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This report is issued irregularly.Inquiries about availability of the reports should be addressed to Research Information

Division, Department of Intellectual Resources, Japan Atomic Energy Research Institute,

Tokai-mura, Naka gun, Ibaraki-ken, 319-11, Japan.

© Japan Atomic Energy Research Institute, 1997

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A Study of Bumup Analysis Method for Irradiation Test of Rock-like Plutonium Fuel Sample Performed at JRR-3

Yoshihiro NAKANO , Hideki TAKANO and Hiroshi AKIE

Excess Plutonium Disposition Fuel Research and Development Team

Tokai Research Establishment

Japan Atomic Energy Research Institute

Tokai-mura, Naka-gun, Ibaraki-ken

(Received June 17,1997)

A study of bumup analysis method for the irradiation test of Rock-like Plutonium Fuel sample

has been performed. The samples have been irradiated in the irradiation hole in the beryllium

reflector of JRR-3. Computer codes used for the analysis are SRAC, ORIGEN2 and MVP. In the

calculation using the SRAC code, fine energy group neutron spectrum of the beryllium reflector is

calculated. Bumup calculation of the irradiation capsule is done using the spectrum as a boundary

source into the capsule. In the bumup calculation for the capsule, two kinds of simple and detailed

models are considered and the results calculated are compared and discussed. Some calculations

are also performed with the ORIGEN2 code and the results are compared with those calculated by

SRAC. The accuracy of the results calculated with the SRAC code is examined as comparing with

those obtained rigorously by the continuous energy Monte Cairo code MVP.

Keywords: Rock-like Plutonium Fuel, Irradiation Test, JRR-3, Bumup Analysis, SRAC .ORIGEN2

MVP, Beryllium Reflector , Neutron Spectrum , Boundary Source


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2.2 SRAC3- 7

2.3 ORIGEN 2 K ££)$?#? ......................................................................................................... 8

3. 14

3.1 SRACffiMJ&fr;UiORIGEN2 (PWR^-f7'7'j) i©tb^ ................................................ 14

3.2 SRAtiES^A^ORIGBN 2 (SRAC5 -Y 'J) tc J; £j$BK©StSSSS .................. 15

3.3 SRACffiMJ&Tf7U<hORIGEN2 (SMCy'Tf? U) Kt; £Xe£j&S©!+S*£H ................. 16

4. ## t ' /cSRAC I: Z 6 # W#f .............................................................................. 20

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5. MVPI:t: 6ltStf..................................................................................................... 23£t#> ............................................................................................................................................... 25

m # ............................................................................................................................................... 26

......................................................................................................................................................... 26

Appendix-A 7? V ........................................ 27Appendix-B 7 gf ...................... 28

iii


Contents

Introduction............................................................................................................................................................ 1

1. Brief Description of Irradiation Test............................................................................................................1

1.1 Irradiation Sample......................................................................................................................................1

1.2 Irradiation Condition................................................................................................................................. 3

2. Analysis of the Irradiation Test....................................................................................................................... 7

2.1 Calculation Code and Neutron Cross Section Library...................................................................... 7

2.2 Calculation with the SRAC Code...........................................................................................................7

2.3 Calculation with the ORIGEN2 Code...................................................................................................8

3. Comparison of the Calculated Results...................................................................................................... 14

3.1 SRAC with Simple Model and ORIGEN2(PWR Library)................................................................14

3.2 Bumup between SRAC with Simple Model and ORIGEN2(SRAC Library)............................. 15

3.3 Xenon Production between SRAC with Simple Model and ORIGEN2(SRAC Library).......... 16

4. Bumup Analysis with the SRAC Code Using Detailed Calculation Model.....................................20

4.1 Description of the Detailed Calculation Model.................................................................................. 20

4.2 Results of the Detailed Calculation..................................................................................................... 20

5. Examination of SRAC Calculation by the MVP Code...........................................................................23

Conclusions...........................................................................................................................................................25

Acknowledgments................................................................................................................................................26

References............................................................................................................................................................. 26

Appendix-A Effect of Neutron Cross Section Library on the Bumup Calculation..............................27

Appendix-B Measured Value of Fluence Monitor and Normalization of Thermal Power

at Calculation............................................................................................................................... 28

IV


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Isotopes Isotopic Abundance

t‘is 21* <. 0.U'

Pu-239 94.33

i'u- MU / ' EB:?

