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FIRSTINTERNATIONAL %ORKSHOP

SEVERE ACCIDENTSAND

THEIR CONSEQUENCES

TABLE OF CONTENTS

Participants

Summary

Overview on Nuclear Power in the USSR

A.J. Gagarinskii

Lesson from Chernobyl: As Viewed Before the Workshop.

J.M. West

Estimating the Temperature of Damaged Reactor Core Components at the Timeof the Explosion

P.A. Platonov, V. I.'apukhin,F.F. Zherdev, V.M. Ma:."kushev, V.M. Bespalov

Extended Analysis of First Phase of Unit IV Accident at Chernobyl NuclearPower Plant

E.O. Adamov, Yu.M. Cherkashov,L.N. Podlazov, A.I. Ionov

Estimation of gualitative Effects of Possible Disturbances in Chernobyl-4

E.O. Adamov, V.V. Vazinger, V.P. Vasilevsky, Yu.V. Mironov,Yu.M. Nikitin, L.N. Podlazov, Yu.M. Cherkashov

Simulation of Reactivity and Neutron Field Changes at Reactor Shutdown

N.N. Andrivshchenko, L.V. Bidoolya, A.C. Kalughin, A.N. Kuzmin

Analysis of Main Circulating Pumps (MCP) Behavior in the Preaccident Periodand the First Phase of the Chernobyl Accident

E.V. Buvlakov, V.A. Babin, L.R. Kevorkov, E.G. Navinskii, V.B. Proklov,A.A. Shpakovskii

(Chernobyl) IV Unit Bubbler Tank: Complex of Post Accident Investigations

M.K. Abdulakhatov, R.V. Arotonj'an, A.B. Blinov, K.P. Chocherov,L.V. Drapchinsky, 8. Ya. Galkin, A.P, Krinitsin, E.M. Pazukhin,

B.F. Petrov, L.A, Pleskachevsky, V.D. Popov

RBMK Fuel Cladding Behavior under LOCA Conditions

O.Yu. Novoselsky, I.I. Kryuchkov, Y. Ya. Abramov

Improvements of Control and Safety System for RBMK type Reactors

E.O. Adamov, Yu. M. Cherkashov, V.V. Kondrat'ev,S.G. Ukharov, R.R. Ionajtis

Operation of NPR with RBMK type Reactors and Measures to Improve Reliabilityand Safety

E.O. Adamov, V.P. Vasi levsky, A.A. Petrov, Yu. M. Chevkashov

R&D Programs for Reliability and Safety Improvement

O. Yu. Novosel'sky, B.A. Gabar aev, E.V. Burlakov, A.A. Kalugin,D.A. Mikhai lov

Neutron-Physical Characteristics of RBMK Type Reactors Before the Accidentand After Implementation of Measures for Increasing Safety

V.P. Borshchev, A.D. Zhirnov, V.D. Nikitin, I.A. Stendock,Yu . M., Cherkashov, E .V . Burlakov, A.V. Krayushki n,

V.S. Romanenko, Yu. A. Tishkin

Feasibi lity Study of Safety of Uranium Graphite RBMK type Reactors

E .O. Adamov, V .P . Vasi lousky, A .A. Petrov, Yu . M. Cherkashov

Theoretical Analysis and Numerical Modelling of Heat Transfer and FuelMigration in Underlying Soils and Constructive Elements of Nuclear Plantsduring and Accident Release from the Core.

R.V. Arutunjan, L.A. Bolshov, V.V. Vitukov. V.M. Golouiznin,A.M. Dykhne, V.P. Kiselev, S.V. Klementowa, I.E. Krayyushkin,A.V. Moskouchenko, V.P. Pisminnii, A.C. Popkov, S.Y. Chernov,

V.V. Chudanov, O.V. Khoruzhii, A.I. Yudin

Multidimensional Calculations support )wo Bang Theory

S. Baron

Lessons Lear. ed about PWR beyond Design Basis Accident

S. Baron

The Fear Evoked by Radiation or Radioactivity and its psycho-sociologicalConsequences: What can be done to improve this si .uation?

S. Pretre

Severe LOCA at Lucens Experimental Nuclear Power Plant in Switzerland onJanuary 21, 1969

J.P. Buclin

Chernobyl -4 Accident Analysis

E.E. Purvis, III, B.W. Spencer

Use of Relative Fission Product Release to Estimate the Fuel TemperatureAfter Chernobyl Accident

A.A. Rimski-Kovsakov

Accident Progression Analysis of Past Severe Accidents

D.J. McCloskey

Ecological and Genetic Consequences of the Chernobyl Atomic Power PlantAccident

V.E. Sokolov, I.N. Ryabov, I.A. Ryabtsev, F.A. Tikhomirov,V.A. Shevchonko, A. I. Taskaev

Cleanup After a Nuclear Accident

J.O. Snihs

The Use of Tradescanti a Clone 01 for Biological Dosimetry in the Region ofthe Chernobyl Nuclear Power Station

I.N. Ni lova, A.B. Semov, A.I. Taskayov, V.A. Shevchenko, N. I. Vavi lov

Strategy for Handling Contaminated Food: A Nordic Proposal

J.O. Snihs

General Recommendations on Intervention Levels and Strategies for Measurements

J.O. Snihs

Analysis of the Efficiency of DNA Repair in Plants for the Genetic Monitoringin the Chernobyl Vicinity

A.B. Syomov, Piti tsyma, S.A. Sergeeva

What is-And What is Not-in NUREG/CR-4214

Niel Wald

Evaluation of Genetic Damage Caused to the Population of Chernobyl Vicinityby Radiation Exposure

V.A. Shevchenko

Genetic Consequences of Chernobyl Accident for the Popul"tion of Gomel andGomel and Mogi lev Regions

G.I. Lazjuk

Techniques to reduce Acute and Chronic Population Doses After MajorRelease of Radionuclides

W. Burkart

Medical Response to Severe Accidents

J.C. Henot.

A Compromise Regarding the Three Different Cultural Responses to the'De Minimis" Question

S, Pretre

Genetic Diso '<).; i'; . ' '' 'u ', <..".<'i i<, the R<.qion u', ih« C' n<'iNuclea<

Y.A. Shevch Lo, l'i, . "n . '.-v.", t..::.flan<aiya, A.V. Chekhovich, B.V. le: tov

the Monitorinu of Seeds of <I.< a e >us Phytoc..nosis in the 30 Ki lomet 'r Zoneof the Cher <booby] Disaster

A.l. Taska:v, t .P. froloua, O.N. Popova, V.A. Shevchen..o

The Role cf'gricultu< al , nu Natur .1 Ecos" tern. in ti:e Internal DoseFormulati<.n in the l<iha!<'taunts o a Contaminated 2ot<e

M.I. Daionov, I.G. Tranikov

Cyto Genetic Effects of Ionizing Radiation in Crepis Tectorum PopulationsGrowing in a 30 Ki <ometer 2Jne of t <e Che; nibyl At <.:<ic Power St <tion

!.. I. Grinikh, Y. Y. She< chenko

Genetic Cor<sequences of Radioactive Contam~natios of Arat>idopsisPopulations

Y. l. At>ramou. 0. M. Iedorenko,V. A. Shevchenko

Photographs

RBMK Fuel Cladding Behavior under LOCk Conditions

0 .Yu . Novoselsky, I.I. Kryuchkov, Y . Ya . Abramov

Improvements of Cont.rol and Safety System for RBMK i.vpe Reactor s

F.O. Adamov, Yu. M. Ci erkasi ov, V.V. Kon<irat'ev,S.G. Ukharov, R.R. onajtis

Op .ration of NPR with RBilK type Reactors and Measures to Improve Reli abi lityand Safety

F..O. Adamov, V.P. Vasilevsky, A.A. Petrov, Yu. M. Chevkashov

RBD Programs for Reliabiiity and Safety Improvement

O. Yu. Novosel'sky, B.A. Gabaraev, E.V. Burl:kov, A.A. Kalugin,D.A. Mikhailov

Neutron-Physical Characteristics of RBMK Type Reactors Before the Accidentand After !mplementation of Measures for Increasing Safety

V.P. Borshchev, A.D. Zhirnov, V.D. Nikitin, I.A. ~tendock,Yu. M., Cherkashov, E.V. Burlakov, A.V. Krayushkin,

V.S. Romanenko, Yu. A. Tishkin

Feasibility Study of Safety of Uranium Graphite RBMK type Reactors

E.O. Adamov, V.P. Vasi lousky, A.A. Petrov, Yu. M. Cherkashov

Theoretical Analysis and Numerical Modelling of Heat Transfer and FuelMigration in Underlying Soils and Constructive Elements of Nuclear Plantsduring and Accident Release from the Core.

R.V. Arutunjan, I .A. Bolshov, V.V. Vitukov., V.M. Golouiznin,A.M. Dykhne, V.P. Kiselev, S.V. Klementowa, I.E. Krayyushkin,A.V. Moskouchenko, V.P. Pisminnii, A.G. Popkov, S.Y. Cher>ov,

V.V. Chu ivanov, O.V. Kh:.ruzhii, A. I. ".udin

Multidimensional Calculations support Two Bang Ti.eory

S. Baron

lessons Learned about PAR beyond Oesign Basis Accii>ent

S. Baron

The Fear Evoked by Radiation or Radioactivity and its psycho-sociologicalConsequences: What can be done to improve this situation?

S. Pretre

Severe LOCA at Lucens Experimental Nuclear Power Plant in Switzet land onJanuary 21, 1969

J.P. Buclin

Chernobyl-4 Accident Analysis

E.E. Purvis, III, B.W. Spencer

Use of Relative Fission Product Release to Estimate t.he Fuel TemperatureAfter Chernobyl Accident

A.A. Rimski-Kovsakov

Accident Progression Analysis of Past Severe Accidents

D.J. McCloskey

Ecological and Genetic Consequences of the Chernobyi Atomic Power PlantAccident

V.E. Sokolov, I.N. Ryabov, I.A. Ryabtsev, F.A. Tikhomirov,V.A. Shevchonko, A. I. Taskaev

Cleanup After a Nuclear Accident

J.O. Snihs

The Use of Tradescanti a Clone 02 for Biological Oosimetry in the Region ofthe Chernobyl Nuclear Power Station

I.N . Nilova, A.B. Semov, A. I. Taskayov, V .A..Shevchenko, N. I. Vavi lov

Strategy for Handling Contaminated Food: A Nordic Proposal

J.O. Snihs

General Recommendations iin Intervention Levels arid Strategies for Measurements

J.O. Snihs

Analysis of the Efficiency of DNA Repair in Plants for the Genetic Monitoringin the Chernobyl Vicinity

A.B. Syomov, Pititsyma, S.A. Sergeeva

What is-And What is Not-in NUREG/CR-4214

Niel Wald

Evaluation of Genetic Damage Caused to the Population of Chernobyl Vicinityby Radiation Exposure

V.A. Shevchenko

Genetic Consequences of Chernobyl Accident for the Population of Gomel andGomel and Mogilev Regions

G.I. Lazjuk

Techniques to reduce Acute and Chronic Population Doses After MajorRelease of Radionuclides

W. Burkart

Medical Response to Severe Accider,ts

J.C. Henot

A Compromise Regarding the. Three Different Cultural Responses to the"De Minimis" Question

S. Pretre

Genetic Disorders in Mice Caught or Exposed in the Region of the ChernobylNuclear Power Station

V.A. Shevchenko, M.D. Pomerantseva, L.K. Ramaiya, A.V. Chekhovich, B.V. Testov

The Monitoring of Seeds of Herbaceous Phytocenosis in the 30 i<ilometer Zoneof the Chernobyl Disaster

A.I. Taskaev, N.P. Froloua, O.N. Popova, V.A, Shevchenco

The Role of Agricultural and Natural Ecosystems in the Internal DoseFormulation in the Inhabitants of a Contaminated Zone

M .I. Balonov, I.G. Tranikova

Cyto Genetic Effects of Ionizing Radiation in Crepi s Tectorun, PopulationsGrowing in a 30 Kilometer Zone of the Chernobyl Atomic Power Station

L. I. Grinikh, V. V, Shevchenko

Genetic Consequences of Radioactive Contamination of ArabidopsisPopulations

V. I. Abramou, 0. M. Iedorenko,V. A. Shevchenko

Photographs

Other Publications used at Workshop

Evaluation of the Genetic Effects of Pu238 "Incorporated into Mice

M.D. Pomevantseva, L .K . Ramaya, V.A. Shevchenko, G .A . Vi lki na,A.M. Lyaginskaya

(Published Mutations Research, 226 (1989) 93-98.)

