nonstationary noise analysis were made using frequency-time spectra based on instantaneous autoregressive (AR) spectra.

The frequency-time spectra are estimated by dividing the sample record into a series of very short sub-recordings which can be considered to be locally stationary. In this study, AR modeling technique based on least squares method is used for estimating the instantaneous spectra because it is possible to analyze a relatively small number of data samples.

First, Four AR methods were tested in regard to fitting to a small size samples using Che test data which were sampled from a sinusoidal analog signal. The result shows that AR method based on the least squares method gives a satisfactory result for estimation of peak frequency but a poor result for estimation of peak height. This is due to the assumption of the stationary, and also because a sufficient number of sampled data which can be considered as stationary is required for the estimation of the statistical parameters.

Next, the estimation tests of the frequency-time spectrum were performed using non-stationary noise data of which stochastic properties change slowly in time. The result shows the followings: - When the length of analysis time window is short regard to the change speed of the time dependent components, the original component can be well estimated,

- When the length is long, the magnitude of the estimated spectral component decreases and the spectral distribution spreads if the original signal has frequency-shifting components. Finally, demonstration test of this method was performed using real

non-stationary data measured at Borssele reactor during the reactor shut-down operation. It was clearly shown that the stochastic properties of several signals change with decrease of coolant temperature and loop

JAERI駒 M 88・221

8.8 A Method of Nonstationary Noise Analysis Using lnstantaneous AR

Spectrum and lts Application to Borssele Reactor Noise Analysis

K. Hayashi and Y. Shinohara

Nonstationary phenomena in reactor noise observed during transient

operati0n such as reactor start-up and shut-down contain very important

information sources which are very usefu1 for diagnosis of nuc1ear power

p1ant. Therefore test and demonstration of a method of nonstationary

noise ana1ysis were made using frequency-time spectra based on instan-

taneous autoregressive (AR) spectra.

The frequency由 time spectra are estimated by dividing the samp1e

record into a series of ver:y short sub-recordings which can be con-

sidered to be 10ca11y stationary. 1n this study， AR mode1ing technique

based on least squares method is uSed for estimating the instantaneous

spectra because it is possib1e to ana1yze a re1ative1y sma11 number of

data samples.

First， Four AR methods were tested in regard to fitting to a sma11

s)ze samp1es using the test data which were samp1ed from a sinusoida1

ana10g signa1. The resu1t shows that AR method based on the 1east

squares method gives a satisfactory resu1t for estimation of peak

frequency but a poor resu1t for estimation of peak height. This is due

to the assumption of the stationary， and a1so because a sufficient num-

ber of sampled data which can be considered as stationary is required

for the estimation of the statistica1 parameters.

Next， the estimation tests of the frequency-time spectrum were per-

formed using non-stationary noise data of which stochastic properties

change s1ow1y in time. The resu1t shows the fo11owings:

-When the length of ana1ysis time window is short regard to the change

spe~d of the time dependent components， the original component can be

well estimat.ed，

-When the 1ength is 10ng， the magnitude of the estimated spectra1 component decreases and the spectral distribution spreads if the

origina1 signal has frequency-shifting components.

Fina11y， demonstration test of this method was performed using rea1

non-stationary data measured at Borssele reactor during the reactor

shut-down operation. 1t was c1ear1y shown that the stochastic properties

of several signa1s change with decrease of coo1ant temperature and 100p

一却5-

J A E R 1 - M 8 8 - 2 2 1

pressure. Furthermore, it was shown that the combined use of the frequency-time spectrum based on AR spectra and the time information functions, i.e. DC signal recordings and the instantaneous mean squares series makes it possible to analyze closely nonstationary data with respect to a time axis.

Reference 1) Hayashi K., Shinohara Y. and Tiirkcan E. : Paper presented for

SMORN-V (1987).

}AER 1・M 88・221

pressure. Furthermore， it was shown that the combined use of the

frequency-time spectrum based on AR spectra and the time information

functions， i.e. DC signal recordings and the instantaneous mean squares

series makes it possible to analyze closely nonstationary data with

respect to a time axis.

Reference

1) Hayashi K.， Shinohara Y. and Turkcan E. Paper presented for

SMORN-V (1987).

J A E R I - M 8 8 - 2 2 1

Y. Shinohara, K. Watanabe, K. Hayashi and K. Nabeshima

The study of sodium boiling detection methods has been being conducted in connection with the Coordinated Research Program on Signal Processing Techniques for Sodium Boiling Noise Detection organized by the IAEA. A detection method has been tested using the boiling acoustic noise signals obtained by the in-pile sodium boiling experiment performed at the Russian fast experimental reactor BOR-60, which we obtained through the IAEA.

The basic principle of the boiling detection method used lies on the assumption that the acoustic noise signals generated by sodium boiling have very sharp pulse-like wave forms whose peak values are relatively higher on the average than those of background noise signals and have dominant frequency components in the region of several tens of kilo Hertz. From the results of preliminary analysis, this assumption is considered to be valid for typical cases of the BOR-60 boiling acoustic noise signals.

In order to improve the signal-to-noise ratio for the purpose of boiling detection, the original signal is first passed through a bandpass filter having a lower cut-off frequency of 24kHz and an upper cutoff frequency of 112kHz to reduce useless background noise components. This process is necessary as the first step of improving the signal-to-noise ratio because otherwise the subsequent nonlinear signal transformation process will become less effective.

The filtered signal is then passed through a squaring circuit to intensify the pulse-like nature of the boiling acoustic noise signal. Although squaring of signal is adopted in the present analysis, any other type of non-linear transformation which can intensify the pulse-like nature of boiling signals may be used.

The squared signal is then passed through a threshold circuit having a certain threshold level to cut off low level signals whose major component can be considered to be the background noise.

The signal thus obtained is passed through a high-pass filter having a cut-off frequency of 48kHz for further reduction of the background noise component and then is squared again to enhance more the signai-to-

]AERI-M 88・221

Y. Shinohara， K. Watanabe， K. Hayashi and K.トiabeshima

甘le study of sodium hoiling detection methods has been being

conducted in connection with the Coordinated Research Program on Signal

Processing Techniques for S出 iumBoiling Noise Detection organized by the

IAEA. A detection method has been tested using the boiling acoustic noise

signals obtained by the in-pile sodium boiling experiment performed at

the Russian fast experimental reactor ~コIR-60 ， which we obtained through

the IAEA.

The basic principle of the boiling detection method used lies on the

assumption that the acoustic noise signals generated by sodium boiling

have very sharp pulse-like wave forms whose pe出 values are relatively

higher on the average than those of background noise signals and have

dominant frequency components in the region of several tens of kilo

Hertz. From the results of preliminary analysis， this assumption is

considered to be valid for typical cases of the BOR四 60boiling acoustic

noise signals.

In order to improve che signal-to-noise ratio for the purpose of

boiling detection， the original signal is first passed through a 凶 nd-

戸 ss filter having a lower cut-off frequency of 24kHz and ar. upper cut-

off frequency of 112kHz to reduce useless background noise components.

This process is necessary as the first step of improving the signal-to-

noise ratio because otherwise the subsequent nonlinear signal

transforr臨 tionprocess will become less effective.

引le filtered signal is then passed through a squaring circuit to

intensify the pulse-like nature of the boiling acoustic noise signal.

Although squaring of signal is adopted in the present ap~lysis ， any other

type of non-linear transformation which can intensify the pulse-like

nature of boiling signals may be used.

The squared signal is then passed through a threshold circuit having

a certain threshold level to cut off low level signals whose I阻 jor

component can be considered to be the凶 ckgroundnoise.

四 esignal thus obtained is passed through a high-pass filter having

a cut-off frequency of 48kHz for further reduction of the 凶 ckground

noise component and then is squared again to enhance more the signal-to-

JAERI-M 88-221

noise ratio. The twice-squared signal is then integrated to generate the feature signal for boiling detection over an averaging time of 1, 10 or 100msec and the integrated value is held constant during the next time interval. This process of reducing useless background noise components and intensifying useful boiling signal components leads to enhanced signal-to-noise ratio for the feature signal.

The following probabilities were calculated. - Probability of a spurious trip which is defined as the probability that a spurious trip signal or a false alarm is generated by the feature signal exceeding the decision boundary when there exists no boiling.

- Probability of missing boiling detection which is defined as the probability that no alarm is generated when there exists boiling.

- Probability of a true alarm defined as the probability that an alarm generated is a true alarm.

- Probability that boiling exists, which is defined as the probability with which a feature signal of amplitude indicates the existence of boiling.

Other parameters such as decision boundary were determined based on the measured feature signal.

From the analysis of the acoustic noise signals, it was concluded that boiling could be detected by acoustic signals using a relatively simple method of signal processing as described in this report. However, in order to improve the reliability of boiling detection, sophisticated methods should be further investigated which can reduce both the probability of a spurious trip and that of missing boiling simultaneously.

Reference

1) Shinohara Y. et al.: "The Result of Sodium Boiling Detection Benchmark Test Using BOR-60 Reactor Noise Data," Paper submitted for the Research Coordination Meeting, November 1987.

JAERI-M 88・221

noise ratio. The twice-squared signal is then integrated to gener-ate the

feature signal for boiling detection over an averaging time of 1， 10 or

100msec and the integrated value is held constant during the next time

interval. This pr∞ess of reducing useless国 ckground noise components

and intensifying useful boiling signal components leads to e山間ced

signal-to-noise ratio for the feature signal.

百lefollowirnr pro凶 bilitieswere calculated.

-Probabili ty of a spurious trip which is defined as the probabili ty that

a spurious trip signal or a false alarm is generated by the feature

signal exceeding the decision boundary when there exists no boiling.

-Probability of missing boiling detection which is defined as the

probability that no alarm is generated when there exists boiling.

-Probability of a true alarm defined as the probability that an alarm

generated is a true alarm.

-Probability that boiling exists， which is defined as the probability

with which a feature signal of amplitude indicates the existence of

凶 iling.

Other parameters such as decision凶 undarywere determined based on

the measured feature signal.

From the analysis of the acoustic noise signals， it was cO/1cluded

that boiling could be detected by acoustic signals using a relatively

simple method of signal processing as descri国 din this rep<コrt. However， in order to improve the reliability of boiling detection， sophisticated methods should be further investigated which can reduce both the

probability of a spurious trip and that of missing boiling

simultaneous1y.

Reference

1) Shinohara Y. et a1.: It叩eResu1t of Sα鉱山nBoiling Detection Benchmark

Test Using 駅東ー60 Reactor Noise Data， It Paper submi tted for the

Research Coordination M'田 ting，Nov，凹ber1987.

-2凶ー

J A E R I - M 8 8 - 2 2 1

8.10 Theoretical Studies of Manipulator Inverse Kinematics S. Sasaki and Y. Shinohara

Obtaining a general solution of the inverse kinematic problem for multi-joint robot manipulators has been considered to be one of the most cumbersome problems in their control. This is because the relationships between the hand and base are expressed in terms of a set of highly non-linear equations including transcendental functions. Needless to say, analytical solutions of such equations cannot be expected except for special configurations of manipulators.

During the last several years, many different methods for this problem have been developed on a world-wide scale to find the way out of deadlock. Among them, the most popular approach is to use the Jacobian matrix on linearization technique. Although this iterative technique is practical to a variety of linkage mechanism analyses, it has inherent numerical problems to the solution characteristics such as initial guess dependence and singularities.

Taking these facts into consideration, three new approaches have been proposed from different standpoints. An underlying concept of the first approach is based on a transformation of kinematic relations into a polynomial with a single joint variable. Namely, the problem of finding the articulated angle of a manipulator is reduced to an algebraic equation. Originally, the order of the equation derived was 24. Irrespective of its complexity of the equation, the joint solutions were entirely exact thanks to the numerical calculation method used. By continued efforts, a marked improvement was made in reducing the order of the equation from 24 to 16 and then to 8. Resulting solution accuracies and computing time are satisfactory, too. The advantage of the presented algorithm is that it allows to find out all arm solutions, compared to the traditional approximation methods.

The second approach is based on the minimization technique of mathematical programming. Directing our attention to features of joint structure of a six-link manipulator, the original kinematic problem was first formulated by a system of equations with four variables determined from the least value for a certain performance index, in which the position and orientation at the end of shoulder is selected so as to match each other between the actual physical situation and its model. The

]AERI-M 88・221

8.10 Theoretical Studies of Manipulator Inverse Kinematics

S. Sasaki and Y. Shinohara

Obtaining a general solution of the inverse kinematic problem for

multi-joint robot manipulators has been considered to be one of the most

cumbersome problems in their control. This is because the relationships

between the hand and base are expressed in terms of a set ot highly

non-linear equations including transcendental functions. Needless to

say， analytical solutions of such equations cannot be expected except

for special configurations of manipulators.

During the last several years， many different methods for this

problem

have been developed on a world-wide scale to find the way out of dead-

lock. Among them， the most popular approach is to use the Jacobian

matrix on linearization technique. Although this iterative technique is

practical to a variety of linkage mechanism analyses， it has 工nherent

numerical problems to the solution characteristics such as initial guess

dependence and singularities.

Taking these facts into consideration， three new approaches have

been proposed from different standpoints. An underlying concept of the

first approach is based on a transformation of kinematic relations into

a polynomial with a single joint variable. Namely， the problem of find-工ng the articulated angle of a manipulator is reduced to an algebraic

equation. Originally， the order of the equation derived was 24. Ir-

respective of its complexity of the equation， the joint solutions were

entirely exact thanks to the numerical calculation method used. By con-

tinued efforts， a marked improvement was made in reducing the order of

the equation from 24 to 16 and then to 8. Resulting s01ution accuracies

and computing time are satisfactory， too. The advantage of the presented

algorithm is that it a110ws to find out a11 arm solutions， compared to

the traditiona1 approximation methods.

The second approach is based on the minimization technique of math-

ematical programming. Directing our attention to features of joint

structure of a six-link manipulator， the original kinematic prob1em was

first formulated by a system of equations with four variables determined

from the least va1ue for a certain performance index， in which the posi-

tion and orientation at the end of shou1der is selected so as Lo 山dtch

each other between the actual physical situation and its model. The

一部9-

J A E R I - M 8 8 - 2 2 1

remaining two variables are determined from constrained conditions. This is the basic idea in the present approach. The accuracies of solutions and convergence speed by this method were amazingly high. As a matter of fact, the present algorithm is available for any type of mechanical manipulator with an n degree of freedom, provided that proper constraints are given in advance.

The third approach was intended to provide direct derivation of individual joint solutions using the knowledge of vector and elementary geometry of the link mechanism, in which formulation of angular displacement by means of rotational ( unit ) vector and simple vector arithmetics are of particular importance. Compared with the traditional transformation method, the present advantages is to permit explicit angular solution for each joint as well as intuitive representation and a better understanding of the configurations of complicated linkage mechanism.

From the results of computer simulations, it is concluded that each approach proposed provides a very powerful technique which leads to theoretically accurate solutions to complicated inverse kinematics problems from completely different angles. Particularly, it should be kept in mind, from the viewpoint of practice, that the second algorithm will make a significant contribution to computing time saving since the convergence process of performance index is normally attained within a few search iterations for an optimal (desired) solution.

References 1) Sasaki S. : JAERI-M 87 039 (1987). 2) Sasaki S. : JAERI-M 87 104 (1987). 3) Sasaki S. rTrans.of the Society of Instrument and Control Eng. 23-3,

23-5, (1987). 4) Sasaki S. :Trans. of the Society of Instrument and Control Eng. 24-2,

(1988). 5) Sasaki S. : JAERI-M 86 122 (1986). 6) Sasaki S. and Shinohara Y. : JAERI-M 87 175 (1987).

]AERI-M 88-221

remaining two variables are determined from constrained conditions. This

is the basic idea in the present approach. The accuracies of solutions

and convergence speed by this method were amazingly high. As a matter of

fact， the present algorithm is available for any type of mechanical

manipulator with an n degree of freedom， provided that proper con-

straints are given in advance.

The third approach was intended to provide direct derivation of

indiviaual joint solutions using the knowledge of vector and elementary

geometry of the link mechanism， in which formulation of angular dis-

placement by means of rotational ( unit ) vector and simple ve己tor

arithmetics are of particular importance. Compared with the traditional

transformation method， the present advantages is to permit explicit an-

gular solution for each joint as well as intuitive representation and a

better understanding of the configurations of complicated linkage

mechanism.

From the results of computer simulations， it is concluded that each

approach proposed provides a very powerful technique which leads to

th~oretically accurate solutions to complicated inverse kinematics

problems from completely different angles. Particularly， it sllould be

kept in mind， from the viewpoint of practice， that the second algorithm

will make a significant contribution to computing time saving since the

convergence process of performance index is normally attained within a

few search iterations for an optimal (desired) solution.

Ref己rences

1) Sasaki S. JAERI-M 87 039 (1987).

2) Sasaki S. JAERI-M 87 104 (1987).

3) Sasaki S. :Trans.of the Society of Instrument and Control Eng. 23-3，

23-5， (1987). 4) Sasaki S. :Trans. of the Society of Instrument and Control Eng・24-2，

(1988) .

5) Sasaki S. JAERI-M 86 122 (1986).

6) Sasaki S. and Shinohara Y. JAERI-M 87 175 (1987).

J A E R I - M 8 8 - 2 2 1

8.11 General Formulation of a Manipulator Linkage Mechanism S.Sasaki

Fundamental to spacial position control of a robot manipulator is a co-ordinate transformation that maps joint co-ordinates into Cartesian co-ordinates. This so-called " direct transformation " can be carried out in a straightforward method by successive multiplication of homogeneous 4x4 matrices based on Denavit and Hartenberg notation, with the result that:

Ao,n = ^ . l ^ ^ * ' ' An-i,n where A denotes the transformation matrix of the fin^l link with respect to the base co-ordinates system and A^ ^ is the one relating the link i and link j. This matrix operation, however, is usually forced to continue a monotonic and dull procedure in placing main reliance on the hand calculation only. Furthermore, its computational complexity increases rapidly with the number of degrees of freedom of the mechanism, with the consequence that manual derivation is liable to commit careless errors in its course.