Pu-241 0.40

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Table 1-2 Composition of Plutonium Sample Disks (mol %)

Zirconia Type Thoria Type

HuO' iiiiiiiiie;; 10

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A1203 65 65

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Table 1-3 Weight of Plutonium Sample Disks (mg)

Pu Sample Type Sample Case Stage Disk Weight Pu02 Weight

1st 179.3 (5 disks) 35.47 (5 disks)

2nd 153.5 (5 disks) 30.37 (5 disks)Thoria Type

3rd 160.2 (5 disks) 31.69 (5 disks)

Sub Total 493.0 (15 disks) 97.51 (15 disks)

1st 147.4 (5 disks) 33.95 (5 disks)

2nd 153.8 (5 disks) 35.42 (5 disks)Zirconia Type

3rd 160.8 (5 disks) 37.03 (5 disks)

Sub Total 462.0 (15 disks) 106.40 (15 disks)

Total 955.0 203.91

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Sample Case Stage Temperature

1st 710

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- 9 -


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Fig. 2-4 Unit Cell Calculation Model of Follower Fuel Plate

— 10 —


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Heat Transferring Medium

Fig. 2-7 One Dimension Calculation Modeling of Irradiation Capsule (Outside the Sample Case)

JAER

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97-051

Fig. 2-8 One Dimension Calculation Modeling of Irradiation Capsule (Inside the Sample Case)

JAER

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97-051


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Table 3-1 Bumup of Pu Sample Calculated with SRAC and ORIGEN2

Pu Sample Code Pu-239 % MWD

Thoria Type - 2 SRAC 35.4 6.15E-3

(lOOOC) ORIGEN2 20.0 3.23E-3

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Table 3-2 Pu-239 Microscopic Cross Section of SRAC and ORIGEN2 (barn)

Sample Code Capture Fission N2N

Thoria - 2 ORIGEN2 5.861E+01 1.062E+02 1.120E-03

(lOOOC) SRAC 9.691E+01 2.091E+02 1.077E-03

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— 14 —


Table 3-3 Fission Release Energy (MeV)

Nuclide SRAC ORIGEN2

Pu-239 %i im 210.63

Pu-241 213.46 211.37

Th-232 194.26 19&42

U-233 200.32 200.98

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Table 3-4 Bumup of Pu Sample Disks Calculated with SRAC and ORIGEN2

Using Pu-239 Microscopic Cross Section of SRAC Simple Model

Pu Sample Code Pu-239 % MWD

Thom Type - i ■ X/Xs k ill!!

<6oot:; OklGENY wsms^mhi^biiiihb

Thoria Type - 2 SRAC 35.4 6.15E-3

(10000) ORIGEN2 33.9 5.82E-3

Thoria Type - ?■ 'WA< :.Y: 3.59B-3

(8ooO^ &33W

Zirconia Type - 1 SRAC 29.1 6.19E-3

(600U) ORIGEN2 28.1 5.90E-3

Zirconia Type- - 2 SKAf '4-i &95B-3

aoooO; URKiEN- 1111111^(1

Zirconia Type - 3 SRAC 26.6 5.17E-3

(800U) ORIGEN2 25.8 4.96E-3

— 15 —


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Table 3-5 Calculated Value of Xe Production and Collected Value at Puncture Test

(Thoria Type - 1)

Pu Sample

Thoria Type -1 (600%:)

Items SRAC ORIGEN2 Collected

Xe Gas Volume (cc) (Release Rate)

0.154(75.3 %)

0.137(84.7 %)

0.116

Xe-131 12.6 13.5 11.9

IsotopicAbundance

Xe-132 18.4 19.4 18.4

(%) Xe-134 25.8 26.8 26.0

Xe-136 43.3 40.4 43.7

— 16 —


Table 3-6 Calculated Value of Xe Production and Collected Value at Puncture Test(Thoria Type - 2)

Pu Sample Items SRAC ORIGEN2 Collected

Xe Gas Volume (cc) 0.174 0.1540.145

(Release Rate) (83.3 % ) (94.2 %)