Genetic Effects of I 131 in Reproductive cells of Male Mice

V.A. Shevchenko, L.'K. Ramaya, M.D. Ponerantseva, A.M. Lyaginskaya

(Published Mutation Research, 226 (1989) 87-91)

Medical Aspects of the Accident at the Chernobyl Nuclear. Power Station-Material Presented at a Conference held at Kiev, May 11 through 13, 1988

Radiation Exposure Near Chernobyl Based on Analysis of Satellite Images(December 1987)

Marvin Goldman and E.A. Warman

Source Terms and Emergency Response - A Post Chernobyl Perspective(September 1988)

E.A. Warman

(TP 87-88/NX/EMR - Stone 8 Webster Engineering Corporation)

Soviet and Far Field Radiation Measurements and an Inferred Source Termfrom Chernobyl (April 1987)

E. A. Warman87-13NX/ENY Stone & Webster Engineering Corporation

Report of the U.S. Department of Energy's Team Analyses of the Chernobyl-4Atomic Energy Station Accident Sequence (November 1986)

DOE/NE-0076 U.S. Department of Energy,Assistant Secretary for Nuclear Energy

Participants

Organizers

E. P. Veli khov, Chai rman, USSR Academy of Sciences and Director,Kurchatov Atomic Energy Institute (KAEI)

L. A. Bolshov, Nuclear Safety Institute (NSI), USSR Academy of SciencesA. J. Gagarinskii, USSR Nuclear Society/YAEIS. V. Shestakov, Vavilov Genetics Institute, USSR Academy of SciencesS. G. Pankratov, S.G., NSIMilton Levenson, Consultant

Group I - Reactor

Chairman - A. K. Kalugin, KAEICo-chairman - W. B. Lowenstein, Electric Power Reseat ch Institute

E. E. Purvis, III, U.S. Department of EnergyR. VS Arutjunian, NSIJ . P. Buclin, S A L'nergie de 1'uest - SusisseV. Drujinin, Research and Development Institute for Power Technology (RDIP)R. B. Duffey, E.G. LG.H. C. Guimbail, EDFE. F. Hicken, GRS

A. V. Krajushkin, KAEIR. Martinelli, ENEA

0. J. Novoselskii, RDIPL. N. Podlazov, RDIPI. A. Stenbock, RDIPF. F. Zherdev, KAEI

Group II - P'lant

Chairman - J. M. Cher kashov, RDIPCo-chairman - J. M. West, Consultant

S. Baron, Brookhaven National LaboratoryP. E. Berghausen, Jr., RH8R

L. Bregeon, Protection and Nuclear Safety Institute (France)L. Burlakov, KAEI

Joe Howieson, AECLR. C. Heider, Sargent and LundyKuklin, VNIPETA. A. Petrov, RDIPB. F. Petrov, Khlokin (Radium) InstituteV. F. Stri zhov, NSIS. G. Ukhavov, RDIPV. P. Vasilevskii, RDIP

Group III - Transport and D'.spersion

Chairman - V. A. Borzilov, Scientific-Manufacturing Organzation "TAIFUN"Co-chairman - E. A. Warman, Stone 5 Webster Engineering Corp.

F. Bosuoli, ENEA

E. K. Gargor, TAIFUNL. M. Hitrov, Ver naskij Geochemistry and Analytical Chemistry Institute,

USSR Academy of SciencesM. P. Kanevskii, KAEII. I. Kpyshev, TAIFUNA lain 1'Homme, Protection and Nuclear Safety Institute (France)D. J. McCloskey, Sandia National LaboratoryW, Schock, K2P Laboratorium for Aerosolphysi k und Fi ltertechni k

V. N. Petrov, Applied Geophysics Institute, USSR Hydrometerology StateCommittee

G. Sege, NRC

E. V. Sobotovich, Geochemistry and Physics Institute, Ukr SSR Ministry ofHealth

V. G. Volkov, KAEI

M. J. Zheleznyak, Kibernetics Institute, Ukr SSR Academy of SciencesV. G. Volkov, KAEI

Group 4 and 5 (combined) - Dosimetry and Consequences

Chairmen-I. A. Li khtarev, Alhunion Radiation Medicine Scientific Center,

USSR Academy of Medical SciencesV. A. Shevchenko, Vavi lov Genetics Institute, USSR Academy of Sciences

Co-chai rmen-L. R. Anspaugh, Lawrence Livermore LaboratoryM. Goldman, University of California, Davis

R. M. Aleyakhin, Agriculture Radiation Protection Dosimetry RadiologyInstitute

W. Burkart, Paul Scherrer InstituteJ. L. Dobrynin, KAEIV. P. Dromin, KAEI

I. V. Fi ljushkin, Biophysics Institute, USSR Ministry of HealthV. A. Kniini kov, Institute of Biophysics, USSP. Ministry of HealthW. Kroger, ISFG. I. Lazivck, Hereditary Illness Institute, Bnl SSR Min',stry of HealthJ. Murray, Uranium InstituteJ. C. Nenot, Institut de Protection et de Surete Nucleai reD. A. Pavlouskii, Biophysics Institute, USSR Ministry of HealthL. A. Pleskachouskii, Khlopkin Radium InstituteS. Pretre, Paul Scherrer InstituteM. D. Pomevantseva, Vavi lov Genetics Institute, USSR Academy of Sciences

A. A. Kimsky-Korsakov, Yhlopkin Radium InstituteI. N. Ryabov, Fauna Evolut.ion Morphology and Ecology InstituteC. Sennis, Italian Commission for Nuclear and Alternative ErergiesA. B. Siomor, Vavilov General Genetics Institutes, USSR Academy of SciencesH. J,. Snihs, National Institute for Radiation Protection (Sweden)V. N. Sorokin, Atomic Energetics Institute, Bol SSR Academy of SciencesA. I. Taskaev, Biology Institute, Komi S;ientific Center, USSR Academy of

SciencesNiel Wald, University of Pittsburg

Other

V. V. Skarlygin, KAE!ST J. Chernyshov, KAEI

E. I. Aslamazian, NSIV. N. Ponomarev, NSIS. A. Shepe'iev, USSRG. P. Kamennov, NSIS. V. Dani lov, USSRG. Kaiser, USSRStankevich, USSRV. I. Kobyliakov, Department of Foreign Affairs, USSP, Academy of SciencesA. B. Ry1 s ki i, KAE IM. S. Veshchonov, NSIEvdokimov, USSRD. I. Beliaev, USSRM. J. Tovbix, USSEM. J. Nestorov, USSRBelaitz, USSR

Press

A. D. de Segonzac, CA M'intercsseD. L. Gvozder, "Poisk" NewspaperV. I. Jan Kulin, "Literatornaya Gazeta" NewspaperG. M. Lvovsky, "Priroda"A. E. Prlnikov, "Moscow News" NewspaperV. Romanenkova, Soviet Union Telegraph Agency

PROGRAM COMMITTEE

Velikhov E.P.

Adamov E.O.

Avdyushin S.I.

Belyaev S.T.Bol'shov L.A.

Borzilov V.A.

ll'in A.V.

- Member of the USSR Academy of Sciences,

Chairman

- D. Sc. (Techn.), Scientific Research and

Design Institute of Power Engineering

of the USSR State Committee for the Use

of Atomic Energy- D. Sc. (Physics and Mathematics), Institute

of Applied Geophysics of the USSR State

Committee for Hydrometeorology- Member of the USSR Academy of Sciences- D. Sc. (Physics and Mathematics), Institute

of Problems of Nuclear Power Safe

Development of the USSR Academy

of Sciences- D. Sc. (Physics and Mathematics), Institute

of Experimental Meteorology'of the USSR

State Committee for Hydrometeorology:- Member of the USSR Academy

of Medical Sciences

Ponomarev-Stepnoy N.N.

- Member of the USSR Academy of Sciences

Rimsky-Korsakov A.A.- D. Sc. (Physics and Mathematics),

V.G. Khlopin Radium Institute of the USSR

State Commitee for the Use of Atomic Energy

Romanenko A.E.

Sokolov V.E.Subbotin V.I.

Sorokin V.N.

Shestakov S.V.

- D. of Medicine, Ministry of Public Health

of the Ukrainian SSR- Member of the USSR Academy of Sciences- Member of the USSR Academy of Sciences- D. Sc. (Techn.), Institute of Nuclear Power

': of the Ukrainian SSR Academy of Science'

Corresponding Member of the USSR

Academy of Sciences

ORGANIZATION COMMITTEE

Bol'shov L.A.

Ponomarev V.N.

Arutyunyan R.V.

Burinsky A.Ya.

Kalugin A.K.

Kushnarev S.V.

Nikitin Yu.M.

Pankratov S.G.

Shevchenko V.A.

- D. Sc. (Physics and Mathematics),

Institute of Problems of Nuclear Power

Safe Development of the USSR Academy

of Sciences, Chairman

- D. Sc. (Physics and Mathematics),

Institute of Problems of Nuclear Power

Safe Development of the USSR Academy

of Sciences- Cand. Sc. (Physics and Mathematics),

Institute of Problems of Nuclear Power

Safe Development of the USSR Academy

of Sciences- Cand. Sc. (Physics and Mathematics),

the USSR Nuclear Society- Cand. Sc. (Physics and Mathematics),

I.V. Kurchatov Institute of Atomic

Energy of the USSR State Committee

for the Use of Atomic Energy- Cand. Sc. (Physics and Mathematics),

the USSR Nuclear Society- D. Sc. (Techn.), Scientific Research and

Design Institute of Power Engineering ofthe USSR State Committee for the Use

of Atomic Energy- Cand. Sc. (Physics and Mathematics),

Institute of Problems of Nuclear Power

Safe Development of the USSR Academy

of Sciences- D. Sc. (Biology), N.I. Vavilov Institute of

General Genetics of the USSR Academy

of Sciences

Address: I.V. Kurchatov Institute of Atomic Energy

Kurchatov Square

Moscow 123182

USSR

Telephone: 196-15-58

196-15-57

Draft

Summary Report

An international workshop on past severe nuclear accidentsand their consequences was held in Dagomys region of Sochi, USSR

on October 30 - November 3, 1989.The plan of this meeting was approved by the USSR Academy

of Sciences and by the USSR State Committee of the Utilization ofAtomic Energy.

While the meeting was held under the umbrella of the ANS-SNS

agreement of cooperation, the meeting was jointly sponsored by thefollowing organizations:

1) Institute of Problems of Nuclear Power Safe Developmentand Institute of General Genetics of the USSR Academy

of Sciences;2) I.V.Kurchatov Atomic Energy Institute and Design

Institute of Power Engineering of the USSR State Committeefor the Use of Atomic Energy;

3) The Soviet Nuclear Society.These five organizations jointly provided for all

arrangements within the Soviet Union, including a premeeting visitto Chernob'yl and a post meeting visit to the laboratory of theRadium Institute ini Leningrad.

This meeting was somewhat different from previousChernobyl meetings and so the events leading up to the meetingare summarized here.

In August of 1988 discussions were started between

Academician E.Velikhov of the USSR arid M.Levenson of the UnitedStates on the question of how to ensure that the maximum amount

of useful information was being obtained from past nuclearaccidents. In most cases of past accidents there has been strongpressure to draw conclusions as soon as possible of "lessonslearned" or causes of accidents have been published beforeall the data has been analysed. It was felt that it might be

worthwhile to have a good technical group review each old accidentto ensure that all important relative information had been

recognized. and that any early conclusions that might have been

1

incorrect were corrected. It was felt that such a review mightbenefit from being multidisciplinary/multinational. It was feltthat normal conference procedures would not produce the desiredresults. In October 1988 three American and one Italianrepresentative met with a Soviet Group assembled in Moscow byAcademician Velikhov. At that meeting it was agreed that LeonidBolshov (USSR) and Milton Levenson (USA) would work together todevelope the first such review.