As an alternative to that, the expressions of recursive relations which are suitable for computer-aided generation of the manipulator kinematics are proposed and applied to a six-link manipulator. Namely, two algorithmic procedures for the above matrix operation were developed in due consideration of the operation sequence between the reference system and the final link or the hand of manipulator. That is,

(D Ai-l,n " Ai-l,iAi,n : ( 2 ) Ao,i = Ao,i-lAi-l,i < i = 1 ' - »n> Matrix AH_-i ^ is known as one representing kinematic features between the two ajacent links. According to appearance of such a conspicuous recurrence, they were re-written into mathematical relationships on an element-by-element basis. This approach is a key point to extract automatically general formulation of kinematics equations. With the aid of these algorithms, the position and orientation of a manipulator hand can be calculated effectively using a computer.

As such, the advantage of the present method is that it allows a general description of kinematic models for all types of manipulator without manual derivation. The validity of the algorithm was verified successfully in comparison with the manually derived expressions.

JAERI-M 88・221

8.11 Genera1 Formu1ation of a Manipu1ator Linkage Mechanism

S.Sasaki

Fundamenta1 to spacia1 position contro1 of a robot manipulatllr is

a co-ordinate transformation that maps joint co-ordinates into Cartesian

co-ordinates. This so-ca11ed 11 direct transformation 11 can be carried

out in a straight[orward method by successive mu1tiplication of

homogeneous 4x4 matrices based on Denavit and Hartenberg notation， with the resu1t that:

Ao，n Ao，IAl，2・・・ An-l，n

where A_ _ denotes the transformation matrix of the fin弓1 link with o，n respect to the base co-ordinates system and A; ・ is the one re1ating

the 1ink i and 1ink j. This matrix operation， however， is usua11y forced

to continue a monotonic and dull procedure in p1acing main reliance on

the hand ca1culation on1y. Furthermore， its computationa1 complexity in-

creases rapid1y with the number of degrees of freedom of the mechanism，

with the consequence that manual derivation is 1iab1e to commit care1ess

errors in its course.

As an alternative to that， the expressions of recursive re1ations

which are suitab1e for computer-aided generation of the manipu1ator

kinematics are proposed and applied to a six-link manipulator. Namely，

two a1gorithmic procedures for the above matrix operation were developed

in due consideration of the operation sequence between the reference

system and the fina1 link or the hand of manipulator. That is，

(1) Ai-1，n = Ai-l，iAi，n (2) Ao，i = Ao，i-lAi-l，i (i=l，…，n)

Matrix Ai-1， i is known as one representing kinematic features between

the two ajacent links. According to appearance of such a conspicuous

recurrence， they were re-written into mathematica1 re1ationships on an

e1ement-by-e1ement basis. This approach is a key point to extract

automatica11y genera1 formu1ation of kinematics equations. With the aid

of these a1gorithms， the position and orientation of a manipu1ator hand

can be ca1cu1ated effective1y using a computer.

As such， the advantage of the present method is that it a110ws a

genera1 description of kinematic mode1s for a11 types of manipu1ator

without manua1 derivation. The va1idity of the algorithm was verified

s"ccessfu11y in compaI'ison with the manua11y derived expressions.

J A E R I - M 8 8 - 2 2 1

8.12 Development of Robotic Remote Handling Technology

H. Usui, Y. Fujii, A. Kumagai, S. Sasaki, J. Shimazaki and Y. Shinohara

As for the robotic remote handling technology for reactor decommissioning, the robotic remote handling system equipped with an amphibious robot manipulator JARM-25 (JAERI Amphibious Robot Manipulator with a load capacity of 25kg) was modified slightly to improve its operability and also to cope with environmental conditions to be considered in its application to actual dismantling work of reactor internals of the JPDR.

In the mockup test conducted in 1986, it was found that the manipulator motion at very low speed required for the plasma arc cutting should be improved. Therefore, the mechanical design of the reduction gears was modified to reduce the backlash. The control system, of the manipulator was also modified to give better control characteristics at its low speed motion as well.

The computer program for the manipulator control was modified to make smoother switch-over between the master-slave and playback modes and thus to improve the positioning accuracy of the slave-arm in the playback mode.

In order to improve the operability during the plasma arc cutting, a special jig for the plasma torch was designed and was attached to the finger part of the manipulator. This jig converts the open-close motion of the manipulator fingers into the longitudinal motion of the plasma torch.

The support frame of the slave side of the system was enlarged so that it can be mounted on the floor of the reactor room above the reactor pit.

The viewing system was also improved so as to make it easier to monitor from different view angles the motion of the manipulator and thus to facilitate positioning operation of the plasma torch attached to the finger part of the manipulator.

Finally, the slave arm of the manipulator was painted with strippable paint to protect it against radioactive contamination. The system was used successfully for dismantling the radioactive reactor

JAERI-M 88・221

8.12 Development of Robotic Remote Handling Technology

H. Usui， Y. Fujii， A. K~阻gai ， S. S錨出i，J. Shimazaki and

Y. Shinohara

As for the robotic remote handling technology for reactor

deconvnissioning， the robotic remote handling system equipped wi th an

剖 Iphibiousrobot manipulator .JARM-25 (JA四 IAmphibious Robot Manipulator

with a load capacity of 25kg) was modified sliihtly to improve its

O戸 rability and also to cope with environmental conditi~ns to be

considered in its application to actual dismantling work of reactor

internals of the JPDR.

In the mockup test conducted in 1986， it was found that the

manipulator motion at very low speed required for the plasma arc cutting

should be improved. Therefore， the mechanical design of the reduction

gears was modified to reduce the backlash.百 econtrol system. of the

mani凹 lator was also modified to give better control characteristics at

its low speed motion as well.

The computer progr創nfor the manipulator control was modified to

make smoother switch-over‘between the master-slave and playback modes and

thus to improve the positioning accuracy of the slave-arm in the playback

In orせerto improve the operability during the plasma arc cutting， a

special jig for the plasma torch was designed and was attached. to the

finger 戸 rtof the s町 lipulator.This jig converts the 0戸 n-close motion

of the manipulator fingers into the longitudinal motion of the pl槌阻

torch.

The support fr町時 ofthe slave side of the system was enlarged so

that i t can be mounted on the floor of the reactor roαn above the reactor

The viewing system was also improved. so田 tomake it easier to

monitor frαn different view angles the motion of the mani夙llator 包叫

thus to facilitate positioning 0伊 rationof the plasma torch attached. to

the finger part of the s町 upulator.

Finally， the slave arm of the manipulator 鴫 S 戸 inted. with

strip戸 ble paint to protect it against radioactive contamination. 百 e

system was used. successfully for dismantling the radioactive reactor

-212ー

J A E R I - M 8 8 - 2 2 1

internals of the JPDR during the period from late January to early March 1988.

On the other hand, a study was made to further improve the operability of the robotic manipulators. In this study, three different type of bilateral master-slave control methods were tested and compared by using position and/or force feedbacks. It was confirmed by the test that the simultaneous use of both position and force feedbacks is very effective for improving the sensitivity of the force acting on the slave arm of the manipulator and thus the operability of the bilateral master-slave mode operation.

As for the robotic remote handling technology for remote maintenance of nuclear fusion reactors, a survey of existing technologies which might be applicable to the fusion reactors has been made. A preliminary study was also started to develop new design concepts for invessel robot manipulators which are to be used in extreme environmental conditions such as high radiation dose rate, high ambient temperature, and high degree of vacuum.

)AERI-M 88・221

internals of the JPDR during the period from late January to ear ly March

αthe other hand， a study was made to further improve the

O戸 rability of the robotic rr凶 ipulators. In thisstudy， three differ~nt

type of bilateral master-slave control methods were tested and compared

by using posi tion and/or force feedbacks. It was confinned by the test

that the simultaneous use of both position and force feedbacks is very

effective for improving the sensitivity of the force acting on the slave

ann of the manipulator and thus the operability of the bilateral master-

slave mode operation.

As for the robotic remote handling technology for remote maintenance

of nuclear fusion reactors， a survey of existing technologies which might

be applicable to the fusion reactors has been made. A preliminary study

was also started to develop new design concepts for invessel robot

manipulators which are to be used in extreme environmental conditions

such as high radiation dose rate， high ambient temperature. and high

degree of vaCU1.Un.

-213一

JAERI-M 88-221

9. Facility Operation and Technique Development

In FY-1987, operations of the Fast Critical Assembly (FCA), the Very High Temperature Reactor Critical assembly (VHTRC) and the Fusion Neutronics Source (FNS) were carr ied out without any major t roubles d is turbing the experiments. The FCA XIV-2 and XV-1 cores were constructed to obtain benchmark data of the HCLWR. Integral experiments were made on the VHTRC-1 and VHTRC-4 to verify the design accuracy of the HTTR. Including the Phase-IIA/IIB experiments on the JAERI/US-DOE collaborative program, the fusion blanket neutronics experiments were carr ied out at the FNS. The integrated operation time of 803 hours , 510 hours and 614 hours was recorded at the FCA, the VHTRC and the FNS, respectively.

An equipment to measure leakage rate of the reactor vessel was newly installed at the FCA to confirm the soundness of the reactor vessel . Renewal work of the control panel and ins t ruments s tar ted a t the VHTRC. The VHTRC operat ing data processing system was improved and used effectively for operation data logging and data analysis of the experiment.

Tritium quantification technique was established by the character is t ic X-ray measurement and the micro-caloriraetry method to improve the tritium handling a t the FNS facility. The Tritium Adsorption Processor (TAP) was operated with no trouble. Beam focus and beam stability of the FNS accelerator was improved result ing in a good transmission of beam.

(Masafumi Nakano)

- 2 1 4 -

]AERI-M 88‘221

9. Facility Operation and Technique Development

In FY-1987， operations of the Fast Critical Assembly (FCA)， the Very High Temperature Reactor Critical assembly (VHTRC) and the

Fusion Neutronics Source (FNS) were carried out without any major

troubles disturbing the experiments. The FCA XIV-2 and XV-1 cores

were constructed to obtain benchmark data of the HCLWR. Integral

experiments were made on the VHTRC-l and VHTRC-4 to verify the

design accuracy of the HTTR. lncluding the Phase-IIA/IIB experiments

on the JAERI/US-DOE collaborative program， the fusion blanket

neutronics experiments were carried out at the FNS. The integrated

operation time of 803 hours， 51.0 hours and 614 hours was recorded at

the FCA， the VHTRC and the FNS， respectively.

An equipment to measure leakage rate of the reactor vessel was

newly installed at the FCA to confirm the soundness of the reactor

vessel. Renewal work of the control panel and instruments started

at the VHTRC. The VHTRC operating data processing system was

improved and used effectively for operation data logging and data

analysis of the experiment.

Tritium quantification technique was established by the

characteristic X国 raymeasurement and the micro-calorimetry method to

improve the tritium handling at the FNS facility. The Tritium

Adsorption Processor (TAP) was operated with no trouble. Beam focus

and beam stability of the FNS accelerator was improved resulting in

a good transmission of beam.

(Masafumi Nakano)

J A E R I - M 8 8 - 2 2 1

9.1 Operation Report of FCA

K.Satoh, A.Ohno, K.Hayasaka, H.Watanabe, S.Fujisaki and M.Nakano

In FY-1987, much labor was expended in activities concerning operation and maintenance of the Fast Critical Assembly (FCA). The FCA XIV-2 and XV-1 cores were assembled to obtain benchmark data on the "High Conversion Light Water Reactor" core. Operations for this purpose were completed as scheduled, without any major problems disturbing execution of the experiments.

Operations of 184 times were carried out in 135 days to perform the experiments. No scram was recorded during the operations. The total operating time was 803 hours and the integrated power 3506.55 Wh. A total of 4232 criticality operations has been recorded at the end of this fiscal year since the first achievement of criticality on 29 April 1967. According to the safety regulation for operation, two days were devoted to the monthly inspection and about 4 weeks to the annual inspection in November. Routine maintenance activity was done in these days to provide maximum operation days for the experiments.

An equipment was installed to check the airtightness of the re ictor vessel. The soundness can be usually examined by measuring leakage rate of the reactor vessel. Fig. 9.1.1 shows the configuration of the measurement system. The leakage rate was deduced from pressure difference between the vessel and the standard container placed in the vessel. This equipment gave us sufficient results to evaluate the soundness in shorter measurement time. As for fuel management, the defects of coating on the surfaces of natural uranium fuel plates were repaired for about 3,000 coupons by spraying colloidal solution of fluorocarbon. Weights of coating on the fuel coupons were obtained from measurements of fuel coupon weight before and after coating. The maintenace activity was taken on the physical protection (P-P) system. The sensitivity and function of the system including the main gate were examined and calibrated. Moreover, a partial modification was made to the procedure for entering the FCA.

JAERI-M 88・221

9.1 Operation Report of FCA

K.Satoh， A.Ohno， K.Hayasaka， H.Watanabe， S.Fujisaki and M.Nakano

In FY-1987， much labor was expended in activities concerning

operation and maintenance of the Fast Critical Assembly (FCA). The FCA

XIV-2 and XV-l cores were assembled to obtain benchmark data on the "High

Conversion Light Water Reactor" core. Operations for this purpose were

completed as scheduled， without any major problems disturbing execution of

the experiments.

Operations of 184 times were carried out in 135 days to perform the

experiments. No scram was recorded during the operations. The total

operating time was 803 hours and the integrated power 3506.55 Wh. A total

of 4232 criticality operations has been recorded at the end of this fiscal

year since the first achievement of criticality on 29 April 1967.

According to the safety regulation for operation， two days were devoted to

the monthly inspection and about 4 weeks to the annual inspection in

November. Routine maintenance activity was done in these days to provide

maximum operation days for the experiments.

An equipment was installed to check the airtightness of the relctor

vessel. The soundness can be usually examined by measuring leakage rate

of the reactor vessel. Fig. 9.1.1 shows the configuration of the

measurement system. The leakage rate was deduced from pressure difference

between the vessel and the standard container placed in the vessel. This

equipment gave us sufficient results to evaluate the soundness in shorter

measurement time. As for fuel management， the defects of coating on

the surfaces of natural uranium fuel plates were repaired for about 3，000

coupons by spraying colloidal solution of fluorocarbon. Weights of

coating on the fuel coupons were obtained from measurements of fuel coupon

weight before and after coating. The maintenace activity was taken on the

physical protection (P-P) system. The sensitivity and function of the

syste即 includingthe main gate were examined and calibrated. Moreover， a

partial modification was made to the procedure for entering the FCA.

JAERI-M 88-221

In connection with safeguards, IAEA and NSB' ' carried out every month inspection under the international treaty. The Physical Inventory Taking (PIT) of the fuels was performed from May 14 to June 12 by means of item counting, weighing and non-destructive assay. IAEA and NSB made the Physical Inventory Verification (PIV) from June 29 to July 3. No anomaly was confirmed.

(*) NSB : Nuclear Safety Bureau

-Reactor vessel •Standard container

w KxH -Va Ive

Ux-(g^<h-f ' Pressure - current converter

Digital manometer

Recorder Current -voltage converter

Fig. 9 . 1 . 1 Configuration of measurement system

- 2 1 6 -

JAERI-M 88・221

In connection with safeguards， IAEA and NSB(骨)carried out every

month inspection under the international treaty. The Physical Inventory

Taking (PIT) of the fuels was performed from May 14 to June 12 by means of

item counting. weighing and non-destructive assay. IAEA and NSB made the

Physical Inventory Verification (PIV) from June 29 to July 3. No anomaly

was confirmed.

(事)NSB Nuclear Safety Bureau

Reactor vessel

:oigitol :man胴 eter

!「-----u¥

ー---Volve

|: J|C町rent切 110ge∞附附

R:ecorder

Fig. 9.1.1 Configuration of measurement sys↑em

-216ー

JAER1-M 88-221

9.2 Operation Report of FNS

J. Kusano, C. Kutsukake, S. Tanaka, Y. Abe and M. Nakano

The operations of the Fusion Neutronics Source (FNS) were performed almost as scheduled in this fiscal year. The integrated operation time was 614 hours . Through February and March 1988, the accelerator operation was stopped for a minor-change of the building and the related work for the air ventilation equipments. As a regular maintenance work, a week long inspection of the control circuit of the accelerator was performed every four months. The improvement work for the accelerator will be described at section 9.4. 0 degree beam line operation

In FY-1987, ten week operation on the 0-degree beam line was performed for the Phase-IIA/IIB experiments on the JAERI/US-DOE collaborative program. During this operation period, the neutron spectroscopy on the Phase-IIA/IIB experiments was made using an NE-213 counter and the new developed proton recoil counter assembled by US-ANL under beam c u r r e n t 10"2 mA operation. The LizO pellets irradiation for the measurement of the zonal tritium production rate were achieved with the beam c u r r e n t of 20 mA operation.

The o ther activit ies of the operation except the collaborative program, the intensive beam operations for the spatial dose ra te distr ibution and spectrum measurements for neutron and gamma-ray in a large concrete cavity, the narrow gap streaming experiment in a type 304 stainless steel assembly were carried out. 80 degree beam line operation

The cross section measurement of the secondary gamma-ray production reaction s ta r ted in this year using the bunched pulse beam operat ions utilizing 2 nano-second pulse width of beam. The intensive beam operations with beam cur ren t of 2 mA on the 80-degree beam line were performed for the activation cross-sect ion measurement and the (n,p) reaction cross-section measurement. To confirm the absolute neutron yield on the 80-degree beam line, the alpha monitor calibration experiment was made using a 45-degree inclined par t icular t a rge t assembly.