Xe-131 12.5 13.3 11.3Thoria Type - 2

(lOOOC) IsotopicAbundance

Xe-132 18.3 19.2 18.5

(%) Xe-134 25.6 26.5 25.1

Xe-136 43.6 41.0 45.1


(Thoria Type - 3)



0.157(2.5 %)

0.139(2.9 %)

0.004

Thoria Type - 3 (800C) Isotopic

Abundance

Xe-131

Xe-132

12.7

18.4

13.6

19.5

12.0

19.4

(%) Xe-134 25.9 27.0 25.7

Xe-136 43.0 39.9 42.9


(Zirconia Type - 1)


Xe Gas Volume (cc) 0.174 0.1550.157

(Release Rate) (90.2 %) (101.3 %)

Xe-131 12.7 13.5 11.7Zirconia Type - 1

(6000 IsotopicAbundance

Xe-132 18.4 19.4 18.4

(%) Xe-134 25.8 26.8 25.4

Xe-136 43.2 40.4 44.5

17 —



(Zirconia Type - 2)


Xe Gas Volume (cc) 0.197 0.1750.146

(Release Rate) (74.1 %) (83.4 %)

Xe-131 12.5 13.3 11.1Zirconia Type - 2

(10000 IsotopicAbundance

Xe-132 18.3 19.2 18.4

(%) Xe-134 25.7 26.5 25.4

Xe-136 43.5 41.0 45.0


(Zirconia Type - 3)



0.145(77.2 %)

0.129(86.8 %)

0.112

Xe-131 12.7 13.6 12.2Zirconia Type - 3

(800O Isotopic Xe-132 18.4 19.5 19.1Abundance

(%) Xe-134 25.9 27.0 25.1

Xe-136 43.0 39.9 43.6

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2.0E-1

1.5E-1

1.0E-1

5.0E-2 -

""l ........ "'I......“I ■ 1■1""0.0E+01.0E-5 1.0E-4 1.0E-3 1.0E-2 1.0E-1 1.0E+0 1.0E+1 1.0E+2 1.0E+3 1.0E+4 1.0E+5 1.0E+6 1.0E+7

Neutron Energy (eV)

Fig. 3-1 Neutron Energy Spectra inside the Irradiation Capsule (2nd Stage Thoria Type Plutonium Sample) Calculated with SRAC Simple Model.

JAER

I-Research

97-051


4. 1 pmFMC«£5!tSPF^

T&z>o wm,5ii, srac-put

fi, >n 7 K £ gag L £ &M T <D Its M-Zhb(SRAC-PU<7)Syt t y '> a > 11#) o f C?, ^<7)Wf/'>3 h VTMmk

i/A/ri-TMtfcEk met*, 7in'/xt-^ (Cor7'f-V) trover, f fi^(^f&gM#$rlE

^lt#£2EB> 88rosiooox:<7)K#eov>tffof:0 lt#f-7;i/trig.4-1e

SC, |tS-e«to*MS^&^)WmLô ft

7;v^>t.*~? (Co) <%#>^mitim#^^ #m L %e-

m-kz*) eitW^Bititfî^L-cwco *<7)'&<7)m%.r~?<D%&uz£t)y

mfemtztvrwzB^tt^-m^sÊL<*<, tBij^r-tfrbiEme#t, tiaMgti, ■7)Vx.y**:~?-ETSe«t &ttttWct. #Lv### mTit, stRe, â^<it#e%of -

-e, Its-e#lbtL^7;v-x>7^ (Co-59) ©iilWIttU 6 j: 9 eit#T<7)mmêm%L^ (Appendix-B

#m o

4. 2 j:^gtm^m

R8&M&& iooot oszEBKêov^-c, Sterne Table

4-ie^-to h v r^-eti, #m%^5i.4%^^ i), T;i/(7)iKn.5#^ ô-cvô Pu-239<D < ? n»fffi«e(iti X «<f ttTJfrW&l.S fettt'DX&V), Z.(Dtztb e«Jt<7)ltSE^* £ < M L tzo 9&±#

LTit,

(1) t VT^K#^ 9^3-7M##oit#e#ê#v\ Hn:)ttifW^llSLfeto

(2) iBiJStIts t <Dmi\At MSTti7;i/x:'y7^-^m<0S-e$)

(3) 7;vo;y7^-^<7)i|iJSS ‘̂bE$l'e^v^BtfI14W^StitiL, iti£*> heLTIt#*e

(0, 5A^#A^tb6o

(l)eov^T(i, ##^-771/(7)#^ &M#^7^(0#^^-CPu-239(0 < ? □ Bmi £ JtEf &

^r;v®icj3it^®rMS(7)Sit'y;V3-7mTv^< bfr%bti2> t

4.