It was felt that the effort should not be done as partof government to government agreements. Various alternatives werediscussed and it was decided to hold the first workshop under theumbrella of the American Nuclear Society (ANS) - Soviet NuclearSociety (SNS) agreement for cooperation. It was decided to try aworkshop format covering five topics: the reactor itself, the restof the power plant, the mechanisms and phenomena of fissionproduct transport, health effect s, and dosimetry. It was agreedthat the first workshop should concentrate on the Chernobylaccident and -"nuld be held in the USSR. A planning workshop washeld at the Worla Laboratory Center in Erice (Sicily) Italy inJanuary 1989 to further explore the concept and developedetails. While it was recognized that seismic matters arean important aspect of any overall nuclear safety program,earthquake played no part in the Chernobyl accident andso was not included in the agenda of the first workshop. It wasdecided to invite two or three American experts, two or threewestern European experts and five or six Soviet experts for eachworking group so that the groups would be small enough to maintainthe working group format. It was not intended to have presenterand listener but rather everyone was expected to be a participant.A trip to Chernobyl was arranged for the non-Soviet participantsso that they could come to the workshop with first handpersonal observations. The Chernobyl visit included the insideof the sarcophagus, viewing the .'nside of the reactor viaperiscope, the control room of unit 1 (at power) and the controlroom of unit 4, the city of Pripyat, and the areas ofecological damage and regrowth.

The overall objective of the workshop is an attempt toreduce uncertainty. It was anticipated that the working groups

would discu s two general themes:1) Understand the technical aspect of the accident:

A - Identify possible initiators and contributorsto severity;

B - Determine the processes and mechanisms of., theaccident sequence itself.

2) Determine the consequences:A - Determine the actual post accident state - where

is the fuel, graphite, etc now;

B - Health and environmental effects;The actual meeting was a mix of presentations and workshop

discussions. Some questions were resolved, that the participantsagreed. In other areas questions that needed resolutionwere identified since either the participants did not agree orelse further data or studies were needed before resolutioncould be expected. The summaries of the working groupconsiderations will be attached.

This is the background that led to the start of. theworkshop. The list of actual participants is attached. Eachworking group had two co-chairmen - one from the ANS and one fromthe SNS. At the first planning session the co-chairmen of theworking group on health effects and the working group on dosimetrydecided to combine their groups and meet as one single working

group. The . eactor working group and the plant systems working

group decided to meet together for a few specific discussions butto meet. separately for most issues.

An assessment of the workshop indicates that most objectiveswere achived. In addition substantial inter-organizational, inter-disciplinary and inter-national interactions and discussions tookplace. One meassure of the assessment of the workshop by theparticipants is that several participants indicated a desire tohave their countries host the next workshop. It should be notedthat participants attendance was essentially 100% throughout the

long working day in spite of the competition offered by Sochi and

the Black See. In some cases the meeting of two working groupstogether resulted in a group a little too large for a good

workshop atmosphere and it is recommended that the next workshop

be slightly smaller in total attendance.

Siiiiiiiiar y uf woi kirig gi oup 3

I t. i s agreed i.t>a t t.l ie pi inc; 1 pa l ( <3l 'III of iod ine wa gaseous,wit.h orily 20'r'n par.t.i< <ilat.e forni. Fst.iniat.es of the tot,al

amoun'f

iodine re leased need fur ther s t.udy, iric iud ing t.liat, wli i cli has iio t,been sub ject t.o t.liis wc>i kst>op.

Serious at t enl. ion was pa i<5 Lo the st.u<'y and t he expei t.evaluat.ion of the rigoi ousiiess of tt>e prognostic: niethods fori-adionuc1 ide behaviour. iri the ace icleri t. zone, tl ie ir pr obab le migrat. i c>n

by air and wat.erways. llew results on l lie se<.ondar.y wind susl>enslor> oft.he radionucl ides fr oin t.lie soi 1, oii t.he pliy™ico-clieniical pr oper.t iesof the deposi ted rad ioiiuc1 icles an<i, iri par t icul ar, c>f t he "hot.par t. icles". 'I'he dist.i" ik>ut ior> c>f var i<>uc fc>rins of deposi teel mat.er i a 1in t he 30-km zone, ai icl tl ie I r> f luei ice c> f'.liose for ms on t he so i 1 andwat er inigr at ion proces . 1'he grc>ul> agi ees„ t hat t.tie i eview of t.hesequestions was very comreliensive.

The est imat es, nia<le l>y t lie gi < >up show t liat. t lie ra< i i <~»>c<c 1 i det i-arisport, and iiiigr al.ioii fr c>in t lie «>nt ainiiiated areas are riegligible at.present., k>ut. dr v<. Ioy>en<eiit. <>f cur i"ec:t. l>i.e<li<=t.ic>ris f'c>;;.i dist.ai>t. fi>t.ui-efurther study 1s r>ecessc>ry.

Rec< >inr. n<3at i or<s:

The fol low i iig i e<.u»><'nctat i ons ai.e macJe in the oi-dei- of'.liediscussion, r athei. t.hari iii t.lieir order of'mport. ance.

A. Sour ce Term.1.Measurement aii<J st,ucly c>f tlie reiiiairi irig long 1 ived r adionuc1 ides „

129e. g. iodirie I,locate<i wi tliiri tlie damage<1 urii t. shoul<l L>e per.foi medto establish the fr'action of t tie core inventory retained in thep 1an t..

2.A more detailed c>ri-.=it e corot ar»iriation map sl>oui<5 k>e developedfor. l.he per iod pr 1 or to clesa< t. i vat. ion. This shoul<J iriclude on-si t edist.r ibut,ioii of graphi te l>lc>cks, etc:, aiicl fuel elemerit.s fragnients inor<Jer to establ ish t.he t.ot.al an<a<crit of cor e clebr is on-sit.e. Higliresolution phot ogr aptis arid clc>siniet.i-y dat.a st>ou ld be ut i 1 ised in theinvest igat. iori.

3.A joint. ar t. i cle summar" is l rig t.he iri for mat. ion on t he sour ce t ermdescribed froin the release and deposit.ion data should be pi"epaired.This work has beeri assighr>ed to 'V. Bor silov, A. Rimski-Knrsakov aridE. Warman.

B.Forms arid t.ranspor t of rad ionuc1 ides.

1.Detailed charact.er.isatic>n of the physico-chemical form- of'.hedeposited i.adionu<" 1 ides should be <Jeveloped, including solubi 1 i t.y,t.he i r t.rans for mat. ion deperid ing in t.he geochemical fact.ors sl i< >uldexpanded.

2.In additiori t.o <Jet,ailed maps displaying radionuclidecont.amination off-site, inaps depict.ing pliysico-chemical forms in soiland wat,er should be developed.

3.The in format. ion eleve lopecl as a r.esul t, of recommendations ( 1. )and (2. ) above, would gr"eat ly enhance the abi 1 i t.y to buid,develope and val idate t.ransport. models. This should be used inpredicting of the radiological situat.ion in tlie contamir>ated areas inthe f'utur e.

S V 6 6 E S T I 0 N Sof the First Internationa] Workshop

on the Severe A< cf~ents and their Consequencesat V. 6. Khlcpfn Radium Inst]fute

Lenfngrad, USSR, November

The members cf the workshop were shown the scienti-fic «ork, done in the fie]d of severe accidents at theRadium Institute and ome discussion followed. This dis-cussion showed, that furtlier study of data from Cherno-hYI accident and its consequence~ will he useful for de-ve]opment of safer reactors and for planning the measu-res to reduce risks in cases of radioactive releases,

The workshop members he]ieve that internationalcooperation of interested organizations fn accomp)]shingthe 1< llowfng tasks would he helpfu] to improvethe operationa] safety

a Joint compilation of ezfstfng data on the Cherno-hy] Source Term should be done , and corre]ated toradionuclide release amounts, ditrihutions and ef-fects on worKers.determination of partic]e size distribution re]atedto distance , including chemical and radionuclidecontent shculd he included int.o Source Term evalua-tion.avalfahility of Chernoby]-0 and other accident dataand specimens for comparative study shou]d be orga-nized to assist in accident scenario developmentand verification.an exchange of specialists on measurement of fa]loutactivities, and ana]Ysis of radionuclide mobilitY invarious soil types and under different natural envi-ronments shou]d he organized for mutual benefit.

WORKSHOP HEMBERS

Hilton Levenson, US

Edward E. Purvis III, US

Marvin Goldman, US

Claudio Sennis, Italy

Iv'.

A. Rimski-Xorsakov, USSR

P~~zzt Eat|em@(tozz) WorkshopG1lll Posit Stt'vcr@ Accll65Elts al12i6 tk)lc;Eri

CGHEKGglLRGKilccs

October 50-Rovember 5, 1989

Sagomqs, Sochi, U.S.S.R.

Summar„'y

Memorandum of Meeting

Draft prepared and sent by Marvin Goldman 13 December 1989

WORKSIIOP SUMMARY

Points of Discussion in Groups 1 8; 2 (reactors and other nuclear plant facilities)

1. Effects and risks of positive reactivity coefficients.

2. I'ositive reactivity effects of control rod motion:

3. Adequate control/safety rod speed for emergency conditions.

4, Consequeiices of multiple pressure tube rupture, including upper lid lifting and release ofradi oacti vi ty.

5. What might be the consequences of a steani line break?

6. How does the list of possible RBMK emergency situations compare to lists for otherreactors? Is it complete?

7. What is the role of confinement or containment for RBMK reactors?

8. Relationship between RBMK rod speed, rod worth, and reactor power.

9. Adequacy of the condensing capacity of the bubbler pool for multiple tube rupture beyond thecurrent design basis.

10, Emergency training of operating personnel and human factors,

11. Fire protection and emergency power.

12. RBMK-process monitoring during abnormal situations.

13. Role of past operating experience, regular and emergency, on P.BMK modifications.

Questions/Recommendations/Conclusions

Some of the discussion items were covered jointly by working groups 1 & 2, but aresummarized separately.

Group 1

The level of detail available and the limited time available meant that in-depth assessments. could not be done. However, it was possible to identify trends and agree on some issues.

1. The efforts to date have reduced the probability of a Chernobyl-type accident, but thereremains a potential for still further increases in safety of the RBMK.

2. What further modifications should be made to the RBMKs to increase stability and to increasethe reliability of containment?

3. There was a consensus that the chief cause of the accident was reactor instability —due both

to design shortcomings and to operating procedures and their violations.

4 3D modeling of the entire recirculation loop will help answer the question of whether theproposed changes are adequate.

5. Can comprehensive initial data concerning Unit 4 operation just prior to the accident be madeavailable for independent analysis —not computer output but original data on neutron flux,coolant flow, temperature, pressure, pump speeds, rod position, etc.?

6. It is important that all information on stage 2 of the accident be made available —distributionof fuel, graphite, condition of rooms inside Unit 4, composition of "slag" or lava, etc.

Group 2

1. The pressure relic< system for failed pressure tubes is being expanded on new RBlv1Ks-what is an adequate size and should it also be backfit to existing RBMK's?

2. Can the program to assure pressure tube integrity be further enhanced?

3. Design basis accident analysis should be expanded to include low-power operation and off-normal conditions.

4. Is the RBMK data base adequate to do a PRA analysis? Would ultrasonic inspection of thestainless steel piping between the core and the steam separators be justified?

5. Is the bubbler pool adequate for multiple tube ruptures or large pipe breaks?

6. Significant improvements in fire protection were discussed. When will all RBMKs bemodified to meet the proposed changes?

Group 3 (Radionuclide Transport): Topics Discussed, Conclusions

I. The total amount of iodine released needs further study and may be substantially larger than

first reported in August 1986.

2. The principal form of iodine transported off site appears to have been gaseous with only 20%in particulate form.

3. The previously reported release of 3 to 4% of the core inventory of non-volatile fissionproducts represented the amount released off-site in the form of aerosols. The fraction of the

core inventory that was ejected in the form of very large aerosol chunks of core debris and

portions of fuel elements deposited on-site and very near the site has not been quantified.