- 2 1 7 -

]AERl・M 88・221

9.2 Operation Report of FNS

J. Kusano， C. Kutsukake， S. Tanaka. Y. Abe and M. Nakano

The operations of the Fusion Neutronics Source (FNS) were

performed almost as scheduled in this fiscal year. The integrated

operation time was 614 hours. Through February and March 1988， the

accelerator operation was stopped for a minor-change of the building

and the related work for the ai1' ventilation equipments. As a

regular mai.ntenance wOl'k， a week long inspection of the control

circuit of the accelerator was performed every four months. The

improvement work for the accelerator will be described at section

O dellree beam line operation

In FY -1987， ten week operation on the O-degree beam line was

performed for the Phase田 IIA/IIBexperiments on the JAERI/US-DOE

col1aborative program. During this operation period， the neutron spectroscopy on the Phase-IIA/IIB experiments was made using an NE-213

counter and the new developed proton recoil counter assembled by US恒

ANL under beam current 10-2 mA operation. The Lizu pellets

irradiation for the measurement of the zonal tritium production rate

were achieved with the beam current of 20 mA operation.

The other activities of the operation except the col1aborative

program， the intensive beam operations for the spatial dose rate distribution and spectrum measurements for neutron and gamma-ray in

a large concrete cavity， the narrow gap streaming experiment in a

type 304 stainless steel assembly were cal'ried out.

8o dellree beam line operation

The cross section measurement of the secondary gamma-ray

production reaction started in this year using the bunched pulse

beam operations utilizing 2 nano-second pulse width of beam. The

intensive beam operations with beam current of 2 mA on the 80-degree

beam line were performed for the activation cross-section

measurement and the (n，p) reaction cross-section measurement. To

confirm the absolute neutron yield on the 80-degree beam line， the alpha monitor calibration experiment was made using a 45-degree

inclined particular target assembly.

-217ー

JAERI-M 88-221

Tritium handling The ta rge t exchange work for a water-cooled target , 10 Ci of

tritium, on the 80-degree beam line was carr ied out safely as a routine work. The exchange work was repeated ten times following to the schedule of the irradiation experiment. The rotating target , 650 Ci of tritium, on the 0-degree beam line was exchanged with a new one for a preparation work of the Phase-IIA/IIB experiments.

The characterist ic X-ray measurement technique was established to quantify tritium amount on the t a rge t and to analyze the tritium contamination. The technique development is detailed in section 9.3.

The total amount of 55 Ci tritium in the exhaust gas from the vacuum system of the accelerator was processed in the Tritium Adsorption Processor (TAP) system. A deterioration of the TAP was observed in case of longer operation required. The process character is t ics were investigated for the relation among the composition of the exhaust gas, the temperature of the catalyzer and the tritium concentration of the accumulated gas. The main cause of the deterioration was estimated as a deterioration of the catalyzer covered with hydro-carbon layer. The modification of the system is planned and it will be executed in FY-1988.

The additional tritium handling room was constructed during February and March 1988 in the FNS building.

- 2 1 8 -

jAERI-M 88・221

Tritiurn handlini!

The target exchange work for a water-cooled target， 10 Ci of tritiurn， on the 80-degree beam line was carried out safely as a

routine work. The exchange work was repeated ten tirnes following to

the schedule of the irradiation experiment. The rotating target，

650 Ci of tritium， on the O-degree beam line was exchanged with a

new one for a preparation work of the Phase-IIA/IIB experiments.

The characteristic X-ray rneasurement technique was estab1ished

to quantify tritium amount on the target and to anaJyze the tritiurn

contamination. The technique development is detailed in section

The total amount of 55 Ci tritium in the exhaust gas from the

vacuum system of the acceJerator was processed in the Tritiurn

Adsorption Processor (TAP) system. A deterioration of the TAP was

observed in case of longer operation required. The process

characteristics were investigated for the relation arnong th~

cornposition of the exhaust gas， the ternperature of the catalyzer and

the tritiurn concentration of the accurnulated gas. The :uain cause of

the deterioration was estimated as a deterioration of the catalyzer

covered with hydro-carbon layer. The modification of the system is

planned and it wil1 be executed in FY-1988.

The additional tritium handling room was constructed during

February and March 1988 in the FNS building.

JAERI-M 88-221

9.3 Tritium Quantif icat ion by Measurement of Charac te r i s t i c X-Rav

C. Kutsukake, J. Kusano, S. Tanaka and Y. Abe

The tritium quantification technique was developed for a handling of the titanium tritide (Ti-T) target being used at the Fusion Neutronics Source (FNS) facility. The quantification was achieved by a characterist ic X-ray measurement method combined with a micro-calorimetry method. The measurable range of the tritium quantification is from 10' 5 Ci to almost 1,000 Ci in this technique.

The main purpose of the development was to investigate the relation between the amount of ta rge t tritium and the neutron yield character is t ics of the target . The secondary purpose was to improve the handling technique for a judgment of a surface contamination by tritium and for an evaluation of the tritium contents in the radioactive wastes.

The quantification technique developed in case of a 10 Ci water cooled tritium ta rge t is as follows; (1) The absolute amount of tritium in a target was measured with

the micro-calorimeter. The sensitivity limit of the micro-calorimeter was about 10 mCi.

(2) The energy spectrum and integral counts of the characteris t ic X-ray from the Ti-T ta rge t were measured by the Si(Li) solid s ta te photon detector and a multi-channel pulse height analyzer system. Fig.9.3.1 shows the X-ray spectrum measured. The titanium and copper character is t ics X-ray and bremsstrahlung X-ray are observed in this figure.

(3) The count of the characterist ic X-ray was calibrated by the absolute value obtained from calorimetric method. Fig.9.3.2 shows the comparison of the X-ray count with tritium values of the calorimetry. After the self-shielding effect correction in the titanium layer, the value of X-ray measurement has good linearity to the calorimetric method.

(4) The measured ta rge t was covered with polyethylene bags or Al-laminated sheet (aluminum coated by thin polyethylene) to protect a contamination of tritium to the instruments. The mass

- 2 1 9 -

JAERI-~r flR-221

9.3 Tritium Quantification by Measur巴mentof Characteristi巴 X-Ra¥'

C. Kutsukake， J. Kusano， S. Tanaka and Y. Abe

The tritium quantification technique was deveJoped for a

handling of the titanium tritide (Ti-T) t.arget being used at the

Fusion Neutronics Source (FNSJ facility. The quantification was

Ilchieved by a characteristic X-ray measurement method combined wlth

a micro-calorimetry method. The measurable range of the tritium

guantification is from 10・ Cito almost 1，000 Ci in this

technique.

The main purpose of the development was to investigate the

relation between the amount of target tritium and the neutron yield

characteristics of the target. The secondary purpose was to improve

the handling technique for a judgment of a surface contamination by

tritium and for an evaluation of the tritium contents in the

radioactive wastes.

The quantification technique developed in case of a 10 Ci water

cooled tritium target is as foJlows;

(1) The absolute amount of tritium in a target was measured with

the micro-calorimeter. The sensitivity limit of the micro-

calorimeter was about 10 mCi.

(2) The energy spectrum and integral counts of the characteristic

X-ray from the Ti-T target were measured by the Si(Li) solid state

photon detectol' and a multi-channel pulse height analyzer system.

Fig.9.3.1 shows the X-ray spectrum measured. The titanium and

copper characteristics X-ray and bremsstrahlung X-ray are

observed in this figure.

(3) The count of the characteristic X-ray was calibrated by the

absolute value obtained from calorimetric method. Fig.9.3.2

shows the comparison of the X-ray count with tritium values of

the calorimetry. After the self-shielding effect correction in the

titanium layer， the value of X由 raymeasurement has good linearity

to the calorimetric method.

(4) The measured target was covered with polyethylene bags or Al-

laminated sheet (aluminum coated by thin polyethylene) to protect

a contamination of tritium to the instruments. The mass

JAERI-M 88-221

attenuation coefficient of the protector were measured in order to correct the X-ray counting.

The X-ray measurement method has the advantages of nondestructibi l i ty, high sensitivity, wide dynamic range, short measuring time and easy handling. This technique is used conveniently to quantify the amount of tritium in a t a rge t before use and after use.

The data of the neutron yield, the tritium consumption, and integral coulomb of incident deuteron beam were accumulated for each Ti-T ta rge t in this fiscal year. Basing on the data, the t a rge t performance has been investigated. Although the neutron yield decreased to the 20% to 50% of the initial value, the tritium consumption ratio only about 10%. Fig.9.3.3 shows the relation of those.

Another use of the characterist ic X-ray measurement, the measurement suppor t s an analysis of surface contamination of tritium. The energy spectrum of character is t ic X-ray informs a chemical form of the contamination source which include Ti-T micro-powder or not. This technique improved a judgment of surface contamination and a classification of radioactive wastes at the FNS facility.

10 5 - •H H •H

Distance 10cm

Co l l ime to r 2mm<j) -

i-> ; 3 1

t 1 0 2

< .v-^''

•J u *\

10 1 '.•r

m° , I | . _ . „ . | . . . . . . . .. ••!*• 5 10 15 20

X-ray Energy (KeV)

Fig. 9.3.1 X-ray spectrum measured of t n ' t i a t e d titanium targe t

- 2 2 0 -

jAER 1・M 88・221

attenuation coefficient of the protector were measured in order

to correct the X-ray counting.

The X-ray measurement method has the advantages of

nondestructibility， high sensitivity， wide dynamic range， short

This technique is used measuring time and easy handling.

convenjently to quantify the amount of tritium in a target before

use and after use.

The data of the neutron yield， the tritium consumption， and were accumulated for integral coulomb of incident deuteron beam

Basing on the data， the each Ti由 Ttarget in this fiscal year.

Although the neutron target performance has been investigated.

yield decreased to the 20% to 50% of the initial value， the tritium

Fig.9.3.3 shows the relation of consumption ratio only about 10%.

Another use of the characteristic X-ray measurement， the

measurement supports an analysis of surface contamination of

those.

The energy spectrum of characteristic X-ray informs a tritium.

chemical form of the contamination source which inc1ude Ti-T micro-

This technique irnproved a judgment of surface powr.ler or not.

contamination and a classification of radioactive wastes at the FNS

facility.

10cm Distance a:l ~ 3

105ト

iLg也~ ~

.・| ココしJ らJ

/ い九吋九.".

・:1 ¥ 」

102ト : ・九七・・

Collimetor 2mmゆ'.-i H

..-1 ト・

10句トー

103ト

μi明白ヨ

HM刊門MM〈

101 ト・『

Energy (KeV) 10

X-ray spectrum measured of

tritiated t;tan;um target

-2却一

Fig. 9.3.1

JAERI-M 88-221

c 5 00

Distance 10cm Coll imetor Zmmt,

self shield corrected

ODservea

2 4 .6 S 10 12 Act iv i ty of Tritium (Ci)

• i g . 9.3.2 The comparison between the characteristic X-ray and the t r i t ium act iv i ty

es O.fcf-

value of end/begining

* Neutron yield ! L T amount (calorimetry) 0 T amount (X-ray)

_i_ _i_ JL.

C 20 40 60 80 100 120 1*0 Total incident current (COUIOITID)

Ratios of the neutron yield and the amount of tritium at the begining of use to those at the end of use as a function of the total current driven into eacn taroet.

]AERI輔 M 88・221

10cm Distance

Co 11 imetor 2mm申900

self shie1d

corrected

/∞serve

The comoarison between the

characteristic X-ray and the

trnium acτ1v1ty

Fig. 9.3.2

--'tz》

A-m)20-OZ255い】

2 ~ 6 8 10 12 Acτivity of Tl・it i um (C i )

vaiue of end/begining

Neutron yield

了間ount (caJorimetry)

amount (X-ray)

ブ引を干→ァ一、〈L.4J

-t、¥、、

。一】司匡

4 0.'1ト

..1.一一一←20 40 OU 80 100 i2-d 140

Total incident current (cou10mo)

• 0.2

Ratios of the neutron yie1d and the amount of tritium

at the begining of use to those at the end of use

as a function of the tota1 current driven into

eacn tarロeτ.

Fig. 9.2.3

jAERI-M 88-221

9.4 An Improvement of Beam Transmission Performance in FNS Accelerator

J. Kusano, C. kutsukake, S. Tanaka, and Y. Abe

In FY-1987, the substantial maintenance activity of the FNS accelerator was paid for increasing the beam cur ren t at the target . The accelerator has been operated for seven years since the installation. The deuterium-ion beam incident upon the ta rge t is going to decrease slightly by a deterioration of both duoplasmatron of two ion sources . This deterioration induces a long operation hour in case of an irradiation experiment with full-power operation. Ambiguity of the operation schedule for an irradiation experiment depends upon the intensity of the ion beam c u r r e n t and the neutron yield character is t ics of the tri t iated titanium targe t .

Several plans were proposed from the experience of the accelerator operation and diagnosis to improve the beam transmission performance. The activities for the improvements and the resul t are itemized as follows: (1) A replacement of the s t rong focusing acceleration tube by

a modified acceleration tube of weaker focusing. The the beam transmission ratio was almost 30% higher than the old one.

(2) A replacement of the high-voltage power supply for the ion beam extraction with a new larger capacity one (the old; 100 kV • 150 mA DC, the new; 80 kV • 200 mA DC). This work improved the stability of the ion beam extraction in case of the heavy duty operation.

(3) A sensitization of the regulation amplifier in the high-voltage control system. This work was achieved by an increase of the amplitude of the regulation amplifier and a timing adjustment of the negative feed-back control system using an additional signal delay technique. The spatial and energy spread of the accelerated ion beam are reduced by this improvement. Almost 10% gain of the beam transmission ratio was observed.

As the resu l t of the improvements, the new ion beam transmission ratio was increased almost 40% compared with the old

- 2 2 2 -

jAERI-M 88・221

9.4 An Improv巴mentof Beam Transmission Performanc巴

in FNS Accelerator

J. Kusano， C. kutsulcake， S. Tanaka， and Y. Abe

In FY-1987， the substantial maintenance activity of the FNS

accelerator was paid for increasing the beam current at the target.

The accelerator has been operated for seven years since the

installation. The deuterium-ion beam incident upon the target is

going to decrease slightly by a deterioration of both duoplasrnatron

of two ion sources. This deterioration induces a long operation hour

in case of an irradiation experirnent with full-power operation.

Arnbiguity of the operation schedule for an irradiation experiment

depends upon the intensity of the ion beam current and the neutron

yield characteristics of the tritiated titaniurn target.

Several plans were proposed from the experience of the

accelerator operation and diagnosis to irnprove the bearn transrnission

perforrnance. The activities for the improvements and the result are

iternized as follows:

(1) A replacement of the strong focusing acceleration tube by

a modified acceleration tube of weaker focusing. The the bearn

transmission ratio was almost 30% higher than the old one.

(2) A replacement of the high-voltage power supply for the ion

beam extraction with a new larger capacity one (the old; 100 kV・

150 mA DC， the new; 80 kV . 200 mA DC). 1'his work improved the

stability of the ion beam extraction in ca!"-e of the heavy duty

operation.

(3) A sensitization of the regulation arnplifier in the high-voltage

control system. This work was achieved by an increase of the

amplitude of the regulation arnplifier and a tirning adjustment of

the negative feed-back control system using an additional signal

delay technique. The spatial and energy spread of the accelerated

ion beam are reduced by this improvement. Almost 10% gain of the

bearn transmission ratio was observed.

As the result of the improvements， the new ion beam

transmission ratio was increased almost 40% compared with the old

JAER1-M 88-221

one. The pulsed beam performance was also improved as achieving minimum pulse width 1.6 nano-second with a stable beam focusing character is t ic .

In addition to the above mentioned, an improvement of the deteriorated duoplasmatron ion sources was also considered about the electrode dimension, the s t ruc tu re of magnet and the cooling mechanism. As the resul t of the consideration, one duoplasmatron ion source will be replaced with a new one August 1988.

A photograph of the new acceleration cube installed is given in Fig.9.4.1

Fig.9.4.1 The new acceleration tube installed.

- 2 2 3 -

JAERI-M 88-221

one. The pulsed beam perfo:-mance was also improved as achieving

minimum pulse width 1.6 nano-second with a stable beam focusing

characteristic.

1n addition to the above mentioned， an improvement of the deteriorated duoplasmatron ion sources was also considered about

the electrode dimension， the structure of magnet and the cooling

mechanism. As the result of the consideration， one duoplasmatron

ion source will be replaced with a new one August 1988.

A photograph of the new acceleration乙ubeinstalled is given in

Fig.9.4.1

Fig.9.4.1 The new acceleration tube instaI1ed.

J A E R I - M 3 8 - 2 2 1

9.5 Operation Report of VHTRC

S. Fujisaki. M. Takeuchi , H. Yoshifuji, T. Ono, M. Seki, K. Kitadate* and M. Nakano

In FY-1987, much labor was expended in activities concerning1 operations and maintenance of the VHTRC (Very High Temperature Reactor Critical Assembly), and consequent performance of operation was achieved in success.

VHTRC-1 and VHTRC-4 cores were assembled to obtain experimental verification for design accuracy of the HTTR.

The operations for this purpose were completed as scheduled without any problem disturbing the execution of the experiments. The fuel management was performed according to the routine procedure.

Operations of 257 times were carried out in 67 days to perform the experiments. The total operation time was 510 hours. Thermal power was 160 watt-hours and critical operation 174 times in this fiscal year.

The integral operation time of 2051 hours has been recorded at the end of this fiscal year since the first achievement of criticality on May 13, 1985.