- 20 -


<b Wes

9.8E+13 (n/cra2 s)£ 1.0E+14 (n/cra2 s)T\ ^Sv^ti^V>0 (2)<D$f%\Z

(3)i:oV'TT3b4^ 7 7 v ?X^-?<7)Co-59<T)RJ&^-eBt#ffihiBl]S1ilh %

ctcz i),

'f;t/T(i#1.5#(:&cTL$cTW&:: LTOTLTV'fc7.0E+13 (n/cm2s)kV^B4»tt7-T£of ^(3)<7)^gtfUHHtJz<7)fflM'?:$>'2fzt&t>ti2>0

Table 4-1 Comparison of Calculated Values between Detailed and Simple Models

Region ItemDetailed Simple (Detailed)

Calculation Calculation /(Simple)

Bumup (Pu-239 %) 51.41 % 35.40 % 1.45

Thoria Type - 2 Total Flux 2.492E+14 1.564E+14 1.59Pu Sample (lOOGC) Pu-239 <7a (b) 304.27 306.05 0.99

Pu-239 Of(b) 208.64 209.14 1.00

Bumup (Pu-239 %) 45.54 % 34.37% 1.32

Zirconia Type - 2 Total Flux 2.278E+14 1.541E+14 1.48Pu Sample (lOOOC) Pu-239 CTa (b) 271.20 289.97 0.94

Pu-239 Cf(b) 185.13 197.76 0.94

Total Flux 2.114E+14Fluence Monitor(Co-59) Thermal Flux 9.793E+13

Total Flux 2.186E+14

1.423E+14(Thoria)

1.54

Sample Case

1.427E+14(Zirconia)

1.53

(Mo)

Thermal Flux 1.030E+14

7.000E+13(Thoria)

1.47

7.000E+13(Zirconia)

1.47

Bumup: Values after 4 cycles irradiation

Other Items: Values before irradiationUnit of Neutron Flux: (n/cm2 s)

- 21 -

ThoriaType Pu Disk

Pellet Case

Sample Case

Fluence Monitor (Co)Pt Case and Fuel Pin Tube

Zirconia Type Pu Disk

Fig. 4-1 Detailed Calculation Model of 2nd Stage Irradiation Capsule (inside the sample case)

J A

ERI-R

esearch 97-051


5. MVPlCcfcSsi-W^^K

MVP3-KfcJflV'T, SRAC^ff&ir^T^Tffl V' t, M"* -v •/* ;V

tl/2Pu-239<7) 5 X £c 1 ^TCM T7I/1 i SRAC t ^ < l@| —6 (Figs.

2-7, 2-8 £#E) o 3^7Clt#JB’7,>’l/ SrFig. 5-li2> :t##^$"Table o :t#kX b V #(i> V'-f tL(7)|tS^-X =1)40,000,00011 X MJTfcSo

Table 5-1 Pu-239 Microscopic Cross Section Calculated with MVP (barn)Cal. Case Total Nu*Fission Fission Elastic SC Capture

ID 'ihur,.> ;300K

3.2893 id+02(1251%)

6.43715d+02(1.286%)

'2 >4/,,= - '#00##0##0

\9610i-viUt; ' .uv.r.--v(1 336%l

ID Thoria-2900K

3.32136d+02(1.307%)

6.48970d+02(1.332%)

2.25272d+02(1.333%)

7.86361d+00(0.875%)

9.84385d+01(1.415%)

;D Thou.3(X)K

3.67949d+02 7.20467d+02(1.177%)

2.50 iiSd+itt(1.178%)

vijj.iju-tUU(0.569%)

GW/a-H)::(1,297%)

3D Thoria-2900K

3.77715d+02(1.223%)

7.37114d+02(1.242%)

2.55917d+02(1.242%)

8.14265d+00(0.601%)

1.13179d+02(1.328%)

ttsmmooK^ b v 7%nn<vi>k7t*r^t3>xjt;er)i'<7)Totemmffi*$t®-fz>t. 1rjv(outfmio%/hl^Tc^nvTii,