4. Fairly detailed discussions were held on radionuclide behavior in the accident zone, including

migrations by waterway and air. New results on secondary suspension and hot particleswere also discussed.

5. It was the consensus of the working group that radionuclide transport/niigration from the

contaminated areas is currently negligible.

QnesrionslRecnmmcndatiniu

l. Analysis of long-lived fission products in debris samples from within the damaged unitshould be used to aid in determining the size of the accident source term. A materials balanceapproach could be utilized to determine how much iodine, cesium, etc. remained within theplant.

2. If it is possible to construct detailed post.-accident contamination maps of the site for theperiod before decontamination started including distribution of graphite, fuel, and buildingdebris, such an activity should be undertaken. Perhaps such a map could be constructedfrom photographs and radiation survey data taken during the first few days. In addition toaiding in transport analysis, such a map would aid in understanding the accident sequence,

3; Members of this working group (as well as other working groups) should take advantage ofthe contacts made and information exchanged to develop joint publications by the workshopparticipants.

4. The presently developed detailed off-site contamination maps should be expanded to includemaps depicting the physico-chemical forms of deposited materials in soil and water? Thiswould not only enhance the understanding of the transport phenomena of the past, it wouldalso aid in predicting transport in the future and in developing and validating transport.models.

Groups 4 R 5 (Dosimetry and Health Effects): Topics Discussed

1. Contributions to total dose of internal and external radioactivity doses.

2. Problems of late health effects predictions from low dose and low-dose rates.

3. What are the criteria to establish "lifetime" dose limits,II4. The contamination of wild animals. How to manage forestry, hunting, and fishing in

contaminated areas?

5. How to apply the methods used in the study of atomic bombing survivors in Japan?

6. Problems of radiation risk communication and perception and of "radiophobia".

7. Effectiveness of various dose-reduction countermeasures.

8. Biological indicators of dose and of radiation injury to humans„ flora, and fauna. Thisincludes genetic studies in selected plant species.

9. Agricultural radionuclide transfer factors and their role in dose reduction and dose prediction.

QttestionslRecommendations

l. A good delineation of functions between various Soviet organizations is urgently needed bypotential collaborators in order to ensure efficient cooperation with foreign countries andindividual experts.

2. Who would do the epidemiological study of all the children with high radioiodine exposure?

3. For model validation cross-calibration and data updating, a joint international reassessment ofthe forecasted dose for the USSR should be performed.

4. Scientific and training exchanges between the USSR and other countries on biomedical anddosimetric problems associated with the Chernobyl accident should be established to advance

technique modifications, validate models for the future, and relate to existing information.

5. A workshop and subsequent work regarding psychological aspects of the Chernobyl accidenthas been recommended.

6. Medical management of combined injuries should be reviewed on an international level tobetter share newer knowledge and to improve medical procedures and options.

7. An international scientific advisory committee dealing with health effects of the Chernobylaccident should be created to assist in planning specific studies, developing health effectsmodels, and interpretating and evaluating data trends.

MEMORANDUM

OF THE FIRST INTERNATIONAL WORKSHOPON PAST SEVERE ACCIDENTS AND THEIR CONSEQUENCES

An International workshop on severe nur,:lear accidents and their consequences was held inthe Dagomys region of Sochi, USSR, October 30 - November 3, 1989.

The plan of this meeting was approved by the USSR Academy of Sciences and by ihc USSRState Committee of the Utilization of Atomic Energy.

While the meeting was held under the umbrella of the ANS-SNS agreement coopera! ion, themeeting was jointly sponsored by the following organizations:

l. Institute of Problems of Safe Development of Nuclew Power (Nuclear Safety Institute),and N.I.Vavilov Institute of General'Genetics of the USSR Academy of Sciences;

2. I.V. Kurchatov Atomic Energy Institute and Design Institute of Power Engineering ofthe USSR State Committee for the Utilization of Atomic Energy;

3. The Soviet Nuclear Society.

These five organizations jointly provided all the arrangements within the Soviet Union,including a pre-meeting visit to Chernobyl and a post-meeting visit to the laboratory of the RadiumInstitute in Leningrad.

Since this meeting was somewhat different from previous Chernobyl meetings, the~entsleading up to the meeting are summarized hei'.

In August 1988, discussions were started between Academician E.Velikhov of the USSR andM. Levenson of the United States on the question of how to ensure that the maximum amount ofuseful information was being obtained from past nuclear accidents. In most cases of past accidents,there has been strong pressure to draw conclusions as soon as possible of "lessons learned" orcauses of accidents have been published before extensive data have been analyzed. I:was felt that itmight be worthwhile to have a good technical group review each old accident to ensure ihatcomprehensive, important relative information had been recognized, including correcting any earlyconclusions that might have been incorrect. It was felt that such a review might benefit from beingmultidisciplinary/multinational. It was felt that normal conference procedures would not produce thedesired results. In October 1988, three American and one Italian representatives met with a SovietGroup assembled in Moscow by Academician Velikliov. At that meeting it was agreed that LeonidBolshov (USSR) and Milton Levenson (USA) would work together to develop the first such review.

It was felt that the effort should not be done only as part of government-to-governmentagreements. Various alternatives were discussed and it was decided to hold the first workshop underthe umbrella of the American Nuclear Society (ANS) - Soviet Nuclear Society (SNS) agreement forcooperation. It was decided to try a workshop format covering five topics: the reactor itself, the restof the power plant, the mechanisms and phenomena of fission product transport, healtli effects, anddosimetry. It was agreed that the first workshop should concentrate on the Chernobyl accident andshould be held in the USSR. A planning workshop. was held at the World Laboratory Center in Erice

(Sicily. Italy) in January 1989 to further explore the concept and develop details. While it was

recognized that seismic matters are an important aspect of any overall nuclear safety program,

earthquake played no part in the Chernobyl accident and so was not included in the agenda of the fir. t

worhshop. It was decided to invite two or three American experts, two or three Western European

experts, and five or six Soviet experts for each working group so that the groups would be small

enough to maintain the working-group fcrtnat. It was not intended to have presenter and listener but

rather everyone was expected to be a participant. A trip to Chernobyl was arranged. for the non-

Soviet participants so that they could come to the workshop with first-hand personal observations.

The Chernobyl visit included the inside of the sarcophagus, viewing the inside of the reactor via

periscope, the control room of Unit 1 (at power) and the control room of Unit 4, the city of Pripyat,

and the areas of ecological damage and regrowth.

The overall objective of the workshop was an attempt to reduce uncertainty in the state or the

'rt of RBMK safety, with identification of potential actions for further improvement of safety. It was

anticipated that the working groups would discuss two general themes:

l. Understand the technical aspect of the accident:

A. Identify possible initiators and contributors to severity;

B.Determine the processes and mechanisms of the accident sequence itself.

2. Determine the consequences:

A. Determine the actual post-accident state of the fuel, graphite, etc. now;

B.Health and environmental effects;

The actual meeting was a mix of presentations and workshop discussions. Some questions

were resolved, that is, the participants agreed. In other areas, questions that needed resolution were

identified since either the participants did not agree or else further data or studies were needed before

resolution could b expected. The summaries of the working-group considerations are attached.

This is the background that led to the start of the workshop. The list of actual participants is

attached. Each working group had two co-chairmen —one from the ANS and one from the SYS. At

the first planning session, the co-chairmen of the working group on health effects and the working

group on dosimetry decided to combine their groups and meet as one single working group. The

reactor working group and the plant systems working group decided to meet together for a few

specific discussions, but to meet separately for most issues.

Points of Discussion

Groups I and 2

1. Effects and risks of positive-reactivity coefficients. More specifically, what are void effects

on reactivity and what are the problems associated with them in operating RBMK reactors?

2. Positive-reactivity effects of control rod motions. What are present day guarantees that no

positive reactivity is possible in the process of control rods being introduced into the core?

3. Is the control and safety rod speed adequate in emergency conditions?

4. Is the multiple rupture of tubes a possible consequence of such an event? This accident mightlead to sharp pressure increase inside the reactor space, with the upper lid being lifted and

radioactivity released.

5. What might be the consequences of a steam line break?

6. How can the list of possible RBMK emergency situations be compared to corresponding listsfor other reactors? Is the former list complete enough?

7. What is the role of confinement or containment for RBMK reactor?

8. What is the relationship between RBMK rod speed, rod worth, and reactor power'?

9. Is the condensing capacity of the bubbler pool adequate for multiple tube rupture beyond thecurrent design basis, for instance, in the conditions of simultaneous ruptures of 3 tubes'?9 tubes?

10. Emergency training of operating personnel and human factors. What is the responsibilityol'he

personnel?

11. Fire protection and emergency power supply.

12. What are the facilities envisaged for RBMK plants monitoring in abnormal situations?

13. Role of past operating experience, both regular and emergency, on RBMK modifications.

The amount of details available and the limited time of the workshop only provided theopportunity to identify trends and to agree on main issues. It is our belief that further cooperation in

specific areas, as well as forming an overall view, would lead to a more complete understanding ofthe situation..

In general, as far as Group 1 is concerned, it was assessed that the efforts being undertakenhave drastically reduced the probability of a Chernobyl-type accident, but there still remains apotential for further increase of the RBMK safety. American and European experts participating in

the workshop offered to assist the USSR in improving RBMK safety, taking part in the supportingresearch and development program.

The conclusion reached by the workshop participants is rather general. To further enhance thc

safety of RBMK, to minimize or practically exclude the regions of reactor instability, and to increasethe reliability of containment warrant further investigations. Specifics such as increasing fuel

enrichment, reducing moderator quantity or lowering moderator efficiency, hardening of the reactor

structures should be studied by RBMK specialists. Five days of workshop discussions, however

intensive they could be, are hardly sufficient to make firm technical recommendations for hardware

improvement or procedure modifications.

Conside.ing the extreme importance of acquiring public support for future RBMKoperations, it may be worthwhile to point out some changes that will be effective in the near future.

Seven Soviet and four foreign presentations were discussed at the sessions of Group 1 and, as

previously mentioned, a few sessions were held together with Group 2. There was enough time left

for thorough discussion of all the presentations, as well as of a number of issues put forward bynon-Soviet colleagues. Foreign participants in the workshop produced results of their own

investigations of Chernobyl accident (E. Purvis, S. Baron, R. Duffey - USA; E. Hicken - West

Germany). H. Guimbail's (France) presentation was devoted to classification of accidents at nuclearplants.

In the Soviet presentations, the first stage of the accident was analyzed using detailed three-dimensional models and RBMK physical characteristics before and after the accident. Th'ubject ofthree presentations was the state of core fragments after the accident and characteristics ot activemasses in some rooms of Unit 4. The visitors appreciated honest and open technical discussionswith their Soviet colleagues. The informality of these discussions facilitated the dialogue greatly.

Principal conclusions drawn from discussions at the workshop, at least as far as its reactorpart is concerned, can bo divided into two subjects: (I) processes going on during the accident and

(II) current safety level of RBMK reactors. Now, let us examine them closely.

I. Processes going on during the accident

l. A consensus was reached at the workshop that the main cause of the accident was reactorinstability -i due both to design shortcomings and violations of operating procedures.

2. High quality of dynamic three-dimensional models presented by Soviet experts wasendorsed. It was proposed that investigations be continued in an expanded form to be able tomodel the entire recirculation loop to answer the question of whether the proposed changesare appropriate or not. !t is advised to pay attention to the necessity of thorough model

testing against the results of carefully planned and executed experiments.

3. There are different approaches in understanding elements of accident process. Iessibility of a

large-scale imbalance of neutron flow (S. Baron), influence of thermohydraulic instability

and pump cavitation on accident process {R.Duffey). Although these details do not change

the general overview of the process and main conclusions on causes of accident, it was

proposed to carefully examine them once again. Besides, can all the initial data concerningUnit 4 operation just prior to the accident be made available —not computer output but

original data on neutron flux, flow, temperature, pump speeds, rod position, etc.? So it was

recommended to bring into order and publish all the initial data concerning Unit 4 ofChernobyl plant before the accident.

4. The second stage of the accident after destruction of the core is investigated to a far lesser

extent. At the same time, it is this stage that determines distribution of the fuel after the

explosion and, consequently, temperature conditions during maximum activity releases.