The maintenance activities of the VHTRC in this fiscal year were as follows: 1) Technical examination of the specifications for renewal of the VHTRC

control panel that installed various instrumentations. 2) Repair of the oil circulating system for the carriage drive

mechanism. 3) The installation of the reactor room monitor with 3 channel ITV

system. 4) The inspections after the overhaul of the control and the safety rod

drive mechanisms and the automatic extinguishing system. The activities relating to the maintenance of the VHTRC were also carried out. Various kinds of graphite rods with slit setting for the activation

foils were made. To improve the data handling operation and experimental data, the core loading pattern drawing program was completed. The specifications and the results of the measurement of the VHTRC operating data processing system will be described in section 9.6.

* (1987 Apr. ~ 1987 Aug.)

]AERト M 88・221

9.5 Operation Report of VHTRC

S. Fujisaki. M. Takeuchi. H. Yoshifuji. T. Ono. M. Seki.

K. Ki tadate* and M. Nakano

In FY-1987， much labor was expended in activities concerning oper-

ations and maintenance of the VHTRC (Very High Temperature Rellctor

Critical Assembly). and cons自quent performance of operation was

achieved in success.

VHTRC-1 and VHTRC-4 cores were assembled to obtain experimental

verification for design accuracy of the HTTR.

The operat ions for this purpose were completed as scheduled wi thout

any prob1em disturbing the execution of the expedments. The fuel

management was performed according to the routine procedure.

Operations of 257 times were carri申dout in 67 d且ysto perform the

experiments. The total operation time was 510 hours. Thermal power

was 160 watt-hours and critical operation 174 times in this fiscal ydar.

The integral operation time of 2051 hours has been recorded at the

end of this fiscal year since the first achi申vement of criticality on

Ms.y 13. 1985.

The maintenance activities of the VHTRC in this fiscal year wero

as follows:

1) Technical examination of the specifications for renewal of the VHTRC

control pa.nel that installed various instrumentations.

2) Repair of the oil circulating system for the carriage drive

mechanism.

3)τ'he installation of the reactor room monitor with 3 channel ITV

system.

4) The inspections after the overhaul oC the control and the saCety rod

driv骨 mechanisms and the automat ic ex:t inguishing system. The

activities relating to the maintenance oC the VlfTRC were a1so carri・dout.

Various kinds oC graphite rods with slit setting Cor the activation

Coils were made. To improve the data handling operation and ex:p・ri-mental data. the core loading pattern drawing program was comp1eted.

The specifications and the results of the measurement of the VH1百C

operating data ~~ocessing system wi11 be described in section 9.6.

禽 (1987Apr. 向 1987Aug.)

J A E R I - M 8 8 - 2 2 1

9.6 Development of VHTRC Operating Data Processing System

S. Fujisaki, M . Takeuchi, T. Ono and M. Seki

The software for the V H T R C operating data processing system which was installed in FY-1986 was improved considerably in FY-1987. The system is usually used for the operation and experimental data analysis. The availability of the data processing system was confirmed by using throughout this fiscal year. The system is composed of the C A M A C system, CPU, key-board, CRT, hybrid recorder, precision digital voltmeter and printer. The operation data consists of 47 kinds of data signals related to the facility. The data signals are processed by the CPU through the C A M A C system and are displayed on the CRT. 1) Indica:ion system of control rods and safety rods are composed of 8 channels.

2 channels : Digital position indication of 2 control rods 6 channels : Full-in or Full-out indication of 6 safety rods

2) Nuclear instrumentation system is composed of 15 channels. 2 channels : Counting indication of neutrons at the start up channel 8 channels : Indication of the current and range signal at

the linear channel 1 channel : Precise measurement data of the linear channel 1 by

precision digital voltmeter 4 channels : Indication of the current and range signal at the safety

channe1 3) Measurement system of the core temperature is composed of 19 channe1s.

17 channels : Indication of the tempearture by thermo element sensors over the whole core

2 channels - Indication of table temperature at the fixed half assembly

4) Measurement system of the table gap is composed of 2 channels. 2 channels : Indication of table gap between Fixed half and Movable

half In addition to these data signals, excess reactivity worth of the core was indicated by the on time calculation using control rod position and calibration curve data. Output data are indicated on the CRT and print out.

All of the data were used for daily operation data logging and for

]AERI-M 88・221

9.6 Development of VHTRC Operating Data Processing System

S. Fujisaki. M. Takeuchi. T. Ono and M. Seki

The software for the VHTRC operating dat& processing syst自m which

was installed in FY-1986 was improved consid 自 ra~ly in FY-1987.

The system is usually used for th自 operationand experimental

data analysis目Theavai labi 1 i ty of the data processing system was con-

Cirmed by using throughout this fiscal year. The system is composed

of the CAMAC systc;>m. CPU. k自y-board. CRT. hybrid recorder. precision

digital voltmeter and printer. The op自rationdata consists of 47 kinds

of data signals related to the facility. The data signals are processed

by the CPU through the CAMAC system and are displayed on the CRT.

1) Indica:ion system of control rods and safety rods are composed

of 8 channels.

2 channels Digital position indication of 2 control rods

6 channels Full-in or Full-out indication of 6 safety rods

2) Nuclear instrumentation system is compos日dof 15 channels.

2 channels Counting indication of neutrons at the start up channel

8 channels Indication of th自 currentand range signal at

1 channel

the linear channel

Precis自 measurement data of the 1 inear channel 1 by

precision digi tal vol tmeter

4 channels Indication of the current and range signal at the safety

chan且e1

3) Measurement system of the core temperature is composed of 19

chan且eIs.

17 channels Indication of the tempearture by thermo element sen-

sors over the whole core

2 channels . Indication of table temperature at the fixed half

assembly

4) Measurement system of the table gap is composed of 2 channels.

2 channels Indication of table gap between Fixed half and Movable

In addition to these data signals. excess reactivity worth of the core

was indicated by the on time calculation using control rod position and

calibration curve data. Output data are indicated on the CRT and print

All of the data were used for daily operation data logging and for

-225一

JAERI-M 88-221

the experiment data analysis. The processed data were used for determination of the control rod worths and the safety rod worths by the period method and the rod drop method. By using the system, the measurement time of the experiment was reduced to 1/3 compared with that of the old technique. The system chart is shown in Fig.9.6.1. An output data of the period method and that of the rod drop method are shown in Fig.9.6.2.

CH.l LINEAR N RANGE CONTROL ROO POSITION(FIXEO)

* (MOVABLE)

SAFETY ROO POSITION CH.2 LINEAR N RANGE SAFETY CH.3.CH.7 RANGE LINEAR H EXT-I RANGE

EXPERIMENT-1 (PULSE) * -2(PULSE) " -3(PULSE) * -l(PULSE)

START UP CH.5(PULSE) START UP CH.6(Pl/LSE)

CH.l LINEAR H (CURRENT) CH.2 LINEAR N (CURRENT) CH.4 LOG N (CURRENT)

SAFETY CH.3 LINEAR N (CURRENT) -SAFETY CH.7 LINEAR N (CURRENT) -LINEAR N EXT-I(CURRENT)

C A M A C S Y S T E M

48 b i t input jate(TTL)

bit input s i te

MKu I6bit Dual port aeaory

100MHz U t c n l n j scaler

SOMhz counter

Scannint ADC

Crate controller

Hybrid recorder Temperature.etc

Precision d i i i t a l voltaeter « r - C H . i U NEA« N (CURRENT)

Printer

Floppy disk

(Micro Computer) floppy disk

Fig.9.6.I Outline of the VIITRC data processing system

- 2 2 6 -

]AERI・M 88-221

the experiment data analysis. Th自 processeddata were used for deter-

mination of the control rod worths and th03 safety rod worths by the

period method and the rod drop method. By using the system. the

measurement time of the experiment was reduced to 1/3 compar自d wi th

that of the old t自chnique. The system chart is shown in Fig.9.6.1.

An output data of the period method and that of the rod drop method

are shown in Fig.9.6.2.

01.1 llNEAR N RA剛克C伽何附lRω 円lSlTI側(FIXED)

閃VA8lE)

sm:rv R加問lSlTION01.2 llHEAR N RANCE SAFETY CH. 3. CH. 7 RANCE llNEAR H 00-1 RA尚E

ロPERIItEIIT-I(円A.SE)'・2(PIJl.SE)'・3(PlJl.SE)'・4(門ιSE)

START ul' CH.5(円んs[)灯~RT UP CH.6(PIJl.SE>

01.1 llNEAR伺〈ωRR印T)α.2 llNEAR N (αJRREHT】01.4似川〈α刷剖T】SA陀rv01.3 LIHEAR H (ClJRR印T)SAF目YCH.7 LIHEAR N (ClJRR町T)llN臥RN以下I(ClJRR印T)

CAMAC SYSTEM

48 bil 1_1，"凶行l)

48 bi t in凹 I'0悼

64Kw 16bi I 加.1同rl聴聞ry

100'他凶白川崎町制er

50Ithz CIMJnter

Scannina ADC

Crate controller

Hybrid reco吋er T ... 開rature.etc

CH.I llH臥置 N(αJ1tR削T)

Fi g.9. 6.1 Out I i ne of the VIITRC da ta process i ng system

J A E R I - M 8 8 - 2 2 1

Measurement of control rod worth by Period Method : B E B : a8fQiS3$Q2£29Bt8B$46ftl7l* E l l # ^ : 0 0 4 w m a j : H J ; "

1988-06-09 :23

CH :1 RANGE :E-08 TS(Sjtifrate): 510.0 (SEC) T F ( M ^ T ) : 70S.0 (SEC) TC(BgSBf) : 180.0 (SEC) TR(_h||B$) : 630.0 (SEC)

100 90

p n I.-.I

80 p n I.-.I 70 E 60 R 50 + i 40 /« 30

20 10 0

- / -"

- « - — " " • ' ' -

1— 1 1 — 1 Y ' f h

0 100 200 300 T 1 M E

500 600 700

10 9 8 7

•\ 6 5

I FIX MOVE POWER

Measurement of safety rod worth by Rod Drop Method

HWSEB : B3fn63^02^29B21B^46B30i* H I 8 M : 002 MVT?+ l>*)\> • El ^ t ' A - . - A : 0.1 (SEC)

10 4 f 5 -<J

n iir-5 -

N 10"-5 -

T 10 -5 -

500 750 T I M E ( S E C )

1988-06-09 11:35 LEFT : 463.2 (SEC) RIGHT: 502.0 (SEC)

Fig.9.6.2 Output data of Period Method and Rod Drop Method

- 2 2 7 -

]AE R 1・M 88・221

F門1easuremen】tof control rod worth by Period ~門1ethod

型望装:露震謁量弘釘喜ぷぷa~噺:!限;渋{露麟覧耳酪酔!や台)片j ?潟;組lt:1{:i:; !;目(空醤酪;;〉行; ;認1罰引it訂:11(:i:;

jJl!!!ii[ ，.//'J11

1988-06-0911:21

膏:t'10VE汗、:P011¥IER

nwnAHリnHU

'ι「EEJ

T 1 M E ( S E C )

門easurementof safety rod worth by Rod Drop Method

測定日:昭和6:3年02月29日21時46分:30秒実験番号:002 角計庁チャンネ)1.': E 1 タイムペース O.1 (SEC)

1 !383-06-0O 11: :~5

LEFT: 463.2 (SEC) PIGHT: 502.0 (SEC)

l 〔1:15 ~"~.，.--ー九九ザ，.'町一~、•.• vr_-..，.. . .....，.，...'

!日ι

250 500 750 T [ t'1 E ( S E C )

!即日 1250

Pig.9.6.2 Output data of Period Meth叫 andRod Drop Met加d

-227ー

J A E R I - M 8 8 - 2 2 1

10. Activities of the Research Committee on Reactor Physics

Y. Kaneko, Y. Ishiguro, T. Nakamura and T. Suzuki

During this fiscal year, there were held two meetings of the Research Committee on Reactor Physics, three meetings each of the Subcommittees on Reactor System and on Fusion Reactor, and two meetings of the Subcommittee on Shielding. In addition, the Seminar on Software Development in Nuclear Energy Research was held in September 1987, in cooperation with the Nuclear Code Committee.

The Research Committee presented one review paper and 9 technical papers at the 30th meeting of the NEA Committee on Reactor Physics (NEACRP) as follows: L-297: Reactor Physics Activities in Japan (October 1986 - September

1987) (Y. Kaneko and K. Shirakata) A-825: Analysis of Chernobyl Reactor Accident (I)

- Nuclear and Thermal Hydraulic Characteristics and Follow-up Calculation of Accident - (T. Wakabayashi et al„)

A-826: Analysis of Chernobyl Reactor Accident (II) - An Examination of the Improvement Measures concerning the Accident of Chernobyl Power Plant -

(T. Wakabayashi et al.) A-827: An Analysis of Reactivity Coefficients of the Chernobyl

Reactor by Cell Calculation (K. Tsuchihashi and F. Akino) A-852: Inherent Safety Performance of a Mixed Oxide Fueled 1000 MWe

Loop type LMFBR under ATWS conditions (K. Yamaguchi, S. Ohta and H. Endo)

A-858: Error and Uncertainty Analysis of Neutron Transport Calculation (K. Furuta, Y. Oka and S. Kondo)

A-857: Fusion Blanket Engineering Benchmark Experiments - JAERI/USDOE Collaborative Program Phase II -

(T. Nakamura) A-845: Uncertainty Evaluation of Void Reactivity Worth in High-

Conversion Light-Water Reactors (Y. Yamaguchi and T. Takeda) A-843: Results and Analysis for FCA Phase-1 Experiment on High

Coversion Light Water Reactor (T. Osugi) A-849: Preliminary Report of HCLWR Cell Burnup Benchmark

Calculations (H. Akie, Y. Ishiguro and H. Takano)

jAERI-M 88-221

Y. Kaneko， Y. 1shiguro， T. Nakamura and T. Suzuki

During this fisca1 year， there were he1d two meetings of the

Research Committee on Reactor Physics， three meetings each of the

Subcommittees on Reactor System and on Fu忌ionReactor， and two meetings

of the Subcommittee on Shie1ding. 1n addition， the Seminar on Software

Development in Nuclear Energy Research was he1d in September 1987， in

cooperation with the Nuc1ear Code Committee.

The Research Committee presented one review paper and 9 technica1

papers at the 30th meeting of the NEA Committee on Reactor Physics

(NEAC即) as fo11ows:

L-297: Reactor Physics Acti.vities in Japan (October 1986 -September

1987) (Y. Kaneko and K. Shirakata)

A-825: Ana1ysis of Chernoby1 Reactor Accident (1)

-Nuc1ear and Therma1 Hydrau1ic Characteristics and Fo11ow-

up Calculation of Accident - (T. Wakabayashi et a1.)

A-826: Analysis of Chernoby1 Reactor Accident (11)

-An Examination of the Improvement Measures concerning the

Accident of Chernoby1 Power Plant -

(T. Wakabayashi et a1.)

A唱 827: An Analysis of Reactivity Coefficients of the Chernobyl

Reactor by Ce1l Calculation (K. Tsuchihashi and F. Akino)

A-852: 1nherent Safety Performance of a Mixed Oxide Fue1ed 1000 MWe

Loop type LMFBR under ATWS conditions

(K. Yamaguchi， S. Ohta and H. Endo)

A-858: Error and Uncertainty Analysis of Neutron Transport

Calculation (K. Furuta， Y. Oka and S. Kondo)

A-857: Fusion B1anket Engineering Benchmark Experiments

ー JAERljUSDOEC011aborative program Phase 11 -

(T. Nakamura)

A-845: Uncertainty Eva1uation of Void Reactivity Worth in High-

Conversion Light-Water Reactors (Y. Yamaguchi and T. Takeda)

A-843: Results and Analysis for FCA Phase-l Experiment on High

Coversion Light Water Reactor (T. Osugi)

A-849: Preliminary Report of HCLWR Ce11 Burnup Benchmark

Calculations (H. Akie， Y. 1shiguro and H. Takano)

J A E R I - M 8 8 - 2 2 1

The 52nd meeting of the Research Committee was held in November 1987 to review the 30th NEACRP meeting where 52 technical papers were presented on 10 topics including 5 new topics; (1) Integral Validation of Recent Delayed Neutron Data (2) Validation of Fission Product Data (in particular for Thermal

Reactors) (3) Physics Aspects of Design Innovation to Increase Inherent Safety

for Fast and Thermal Reactors (A) Fusion Blanket Experiments. Comparison of Measurements and

Calculations (5) Reactor Physics Issues Related to Intermediate Spectra Reactors

(Experiments, Burn-up Related Problems, Eventual Design Features) (6) Recent Results from Operating Reactors (7) Calculations and Measurements of Void Coefficients on Thermal and

Epithermal Lattices (8) Uncertainties in Reactivity Feedback Coefficients in Fast

Reactors (9) Reactivity Effects of Fuel Fragmentation in Light Water Cooled

Reactors (10) Data Testing in the USSR

In the 30th meeting of NEACRP, Japanese delegates proposed a specialist meeting on HCLWR cell burn-up benchmark calculations. In addition they proposed a Tritium Production Rate (TPR) benchmark experiment. Discussions were made also for the possibility of the 3-dimmensional transport calculation benchmark.

The Subcommittee on Reactor System held the 22nd meeting in June 1987 to discuss the papers to be submitted to the 30th NEACRP meeting. In the 23rd meeting held in November, review was made on the papers presented at the 30th NEACRP meeting. Summaries were also reported on the results of "NEACRP High Conversion Light Water Reactor Lattice Burnup Benchmark Calculation." In the 24th meeting held in March 1988, most of the discussions were concentrated on the topics of the experiment of JUPITER Phase 3 and the VHTRC experiment.