$msi i9

x^btz*>k.%7Lbtib0 z<DZtfrt>, SRAC-e(7)###0@t#^(±, HE<7)S^EWii

/J'fNffi LtV'4 6 oPuK#m&Kli, 300K h 900K02^ - vUC-O V'TStS£ o £^

9H. l^TC^T^K 3'<k7t*r)VjtK~k£ < &W„ Pu-239<0«t*gfl&i&*94%fcft<, K

- 23


45.0 (mm)

Be Reflector

Outer Pipe Reflection Pipe Heat Transfer Medium

(a) Outside the Sample Case

27.26Fuel Pin Tube ■

Pellet Case

Pellet Case y^^Pt Case

y" Fuel Pin Tube

(b) Inside the Sample Case(c) Vertical Cross Section of the Sample Case

Fig. 5-1 3-D Calculation model of 2nd Stage Irradiation Capsule for MVP

— 24 —


*£#>

JRR-3~CM<D$ft>titz h V TMd3 t 3 - o

SRAC3- K*Cli, 6(7)4:%?-A#m^mV'th%?-####*?'> 3 'simmz&RLxfmiffvtio

MA##^^6JRR-3^V V WWi, Jz*fr3r-72&<7)2^7ci£ffc&k X h jRR-3'J^'L'ft^:£frv'jfcfc tz«,

pu-239^5 M%*rn'twm*r)wffltz±£%mû%fr-otz0

tzfz U f@i,It#*T7W::is^Tj±,^ - <7) $J * £ Simc BS L T # ^ ft /ztiTe ti & ri' o £ * tz, i ^

fiNffllt# ^weti, 7)1/3: t:MLT#ibft£, 7 L

#tr h v Tômtzov^Tii, It#* r;v fre 5 ? n EMEtc lit A, t*m* %Its*r* > -e *> +5*&lt#Emmbtiz>z.ttf¥

MZtiZo

0RIGEN23" KtZ.t&ftS'eift 3- KtzSB*^S£ ft* V'£pWRffi<7) 7/f T'y V L^#^t:(i, Pu-239<7)5 Cftli.J®#t Wt>ftZzJRR-3omm^V 'J PWR^'^-L'X^^ Mui *) o

SRAC-eSfS L £Pu-239 <7)5 7n EWE £ L T 0RIGEN2t: i 6 If# £fr o t: ti xsrac t mm<n9m%#*m t>ft*o

MVP3- Ktzj: tu SRAC<7)#t#t T)l/(7)^^%(7)#E^1f-3 ô f <7)ig^ 1 #DC*-77W7) Pu-239 5 ? n EWE W£%* t' )V i: D & lo%@$/h $^ShôTv^^h t5?W -o tz0 Ztia, l^l

1^7Ctf)l/Tli2<7)%^$r#g-e# ^V^*h#x_^ft6o CcO^d'Wft SRACt:i6#mm<7)@+#mmii, nmvmTzm*^<â/MF*uv^

— 25 —


m »

<, ztztbZMu

SRAC3- K<ateJflfc*fco-CIJ,a & tc, 3- K(7)##%g&#tffc TWz/f iiUo 0RIGEN23- fcfcfc o

Tt±, ^ - muimm, ^mu-f-A •^ihtzT FsU **V'fc£§* LtZo MVP3- K«C^LTIi, • fiSSS

S'i&ffi&fcffcoTV'fcrtf#* Lfco

^<7)ifeki>< <D1] + frb*r%m*\'tztz%tLtz0 ^$$(iS<SSWcL^t-o

##£6*

1) H. Akie, et. al„ " A NEW FUEL MATERIAL FOR ONCE-THROUGH WEAPONS PLUTONIUM

BURNING ".Nuclear Technology, vol. 107, p.l82(1994).

2) mt, m, " srar/W' -•> am • mmm$t<nmkn, [1997$#^EhSSII^#G45, B$mfll##(1997).

3) lljffl, #, " JRR-3M?(DBB##it#", Fl997$#^$^J EtSSlI^#

G49, B#:Mf^#R(1997).

4) #, "SRAC95;M^lt#3-KvXf A", JAERI-Data/Code 96-015(1996).