Therefore, it is important that the information concerning stage 2 of the accident be made

available —distribution of fuel, graphite, conditions of rooms inside Unit 4, composition of'slag'r lava, etc. Thus it was proposed, first, to bring this data into order and, second, to

establish international collaboration on analysis of fragments of destroyed core (graphite,

fuel, zirconium tubes).

10

II. Current safety level of RIIMK reactors

1. Steam void-reactivity coefficient.

lt was stated that the value +038, which is planned to be achieved after the end of fuelenrichment reload of 2.4%, ensures negative power reactivity coefficient in analyzed operationmodes of reactor. It is demonstrated even now that the part of fuel assemblies with enrichment 2.4%amounts to 40-75%. It is recommended to analyze possible changes in reactor characteristics aftermeasures taken, in order to avoid undesirable side effects. lt was proposed to examine the possibilityof decreasing the steam effect value. The attention is drawn to the necessity of appropriate testing ofthree-dimensional program because a number of eflects (e.g., dehydration of hot reactor) can bestudied only by calculations.

2. Control and protection system.

It was pointed out that constructional changes of the rod, increase of control elementsresponse rapidity, and separation of quick emergency protection group (2.5 sec is now the totalduration of rods insertion) substantially improved reactor safety. Whatever currently regarded initialevents can be, abuse of operation modes included, modern control and protection system promotesreliability of scram.

3. Reactor stability.

As was reported by RBMK experts at all nuclear plants using RBMK systematicmeasurement of parameters, determining reactor stability is completed; (steam void-reactivitycoefficient, fast power reactivity coefficient, characteristic time of neutron field deformation in theabsence of automatic, and manual effect on power distribution). Diminishing of steam-reactivityeffect caused significant increase of reactor stability.

4. In-pile control systems.

The issue needs further investigation since the group did not have enough time for itsdiscussion.

Group 2

The operating RBMK reactors are currently equipped with steam relief capacity capable ofcoping with a simultaneous rupture of three pressure tubes. Pressure relief systems with a capacitythree times higher have been designed and are implemented now in the new RBMK reactors. Yet the

question arises: what is the adequate size of such a system and should it also be backfit to existingR BMK's?

Multiple tube ruptures exceeding the capacity of the relief system is the most severe event forRBMK reactors. Technical means of preventing causes of multiple failure such as thermal-hydraulic

instability or local excessive increase in power in a series of tubes, are provided. The channel tubes

are regularly checked in operation. A testing device designed as a RBMK core. fragment is under

construction in the USSR. The pressure tubes'ehavior will be investigated on this experimental

stand, one of the tubes being ruptured. All these provisions make the probability of a major accident

rather low. However, in view of the severity of the consequences of this event, all the causes of

such a situation should be analyzed time and again, in order to identify all the possible requirementsfor further improvements. In any case, steam relief systems of the operating reactors should bemodified to the capacity of the design implemented in the new units.

R&D program for the assurance of pressure tube integrity should be pursued actively, asindicated, taking existing information into account.

The initial list of design initial accident events for RBMK power plants was drasticallycorrected after the Chernobyl accident in coordination with the regulatory authorities and nowincludes nearly 50 typical situations of the following types:

- reactiviiy deviation accidents;

- failures in core cooling;

- ruptures in circulation loop pipe lines, steam lines and water-supplied pipe lines:

- equipment failure;

- other accidents (fire, water, flood, etc.).

A good number of the initial events were analyzed and safety of the reactor, equipped with

the design safety systems, was proved. Comparison with the similar list of the US nuclear powerplants would be useful.

Attention should be paid to a more detailed analysis of the low-power level (3-5 %)operation, off-normal conditions, and even accidents included neutron flux and power distortionsthat accompany the accident development should be considered.

Despite low probability, accident analysis should exclude the first signal validity in operatingand emergency protection. This is a provision to be incorporated in the reactor emergency protectionrequirements and in the safety system parameters.

1. Containment

A containment was not used on RBMK plants. Instead an accident localization system was

composed of high strength boxes that contain the major circulation equipment.

Some smaller pipe lines above the reactor are located outside the localization system. The

upper lines are made of stainless steel and are accessible for ultrasonic examination to determine the

piping status.

The workshop recommended that a Probabilistic Risk Analysis (PRA) be performed and

evaluated along with recommendations of "Basic Safety Principles for Nuclear Power Plants"-Safety Series 75, INSAG-3.

In analyzing emergency situations involving automatic reduction of the reactor power or its

shut-down, detailed research should be made not only of the accident process itself, but also ofevents that may accompany it. Technical provisions should be adopted to prevent switching off ofthe emergency protection by an operator. Access to emergency protection control panels in all

operating RBMK NPPs is safeguarded by two code keys.

12

2. Bubbler Pool

The pool is designed for steam condensing in case of a leakage in circulation pipe linesincluding a rupture of the largest pipe line (pressure collector). The ~ater volume in the pool isapproximately 3500 m3 whereas air volume may reach 13000m3. Bypass valves are designed on areserve basis to take into account a failure of one or several valves. The bubbler pool should bechecked to verify that it has the capacity to accommodate simultaneous (multiple) tube rupture orlarge pipe break. The bubblers should be tested on a large test stand.

3, Fire Protection

To provide for a higher NPP, fire safety special regulations "Fire Safety for NPP" (BCH-86)were introduced and coordinated with the state fire inspection authorities. New plants are designedand concrete measures are being taken to modernize construction and operation of NPPs in

accordance with the regulator requirements, all this to be completed within NPP reconstruction

p riod till 1993.

According to the regulations:

automatically operating fire fighting systems should be installed in the safety system roomsindependently of the plant fire fighting system;

non inflammable cables are developed and being introduced in the vital safety systems;

cable lines in different safety channels should be spaced properly. Special fire protectionlayers should be provided for in operating units containing overlapping cable lines in safety

systems and in other cases that do not meet ~Jie regulation requirements.

other organizational and technical requirements are intel'ded to prevent a fire or to limit thefire zone (fire protection partitions at a distance oi 70 m maximum in cable channels, 1.5hour

fire resistance for safety systems or vital safety equipment).

foam glass roof covers for NPP turbine halls have been worked out.

Group 3

The section work in Group 3 included the discussion of a number of basic questions, relating

to the behavior of radionuclides in the NPP emergency zone and outside (in the USSR and abroad).The main subjects were:

- the source term,

- atmospheric transport ai d sedimentation;

- secondary contamination effects and migration processes;

- physico-chemical forms and "hot particles".

In the discussion of the source term the results and data on direct release of radionuclidesfrom the reactor, on the present composition of remaining fuel, data on the fallout on the USSRterritory and abroad, and the tnathematical modelling were brought together and discussed.

Such multilateral discussion permittni a more detailed description of the source:

in gaseous-aerosol phase (without the noble gases) the atmospheric release, which causedthen the environment contamination was about 3-4/o of the core activity inventory;

in the first days after the accident the release reached tens of megacuries per day, in June1986- several thousand curiesperday, in the fall of 1986- tens of curies per day, atpresent- up to one curie per day;

up to 25% of Cs-137 core inventory was released to the atmosphere, and about one half ofthat was deposited in the USSR;

the previously reported release of 3 to 4% of the core inventory oi non-volatile fissionproducts represented the amount released offsite in the form of aerosols. The fraction of thecore inventory which was ejected in the form of very large aerosol chunks of core debris andpotions of fuel elements which were deposited on-site and very near the site has not beenquantified;

the iodine release has been variously estimated to range from 20 to 60% of the core inventory(the Soviet estimate is 20%). It is agreed that the principal form of iodine was gaseous, with

only 20% in particulate form. Estimates of the total amount of iodine released need furtherstudy, including V/estern analyses which had not been included in this work" hop.

Serious attention was paid to the study and the expert evaluation of the rigorousness of theprognostic methods for radionuclide behavior in the accident zone, their probable migration by airand waterways. New results were reviewed on the secondary wind suspension of the radionuclidesfrom the soil and the physico-chemical properties of the deposited radionuclides, with particularattention to the "hot particles". The distribution of various forms of deposited material in the 30-kmzone, other regions of the Ukraine and Byelorussia, and the infiuence of those forms on the soil and

water migration process were discussed in some detail. The group agrees that the review of these;questions was very comprehensive.

The estimates presented to the group which show that the radionuclide transport and

migration from the offsite contaminated areas are negligible at present appeared quite reasonable.However, further study is necessary for development of correct predictions for the distant future.

Recommendari ons

The following recommendations have been made (following the order of discussion, rather

than in their order of importance).

A. Source Term

Al. Measurement and study of the remaining long-lived radionuclides located within the

damaged unit should be performed to establish the fraction of the core inventory retained in

the plant.

A2. Detailed on-site contamination maps should be developed for the period prior todecontamination. This should include on-site distribution of graphite blocks, core debris,fuel fragments, etc., in order to establish the total amount of core deposited on-site. Highresolution photographs and dosimetry data should be utilized in the investigation.

A3. Workshop members of this working group (as well as other working groups), should takeadvantage of the contacts made to develop joint publications.

B. Forms and transport of radionuclides.

Bl. Detailed characterization of the physico-chemical forms of the deposited radionucl idesshould be developed, including solubility. Characterization of their transformationdepending on the geochemical factors should be expanded.

B2. In addition to detailed maps displaying off-site radionuclide contamination, maps depictingphysico-chemical forms in soil and water shou',d also be produced.

B3. The information developed as a result of recommendations (Bl.) and (B2.) above, wouldgreatly enhance the ability to build; develop and validate transport models.

This should be used in predicting the radiological situation in contaminated areas in the future.

Groups 4 and 5

Groups 4 and 5 had to meet in the same room, so the four co-chairmen decided to combinetheir efforts and to hold joint sessions. This was fortunate for all concerned, as many presentationswere overlapping in context between the two groups. As a result, almost 4 of the 5 days were spenton presentations and discussions of individual papers, and less time was available for generaldiscussion. The groups did identify some important problems that should be the subject of futurework and international cooperation. It was agreed that we should take maximum advantage of the

tragic accident at Chernobyl. Unlike the subjects of other working groups, much of the work in theareas of dosimetry and health effects still remains to be completed.

The first presentation was given by LA. Likhtarev on the subject of the Dosimetric and

Epidemiological Problems of the Chernobyl Accident. Dr. Likhtarev headed a group in theUkrainian Republic providing both short-term and long-term dose prognosis, with the final endpoint

being the forecast of the 70-y dose comn'iitment. The conservative nature of the forecasts was

discussed, partly as a matter of philosophy. One point made was that forecasting a dose that is lowerthan that which actually occurred was an unacceptable mistake, and the conservative factor of the

forecasts made following the accident has been between 2 and 10. The 12-rem forecast dose for the

next 36 hr in Pripyat submitted to the IAEA was conservative by a factor of 5 to 10. The forecast ofthe external dose for the first year was very accurate, with a conservative factor of 1.5. For internal

doses, the actual doses were from 5 to 10 times lower than forecast due to countermeasures; for the

second and third years, the conservative factor is from 2 to 3. A question raised is what dosesshould be used for epidemiology-the forecasted doses or the actual?

Another issue raised by Likhtarev was what should be the reference point for a health study?

It was suggested that it should be the general state of health, as much of the overall stress is not due

to radiation but to the accident itself. The way the person is feeling is an important parameter of

15

healtli, as even today there is a tremendous rcac ion rcgaiding thc way people feel followirig tliI:.

accident.

In response lo questions, it was stated that dc>se iecoiistruction» t'or 22,000 people have beendone. Some of the tools have involved mcasuri.rg the therrnoluminescence of'sugar in abandonedhomes and of building materials. Some autopsies liave also been done, and no hot particlr.s havebeen found in autopsy materials.

Another question specific;illy related tc th~'. dose to ihe thyroid was, was tire average dose tothe thyroid of 200,000 children aged 0 to 7 yc;irs really 1 Gy'? l.ikhtarev replied that thi» was amistake. The people wlio suffered the most v;ere tire children who received heavy doses to thethyroid «nd the young people taking part in the cleanup. f or tlie dose to children's tliyroid in the 8regions of the Ukraine, there were 6000 ch:ldren with doses ot n)ore than 2(N rem and 5(X) childrenwith doses of more than 1000 rem. Some early effects noted were high levels of thyroxine andhyperthyroidism in the first year following the accident.