The Subcommittee on Fusion Reactor held three meetings during this term. The 23rd meeting in June 1987 was held concurrently with a part of the 3rd JAERI/USDOE workshop on Fusion Blanket Neutronics. Discussed items are preparation of the papers for the 30th NEACRP meeting and a proposal for the test of JENDL-3T by integral experiments. At the 24th

]AERI-M 88・221

百le52nd meeting of the Research Committee was held in November

1987 to review the 30th NEACRP meeting where 52 technical papers were

presented on 10 topics including 5 new topics;

(1) Integral Validation of Recent Delayed Neutron Data

(2) Validation of Fission product Data (in particular for Thermal

Reactors)

(3) Physics Aspects of Design Innovation to Increase Inherent Safety

for Fast and Thermal Reactors

(4) Fusion Blanket Experiments. Comparison of Measurements and

Calculations

(5) Reactor Physics Issues Related to Intermediate Spectra Reactors

(Experiments， Burn-up Related Problems， Eventual Design Features)

(6) Recent Results from Operating Reactors

(7) Calculations and Measurements of Void Coefficients on Thermal and

Epithermal Lattices

(8) Uncertainties in Reactivity Feedback Coefficients in Fast

Reactors

(9) Reactivity Effects of Fuel Fragmentation in Light Water Cooled

Reactors

(10) Data Testing in the USSR

In the 30th meeting of NEACRP. Japanese de1egates proposed a

specialist meeting on HCLWR cell burn-up benchmark calculations. In

addition they proposed a Tritium Production Rate (TPR) benchmark

experiment. Discussions were made also for the possibility of the 3-

dimmensional transport calculation benchmark.

The Subccmmittee on Reactor System held the 22nd meeting in June

1987 to discuss the papers to be submitted to the 30th NEACRP meeting.

In the 23rd meeting held in November. review was made on the papers

presented at the 30th NEACRP meeting. Summaries were also reported on

the results of "NEACRP High Conversion Light Water Reactor Lattice

Burnup Benchmark Calculation." In the 24th meeting held in March 1988，

most of the discussions were concentrated on the topics of the experi-

ment of JUPITER Phase 3 and the VHTRC experiment.

The Subcommittee on Fusion Reactor held three meetings during this

term. 百le23rd meeting in June 1987 was held concurrently with a part

of the 3rd JAERI/USDOE workshop on Fusion Blanket Neutronics. Discussed

items are preparation of the papers for the 30th NEACRP meeting and a

proposal for the test of JENDL-3T by integral experiments. At the 24th

-229一

J A E R 1 - M 8 8 - 2 2 1

meeting in October 1987, the results was reported and discussed of integral test on JENDL-3T for the nuclides used in fusion reactor. Other topics were the report on the 30th NEACRP meeting including the proposal for intercomparison of measuring techniques on tritium production rates, and on a proposal of IAEA benchmark calculations on lead sphere experiments. The 25th meeting in February 1988 dealt with the report on the US-Japanese Universities Workshop on Meutronics for Tritium Breeding Blanket along with the present status of the ITER project and its relevance to the activity of the Subcommittee.

The Subcommittee on Shielding held the 24th meeting in August 1987 to discuss the results of the second meeting on the NEACRP-CSNI cask benchmarks, the analysis of the ASPIS water benchmark experiment and the analyses of the JASPER experiments. At the same time, the announcement was made on the Seventh International Conference on Radiation Shielding (ICRS7) to be held at Bournemouth in September 1988. In the 25th meeting held in February 1988, discussions were concentrated on the technical contents of the 31 abstracts of papers submitted to the ICRS7 to prepare for the paper selection meeting of the International Program Committee held on 25 and 26 February 1988 in London.

Reference

Nuclear Code Committee and Committee on Reactor Physics: "Proceedings of the Fourth Seminar on Software Development in Nuclear Energy Research," JAERI-M 87-199 (1987) (in Japanese).

JAERI-M 88・221

meeting in October 1987， the results was reported and discussed of

integral test on JENDL-3T for the nuclides used in fusion reactor.

Other topics were the report on the 30th NEACRP meeting including

the proposal for intercomparison of measuring techniques on tritium

production rates， and on a proposal of IAEA benchmark calcu1.ations on

lead sphere experiments. The 25th meeting in February 1988 dealt with

the report on the US-JapaneGe Universities Workshop on Neutronics for

Tritium Breeding Blanket along with the present status of the ITER

project and its relevance to the activity of the Subcommittee.

The Subcommittee on Shielding held the 24th meeting in August 1987

to discuss the results of the second meeting on the NEACRP-CSNI cask

benchmarks， the analysis of the ASPIS water benchmark experiment and

the analyses of the JASPER experiments. At the same time， the an-

nouncement was made on the Seventh International Conference on

Radiation Shielding (ICRS7) to be held at Bournemouth in September

1988. In the 25th meeting held in February 1988， discussions were con-

centrated on the technical contents of the 31 abstracts of papers

submitted to the ICRS7 to prepare for the paper selection meeting of

the International program Committee held on 25 and 26 February 1988 in

London.

Reference

Nuclear Code Committee and C01lllllittee on Reactor Physics: "Proceedings

of the Fourth Seminar on Software Development in Nuclear Energy

Research，" JAERI-M 87-199 (1987) (in Japanese).

J A E R I - M 8 8 - 2 2 1

Publication List

1. Nuclear Data and Grc:p Constants - Benchmark Tests of JENDL-3T -

(1) Hayashi S. , Kimura I., Kobayashi K., Yamamoto S., Nishihara H., Mori T., Kanazawa S. and Nakagawa M.: "Measurement and Analysis of Fast Neutron Spectra in Iron, Nickel, and Chromium," J. Nucl. Sci. Technol., 24_, 702 (1987).

(2) Takano H. and Kaneko K.: "Benchmark Tests of JENDL-3T for Thermal Reactor and High Conversion Light Water Reactor," JAERI-M 88-065, pp.119-135 (1988).

(3) Takeda T., Takamoto M., Takano H. and Hasegawa A.: "Benchmark Test of JENDL-3T on Fast Reactors," JAERI-M 88-065, pp.135-148 (1988).

(4) Nakazawa M., Hasegawa A. et al.: "Proposals on Post-JENDL-3 Activity Programme for JNDC", JAERI-M 87-025, pp.9-13.

(5) Hasegawa A.: "For the Enhancement of the Utilization of Nuclear Data Prepared by JNDC," ibid., pp.33-39.

(6) Hasegawa A.: "A Design Study of the Consolidated Data Bank for the Evaluated Nuclear Data Library," JAERI-M 87-181.

2. Theoretical Method and Code Development (1) Mori T., Sasaki M. and Nakagawa M.: "ANISN-DD: One-dimensional

Sn Transport Code Using Multi-group Double-differential Form Cross Sections," JAERI-M 87-123 (1987).

(2) Sasaki M., Sato W., Nakagawa M. and Mori T.: "A Users' Guide of a Plotting Program PLTJOINT," JAERI-M 88-036 (1988).

(3) Tsuchihashi K. and Akino F.: "Analysis of Reactivity Coefficients of Chernobyl Reactor by Cell Calculation," J. Nucl. Sci. Technol., 24_, 1055 (1987).

(4) Miki R., Itoh T. and Tsuchihashi K.: "Critical Experiment and Analysis on Thorium Test Assembly in UTR-KINKI," JAERI-M 88-065, pp.460-466 (1988).

(5) Nishida T., Nakahara Y. and Tsutsui T.: "Analysis of the Mass Formula Dependence of Spallation Product Distribution," JAERI-M 87-088 (1987).

(6) Nishida T. and Nakahara Y.: "Analysis of Produced Nuclei and Emitted Neutrons in Nuclear Spallation Reaction," Kerntechnik,

JAERI-M 88・221

Publication List

1. Nuclear Data and Gro・:pConstants

-Benchmark Tests of JENDL-3T -

(1) Hayashi S.， Kimura 1.， Kobayashi K.， Yamamoto S.，

Nishihara H.， Mori T.， Kanazawa S. and Nakagawa M.: "Measure-

ment and Analysis of Fast Neutron Spectra in Iron， Nicke1， and

Chromium，" J. Nucl. Sci. Techno1.， 24， 702 (1987).

(2) Takano H. and Kaneko K.: "Benchmark Tests of JENDL-3T for

Thermal Reactor and High Conversion Light Water Reactor，"

JAERI-M 88-065， pp.119-135 (1988).

(3) Takeda T.， Takamoto M.， Takano H. and Hasegaw.l A.: "Benchmark

Test of JENDL-3T on Fast Reactors，" JAERI-M 88-065， pp.135-148

(1988) •

(4) Nakazawa M.， Hasegawa A. et a1.: "Proposals on Post-JENDL-3

Activity Programme for JNDC"， JAERI斗187-025， pp.9-13.

(5) Hasegawa A.: "For the Enhancement of the Utilization of Nuclear

Data Prcpared by JNDC，" ibid.， pp.33田 39.

(6) Hasegawa A.: .~ Design Study of the Consolidated Data Bank for

the Evaluated Nuc1ear Data Library，" JAERI-M 87-181.

(1) Mori T.， Sasaki M. and Nakagawa M.: "ANISN-DD: One-dimensional

Sn Transport Code Using Multi-group Double-differential Form

Cross Sections，" JAERI-M 87-123 (1987).

(2) Sasaki M.， Sato W.， Nakagawa M. and Mori T.: "A Users' Guide of

a Plotting program PLTJOINT，" JAERI-M 88司 036 (1988).

(3) Tsuchihashi K. and Akino F.: "Analysis of Reactivity

Coefficients of Chernobyl Reactor by Cell Calculation，" J. Nucl.

Sci. Technol.， 24， 1055 (1987).

(4) Miki R.， Itoh T. and Tsuchihashi K.: "Critical Experiment and

Analysis on Thorium Test Assembly in UTR-KINKI，" JAERI-M 88-065，

pp.460-466 (1988).

(5) Nishida T.. Nakahara Y. and Tsutsui T.: "Analysis of the Mass

Formula Dependence of Spallation Product Distribution，" JAERI-M

87-088 (1987).

(6) Nishida T. and Nakahara Y.: "Analysis of produced Nuclei and

Emitted Neutrons in Nuclear Spallation Reaction，" Kerntechnik，

JAERI-M 88-221

5 0 , 193 ( 1 9 8 7 ) .

(7) Nishida T. and Nakahara Y.: "Analysis of Transmutation of Transuranic Wastes by Nuclear Spallation Reaction," NEANDC(J) 127/U INDC(JPN) 113/L JAERI-M 88-065, p.246 (1988).

(8) Kanno I.: "Double-energy Double-velocity Measurement System for Fission Fragments and its Application," JAERI-M 87-173 (1987).

(9) Kanno I.: "Models of Formation and Erosion of a Plasma Column in a Silicon Surface-barrier Detector," Rev. Sci. Instrum., 58, 1926 (1987).

3. Reactor Physics Experiment and Analysis (1) Okajima S., Narita M. and Kobayashi K.: "Simple Determination

of Low Power on Reflected Reactors Using the Feynman-a Method," Ann. Nucl. Energy, j ^ (12), 673-676 (1987).

(2) Okajima S., Maeda A. and Mukaiyama T.: "Calculation Code of Heterogeneity Effects for Analysis of Small Sample Reactivity Worth," JAERI-M 88-035 (1988).

(3) Yamane T., Akino F. and Kaneko Y.: "A Simple Method for Reactivity Determination Based on Integral Version of Pulsed Neutron Area-Ratio Method," Proceedings of International Seminar on Nuclear Criticality Safety, October 19-23, p.307 (1987).

(4) Yasuda H., Akino F., Takeuchi M. and Kaneko Y.: "Measurements of Doppler Effect of Coated Particle Fuel Rod in SHE-14 Core Using Sample Heating Device," J. Nucl. Sci. Technol., 24_, 431 (1987).

(5) Yasuda H., Akino F., Yamane T., Yoshihara F., Kitadate K., Yosifuji H., Takeuchi M., Ono T. and Kaneko Y.: "Construction of VHTRC (Very High Temperature Reactor Critical Assembly)," JAERI 1305, (1987) (in Japanese).

4. Advanced Reactor System Design Studies (1) Takano H. and Akie H.: "Effect of Transplutonium Nuclides on

Burnup Reactivity Changes in HCLWRs," J. Nucl. Sci. Technol., 24, 501 (1987).

(2) Akie H. and Takano H.: "Effect of Fission Gas Release on Burnup Characteristics and Void Reactivity in HCLWRs," J. Nucl. Sci. Technol., 24, 668 (1987).

JAERI-M 88・221

50， 193 (1987).

(7) Nishida T. and Nakahara Y.: "Ana1ysis of Transmutation of

Transuranic Wastes by Nuclear Spallation Reaction，" NEANDC(J)

127/U INDC(JPN) 113/L JAERI-M 88-065， p.246 (1988).

(8) Kanno 1.: "Doub1e-energy Doub1e-ve10city Measurement System for

Fission Fragments and its App1ication，" JAERI-M 87-173 (1987).

(9) Kanno 1.: "Mode1s of Formation and Erosion of a P1asma Column

in a Si1icon Surface-barrier Detector，" Rev. Sci. Instrum.， 58，

1926 (1987).

3. Reactor Physics Experiment and Ana1ysis

(1) Okajima S.， Narita M. and Kobayashi K.: "Simp1e Determination

of Low Power on Ref1ected Reactors Using the Feynman-αMethod，"

Ann. Nuc1. Energy， 14 (12)， 673-676 (1987).

(2) Okajima S.， Maeda A. and Mukaiyama T.: "Ca1cu1ation Code of

Heterogeneity Effects foI' Ana1ysis of Sma11 Samp1e Reactivity

Worth." JAERI-M 88-035 (1988).

(3) Yamane T.， Akino F. and Kaneko Y.: "A Simp1e Method for

Reactivity Determination Based on Integral Version of Pul sed

Neutron Area-Ratio Method，" Proceedings of Internationa1 Seminar

on Nuc1ear Criticality Safety， October 19-23， p.307 (1987).

(4) Yasuda H.， Akino F.， Takeuchi M. and Kaneko Y.: "Measurements

of Dopp1er Effect of Coated Particle Fue1 Rod in SHE-14 Core

Using Sample Heating Device，" J. Nucl. Sci. Technol.， 2生， 431

(1987) .

(5) Yasuda H.， Akino F.， Yamane T.， Yoshihara F.， Kitadate K.，

Yosifuji H.， Takeuchi M.， Ono T. and Kaneko Y.: "Construction

of VHTRC (Very High Temperature Reactor Critica1 Assembly)，"

JAERI 1305， (1987) (in Japanese).

4. Advanced Reactor System Design Studies

(1) Takano H. and Akie H.: "Effect of Transp1utonium Nuc1ides on

Burnup Reactivity Changes in HCLWRs，" J. Nucl. Sci. Techn01.，

24，501 (1987).

(2) Akie H. and Takano H.: "Effect of Fission Gas Re1ease on Burnup

Characteristics and Void Reactivity in HCLWRs，" J. Nuc1. Sci.

Techn01.， 24，668 (1987).

J A E R I - M 8 8 - 2 2 1

(3) Nakano M., Takeda T. and Takano H.: "Sensitivity Analysis of Cell Neutronic Parameters in High-Conversion Light-Water Reactors," J. Nucl. Sci. Technol., 24_ (8), 610-620 (1987).

(4) Takano H., Kaneko K. , Akie H. and Ishiguro Y.: "The Effect of Fission Products on Burnup Characteristics in High Conversion Light Water Reactors," Nucl. Technol., 80, 250 (1988).

(5) Okumura K., Ishiguro Y. and Tsuchihashi K.: "On the Calculation Methods of Control Rod Worth for High Conversion Light Water Reactors," J. Nucl. Sci. Technol., _25, 318 (1988).

(6) lukeda T., Ishiguro Y. , Yoshida H. , Murao Y., et al.: "Present Status of Research Activities on High Conversion LWR in Japan," J. Atomic Energy Society of Japan, 23_, 1060 (1987) (in Japanese).

(7) Sako K.: "Concept and Feasibility of an Integrated Pressurized Water Reactor Incorporating a Poison Tank," Proc. of 25th Symposium on "The Advance Line of Power and Energy Technologies", Japan Society of Mechanical Engineers, Tokyo, Japan (Nov. 1987) (in Japanese).

(8) Sako K.: "Prospects of Increasing Passive Safety and Conversion Ratio in Light Water Reactors," Proc. 1988 National Symposium on Atomic Energy, Tokyo, Japan (Feb. 1988) (in Japanese).

(9) Asahi Y. and Wakabayashi H.: "improvement of Passive Safety of Reactors," Nucl. Sci. Eng. , 96 , 73-84 (1987).

(10) Ise T.: "R&D Trends in Small and Medium Sized Next Generation Reactor - Inherently Safe Light Water Reactors -," Proc. the Forth Seninar of Software Development in Nuclear Energy Research, JAERI-M 87-199, pp.49-138 (1987) (in Japanese).

(11) Ise T., Yamada M. and Hattori S.: "Trends in Research and Development of Innovative Small and Medium Sized Power Reactors - Mainly on Inherently Safe Reactors -," J. At. Energy Soc. Japan, 30, 104 (1988) (in Japanese).

5. Fusion Neutronics (1) Oyama Y., Tanaka S., Tsuda K., Ikeda Y. and Maekawa H.:

"A Small Spherical NE213 Scintillation Detector for Use in In-Assembly Fast Neutron Spectrum Measurements," Nucl. Instr. Meth. A256, 333 (1987).

(2) Oyama Y. and Maekawa H.: "Measurement and Analysis of An

JAERI-M 88・221

(3) Nakano M.， Takeda T. and Takano H.: "Sensitivity Analysis of

Cell Neutronic Parameters in High-Conversion Light-Water

Reactors，" J. Nucl. Sci. Technol.， 24 (8)，610-620 (1987).

(4) Takano H.， Kaneko K.， Akie H. and Ishiguro Y.: "The Effect of

Fission Products on Bumup Characteristics in High Conversion

Light Water Reactors，" Nucl. Technol.， 80， 250 (1988).