5) A. G. Groff, "A User's Manual for the ORIGEN2 Computer Code", ORNL/TM-7175(1980).

6) s*ie, #, " Mvp/cMvp ^ ^ff>T*7Vnn- K", JAERI-Data/Code 94-007(1994).

7) " BRF-24H* ^ "/-tr/l/ F/MESM^:", %#(1996).

8) $>m, #, " f 3-) Tl997^#^

EhSSIIMGSI, B$mf^#^(1997).

9) (Ed.) K. SHIBATA, et. al., ” CURVES AND TABLES OF NEUTRON CROSS SECTIONS IN

JENDL-3.2", JAERI-Data/Code 97-003(1997).

10) ?im, #, " ^ss-z/i/ h --> a-b#(7)hbWM", A## ri997##</)#&j

EhSIIIPWGSO, B#:mf^#&(1997).

- 26 -


Appendix-A y4 7y V & k 3 #k(:3 L' ?

SRACC Z 19 , JENDL-3.2, JENDL-3.1, JENDL-2£$V'TPu-23901#ff®f £3c«!>TJfc$ t tz0 ifz, ZVMWffiZMMIX, ORIGEN2H X.bl^rA 'p Wco^tDBMtSSrfivx #%

Table A-l y 4 7y') il £

LibrarySRACT$36 <m 0RIGEN2T<7)#%%

(1 74 ^^,Pu-239%)Capture Fission N2N

JENDL-3.2 9.691E+01 2.091E+02 1.077E-03 9.86

JENDL-3.1 9.698E+01 2.090E+02 1.139E-03 9.86

JENDL-2 9.642E+01 2.179E+02 3.432E-03 10.10

JENDL-3.2 kJENDL-3. IT (i(i <k k> <k,H(i&V>0 t tz, JENDL-2(i, 7 4

y? V Z 0 Ztilzk iotv^0

27


Appendix-B

PuK#0m^^f 6 g%T.muk, x 'o'jffrtitz1)0 7nsx-yx*-?<Dtfnn. m

o >IO(B-1)jU2> 4^4" f h ^r## L,&IZ£Z7JUX.>A^~?1%M (fit X.ifCo-59) 0##&#%f6 Z. 2: 1: 2 o T#&fL&o

<b = — x — Nn a„

1 1 X---- X —X

Ka e-Yv

xA • frn , y 1 ~ GXp(-A • ij),

l-exp(-Alm) # exp (A-fe),(B-l)

::t, ±5t£S2m>^-*41t2l£B$tl£:^;l/b*-?^TIj:*X 59Co (n,y) 60CoKS ■C4. £ h 60Cov6$, «15.27#T%m ttl. 173Me VO yiCoi'T IFffif & 0

: coom^-m 2.53E+15

<?. : Co-590Mlff®S 37 (barn)

a : Co^OCo-590#%)t 1.0

* : 1.173MeV07BSt8W 52.1 (cps)

e : 1.173MeVO 7 (IISSES 200 mm) 2.7E-4

yr : Co-6001.173MeVO 7 0.999

A : Co-600 4.1698E-9 (s ')

L : immimmm 300 (s)

A : Table B-l

4 : Table B-l

#"4-^ f IV "CO MM I, ^ <tiPB#r5 tc £Table B-lK/J<to

Table B-l &"*M ^ JVO 7;Vi>7^-^ t

No. of Irradiation Cycle( 1=)

Irradiation Time (s)

hCooling Time (s)

*cyi 1 - exp(-A • 1,); m exp(Afc)z

2165400 54862080 7.15 IE-3

2 2152080 40346880 7.550E-3

1906200 57322880 6.776B-3

4 2165400 34298880 7.791E-3

- 28 -


it:, Co-600#m%CMf.K+MV'#"?, (B-l)âS6^ti,

A L

l-exp(-Arm)= 1.0

h&&„

(f> = 7.07E + 13 (n/cm2s)