I.V. Fi%uslr/in presented a paper on the Optimization of tne iVlethr)d of 13iological DosiinetryUsing Lymphocytes from Pcriphcral 13lood of Man. The main topic discussed was the great value ofbiological dosimeters and tire use of detecting dicentrics and rings in circulating lymphocytes. '1'his isan accepted technique for homogeneous irradiatioll of ttle body, but thc diagnostic value drops andthere are problems with inhomogeneous irradiation. Dr. Filyushkin stated thai tliey had taught theircomputer to unfold the dose distribution for inhornogeneous irradiation from examination of thedicentrics and other abnormalities The lower limit of dose is now 1 Gy, and, of course the techniquedoes not work for long-term exposure.

1'.JV. Sorolrin gave two short talks. One was on the Mobile Fraction of Fission Nuclides.This work involved a study of human hair sampled during the autumn of 1988. Sorokin reportedthat they found 0.5 to 1 Hq/kg of alpha-emitters in hair in Minsk, and 800 to 1000 Hq/kg in thesouthern area of Byelorussia. The second paper concerned the Determination of Alpha Particles; this

involved the study of cattle tissue. He reported 8,500 hot particles per kg of lung tissue. Thisamounted to 1 x10 Ci/kg of lung tissue (total alpha, beta, and gamma activity) and 1 x10 7 to lx10-~ Ci/kg of meat. These results were received with considerable skepticism, and it was agreed that

confirmation was needed.

O.A. Pavlovs/'y presented a talk on the Radiological Consequences of the Chernobyl

Accident for the USSR Population. Radiation doses were significant in regions far from the plant

within the European regions of the USSR. Calculations of dose commitment were based upon

measurem.nts of external gamma radiation dose rate and measurements of radionuclides in air, soil,

milk, meat, and vegetables. Anoth- r factor in the calculations was ar, environmental half time of 14

years for cesium. The calculated lifetime collective effective dose equivalent is 326,500 person-Sv,

of which 60% is projected to arise from external gamma exposure from deposited fallout.

M.l. Balonov gave a talk on Characteristics of Effective Countermeasures to Protect

Populations in Regions of Severe Radiation Accidents; this related to experience in thc Bryansk

region of the RSSR. About 2000 km were contaminated with >50 Ci/km2 of Cs and about

100,000 people live in this area. Extensive measures were taken here to reduce radiation dose to the

people. These included the removal of contaminated soil to reduce external exposure, the

decontamination of areas, and control of foodstuffs. The overall effect of these measures was a

16

reduction of the dose by a factor of 2, with much greater reductions being achieved at later timeswhen the measures were fully implemented.

V.F. Demin discussed problems in forecasting the health effects for the exposure of a largepopulation to relatively small doses. Primary concerns were the use of the absolute or the relativerisk models and the problem of determining the magnitude of a low-dose-rate reduction factor in

calculating the risk. These projections of risk are a real problem in terms of balancing them againstvery real risks, such as malnutrition from avoiding foods due tn fear of exposure to low doses ofradiation.

L.R. Anspuugh gave a talk on Methods of f.orecasting or Reconstructing Radiation DosesFollowing Severe Accidents. The methods discussed had b«en developed for the reconstruction olradiation doses to people living downwind of tire U.S. Nevada Test Site, where nuclear weapon-related devices had been tested in the atmosphere during the 1950s. The methods used proved to bequite general and were applied to the situation throughout the northern hemisphere following theChernobyl accident. Some important issues in such forecasts (over the very long tetm) are the rate ofdecrease with time of the external exposure rate, which depends upon the rate of vertical migration ofcesium in soil; proper values for the shielding and occupancy factors; the fraction of depositedmaterial that is retained on vegetation; the lone-term transfer of cesium to meat and fish; the effect ofthe consumption of minor foods like mushrooms, tea, and herbs that may have been contaminated torelatively high levels; and the effectiveness of countermeasures.

IV. Burkarr gave a talk on Techniques to Reduce Acute and Chronic Population Doses afterMajor Releases of Radionuclides. One of the concepts used by the Swiss is to educate the public sothat it can take care of itself during early times following an accident. It is also important thatpsychological stress be avoided as much as possible by rapidly providing realistic data to the publicfollowing an accident.

Brief summaries were presented of some of the health-effects-related research now underwayin the USSR in the aftermath of the Chernobyl accident. Environmental monitoring is continuing in

many areas where contamination levels warrant continued surveillarice.

'Some of the preliminary findings fKnizhnikov J suggest that the effective Cs half-life in milk

may be about 2-7 years, rather than the 14 years originally forecast. It was reported that the state ofhealth of the exposed populations is the most difficult aspect of the entire problem. There is a high

level of polarization between the public and the experts; this is exacerbated by the continuing media

reports of leukemias and cancers, of deformed animals, and of increased infant mortality. A

comprehensive data base has yet to be prepared that would permit a scientifically solid assessment ofthe allegations to be prepared. This is obviously a most significant objective for the scientific

community. Some early reviews of health statistics suggest that, for example, the infant mortality

rates in Gomel and Mogilev actually decreased by about 10-20% over pre-accident levels. There is no

clear understanding yet as to whether morbidity rates are cnanging significantly in the areas ofgreatest concern, and whether the closer post-accident scrutiny has influenced the data base. There

have been reports of increased levels of multifactorial diseases, and an increase of voluntary

abortions after the accident. These and ether questions should be the subject of objective review. The

social and psychological impacts on the affected populations continue to be one of the major areas ofpublic concern.

The "liquidators" (i,e., cleanup crews) received significant doses /Lazyu/'J. Early on, some

25,000 persons received doses estimated at 50-70 rem, plus some 250,000 whose doses have heen

estimated at 5-25 rem. There is still no complete validation nf the extent and accuracy of these doses.

It was also noted /Wald j that in a child/mother study of the population near the Three-Mile-

Island accident that the mothers were (and still are) disturbed and that psychological stress was

measured in the children as well. These reactions seem to be the product of the fear of adverse health

consequences, but not necessarily the result of having actually been exposed. He also gave a brief

summary of the new U.S. Nuclear Regulatory Commission radiation health-effects model

(NUREG 4214).

A psychological approach to underst mding and coping with radiation lears, "radiophobia",

was presented by Prcirre, along with suggested mitigation strategies employing educators, medical

and radiation specialists. This topic was discussed at several points during the week, and its singular

importance was agreed upon.

Regarding medical management of heavily exposed persons, it was suggested /Nenorj that at

less than 1 Gy patients can bc handled on an "out-patient" basis, that from 1-2 Gy, general follow-up

is needed, and those receiving 2-4 Gy need strict hematological follow-up. Intensive hematological

care is required for anyone receiving over 4 Gy. Early accurate dosimetry is essential, and a major

goal is the development of cheap, rapid biological dosimeters; the micronuclei technique may be a

worthy candidate.

Reports of the protective actions used by some countries were reviewed /Snihs J. 1here is no

universally accepted set of standards in this area, and where some of the standards existed, they were

changed in response to public pressure rather than because of newer knowledge. It was mentioned

that th" only criterion should be net benefit to the people and to society; further, the possibly

forthcoming ICRP dose limit reductions from 0.5 to 0.1 rem/yr for the public do not pertain to

accidents. One development in the USSR is the implementation of a 70-year, 35 rem limit for those

living in Cs-contaminated areas (Alexak/n'nj. Some are arguing for a 7 rem/70 year limit, even

though this is close to the natural background level. The cleanup cost had been estimated at about 4

billion rubies (some say 8 billion), but the Byelorussians have requested 17 billion for further work.

(This may exceed the entire budget of the Republic.)

A series of Soviet presentations concentrated on possible genetic implications of the accident

at Chernobyl. Pon~ranrseva summarized studies of rodents captured in the contaminated zones. No

"giant rodents" have been seen. Breeding trials indicated that dominant lethals were not evident, that

doses of 20 to 350 mrem/hour induced sterility, and that high exposures may increase the frequency

of translocations. Taskayev has studied plants exposed to 0,2 to 200 mR/hr and found some

morphosis in plants within the 10-km zone. Shevchenko showed some photos of somatic changes in

irradiated plants and suggested that doses in excess of 0.1 rad/day induced genetic consequences.

About 80% of the rodents died in and around the "red forest." Survivors demonstrate histologic

changes in pancreas and liver. Blood cell analyses on persons receiving higher doses (0-30 rem)

have been evaluated for cytogenetic scoring. The threshold for detecting change is about 4 rem.

Siornov used Tradescanria as a biological dosimeter in the Chernobyl area (exposed to 0.35 to 15

mR/hr and found good agreement with data from past controlled studies. Goldman showed the

progression of radiation effects on the nearby pine forest (the "red forest") as determined by remote

sensing from a Landsat satellite and analysis of resulting infrared images. Ryabov summarized someof the findings on flora and fauna in the close in area. Because of the absence of people in theevacuated zone, some of the herds of boar and deer have actually expanded several-fold. Fish caughtfrom the reactor cooling pond contain about 10 pCi t37Cs/kg.

General Conclusions

1. The methods used by scientists from the USSR and other countries to forecast potentialradiation doses appear to be quite similar, but there are some important differences in tho

choice of parameter values. The choice of these parameter values is of major importance in

forecasting radiation doses over very long time periods.

All present were concerned about the forecasting of health effects that might result from theexposure of a large population to radiation delivered at low dose rates and which would havea major dose component arising from internal exposure. The USSR population exposedfollowing the accident at Chernobyl is a unique resource that should be studied to address thisbroad and important question.

3. The problem of radiophobia is a very real one, and the public seems to l.ave adopted an

irrational approach in seeking to avoid even minuscule risk from radiation while acceptingmuch larger ririks from other sources. Efforts are needed to reach a community consensuson what risks from radiation exposure are acceptable under various conditions.

4. Based upon the best possible forecast models and the criteria of risk acceptance under various

conditions, a range of acceptable radionuclide concentrations should be developed for anyfuture accident.

5. Countermeasures can be very effective in reducing the population exposure. However, thecountermeasures can be very expensive and some of them also carry their own biologicalrisk. Acceptable and practical procedures are needed for optimizing radiation protection in

terms of balancing cost and risk.

6. In terms of radiation dose reconstruction, biological and other natural dosimeters are very

important. There is a need for the cross calibration of physical and biological dosimeters and

for the development of faster, more sensitive biological dosimeters.

7. Agricultural radionuclide transfer factors have been studied both in the field and in the

greenhouse. Typically transfer factors are much greater in the field, presumably due to the

influences of resuspension, rain splash, and other contaminating events. Likewise, some

countermeasures are much less effective in the field'han they are under greenhouse

conditions.

Recommendations

Considerable discussion centered on the real problems of communication to the public and to

authorities of what the correct information was available relative to the accident, as contrasted to what

was actually being done to contain and mitigate the consequences of the accident. It was agreed that

these are two separate tasks, and that a clear distinction was needed. The following recommendations

are each important, and a priority listing should include consideration of those studies in whichimportant information would be lost by delayed implementation. Resource limitations also play a roleand it was the consensus that high quality studies and validation were more important than ntorecomprehensive masses of data of lesser completeness.

There was general agreement that the most important need was for the conduct of credibleepidemiological studies of the population affected by the Chernobyl accident. It is essentialthat very focused and carefully designed studies of selected populations be started quite soon.Much might be learned of the risk associated with the exposure to substantial doses at

relatively low dose rates and where a major fraction of the dose is from the ingestionol'adionuclides.The conduct of a proper study would be greatly facilitated by a transfer of

technology from U.S. and Japanese scientists involved in the study of the Japanese survivors

of the atomic bombings. This technology transfer might be accomplished by extensive visitsof U.S, and Japanese scientists to the USSR and/or by extensive visits of USSR scientists tothe U.S. and Japan.

In order to assist in and enhance the credibility of epidemiological studies to be conducted,creation of an international advisory committee should be considered. The study designsshould include consideration of children who received high doses, of xvork«rs who receiveddoses after the accident, and of selected populations of the country including evacuees.