(5) Okumura K.， Ishiguro Y. and Tsuchihashi K.: "On the Calculation

Methods of Control Rod Worth for High Conversion Light Water

Reactors，" J. Nucl. Sci. Technol.， 25， 318 (1988).

(6) 'l<.lkeda T.， Ishiguro Y.， Yoshida H.， Murao Y.， et al.: "Present

Status of Research Activities on High Conversion LWR in Japan，"

J. Atomic Energy Society of Japan， 29， 1060 (1987) (in Japanese).

(7) Sako K.: "Concept and Feasibility of an Integrated Pressurized

Water Reactor Incorporating a Poison Tank，" Proc. Qf 25th

Symposium on "The Advance Line of Power and Energy Technologies"，

Japan Society of Mechanical Engineers， Tokyo， Japan (Nov. 1987)

(in Japanese).

(8) Sako K.: "Prospects of lncreasing Passive Safety and Conversion

Ratio in Light ¥Jater Reactors，" Proc. 1988 National Symposium on

Atomic Energy， Tokyo， Japan (Feb. 1988) (in Japanese).

(9) Asahi Y. and Wakabayashi H.: "lmprovement of Passive Safety of

Reactors，" Nucl. Sci. Eng.，笠， 73-84 (1987).

(10) Ise T.: "R&D Trends in Small and Medium Sized Next Generation

Reactor - lnherently Safe Light Water Reactors -" Proc. the

Forth Seninar of Software Development in Nuclear Energy Research，

JAERI-M 87-199， pp.49-138 (1987) (in Japanese).

(11) Ise T.， Yamada M. and Hattori S.: "Trends in Research and

Development of Innovative Small and Medium Sized Power Reactors

-Mainly on Inherently Safe Reactors -" J. At. Energy Soc.

Japan， 30， 104 (1988) (in Japanese).

(1) Oyama Y.， Tanaka S.， Tsuda K.， Ikeda Y. and Maekawa H.:

"A Small Spherical NE213 Scintillation Detector for Use in ln-

Assembly Fast Neutron Spectrum Measurements，" Nucl. lnstr.

Meth. A256， 333 (1987).

(2) Oyall回 Y. and l-包ekawaH.: "Measurement and Analysis of An

J A E R 1 - M 8 8 - 2 2 1

Angular Neutron Flux on A Beryllium Slab Irradiated with Deuterium-Tritium Neutrons," Nucl. Sci. Eng., 9_7, 220 (1987).

(3) Oyama Y.: "Fusion Neutron Shielding Benchmark Ploblem," JAEKl-M 87-203, pp.140-150 (1987) (in Japanese).

(4) Ikeda Y.: ibid, pp.209-219. (5) OishiK.: ibid, pp. 220-230. (6) Ikeda Y., Konno C , Oishi K., Nakamura T. , Miyade 11., Kawade K.,

Yamamoto H. and Katoh T.: "Activation Cross Section Measurements for Fusion Reactor Structural Materials at Neutron Energy from 13.3 to 15.0 MeV Using FNS Facility," JAERI 1312 (1988).

(7) Ikeda Y. and Youssef M. Z.: "Two-Dimentional Cross Section Sensitivity and uncertainty Analysis for Tritium Production Rate in Fusion-Oriented Integral Experiments," Fusion Technol. 1_3_ (4), 616 (1988).

(8) Maekawa H.: "Proceeding of The 1987 Seminar on Nuclear Data," JAERI-M 88-065, pp.198-211 (1988).

(9) Oishi K., Ikeda Y., Tomioka K. and Nakamura T.: ibid, pp.388-398 (1988).

(10) Maekawa H., Ikeda Y., Oyama Y., Yamaguchi S., Tsuda K., Fukumoto T., Kosako K., Yoshizawa M. ;i.ii! Nakamura T. : "Benchmark Experiments on A 60 cm-Thick Graphite Cylindrical Assembly," JAERI-M 88-034 (1988).

(11) Kosako K., Oyama Y., Maekawa H. and Nakamura T.: "Neutron Cross Section Libraries for Analysis of Fusion Neutronics Experiments," JAERI-M 88-076 (1988).

6. Radiation Shielding (1) Ise T., Narita H. and Itoh Y.: "Evaluation of Trial Design

Studies for an Advanced Marine Reactor (7) Shielding (I): Evaluation of Shielding Design," JAERI-M 88-055 (1988) (in Japanese).

(2) Nakashima H., Tanaka S. and Maekawa H.: "Experiments and Calculations of 14 MeV Neutron Streaming through Multi-Laye: • d Slit Assembly," J. Nucl. Sci. Technol., 24 (8), 601 (1987).

(3) Sakamoto Y., Tanaka S. and Harima Y.: "Data Library of Gamma-Ray Buildup Factors for Point Isotropic Source," JAERI-M 87-210.

(4) Tanaka S. et al.: "Annotated Bibliography and Discussion of Gamma-Ray Buildup Factors," Appl. Rad. Isot. 39 (3), 241 (1987).

-234 -

]AER1-M 88・221

Angular Neutron Flux on A Bery11ium S1ab Irradiated with

Deu tcrium-Tritium Neu t rons，" Nuc 1. Sc 1. Eng.， 97， 220 (1987).

(3) Oyall温 Y.: "Fusion Neutron Shielding Benchmark Ploblem，'・.JAERI-N

87-203， pp.140-150 (1987) (in Japanese).

(4) Ikeda Y.: ibid， pp.209-219.

(5) Oishi K.: ibid， pp. 220-230.

(6) Ikeda Y.， Konno C.， Oishi K.， Nakamllra '1'.，ト1iyade11.， Kawade K.，

Yamamoto H. and Katoh T.: "Ac tiva"ion Cross Scc tiol1ト'leaSllH'-

ments for Fusion Reactor Structllral Nateri礼Isat Neutrol1 ~l1 er民V

from 13.3 to 15.0トleVUsing FNS Faci1ity，" .JAERl 1312 (1988).

(7) Ikeda Y. and Youssef M. Z.: "Two-Dimentional Cross Section

Sensitivity and Uncertainty Analysis for Tritium Production Ratc

in Fusion-Oriented Integral Experiments，" Fusion Techn01. 13

(4)， 616 (1988).

(8) Maekawa H.: "Proceeding of The 1987 Seminar on Nuclear Data，"

JAERI-M 88-065， pp.198-211 (1988).

(9) Oishi K.， Ikeda Y.， Tomioka K. and Nakamura T.: ibid， pp. 388-

398 (1988).

(10) Maekawa H.， lkeda Y.， Oyama Y.， Yamaguchi S.， Tsuda K.， Fukllmoto

T.， Kosako K.， Yoshizawa M. ::.".! Nakamura T.: "Benchmark Experi-

ments on A 60 cm-Thick Graphite Cylindrical Assembly，" JAERI-M

88-034 (1988).

(11) Kosako K.， Oyama Y.， Naekawa H. and Nakamura T.: "Neu tron Cross

Section Libraries for Analysis of Fusion Neutronics Experiments，"

JAERI-N 88-076 (1988).

6. Radiation 5hielding

(1) Ise T.， Narita H. and Itoh Y.: "Evaluation of Trial Design

Studies for an Advanced Narine Reac tor (7) Shielding (1):

Evaluation of 5hielding Design，" JAERI-M 88-055 (1988) (in

Japanese).

(2) Nakashima H.， Tanaka 5. and Naekawa H.: "Experiments and

Calculations of 14 MeV Neutron 5treaming through Multi-Laye:' d

51it Assembly，" J. Nucl. Sci. Technol.， 2主(8)，601 (1987).

(3) 5akamoto Y.， Tanaka 5. and Harima Y.: "Data Library of GallUl1a-

Ray Buildup Factors for Point Iso!:ropic Source，" JAERI-N 87-210.

(4) Tanaka S. et a1.: "Annotated Bibliography and Discussion of

GallUl1a-Ray Buildup Factors，" App1. Rad. Isot. 39 (3). 241 (1987).

J A E R I - M 8 8 - 2 2 1

(5) Oka Y., Sasamoto N. et al.: "Fusion Neutron Shielding Benchmark Problems," JAERI-M 87-203.

7. Reactor and Nuclear Instrumentation (1) Wakayama N. et al.: "Control and Instrumentation of Nuclear

Power Plants," J. Soc. Instrument Control Eng. 26 (8) (1987) (in Japanese).

(2) Terada H. et al.: "Development of Fuel Failure Detection System for a High Temperature Gas Cooled Reactor (II)," IEEE. Trans. NS-35 (3) (1988).

(3) Katagiri M., et al.: "Manufacture of Mercuric Iodiode (Hgl2) Nuclear Radiation Detector and its Application," J. At. Energy Soc. Jpn. 29 (8) (1987) (in Japanese).

(4) Kakuta T. et al.: "Radiation Resistance of Optical Fibers," IEEJ, EIM-87-128 (1988) (in Japanese).

(5) Kakuta T. et al.: "Development of Data Base on Radiation Hardened Equipment and Materials," IEEJ, EIM-87-148 (1987) (in Japanese).

(6) Kakuta T., Nakase Y. et al.: "Compilation of the Radiation Hardened Data on Dielectric and Insulating Materials," IEEJ, EIM-87-144 (1987) (in Japanese).

(7) Ara K., Yagi H., Oomura T., Oomine M. et al.: "Formation of Magnetic Gratings by Beam Irradiation, —on Non-Magnetic Stainless Steel—," IEEJ, MAG-87-90 (1987) (in Japanese).

(8) Ara K., Yagi H., Oomura T., Oomine M. et al.: "Formation of Magnetic Gratings by Beam Irradiation, —on Carbon Steel, Invar, Ferromagnetic Stainless Steel—," IEEJ, MAG-87-91 (1987) (in Japanese).

(9) Yagi H., Kakuta T. and Ara K.: "Gamma-ray Irradiation Tests of Sensor Devices," Trans, of Robotic Sensor Symposium, 2007 (1988) (in Japanese).

(10) Sakai E. et al.: "Gase Nuclear Particle Detectors" International Workshop "Compound Semiconductors for Room Temperature X-ray and Nuclear Detectors," Davos, Sept. 1987, Nucl. Instrum. Method in Physical Research (to be published).

(11) Sakai E. et al.: "Average Energy Needed to Produce an Electron-Hole Pair in GaSe Nuclear Particle Detectors," IEEE. Trans. NS-35(1) (1988).

JAERI-M 88・221

(5) Oka Y.， Sasamoto N. et a1.: "Fusion Neutron Shie1ding Benchmark

Prob1ems，" JAERI-N 87-203.

7. Reactor and Nuc1ear Instrumentation

(1) Wakayama N. et a1.: "Contro1 and Instrumentation of Nuclear

Power Plants，" J. Soc. Instrument Contro1 Eng. 26 (8) (1987)

(in Japanese).

(2) Terada H. et a1.: "Development of Fue1 Failure Detection System

for a High Temperature Gas Cooled Reactor (II) ，" IEEE. Trans.

NS-35 (3) (1988).

(3) Katagiri N.， et a1.: "Nanufacture of Nercuric Iodiode (Hg12)

Nuc1ear Radiation Detector and its App1ication，" J. At. Energy

Soc. Jpn. 29 (8) (1987) (in Japanese).

(4) Kakuta T. et a1.: "Radiation Resistance of Optical Fibers，"

IEEJ， EIN-87-128 (1988) (in Japanese).

(5) Kakuta T. et a1.: "Development of Data Base on Radiation

Hardened Equipment and Nateria1s，" IEEJ， EIN-87-148 (1987) (in

Japanese).

(6) Kakuta T.， Nakase Y. et a1.: "Compilation of the Radiation

Hardened Data on Dielectric and Insulating Nate~ia1s ，" IEEJ，

EIN-87-144 (1987) (in Japanese).

(7) Ara K.， Yagi H.， Oomura T.， Oomine N. et a1.: "Formation of

Nagnetic Gratings by Beam lrradiation，ーーonNon-Nagnetic

Stain1ess Stee1--，" IEEJ， NAG-87-90 (1987) (in Japanese).

(8) Ara K.， Yagi H.， Oomura T.， Oomine N. et a1.: "Formation of

Nagnetic Gratings by Beam Irradiation，ーーonCarbon Steel， Invar，

Ferromagnetic Stain1ess Stee1--，" IEEJ， NAG-87-91 (1987) (in

Japanese).

(9) Yagi H.， Kakuta T. a.nd Ara K.: "Gamma-ray Irradiation Tests of

Sensor Devices，" Trans. of Robotic Sensor Symposium， 2007 (1988)

(in Japanese).

(10) Sakai E. et a1.: "Gase Nuc1ear Particle Detectors" Interna-

tional Workshop "Compound Semiconductors for Room Temperature

X-ray and Nuc1ear Detectors，" Davos， Sept. 1987， Nuc1. lnstrum.

Method in Physica1 Research (to be published).

(11) Sakai E. et a1.: "Average Energy Needed to Produce an Electron-

Ho1e Pair in GaSe Nuc1ear Particle Detectors，" IEEE. Trans.

NS-35(1) (1988).

J A E R I - M 8 8 - 2 2 1

8. Reactor Control, Diagnosis and Robotics (1) Suzudo T.: "Development of Identifier Programs Using Model

Reference Adaptive System," JAERI-M 87-073 (1987). (2) Shinohara Y.: "Application of an AI Method to Optimal Reactor

Control Problem," ANS Topical Meeting on Artificial Intelligence and Other Innovative Computer Applications in the Nuclear Industry, (1987).

(3) Nabeshima K. et al.: "Development of Dynamic Analysis Code for Nuclear Ship Propulsion Reactor," Fall Meeting of At. F.nergy Soc. of Japan, (1987).

(4) Shimazaki J. et al.: "Nonlinear Dynamics Simulation in CAD System for Control System Analysis of Nuclear Plant," Fall Meeting of At. Energy Soc. of Japan, (1987).

(5) Kusunoki T. et al.: "Development of Plant Dynamic Analysis Code for Integrated Self-Pressurized Water Reactor (ISPDYN), and Comparative Study of Pressure Control Methods," JAERI-M 87-213 (1987).

(6) Fujii Y. et al.: "Design and Operational Experience of the Man-Machine Interface of a Fully Computerized Control System," IAEA Conference on Man-Machine Interface in the Nuclear Industry, (1988).

(7) Shinohara Y. et al.: "The result of Sodium Boiling Detection Benchmark Test Using B0R-60 Reactor Noise Data," IAEA CRP Meeting (1987).

(8) Shinohara Y. et al.: "Detection of In-core Sodium Boiling by an Acoustic Method," Fall Meeting of At. Energy Soc. of Japan, (1987).

(9) Hayashi K. et al.: Study on the Goodness of System Identification Using Multivariate AR Modeling," SMORN-V (1987).

(10) Hayashi K. et al.: A Method of Nonstationary Noise Analysis Using Instantaneous AR Spectrum and Its Application to Borssele Reactor Noise Analysis," SMORN-V (1987).

(11) Hayashi K. et al.: "Development of Nonstationary Reactor Noise Data Recording System," Annual Meeting of At. Energy Soc. of Japan, (1988).

(12) Fujii Y. et al.: "Development of Remote Handling Technology for Reactor Dismantling (7)," Annual Meeting of At. Energy Soc. of Japan, (1988).

]AERI-M 88・221

8. Reactor Contro1. Diagnosis and Robotics

(1) Suzudo T.: "Deve10pment of Identifier Programs Using Model

Reference Adaptive System，" JAERI-H 87-073 (1987).

(2) Shinohara Y.: "Application of an AI Hethod to Optima1 Reactor

Control problem，" ANS Topical Meeting on Artificia1 lntelligence

and Other Innovative Computer App1ications in the Nuclear

Industry. (1987).

(3) Nabeshima K. et a1.: "Development of Dynamic Analysis Codt- for

Nuc1ear Ship Propu1sion Reactor." Fall Heeting of At. E:lergy

Soc. of Japan. (1987).

(4) Shimazaki J. et a1.: "Non1inear Dynamics Simulation in CAD

System for Contro1 System Ana1ysis of Nucle:u P1ant，" Fall

Heeting of At. Energy Soc. of Japan， (1987).

(5) Kusunoki T. et a1.: "Development of P1ant Dynamic Analysis Code

for Integrated Se1f-Pressllrized Water Reactor (ISPDYN). and

Comparative Study of Pressure Contro1 Hethods，" JAERI-M 87-213

(1987) .

(6) Fuj ii Y. et a1.: "Design and Operationa1 Experiene;c of the Nan-

Machine Interface of a Fully Computerized Contr01 System." IAEA

Conference on Man-Machine lnterface in the Nuc1ear lndustry，

(1988) .

(7) Shinohara Y. et a1.: "The result of Sodium soiling Detection

Benchmark Test Using BOR-60 Reactor Noise Data，" IAEA CRP

Heeting (1987).

(8) Shinohara Y. et a1.: "Detection of In-core Sodium Boi1ing by

an Acoustic Method，" Fa11 Meeting of At. Energy Soc. of Japan，

(1987) .

(9) Hayashi K. et a1.: Study on the Goodness of System Identifica-

tion Using Multivariate AR Mode1ing，" SMORN-V (1987).

(10) Hayashi K. et a1.: A Method of Nonstationary Noise Analysis

Using Instantaneous AR Spectrum and Its Application to Borsse1e

Reactor Noise Ana1ysis，" SMORN-V (1987).

(11) Hayashi K. et a1.: "Develo?meIlt of N()nstationary Reactor Noise

Data Recording System，" Annua1 Heeting of At. Energy Soc. of

Japan， (1988).

(12) Fujii Y. et a1.: "Deve10pment of Rt:mote Hand1ing Technology for

Reactor Dismant1ing (7)，" Annua1 Meeting of At. Energy Soc. of

Japan， (1988).

J A E R I - M 8 8 - 2 2 1

(13) Yagi H. et al.: "Development of Remote Handling Technology for Reactor Dismantling (8)," Annual Meeting of At. Energy Soc. of Japan, (1988).