## y SBfIMiJSmtix - O7.07E+13riWttTi£Oi&Jîl t 2 fl, SRACftlrTztiu, t l

20fc*K, Co-590#%M#(rj:<). 0 $l^#^)0RTMv^fLT#^:##'e0#037barnt#MLTV'6«,

LfrLtctfb, ÊOjKWTIi, n/s’;v b*-?o* K5it«)E$ $U^$tLtzymtmmfrbmM^-s-frzmmtzz. 20*0, 4-Eoigijs^m*

f & 2 a #. 2 2 $TOMr^*^:7.07E+13h JElv^tfiIEL< (i,

StMbo3n££BL£0 (B-l)ôWmi:. CoOM^H&iV CoÔCo-590#%jta$-#lf6 ^(B-2)^:^^ 0, 2 tlii >

Co-590 KE# £ t%Z> o

(/)N0aaa =K

e-Yyx L t y 1 ~ exp(-A • r,);

l-exp(-Afm) tf exp(A-fc),(B-2)

2 0RS^(i> 6fL^: y m#+m$L < MLT#6ft6,

£/iu y mstmômm hco-eoo^m^ômwaomm, ^ z

^(B-i)^#mo|#(:Mvâm(:Z3mm^Â/'ro6« MStS^r^-eti, 7)Vx.yx*fo##&mif;52,b&<#mwt:%')2*&o ^2-e, m-#oiî:#^fi&7

)Vx.y7*~9 mi V'l ,

mm^r)v\zxmnît, v*~? (co-59) o#%#agm#m+#:-c& ^.

- 29 -


Fig. (Co-59) O72#O!m##<0#%#4XBf@#2:,h ;v5r^-t0 ,I<7)72$p8fr®fl£ =*/<;!/ b^E- h

<7flitoto/ = 1.236E + 2 (barn)

ifc, ## (0.602 eV^T) COV'-C<%)«&##-*'&

<7a.,* = 2.539E +1 (barn)

btuho c:tui> SMv^TWcS (=37 barn) (Zjt^T 9 /b S < * o TV'* 0 £*Ui,

co-59<%^#mmm#^ b^ r v>

fibo z.<nz.bfrbi>, f 4̂:& *<d

r-zrt iziEEtzits 5 KtmmmK&emf a c t c t timmt®,

— 30 -

Mic

rosc

opic

Abs

orpt

ion

Cro

ss S

ectio

n (b

am)

Co-59 Microscopic Absorption Cross Section Calculated by SRAC-PIJ with Fine Geometory (IGT=11), 2nd Case (temp=1000 C),

Temp of Cal: Co-59=1140K , Pu=1270K Neutron Spectra of Co-59 Flux Monitor Position

------ Co-59 Micro Abs. XS Flux

1.0E+01 .OE+3

1.0E-11 .OE+2

1.0E-21.0E+1

1.0E-31.0E+0

1.0E-41.0E-1

1.0E-51.0E-2

1.0E-61.0E-3

1.0E-71.0E-41.0E-5 1.0E-4 1.0E-3 1.0E-2 1.0E-1 1.0E+0 1.0E+1 1 .OE+2 1 .OE+3 1.0E+4 1.0E+5 1.0E+6 1.0E+7

Neutron Energy (eV)

Fig. B-l Microscopic Absorption Cross Section and Neutron Energy Spectrum of Cobalt Fluence Monitor

Neu

tron F

lux

per L

ethe

rgy

(arb

itrar

y un

it)

(SI) 6

*1ft g * se ^ g w 12 #

6 5 y - h /!/ m ft, B$, 0 min, h, d

H e + o g 7 A kg It, #, »

«S r=i 8> s V y h 1, L

e m 7 y *< T A b y t7 /u t' y K # T- f & v eV

% « a t tv mol uit $ •h y t 7 cd

¥ B ft 7 -^7 y rad 1 eV = 1.60218x 10""J

A f* ft xr 7 i?T y sr 1 u= 1.66054 x 10"" kg

*3 @W©8#5rt>-3SIfflMto

a g » K<f(6© SI $(6 it£z&m

@ m St 'X /V y Hz s'1 g * 12 ft

ti •=■ a - b y N m-kg/s2 *y?'Xln-i Ae -n , £ % y>" X A Pa N/m2 y>' - y bi7<e-, tt$, me ^ - J N-m v* — vt/ barX $ . tt w sfc 7 y 5 W JZs ft to Galtit. a # ^ — o y C A-s * yi U - Cise, «e. esTi jK /u h V W/A u- y h ft y R# * s e 7 t 7 K F C/V 7 K rada m ffi 6t A - A n V/A y a rem3 y g 9 9 y x V - * y X s A/V« $ 0 JL — Wb Vs 1 A- 0.1 nm-10 10 me m « ® X X 7 T Wb/m2 1 b=100 fm2=10'2" m24 y g ? f y % 'x y U H Wb/A