Conduct of a credible epidemiological study requires the best possible evaluation of the dosereceived by theaffected individuals. This can be facilitated by the joint consultation ofUSSR scientists and those from other countries on both physical and biological methods ofdose recons true tion.

O»e means of achieving agreement on methods of physical dose reconstruction would be a

joint reassessment of the collective dose to the USSR populace. This task can can startimmediately.

Much of the radiation dose from the Chernobyl accident has yet to be delivered. I'he dosethat actually will be delivered is being forecast based upon models for the long term behaviorof cesium in agricultural, urban, and native ecosystems. There is an urgent need to validatethese models by documenting the actual long term behavior of cesium in these differentenvironments. This can be greatly facilitated by an exchange of scientists among the USSRand other countries with established experts in this field.

Technology transfer and consultation should be implemented on methods of biologicaldosimetry. Further research is needed and should be conducted in this area, so that more

sensitive, reliable and accurate methods can be developed.

Joint work should be done to define what levels of risk from radiation are acceptable under

different conditions. Levels of acceptable risk must then be translated to levels of dose and ofconcentrations of radionuclides in food.

In the event of future accidents, the methods of dose prognosis are of great importance. Joint

efforts in this area are highly desirable. As noted above, a useful and practical exercisewould be a joint reassessment of the dose to the USSR population from the

Cl;ernobyl accident.

20

9. Great concern was expressed about radiophobia, the greater than justified fear of radiationfollowing the accident. It was agreed that an international workshop on this topic should beheld, and that this workshop must include psychologists and journalists. Radiophobia wasalso seen following the accident at Three Mile Island in the U.S.

10. There are many individuals and organizations involved in dosimetric and health-effectsstudies in the USSR. It is confusing to foreign visitors in that lines of responsibility are notobvious and that responsibilities may be limited geographically. It would be helpful in

expediting interactions if an inventory of scientists involved in the post Chernobyl studiescould be prepared. Such an inventory should include the nature of their study, their parentorganization, and their responsibility within the framework of scientific discipline and

geographical coverage.

11. A health-effects model should be developed in conjunction with the epidemiology studiesrelated toChernobyl. This model should include resolution of problems of relative andabsolute risks, dose-rate amelioration factors, non-homogeneous exposure, age and theunique problems of thyroid and neonatal exposure. Quantification of uncertainties and asensitivity analysis should be integrated with the model. The role of theoretical approaches tosuch modeling should also be included.

12. The accuracy and ultimate sensitivity of remote-sensing techniques as a means to quantify anddelineate radiation effects in plants should he determined using the Chernobyl data bases.

13. Medical management of combined injuries should be reviewed and recommendations forclinical treatment and evaluations should be developed.

14. Evaluation of the qualitative and quantitative aspects of hot particles in the nvironment and in

humans should be undertaken.

15. Development of critical studies of genetic dose response in specific plant species should beconsidered.

Finally, the group was very pleased at the level of communication and regards the Dagomys

workshop as a good start of an orderly process rather then a single event. The same opinion was

expressed by nearly all the workshop participants.

NUCLEAR POWER

IN THE USSR ZNZRGY PROGRAM~

PRO 5 CONS NUCLEAR POWER

FOR:

RESOURCES 5 CONSUMPTION 'ZZRRITORIAL DISPROPORTIONS

LARGE SHARE OF OIL AND GAS IN ELECTRICITY GIZERATION

NEGATIVE ECOLOGICAL ASPECTS OF COAL P~JSTS

HIGH POWER INDUSTRY DIWZIDPMENT RATE

AGAINST!

INCREASED SAFETY REQUIB~WTS

RAISE IN COSTS

LONGZR CONSTRUCTION SCHZDULES

PUBLIC OPINION

"CHERNOBYL ACCIDEt'lT. CONSEQUEtJCES ANALYSYS AIG) MITIGATION"

PROGRAJNE

"E'JTOMBMENT" STUDIES

FUEL LOCATION AND STATUS

SAFETY ASSURAt JCE

FUEL COMPOSITION AN) BEHAVIOUR DYNAMICS

STRUCTURES AND MATERIALS BEHAVIOUR

METHODS RJD TECHNIQUES TO STUDY FUEL AND MATERIALS

COMPUTER MODELLING OF '. M ACCIDE1'JT PROCESS

2. RADIOACTIVE RELEASE STUDIES

HOT PARTICLES

CAESIUM "SPOTS"

3. RAD IOECOLOGI GAL CO1'JSEQUENCES

RADIONUCLIDE BEHAVIOUR IN THE ENVIRONMENT

RADIONUCLIDE MIGRATION MODELS

ECOLOGICAL CONSEQUENCES OF EPlIHONMENTAL CONTAMINATION

4. 1'CBH DIAGNOSTIC METHODS A1'JD TECHNIQUES

FIELD SPECTROMETRY

EXPRESS RADIOCHEMISTRY

NUCLEAR RADIATION VISUALIZATION

!

4-

~i'~r.'i a'

1 C

I

i'rj„.'SI

eriv»rA

4

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't,

OUTSIDE THE USSR TERRITORY

AT TH."". USSR TZRRIiORY

I

OFF SITE TERRITORY

( 80 KM AROUND )

I

1

I

~ CHiRNOBYL NPP

SITZ''iSIDZ

THE 4-th UNITY

THE DISTRIBUTION OR FUEL WHICH WERE

MQZASZD FROM 4-th UNITY REACTOR CORE

STRUCTURE OF FUEL EXTRACTION ANDPRIMARY ENERGY GENERATION

100%- ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ P ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ \ \ \ ~~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~~ ~

90~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

OTHERS ""~ ~ ~~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~~ ~ ~ I ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~

HYDRONUCLEAR

~~~/~ ~~ ~~ ~

~ ~ ~ ~ ~~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~

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~ ~ ~ ~ ~ ~ ~ ~ ~ ~ I ~ ~~ ~ ~ ~ ~~

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50»

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193Q 1940 1990 2000

STAGEI THIRD STAGE'

I

'1950 1960 1970 1980

FIRST STAGE I iSECOND

2010

5 oO

CllQl

—Oz

me 200~~300

12

O O5 3m 33 tll

Q 0z2

I

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( rm r lO

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200

425

——417417—=417

wNe% ~417r~ 4401590

470

~~~~012440

1420

5000

1000 ~

Data about environmental contaa~4ion by Cs.

Total amount is (1.5-2)VER.

ht the territory of USSR(exept of territory of Chernobyl NFZ)

mast of fallout is 40K.

Sarface of territories with density of contamination (5 f5) LKi/1mP

agan>ximately 2.1*"0 knP.

Surface of territories with density of contamination more than15 lKi/laP!

~e'ussia SPSR -Zest OMRr Ukrain SSR -1.5*10knP

Rr Belorussia SSR-7*10 laP

Qthe territory of Europe(exempt of territory of USSR) amount ofbeaut is 35%.

MIUm density in the "spots":