(14) Sasaki S.: "Inverse Kinematics Algorithm for a Six-Link Manipulator Using a Polynomial Expression," J. Soc. Instr. Control Engineers, 23,, 485-490 (1987).

(15) Usui H. et al.: "Development of Robotic. Remote Handling Technology," Workshop Meeting on Reactor Decommissioning, (1987).

(16) Sasaki S.: "An Improved Method of Inverse Kinematics Calculation for a Six-Link Manipulator," JAERI-M 87-104 (1987).

(17) Sasaki S.: "A New Representation for Kinematics of Manipulator," JAERI-M 87-125 (1987).

(18) Usui H. et al.: "Development of Remote Handling Technology for Reactor Dismantling (9)," Fall Meeting of At. Energy Soc. of Japan, (1987).

(19) Sasaki S. et al.: "Kinematic Analysis of a Six-Link Manipulator Based on Vector, Elementary Geometry and Polynomial Expression," JAERI-M 87-175 (1987).

(20) Shinohara Y.: "R&D of Nuclear Robotics at JAERI," Robot (1988). (21) Fujii Y. et al.: "Development of Remote Hanlding Technology for

Reactor Dismantling (10)," Annual Meeting of At. Energy Soc. of Japan, (1988).

(22) Sasaki S.: "Inverse Kinematics Algorithm for a Six-Link Manipulator Using an Improved Polynomial Expressions," J. Soc. Instr. Control Engineers, 24, 204-206 (1988).

(23) Sasaki S.: "General Expression of Manipulator Kinematics," JAERI-M 88-018 (1988).

(24) Usui H. et al.: "Robotic Remote Handling Technology," JPDR Workshop Meeting (1988).

(25) Shinohara Y.: "Periodical Inspection and Its Automation in Nuclear Industry," J. Soc. Instr. Control Engineers, Yl_ (1988).

9. Activities of the Research Committee on Reactor Physics (1) Nuclear Code Committee and Committee on Reactor Physics:

"Proceedings of the Fourth Seminar on Software Development in Nuclear Energy Research," JAERI-M 87-199 (1987) (in Japanese).

jAERI-M 88-221

(13) Yagi H. et a1.: "Deve10pment of Remote Handling Techn010gy for

Reactor Dismant1ing (8)，" Annua1 Heeting of At. Energy Soc. of

Japan， (1988).

(14) Sasaki S.: "Inverse Kinematics A1gorithm for a Six-Link

Hanipul斗torUsing a P01ynomj弓i.Expression，" J. Soc. lnstr.

Contr01 Engineers， 23， 485-490 (1987).

(15) Usui H. et a1.: "Development of Robotic Remote f1andling

Techn010gy，" ¥oJorkshop Nee ting on Reac tor Decommissioning， (1987).

(16) Sasaki S.: "An Improved ~lethod of fnverse Kinematics Calcula-

tion for a Six由LinkトlanipuIator，" JAERlート187-104 (1987).

(17) Sasaki S.: "A New Representation for Kinematics of ~lanipulator ，"

JAERI-H 87-125 (1987).

(18) Usui H. et al.: "Deve10pment of Remote Handling Techn010gy for

Reactol-Dismant1ing (9)，" Fa11 Neeting of At. Energy Soc. of

Japan， (1987).

(19) Sasaki S. et a1.: "Kinematic Ana1ysis of a Six-Link Hanipulator

Based on Vec tor， E1ementary Ge，)metry and P01ynomia1 Expression，"

JAERI-N 87-175 (1987).

(20) Shinohara Y.: "R&D of Nuclear Robotics at JAERI，" Robot (1988).

(21) Fujii Y. et a1.: "Development of Remote Han1ding Technology for

Reactor Dismant1ing (10)，" Annua1 Heeting of At. Energy Soc. of

Japan， (1988).

(22) Sasaki S.: "lnverse Kinematics A1gorithm for a Six-Link

Hanipu1ator Using an Improved P01ynomia1 Expressions." J. Soc.

Instr. Contr01 Engineers， 24， 204-206 (1988).

(23) Sasaki S.: "Genera1 Expression of Hanipu1ator Kinemettics，"

JAERI-H 88-018 (1988).

(24) Usui H. et a1.: "Robotic Remote Handling Technology，" JPDR

Workshop Heeting (1988)，

(25) Shinohara Y.: "Periodical Inspection and Its Automation in

Nuclear Industry，" J. Soc. Instr. Control Engineers， 27 (1988).

(1) Nuclear Code Committee and Committee on Reactor Physics:

"Proceedings of the Fourth Seminar on Softw，品reDevelopment in

Nuclear Energy Research，" JAERI-H 87-199 (1987) (in Japanese).

一237ー

J A E R I - M 88-221

Author Index

ABE, Yuichi AKIE, Hiroshi AKINO, Fujiyoshi ASAHI, Yoshiro ASAMI, Susumu ARA, Katsuyuki

BENNETT , Edgar F.

FUJII, Yoshio FUJISAKI, Shingo FUKUDA+, Katsuo

GOTOH, Hiroshi

HARUYAMA, Mitsuo HASEGAWA, Akira HAYASAKA, Katsuhisa HAYASHI, Koji HIRAOKA, Toru ICHIKAWA, Hiroshi

* IDO , Masaru IIJIMA, Susumu IKEDA, Yuj iro

ISE, Takeharu ISHIGURO, Yukio ISHIKAWA , Tomoyuki ITOH, Hirokuni ITOH, Hiroshi ITOH+, Yasuyoshi KANAl' , Yasuji KANEKO , Kunio KANEKO, Yoshihiko KANNO, Ikuo KATAGIRI, Masaki KAWASAKI , Hiromitsu

9 . 2 , 9 . 3 , 9 .4

4 . 4 , 4 . 5

3 . 1 , 3 . 2 , 3 . 3 , 3.4

7 . 2 , 7.6

9 . 1 , 9 . 5 , 9 .6

7.4, 7.7

7.4 1 .3 , 1.4, 1.9, 2 . 4 , 6 .1 3 . 5 , 9 . 1

8 . 5 , 8 . 6 , 8 . 7 , 8 . 8 , 8.9

4 . , 4 .15

4 . 1 0 , 4 . 1 1

3 . 7 , 4 .16

1.6, 5 . 1 , 5 . 3 , 5 . 5 , 5 .6

4 . 8 , 6 .3 , 6.5

2 . , 4 . 1 , 4 . 2 , 4 . 3 , 4 . 4 ,

1 . 1 , 1 .2, 4 . 3

Foreword, 3 . 2 , 3 . 3 , 3 . 4 ,

]AERI-M 88・221

Author Index

ABE， Yuichi 9.2， 9.3， 9.4

AKIE， Hiroshi /..4， 4.5

AKINO， Fuj iyoshi 3.1， 3.2， 3.3， 3.4

ASAH1， Yoshiro 4.9

ASAMI， Susumu 7.3

ARA， Katsuyuki 7.2， 7.6

* BENNI~TT Edgar F. 5.1

FUJII， Yoshio 8.12

FUJISAKI， Shingo 9.1，9.5，9.6

FUKUDA+， Katsuo 7.3

GOTOH， Hiroshi 7.4， 7.7

HARUYAMA，トlitsuo 7.4

HASEGAWA， Akira 1.3， 1.4， 1.9，2.4，6.1

HAYASAKA， Katsuhisa 3.5， 9.1

HAYASHI， Koj i 8.5，8.6，8.7，8.8，8.9

HlRAOKA， Toru 4.， 4.15

ICH1KAWA， Hiroshi 4.10，4.11 犬

1DO ， Masaru 4.11

IIJIHA， Susumu 3.7， 4.16

IKEDA， Yujiro 1.6， 5.1， 5.3， 5.5， 5.6， 5.7， 5.8，

1SE， Takeharu 4.8， 6.3， 6.5

1SHIGURO， Yukio 2.， 4.1， 4.2， 4.3， 4.4， 10.

* ISHIKAWA ， Tomoyuki 1.8

1TOH， Hirokuni 7.5

ITOH， Hiroshi 7.1

ITOH+， Yasuyoshi 6.3 ，~

KANAI ， Yasuji 6.6

* KANEKO ， Kunio 1.1， 1.2，4.3

KANEKO， Yoshihiko Foreword， 3.2， 3.3， 3.4， 10.

KANNO， Ikuo 3.1

KATAGIRI， Hasaki 7.5 大

KAWASAKI ， Hiromitsu 6.5

-238ー

J A E R I - M 8 8 - 2 2 1

KITADATE , Kenji KOBAYASHI+, Noriaki KONNO, Ghikara

KOSAKO , Kazuaki KOTEGAWA, Hiroshi KUMAGAI, Akio KUSANO, Joichi KUSUNOKI+, Tsuyoshi KUTSUKAKE, Chuzo

MAEKAWA, Hiroshi MARUO + Takeshi MATTAS , Richard F MORI, Takamasa MUKAIYAMA, Takehiko

NABESHIMA, Kunih iko NAGATANI , Mutsumi NAITO+, Yoshitaka NAKAGAWA, Masayuki NAKAHARA, Yasuaki NAKAJIMA , Hiroo NAKAMURA, Tomoo NAKANO, Masafumi NAKASHIMA, Hiroshi NARITA , Hideo NARIYAMA , Nobuteru NEMOTO, Tatsuo NISHIDA, Takahiko

OBATA+, Masahiro OBU, Makoto OHNO, Akio OGAWA+, Toru OIGAWA, Hiroyuki OISHI , Koji OKAJIMA, Shigeaki OKUMURA, Keisuke

3.3, 9.5 7.3 1.6, 5.1, 5.3, 5.5, 5.6 5.10, 6.2 1.6, 1.7, 1.8, 5.4, 5.5 6.1 8.12 9.2, 9.3, 9.4 8.1 9.2, 9.3, 9.4

1.6, 1.7, 1.8, 5.1, 5.3 6.5 5.1 1.5, 2.4, 5.2 4.12

8.1, 8.2, 8.7, 8.9 3.5 4.10, 4.11 1.5, 2.4, 5.2 2.1, 2.2, 2.3 4.8 1.6, 5., 5.1, 5.4, 6.2, 3., 9., 9.1, 9.2, 9.5 2.4, 6.1 1.8 6.6 3.7 2.1, 2.2

7.3 3.7 3.5, 9.1 4.12, 4.13 3.5 5.5, 6.2 3.6, 3.8 4.1, 4.2, 4.3

Noriaki

Chikara

，，

F仏EKAWA， Hiroshi

NARUO+ Takeshi

* MATTAS ， Richard F.

NORI， Takamasa

MUKAIYAMA， Takehiko

NABESHIHA， Kunihiko

* NAGATANI ， Mutsumi

NAITO+， Yosh i taka

NAKAGAWA， Hasayuki

NAKAHARA， Yasuaki

* NAKAJINA ， Hiroo

NAKAMURA， Tomoo

NAKANO， Hasafumi

NAKASHlMA， Hiroshi

* NARITA ， Hideo 世

NARIYA1-'.A， Nobuteru

NEMOTO， Tatsuo

NISHIDA. Takahiko

OBATA +， Hasahiro

OBU， Makoto

OHNO， Akio

OGAWA+， Toru

OIGAWA， Hiroyuki 士

OISHI ， Koji

OKAJlMA， Shigeaki

OKUMURA' Keisuke

JAERI~M 88幽 221

3.3， 9.5

1.6， 5.1， 5.3， 5.5， 5.6， 5.7， 5.8，

5.10， 6.2

1.6， 1.7， 1.8，5.4，5.5

9.2， 9.3， 9.4

1.6， 1.7， 1.8， 5.1， 5.3， 5.4， 5.9

1.5， 2.4， 5.2

8.1， 8.2， 8.7， 8.9

4.10，4.11

1.5， 2.4， 5.2

2.1， 2.2， 2.3

1.6， 5.， 5.1， 5.4， 6.2， 10.

3.，9.，9.1，9.2，9.5

2.4， 6.1

2.1， 2.2

3.5，9.1

4.12， 4.13

5.5， 6.2

3.6， 3.8

4.1， 4.2， 4.3

JAER1-M 88-221

ONO, Toshihiko OSAKABE , Masahiro OSUGI, Toshitaka OYAMA, Yukio

PORGES , Karl G.

SAITO*, Jun SAKASAI, Kaoru SAKURAI+, Kiyoshi SAKURAI, Takeshi SAKO, Kiyoshi SASAKI*, Makoto SASAKI, Shinobu SASAMOTO, Nobuo SATO , Osamu SATO , Wakaei SATOH, Kunio SEKI, Masakazu SHIMAZAKI, Junya SHINOHARA, Yoshikuni

SUZUKI, Katsno SUZUKI, Makoto SUZUKI, Tomoo

TAHARA , Yoshihisa TAKADA, Hiroshi TAKANO, Hideki

* TAKEUCHI , Kiyoshi TAKEUCHI, Motoyoshi TAKEUCHI, Yoshitaka TAKIZUKA+, Takakazu TANABE+, Fumiya TANAKA, Shigeru TANAKA, Shun-ichi TERADA, Hiromi TERANISHI , Kazuo

3 . 1 , 3 . 3 , 9 . 5 , 9 .6

4 . 1 2 , 4 .15

3.5, 3 .6, 3.8 1.6, 1 .7 , 1 .8 , 5 . 3 , 5 . 5 , 5.10

7.1 1.8 3.6, 3.7, 3.8 4.6, 4.7, 4.15 2.4 8.10 , 8.11, 8 .12 1.8 1.8 3.6 3.5, 9.1 9.5, 9.6 8.1, 8.2, 8.4 , 8. 12 8., : 8.1, 8.3, 8.7 , 8.8, 8.9, 8. 10, 8.12 8.4 7.4 2.4, 6., 6.1, 6.4 , 6.6, 10.

3.6, 3.7 2.3 1., 1 L.l, 1.2, 4.4; , 4.5, 4.10, 4 .11 4.12, , 4.14, 4. 15, 4.16 6.6 3.1, 3.2, 3.3, 9.! i, 9.6 7.4 4.12 4.8 9.2, 9.3, 9.4 2.4, 6.1 7.3 7.7

- 2 4 0 -

ONO， t

OSAKABE ，

OSUGI，

OYAMA，

PORGES ，

Scusss * SASAKI ，

SAS必(1，

SASAMOTO，

* SATO ，

SATOH，

SEKI，

SHlMAZAKI，

SHINOHARA，

SUZUKI，

* TAHARA ，

TAKADA，

TAKANO，

Toshihiko

Masahlro

Toshitaka

Karl G.

Kiyoshi

Takeshi

Kiyoshi

Makoto

Shinobu

Wakaei

Masakazu

Yoshikuni

Katsno

Makoto

Yoshihisa

Hiroshi

Hideki

* TAKEUCHI ， Kiyoshi

T~KEUCHI ， Motoyoshi

TAKEUCHI， Yoshitaka

TAKIZ11KA+， Takakazu

TANABE+， FUII11ya

TANAKA， Shigeru

TANAKA， Shun-ichi

TERADA， Hiromi

* TERANISHI ， Kazuo

)AERI-M 88・221

3.1， 3.3， 9.5， 9.6

4.12，4.15

3.5， 3.6， 3.8

1.6， 1.7， 1.8， 5.3， 5.5， 5.10

3.6，3.7， 3.8

4.6， 4.7， 4.15

8.10，8.11，8.12

3.5， 9.1

9.5， 9.6

8.1，8.2，8.4，8.12

8.， 8.1， 8.3， 8.7， 8.8， 8.9， 8.10，

2.4， 6.， 6.1， 6.4， 6.6， 10.

3.6， 3.7

1.， 1.1， 1.2， 4.4，4.5， 4.10， 4.11，

4.12， 4.14， 4.15， 4.16

3.1， 3.2， 3.3， 9.5， 9.6

9.2， 9.3， 9.4

2.4， 6.1

-240ー

J A E R I - M 88-221

TOMIOKA , Kazuyuki TSUDA, Koichi TSUTSUI, Tsuneo

UEKI , Kohtaro USUI, Hozumi

WAKAYAMA, Naoaki WATANABE, Hideaki WATANABE, Koichi

YAGI, Hideyuki YAMADA, Masaharu YAMAGISHI, Hideshi YAMAGUCHI, Seiya YAMAJI*, Akio YAMANE, Tsuyoshi YASUDA, Hideshi YOKOMURA+, Takeyoshi YOSHIDA, Hiroshi YOSHIDA, Hiroyuki YOSHIFUJI, Hisashi YOSHIMURA, Youichi

5.5, 6 .2 1.6, 5 .1 2.1, 2 .2

1.8 8.12

7., 7. 1, 7. 2, 7. 3, 7.4, 7. 5 9.1 8.5, 8, .9

7.6 7.2 7.1, 7, .4 1.6, 1. 7, 5 .1, 5 .3, , 5.4, 5 6.4 3.3, 3. ,4 3.3 8.1 7.3 4.16 3.1, 3. 3, 9 .5 7.4

+ Contributors attached to JAERI but not to Department of Reactor Engineering.

* Contributors attached not to JAERI.

)AE R 1‘M 88・221

* TOMIOKA ， Kazuyuki 5.5， 6.2

TSUDA， Koichi 1.6，5.1

TSUTSUI， Tsuneo 2.1， 2.2

* UEKI ， Kohtaro 1.8

USUI， Hozumi 8.12

WAKAYAMA， Naoaki 7.，7.1，7.2， 7.3， 7.4， 7.5

WATANABE， Hideaki 9.1

WATANABE， Koichi 8.5， 8.9

YAGI， Hideyuki 7.6

YAMADA， Masaharu 7.2

YAMAGISHI， Hideshi 7.1， 7.4

YAMAGUCHI， Seiya 1.6， 1.7， 5.1， 5.3， 5.4， 5.10 古

YAMAJI ， Akio 6.4

YAMANE， Tsuyoshi 3.3， 3.4

YASUDA， Hideshi 3.3

YOKOMURA+， Takeyoshi 8.1

YOSHIDA， Hirosili 7.3

YOSHIDA， Hiroyuki 4.16

YOSHlFUJI， Hisashi 3.1， 3.3， 9.5

YOSHlMURA ， Youichi 7.4

+ Contributors attached to JAERI but not to Department of Reactor

Engineering.