1 bar=0.1 MPa= 10‘Pa-b yu y A 3. fi ft -b y|y y 7 x® "C

it $ yu — y y lm cd - sr 1 Gal=l cm/s2 == 10~2m/s2

m ® yi/ 7 X lx lm/m2 1 Ci=3.7x]0"Bq

tt w *6 X i |y ;u Bq s"1 1 R=2.58xl0"*C/kg

R K m e g lx 4 Gy J/kg 1 rad = 1 cGy = 10" Gy

« e s e •y - ^ tv h Sv J/kg 1 rem= 1 cSv= 10 zSv

*s simmsfcSt mails 12 4

10" x 7 A- E10" -i 9 p10'2 T 7 T10' +' ft G10' y ft M10' * a k102 XX 7 h h10' r A da

10"1 T •> d10"2 t y f c10"' 5 V m10"' 7 4 7 D M10"’ a y n10"'2 e 3 p10"“ 7 i X 5 f10"" 7 h a

(E)

1. * i - 5 ii rsie*<i»j $5i$, aiemamm iggs^fiiiicAt. levfcACf 1 u©«tli CODATA © 1986^JtS?

illcJ: r>tz„

2. *4 IC|ii6E. y y h, T-/U,

ETi'5»mS®9ito^©x5<:CTliWiSLfco

3. bar li, JISTIi**©E^t#4yfia

AKRO# 2 ©A f 3 V -IctfSShTV 4„

4. ECI»flSfa*^tt^TIibar, barntA

U* FlfilE®W(4j mmHg £r8 2 ©A A A 9

-icAnri'^o

* * *

ti N(=10'dyn) kgf Ibf E MPa( = 10 bar) kgf/cm2 atm mmHg(Torr) lbf/in2(psi)

1 0.101972 0.224809 1 10.1972 9.86923 7.50062 x 10' 145.038

9.80665 1 2.20462 A 0.0980665 1 0.967841 735.559 14.2233

4.44822 0.453592 1 0.101325 1.03323 1 760 14.6959

ti ® 1 Pa-s(N-s/m2)= 10P(tf 7 X)(g/(cm-s)) 1.33322 x 10"' 1.35951 x 10"' 1.31579 x 10"' 1 1.93368 x 10"1

ftti® 1 m2/s= 10'St( x h - 7 x) (cm2/s) 6.89476 x 10"' 7.03070 x 10"2 6.80460 x I0"2 51.7149 1

J( = 10’ erg) kgf*m kW* h cal(lteE) Btu ft • Ibf eV

1 0.101972 2.77778x 10"2 0.238889 9.47813 x 10"* 0.737562 6.24150 x 10"

9.80665 1 2.72407 x 10"' 2.34270 9.29487 x 10"' 7.23301 6.12082x 10"

3.6 x 10' 3.67098 x 10s 1 8.59999 x 10' 3412.13 2.65522 x 10' 2.24694 x 102‘

4.18605 0.426858 1.16279 x 10"' 1 3.96759 x 10"' 3.08747 2.61272x 10"

1055.06 107.586 2.93072 x 10"' 252.042 1 778.172 6.58515 x 1021

1.35582 0.138255 3.76616 x 10"T 0.323890 1.28506 x 10 2 1 8.46233 x 10"

1.60218 x 10 " 1.63377 x 10"20 4.45050 x 10"2‘ 3.82743 x 10"21 1.51857 x 10"22 1.18171 x 10" 1

1 cal = 4.18605 J(lteE-)

= 4.184 J

= 4.1855 J (15 *C)

= 4.i868j(aiemmm)

aw ips= 75 kgf-m/s

= 735.499 W

Bq Ci 1 Gy rad 8 C/kg R «ey

Sv rem

1 2.70270 x 10 " 1 100 i 3876 i 100#6

3.7 x 10" 1a

0.01 1e

2.58 x 10" 1#

0.01 1

(86$F12fl26BSE)

JAERI-Research—97-051 - OSTI.GOV

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