* Contributors attached not to JAERI.

Appendix I Department of Reactor Engineering Organization Chart

Director (Vacant)

Deputy Director : Y. Kaneko*

Research Committee on Reactor Physics

<Sub-Com»1ttees> . ,—.... ( Reactor System i Shielding ; Fusion Reactor

Advanced Reactor Assessment Tew

T. Hiraoka (Head)

K. Sako

T. lse

H. Takano

T. Hukaiyam*

r. Asahl

H. lchikaxa

T. M a i t o ' 1 '

T. Furuta

T. Tiklzuka - ( 3 )

(1) (2)

F. Tanabe1

T. Ogiwa (4)

Reactor System Lab.

Ishlguro (Head)

Y. Nakahara

H. Nakagawa

V. Taj i

T. Nishida

T. Tsutsui T. Mori

H. Akie

K. Okuuira

Fast Reactor Physics Lab.

H. Yoshlda (Head)

M. Dbu

S. lijima T. Osugl

T. Neaoto

S. Okajina

K. Sakurai

H. Ofgan

Thermal Reactor Physics Lab.

kaneko (Concurrent)

K. Tsuchihashi

H. Vasuda

F. Akino

N. Mori

T. Yamane

I . Xanno

(Ho.): Original organization of the shared-time members

(1) Department of fuel Safety Research

(2) Department o f High Temperature Engineering

(3) Department of Reactor Safety Research

(4 ) Department o f Fuels and Mater ia ls Research

(5) Department of Power Reactor Projects

(6) Department o f Large Tolumak Research

(7) Department of Engineering Services

* Acting Director

Reactor Instrumentation Lab.

. Wakayama (Head)

K. Ara

H. Gotoh

H. Yagi

M. Katagiri

H. Yamada

T. Kakuta

H. Yamagishi

M. haruyatna

H. I ton

K. Sakasal

Y. Takeuchi

Y. Yoshimura

S. Asami

M. Suzuki

H. I toh

Reactor Control Lab.

Shinohara (Head)

T. Fujisawa

J . Shioazaki

I I . Usui

K. Uatanabe

Y. F u j i i

S. Sasaki

K. Hayashi K. Nabeshima

A. Kunagai

K. S u z u k i ( 5 )

Fusion Reactor Physics Lab.

T. Nakamura (Head)

H. Maekawa

Ikeda Oyama Yamaguchi

Kosako Sen(«)

Shielding Lab.

(Kead)

S. Tanaka

N. Sasaaoto

A. Hasegawa

H. Kotegaw

H. Kakashtma

Reactor Physics F a c i l i t y Operation Div.

hakano (General Manager)

N. Watanabe

S. Fujisaki

A. Ohno

K. Satoh

H. Yoshifuj i

H. Takeuchi

J . Kusano

C. Kutsukake

S. Tanaka

T. Ono

Y. Abe

k. Hayasaka

H. Seki

Y. Kanega1

I . Takahashi 1 7 '

T. Shi ina' '

, ( 7 )

a s h i (

Administrative Services

K. Toriutri {Administrative

Manager) T. No garni K. Koike T. Yamashita K. Konori H. Kogure

K. Moriguchi

H. Toshida

S. rtomiyama

Department of Reactor Engineering Appendix I

Organization Chart

Depattlnent of Reactol" Engll':eenng

Director (Vac::ant)

Deputy Bfrector:1hneMe '* Acting Oirector

ReSllrth Co嗣 Itteeon Reo川ctorPhyslcs

< 5ub-COI圃 Ittees > 一一一一一一一一日一一・・・・・一一一一一-一一一-Relctor 5yst四 5h1eldlng Fuslon Reactor

】〉明間両日

SF日∞

Admi.nistrative S.同 ices

K. Tod回 i(Adlni ni str.ti ve

Hono9Or)

T. Nog掴 t

IC. Koike

T. Y~5hitJ

K. Ko同ri

H. Kogu同

ReactDr Physic5 hcility Ope:-ation Div.

H: NO旬開(6on...l lianager)

H. Watana~

S. Fujisoki

A.日hr.心

K. Sitoh

H. Yoshifuji

阿. Tok.uchi

J.. KuSino

C. Kutsuk.k.

S. Tinoko

T. Ono

1 晶e

K. Hiy..oki

H. Seki (7) Y. Ka開明

1. T・kihlshi(1) (7)

i. Sh1fno

T. S岡山3{陶ad)

s. Tanaka

N. S.S...，to

A. t匂Segilf'

H. Kotega ....

H. Naklshh同

Fusion Reactot Physics Lob.

T.. Nakamura (Heod)

H. liaekawa

Y. Ikeda

Y. Oy.旧

S. Yomaguchi

K. Tsuda

c. Konno

K. Kosako

Y. S.ki(6)

Reactor Contro1 lab.

Shi問 hara(Ilead)

T. Fujiso同

J. Shi回 zaki

tt. Usui

K. W.tonobe

Y. Fujii

S. Sos.ki

K. Hayoshi

K. NObeshi副

A. K咽 O9Oi(5)

K. Slttukl

Reactor Instru-mentation lab.

N. Wakay四 a(Head)

K. Ara

H. Gotoh

H. Y'91

M. katagid

H. Yomod.

T. Kakut・H. Y....glshl

H. Haruy.・noH. Itoh

K. S.hs・1Y. Tak.uchi

Y. Yoshi・Urls. As嗣 t

H. Suzukl

H. Itoh

Tllermal R.actor Phys;cs lab.

Y. Kan.kO (Concurr.nt)

K. Tsuchihlshi

H. Yasuda

F. Akino

N. Morl

T. Y岬 an.

1. Kanno

Fut R.actor Physics L.b. -H. Yoshlda

{憎ad)

M. Obu

5. Iljl嗣

T. Osugi

T. H倒聞to

S. Okljl副

K. 5lkur.i

H. Ol9Owo

R..ctor 5yst凹

Y. Ishiguro (Heod)

Y. N.kohlr.

H. H.kagaw.

Y. Tajl

T. Hlshld.

T. Tsutsul

T. Morl

H. Akle

K. Ok圃 ur・

Ady.nc.d R.・ctorAssesSl間同 T.醐

T. Hlraok. {陶.d)

K. S.k。T. Ise

H. Tak.no

T. Muk.ly酬 a

Y. Auhl

H. Ichlkaw・Y. H・Ito(l)

T. furuta(1)

T. Taklzuk. (2

f. Ta同国(3)

T. Ogaw.(・}

記令。

i魁 L主組担p1orQ・nlzatlon0' the山内d引間眠曲ers

(1) Oep・r回ent01 f叫 15.f.ty Rese.rch

(2) Oep・r臨時。1Hlgh T，個師同tu同 Englnetrlng

(3) Oep・F旬睡nt01 R..ctor 5・1.tyR.selrch

い)Oep・r回 ntof F国1s.nd Materia 1s R.¥urth

(5) Oepartlle叫 of'0同 rReocto~ Projects

(5) Oep・r同 nt01 Lar90 Tok醐 k量esel.rch

(7) O.p.rllllent of Engl n..r1"g So内 ices

J A E R I - M 8 8 - 2 2 1

Appendix II Abbreviations

FCA : JFast Critical Assembly

The FCA is a split-table type facility of horizontal matrix structure designed for studying nuclear characteristics of fast reactor. The construction of the FCA was started in 1965 and the first core went critical on 29th April, 1967. The main features of the facility are summarized as follows:

Type : Split-table type of horizontal matrix structure Size : 2.8m x 2.8m x 1.3m (each half assembly) Fuel : Enriched uranium and plutonium

(Plate type ) Other material : Sodium, stainless steel, aluminum oxide

(AI2O3), polystylene etc. (Plate type)

Maximum power : 2 kW Assembly name constructed : FCA I ~ FCA XV

Critical experiments were made using enriched uranium cores in 1960s for investigating basic characteristics of fast reactor cores. Mock-up experiments were extensively made in 1970s for the Fast Experimental Reactor JOYO (initial critical in 1977) and the Prototype Fast Breeder Reactor MONJU (under construction). Recently, the main subjects are the investigation of the core characteristics of an axially heLerogeneous large fast breeder reactor and the core physics study on a high conversion light water reactor.

FNS : Fusion K[eutronics jJource

The FNS is an accelerator based D-T neutron source installed for the purpose of investigating the neutronics on the D-T fusion reactor blanket and shielding. It provides following three functions to meet experimental requirements: 1) High intensity DC point source 2) DC point source with wide variation of neutron yield rate 3) Pulsed neutron source

JAERI-M 88司 221

Appendix 11 Abbreviations

FCA Fast Critical全ssembly

The FCA is a split田 tabletype faci1ity of horizontal matrix

structure designed for studying nuclear characteristics of fust

reactor. The construction of the FCA was started in 1965 and

the first core went critical on 29th April， 1967. The main

features of the facility are summar.ized as follows:

Type Split-table type of horizontal matrix structure

Size 2.8m x 2.8m x 1.3m (each half assembly)

Fuel Enriched uranium and plutonium

(Plate type )

Other material Sodium， stainless steel， aluminum oxide

(A1203)， polystylene etc.

(Plate type)

Maximum power 2 kW

Assembly name constructed FCA 1 ~ FCA XV

Critical experiments were made using enriched uranium cores in

1960s for investigating basic characteristics of fast reactor

cores. Mock-up experiments were extensively made in 1970s for

the Fast Experimental Reactor JOYO (initial critical in 1977)

and the Prototype Fast Breeder Reactor MONJU (under construc-

tion). Recently， the main subjects are the investigation of

the core characteristics of an axially heLerogeneous large fast

breeder reactor and the core physics study on a high conversion

light water reactor.

FNS Fusion Neutronics Source

The FNS is an accelerator based D-T neutron source installed

for the purpose of investigating the neutronics on the D-T

fusion reactor blanket and shielding. 1t provides following

three functions to meet experimental requirements:

1) High intensity DC point source

2) DC point source with wide variation of neutron yield rate

3) Pulsed neutron source

J A E R l - M 8 8 - 2 2 1

The D-T neutrons are generated via 3T(d,n)'*He reaction. There are two beam lines; one is so called 0° line for high current operation, and the other is so called 80° line for rather low current operation. The major specifications of the FNS accelerator are shown in the following Table.

< Table >

Items 0° 80°

• Beam current > 20 inA 3 mA • Beam size < 15 mm $ < 15 nun f • Pulse width < 2 ns • Frequency 2 MHz • Peak current 40 mA • Target assembly Rotating Stationally

(Water cooled) (Water cooled. Air cooled)

• Amount of 3T < 1000 Ci 10 Ci • Neutron Yield 4 * 10 1 2/s 5 x 10 i x/s

The major experimental subjects are as follows: 1) Tritium production rate in the various blanket configura

tions 2) Nuclear heating rate in the structural materials 3) Shielding performance for D-T neutrons in the various

structure configurations 4) Induced effects on the structural materials by D-T neutrons

HCLWR : jligh Conversion Light Water jleactor

HTTR : High Temperature engineering Test jteactor

JAERI is to construct and operate the HTTR, to carry out the necessary R & D for establishing and upgrading the HTGR (Hight Temperature Gas-cooled Reactor) technology basis, and to conduct various innovative basic researches on high-temperature technologies such as ceramics and fusion reactor materials. The HTTR consists of a core of 30 MWt, a main cooling circuit, an auxiliary cooling circuit and related components. The reactor pressure vessel is 13.2m high and 5.5m in diameter and contains the core, graphite reflectors, metallic core support structure and radial restraining devices.

]A E R (・ M 88・221

The D-T neutrons are generated via 3T(d.n)句1Iereaction.

There are two beam 11nes; one is so called 00 11ne for higト

current operatlon， and th巴 otheris 50 called 800 1 ine for

rather low current opcration. The major specifications of

the FNS acce]erator nre shown in the following Table.

• Beam current • Beam size • Pulse width • Frequency • Peak current • Target assembly

. Amount of 3T • Neutron Yield

< Table;>

> 20 mA < 1S mmゆ

Rotating (Water cooled)

< 1000 Ci 4 x 1012

3mA < 15問中< 2 ns

2 MHz 40 mA

Stationally (Water cooled. Air cooled)

10 Ci 5x l011js

The major experimental subjects are as follows:

1)τritium production rate 1n the various blanket conf1gura-

2) Nuclear heating rate in the structural materials

3) Shielding performance for D-T neutrons in the various

structure configurations

4) lnduced effects on the structural materials by D-T neutrons

HCLWR 旦19h旦onversion~ight Water旦eactor

HTTR 旦19h1e町 eratureengineering Test旦eactor

JAERI is to construct and operate the HTTR. to carry out the

necessary R & D for establishing and upgrading the HTGR (Hight

Temperature Gas-cooled Reactor) technology basis， and to con-

duct various innovative bas1c researches on high-temperature

technologies such as ceramics and fusion reactor materials.

The HTTR consists of a core of 30 MWt， a ma1n cooling circuit，

an auxiliary cooling circuit and related components. The

reactor pressure vessel is 13.2m high and 5.5m in diameter

and contains the core， graphite reflectors， metallic core

support structure and radial restraining devices.

J A E R I - M 8 8 - 2 2 1

< Table> Specification of HTTR

Thermal power 30 MW Outlet coolant temperature 850°C/950°C Inlet coolant temperature 395°C Fuel Low enriched UO2 Fuel element type Prismatic block Direction of coolant flow Downward-flow Pressure vessel Steel Number of main cooling loop 1 Heat removal IHX and PWR(parallel loaded) Primary coolant pressure 4 MPa Containment type Steel containment Plant lifetime 20 years

JENDL-3T : A temporary file of Japanese Evaluated Nuclear Data Library, version 3 released in 1988.

JENDL-3PR1 : A preliminary version of JENDL-3 including eight nuclei 6Li, 7Li, 9Be, 1 2 C , 1 6 D , Cr, Fe and Ni compiled until the end of 1983.

JENDL-3PR2 : In JENDL-3PR1, a part of the data for 6Li, 7Li and 1 2 C were revised on March 1985.

MOX : Plutonium-Uranium mixed oxide formerly written by Pu02-U02

TRU : Trans-Uranium

VHTRC : \ ? e r v High 2 e m P e r a t u r e Reactor Critical assembly

1. Purpose Neutronics design verification of Very High Temperature Reactor

2. Main features of VHTRC Type : Split table type of hexagonal graphite block

structure

JENDL-3T

JENDL-3PRl

JENDL-3PR2

JAERI-M 88・221

<: Table:> Specification of HTTR

Thermal power 130 MW

Outlet coolant temperature 1850oC/950oC

1nlet coolant temperature 13950C

Fuel ILow enriched U02

Fuel element type IPrismatic block

Direction of coolant flow IDownward-flow

Pressure vessel ISteel

Number uf main cooling loopll

Heat removal IIHX and PWR(parallel loaded)

Primary coolant pressure 14 MPa

Containment type ISteel contajnment

Plant lifetime 120 years

A temporary file of Japanesp. Evaluated Nuclear Data

Library， version 3 released in 1988.

A preliminary version of JENDL-3 including eight nuclei

6Li， 7Li， 9Be， 12C， 16D， Cr， Fe and Ni compiled until the

end of 1983.

6.， 7. _ _ ~ 12 1n JENDL-3PR1， a part of the data for oL1， 'L1 and ~'C

were revised on March 1985.

MOX Plutoniun.-Uranium mixed oxide formerly written by PU02-U02

TRU Trans-Uranium

VHTRC !ery旦ighlemperature !eactor Eritical assembly

1. Purpose

Neutronics design verification of Very High Tempera-

ture Reactor

2. Main features of VHTRC

Type Split table type of hexagonal graphite block

structure

J A E R I - M 88-221

Size : Height ; 2.4 m Length ; 2.4 m (Each half)

Fuel : Coated particle fuel compact (Pin in block) <Size> 36! 18mm outer, inner diameters, 36mm height <Coated particle> 960pm overall diameter (600 m UO kkernel coated with pyrolitic carbon and SiC) < Enrichment> 2, 4 and 6 %

Moderator and reflector : Ciaphite Core temperature : Room temperature to 210°C by

electric heaters Maximum power : 10 W Auxiliary equipments :

(1) Sample heating device (Up to 800°C) (2) Pulsed neutron source

3. Main iteirj of experiments (1) Critical mass (2) Reactivity worth of control and burnable poison

rods (3) Power distribution (4) Temperature coefficient of reactivity of the

whole core up to 200°C (5) Doppler effect of a sample fuel rod up to 810°C (6) Cell spectram indices

)AER 1・M 88幽 221

Size Height 2.4 m

Length 2.4 m (Each half)

Fuel Coated particle fuel compact (Pin in block)

くSize> 36. 18mm outer. inner dial1'.w ters •

36mm height

くCoatedpartic1e> 960いmoverall diameter

(600 m UO kkernel coated with pyrolitic

carbon and SiC)

くEnrichment> 2. 4 and 6χ

Moderator and reflector G~aphite

Core temperature Room temperature to 2100C by

electric heaters

Maximum power 10 W

Auxiliary equipments

(1) Sample heating device (Up to 80QoC)

(2) Pulsed neutron source

3. Main ite~3 of experiments

(1) Critical mass

(2) Reactivity worth of control and burnable poison

(3) Power distribution

(4) Temperature coefficient of reactivity of the

whole core up to 2000C

(5) Doppler effect of a sample fuel rod up to 810Q

(6) Cell spectram indices

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