PrSw eden —2 Ki/kP;P~iustria,NI.Cer;may and. Switzerland — (1-1.5)Ki/EaP

PASSIVE COOLFD

EXCI IANG F Il

PASSIVE COOL

sara vI

CUMULATOll TANKI

IQD DINE

F ILT E'l

I''Y AEnOSOLl

~ I I l [ [~ F ILTEIl

rnESSUIlIZEn

- STEAM GENI

'I EBATOIlI

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////////STACK

STEAM GENEIlATO

MAIN

CIRCULATION PUM

...,l5iii~ II

~~~"i%<"i7~ ~

WiiiiiiiZiiiiE/ill

I I ~PPPPIPIPYi I

~ COIlECATCfIFR

TECHNICAL MEASlJKZ TO IMPROVE VVER-88 SAFETY

DECREASED PROBABILITY OF REACTOR CORE DISRUPTION

NUMBER 3F CONTROL RODS 61 + 121

240 ~ 100 C

FAST BORON INTRODUCTION INTO PRIMARY COOLAiiT

PASSIVE HEAR REMOVAL FROM PRIMARY CIRCUIT

CONTAINMEF1 OF LEAK FROM PRIMARY TO SECONDARY CIRCUIT

NEW DESIGN OF MAIN CIRCULATION PUMP SEALINGS

SEVERE ACCIDENT CONSEQUENCES MITIGATION

HYDROGEN MONITORING AND IGNITION

FILTERED RELEASE OF STEAM AND GAS MIXTURE

CORE CATCHER

SAFETY IMPROVEKENT

OF ZXIS'IING NUCLEAR POWER PLANTS

SAFETY ZAPROVEMENT HZP~URES

GE6RA~i MEASURES:

PERSONNEL: TRAINING'IMJLATORSy EXAMS

OPERATION AND REGlP ATORY DOCU?KNTATION

ADDITIONAL ANALYSIS OF DESIGN-BASIS AND SEVERE ACCIDENTS

EMERGENCY PL BRING

RB?K REACTORS:

DECREASE OF VOID EFFECT

FAST-ACTING EKZRGZNCY PROTECTION SYSTEM

ADDITIONAL 3%ERGENCY PROTECTION SYSTEM AIJBMS

DIAGNOSTICS OF METAL, ZQUPKENT AND KP'tJTRONS

VVZR REACTORS:

DIAGNOSTICS OF THE VESSEL, TUBE AND STEAM GkÃIZATOR METAL

REACTOR VESSRI RADIATION EMBRITTt'>~Vl AND ANNIMING

TRANSITION TO T?GKf.'-YEAR FUEL CYCLE

NUCLEAR POWER Ql THE USSR

INSTALLED CAPACITY (APRIL 1989) 34,9 GW(e)

IN OPERATION UNDER CONSTRUCTION

No OP UNITS CAPACITY

GW(e)

No OP UNITS CAPACITY

G7I (e)

VVER-440 3.8

VVZR-1000 14 14.0 25 25.0

RBMK-1 000 13 13.0 2.0

RBMK-1500 3.0 1.5

BN 0.8 1.6

AST 2.0 (thermal)

OTHER 0,3

NUCLEAR POWER AND PUBLIC OPINION

STATUS!

NZGAT'IVE ATTITUDE AS A RESULT OF 'DiZ CHERNOBYL ACCIDENT

iUINSK ASTe FRASNODAR NPPs IGNALINA NPPi KR'KY ASTe

VORONZZH AST... DEMONSTRATIONS, LZlJ.'KHS, PUBLICATIONS)

FEATURE: OPZNESS, LACK OF 3Ãl'ERACTION EXP.rAIENCE

AIM:

FORMATION OF POSITIVE ATTITUDE TO NUCLFM POWER SINCE

THIRE IS NO ALTERNATIVE

APPROACHES:

PUBLIC OPINION ANAIYSIS

LEGAL FOUNDATION FOR NUCLEAR POWER

FORMA'» iON OF ATTITUDE

MASS MEDIA (TV y RADIO ~ PUBLICATIONS )

DISCUSSIONS

EXIBITIONS

INFORMATION

'NT: 3-I21DUSTRIAL PUBLIC RELATIONS COUNSIL

PUBLIC ORGANIZATIONS (NUCLEAR SOCIETY)

INTERNATIONAL COOPERATION

NEW GYRATIONOP NUCLEAR POWER PLANTS

IhPROVZD SAFETY NUCLEAR POWER PLANT DESIGNS WITH BEJ.J.'E8 TECHNICAL

AND ECONOMIC PARAK."ZERS

VVZR:

VVZR-.92 DESI N = VVER-88 + (POWER INCREASE, LlszxIME,FUEL CYCLE, ADDITIONAL SAFETY)

VVZR-500 DESIGN

AST:

DISTRICT HZATING

BN:

BREEDING, Na

HTGR: ELECTRICITY, HEAT/STEAM, PROCESS HEAT;

INHERENT SAFETY, He

MODULAR VGM DFSIGN (250 MW (t) )

INTEGRAL VG-400 DESI(N (1000 MW (t)) IN PRRSTRESSZDCONCRETE

SMALL POWER NUCIZAR PLANTS (50-100 MW (e) ) FOR REMOTE REGIONS

LESSONS FROM CHERNOBYL

AS VIEWED BEFORE WORKSHOP

John M.West

In my opinion the following sequence of major events connected

with the accident at Chernobyl-4 is resonable and consistent with the

facts.(1) Prior to closing the inlet valve to the turbine to start the

experiment, the reactor was operating with very little steam in the

fueled tubes because of low power level and very high coolant flow.

The reactor also had a highly positive steam void coefficient of

reactivity at the start of the experiment.

The control and safety rods were positioned such that they could

not subtract reactivity rapidly and might even add reactivity during

their initial inward movement.

(2) After the closing of the turbine inlet valve, those main

coolant pumps which were connected to the generator began to coast

down. The decreasing reactor coolant flow resulted in additional

steam formation in the reactor, an increased reactivity, and a higher

power level. This was a rather classic case of positive power

coefficient - with an increase in power producing still more power,

etc. However, in the case of Chernobyl-4 this situation was aggravated

by the fact that the coolant flow was decreasing because of the

coostdown of the pumps, plus the fact that as steam content increased,

the hydraulic resistance increased, thereby decreasing the coolant

flow more rapidly then would have occured from pump coastdown alone.The above factors caused the power leveJ to increase

substantially, but apparently not enough to cause melting of the fuel.The automatic control system helped limit 'e magnitude of the

increase in reactor power level and he reactor had not reached prompt

criticality.(3) When the coolant flow decreased to a set point, at least

those four pumps which were drawing their electricity from the

generator lost their power througn an automatic disconnect. A very

rapid decrease in coolant flow then occured, with a consequent sudden

increase in steam content of the coolant and a large increase in

reactivity, making the reactor supercritical on prompt neutrons alone.

The power level rose very quickly under the prompt criticalcondition, creating more steam quickly and increasing the pressure in

the fueled tubes beyond their rupture strength.

The sudden increse in pressure within the coolant tubes caused

the main coolant check valves to close. This stopped coolant flow

altogether and caused the steam valume fraction in the core toincrease even faster - and probably to a very high fraction.

(4) The rupture of coolant tubes allowed water/steam to be

injected into the hot graphite, providing a pressure which lifted the

cover above the reactor and ruptured all coolant tubes at their up

ends. Fuel and graphite were ejected from the reactor as the cover was

blown off. At this stage the reactor probably became subcritical,although this is not known for sure.

(5) As the pressure in the fueled tubes relieved itself through

the ruptured upper tube ends, the check valves in the major coolant

inlet lines opened, allowing water to enter the tubes. The

vaporization of this water by molten and fragmented fuel caused

another pressure pulse wich propelled additional fuel into the

atmosphere. This could have been the second explosion which people

reported hearing.

Several changes have been identified by USSR experts which should

go a long way toward assuring that a Chernobyl-type accident will

never occur again. Increased U-235 content of the fuel and use of

poison within the reactor should make the steam void coefficiant of

reactivity much less positive (hopefully negative) at all operating

conditions. Safety rods will drop much faster. Control rods and safety

rods will never add reactivity, even during the initial parts of their

insertion. Instrumentation will give a much better picture of

conditions throughout the reactor than heretofore. Simulations will be

used to show dynamic conditions. Procedures will be improved, and

adherence to procedures will be mandatory. Training of operators will

be improved. All of these changes are worthwhile.

A major concern I still have is that all possible types of

accidents have not yet, to my knowledge, been analysed More attention

should be given to coolant tube ruptures and breaks in the coolant

outlet pipes, as well as the steam drums. All credible failures and/or

malfunctions in the external hardware should be analysed to determine

whether they could trigger a serious accident and, if so, what can be

done to prevent the failure or to mitigate the consequences of the

falure. Similiar analyses should be made of all instrumentation and

control systems.

I am also greatly concerned that "containment" is not provided

against ruptures of hardware in the exit coolant lines from the

reactor. Providing such containment for reactors of the Chernobyl type

appears to be difficult, but this problem warre»ts a iot of thought.

I wonder whether a number of things wich perturb the reactor can

be slowed down while still having a practical power plant. Can the

capability of control rods for adding reactivity be decreased Can

the speed with which feed water valves operate be decreased: Can the

speed of change in power level be reduced?

How many pumpes were disconnected from their power supply during

Chernobyl accident? How did this disconnect occur? Can this occur

sometime again when it would be hazardous?

How does the safety of the 1500MWe version of the RBMK reactorcompare with the 1000 MWe version?

In my opinion, procedures should call for scramming a reactorwhenever there io the slightest doubt about safety. The reactor should

be scrammed when the inlet value to a turbine suddenly closes. There

are other examples where a scram should be initiated rather than

merely a reduction in power.

I would appreciate a .discussion of the pressure suppression

systems on RBMK reactors.

Comments on the sequence of events I have postulated for the

accident would be appreciated. I have many more questions, some of

wich will probably be answered in the papers which are to be presented

in the workshop.

I suggest that as an overall objective the RBMK reactor be

designed and operated such that prompt reactivity is impossible

despite any equipment failures or human errors. A further objective

should be to reduce the probability of an accident producing

significant fuel damage to an extremely low value and to prevent the

escape of harmful amounts of radioactivity to the environment.

IST INTERNATIONAL MORKSHOP ON SEVERE ACCIDENTS AND THEIR CONSEQUENCES

Estimating the Temperature of Damaged Reactor Core Componentsat the Time of Explosion

Authors

P.A. PlatonovV.I. KarpukhinF.F. 2herdev

V.H. NarkushevV.N. Bespalov

Studies were made of the physicomechanical properties of the graphite, as wellas of the distributions of the concentration of radioactive isotopes over thecross section of fragments of blocks of masonry, sampled at the plant site. ofthe core of unit No 4 of the ChAES [Chernobyl Nuclear Power Plant).Considerable mass transfer of fission isotopes from the surface into theinterior of the graphite was detected.

The distributions of the concentration of fission fragments were treated asthe consequence of thermal-activated processes of the migration of cesiumisotopes and fuel aerosols. In so doing it was assumed that the masonryfragments ejected from the reactor at the time of the explosion initially hada high temperature and surface fouling by the dispersed fuel, and cooled offin an atmosphere containing radioactive aerosols in considerable concentration.

Studies were made of the parameters of the diffusion of cesium in the graphiteof an RBMK [uranium-graphite channel-type reactor] in the temperature range of

.900 to 2000 'C.

The results obtained were used for estimating the maximum temperature of thereactor's masonry blocks at the time of the explosion within the framework ofthe assumption of mechanisms for the transfer of cesium in the graphite bymeans of diffusion and thermal transpiration.

Results of Study of Graphite of Masonry of Damaged Reactor Core

During 1986 associates of IAE [Institute of Atomic Energy imeni I. V.Kurchatov] made several visits to the site of the ChAES for the purpose ofcollecting fragments'of core structures of unit No 4. The accumulated

observations and ar. analysis of the evidence of eyewitnesses made it possibleto note certain particular features of the explosion.

Fra m nts of graphite blocks collected on roofs were deeply fused into theasp act, which in a number of cases acquired a characteristic vitreouscarbonized structure directly around the graphite.

At the time when the platforms of the vent stack were being cleaned up, blockswere discovered that were "welded onto" the steel floor (pig iron forms in thearea of contact of graphite with steel with T 0 1000 'C ).A piece of a TK [fuel assembly] discovered on the roof of the gas cylinderSAOR [emergency core cooling system] building included a lower end transitionpiece and was identified by number as belonging to the peripheral region ofthe reactor.

Small pieces of graphite that fell near the south wall of the turbine buildingwere hot approximately 20 minutes after the explosion (according to theevidence of firefighters).

Witnesses of the explosion and firefighters indicated the glowing of fragmentsof graphite ejected onto roofs.

l(1

Taking into account the fact that the temperature of the~(reactor's graphitemasonry directly before the explosion was approximately 300 'C. the factsenumerated made one concentrate one's attention, in studying the collectedgraphite fragments, on attempts to reconstruct their temperature at theinstant of the explosion.

A plan of part of the building structures of the ChAES adjacent to the damagedunit is presented in fig 1. with the places where masonry specimens werecollected marked. Sketches of certain fragments and a reconstruction of theirposition in a full-size block are presented in figs 2 and 3. The surfaces ofthe pieces of graphite had visible traces of high-temperature corrosion,although the probability of sooting cannot be ruled out. The gamma radiationexposure dose rate at a distance of 0.5 m from the fragments did not exceed 1to 5 R/h (according to the status for June-October 1986).

About 60 percent of the graphite fragments represented parts of short blocksor of plugs of the reactor's reflector. In almost all the remaining cases(with the exception of a block with a fire-contact ring wedged in it) thequestion of the location of the original block in the reactor remained open.The following studies were conducted in order to peg fragments of blocks totheir location in the reactor.

As is known, reactor graphite changes its physicomechanical properties underthe effect of fast neutron irradiation, and the character of this processdepends on the temperature, Hence, .knowing the distribution of thetemperature and neutron flux over the height and radius of the core and thephysicomechanxcal properties of the ejected graphite, it is possible to try toisolate the probable region of the reactor to which a given fragment belongs.However, the actual situation is complicated by several facts:

f I

I)1 0I<

Area of principaleiection

Figure l. Diagram of Buildings Adjacent to Damaged Unit of ChAES:

The numbers indicate the places where graphite fragments

were collected.

Block 10

Block 12 Block 13

Figure 2. Sketch of Preparation of Fragments of Blocks WhenReference Diffusion Specimens Were Prepared

Block 14 Block 15

Sketch of Preparation of Fragments of Blocks When ReferenceDiffusion Specimens Were Made

Reconstruction of Graphite Blocks According to Location of Fragmentin Full-Size Block

Figure 3.

The natural spread in the values of the initial properties of the graphite.

The smallness of the accumulated fluence, which results in only a slightchange in the majority of the parameters of a material.

The possible annealing of the change in the material's properties, if thematerial's temperature reached a level of 1000 to 1500 'C in the process ofthe accident.

For unit No 4 of the ChAES, which had worked for two years, the maximumfluence to the material of the core masonry equaled 1.5 ~ 10» cm

'E

> 0.18 MeV). This value was substantially lower (approximately5 ~ 10» cm e) for blocks of the reflector. Here the temperature of .thereflector's blocks was in the range of 200 to 300 'C, and in the case ofblocks of the core was approximately 400 'C for the inside surface andapproximately 500 to 600 'C for the outside surface. The values presented intable 1 constitute the expected changes in certain structural physicalcharacteristics of the graphite.,

Table 1. Change in Certain Structural Mechanical Properties of RBMK

Graphite Under Irradiation

InitialValues Reflector, T = 350

to 450 'C, F =~ 0.5 ~ 10~'m

Expected ValuesInside Surfaceof Core, T 350to 450 'C, F ~

1.5 ~ 10» cm ~

Outside Surfaceof Core, T = 500to 600 'C, F == 1.5 ~ 10~'m

CutoutDirec-tion

Resist-ivity,

10-e0 m 10 13 22+ 2 29+ 3 30+ 3 3S+4 2Q» 3 31 «3

Coeffi-cientof ther-mal ex-pansion,

10-eK

4.5«0.1 5.8i0.1 4 '+0. 1 5.9~0 1 4.6+0.1 6.0+0.1

Young'smodulus,GPa

10,1+0.8

7.0+0.6 12,0« 8.8+0,7 9,8+0.8 6,5+0,11,0

In order to study the actual state of the material, specimens were cut outfrom fragments of blocks in directions parallel and perpendicular to theitem's molding axis and studies, were made. of certain of their

physicomechanical properties. The values of the expected changes inproperties of the material and their actual difference from the initial valuesaveraged in terms of a great number of lots of RBHK graphite are compared intable 2. It is obvious from the results that a partial or complete return ofthe material's properties to the initial state took place.

Table 2. Comparison of Expected and Actual Changes in Properties of Graphiteof Blocks of Damaged Reactor

Physical Parameter ResistivityPercentage Change Expected ActualCutout

J «(>Block 1

120~ 120+ 40 75200 170

Block 2 150 170

Young's NodulusExpected Actual

Zb~db 15 76 ZO 4

ac 50 30

Block 3

Block 4

Block 5

Block 6

Block 7

120 120 0

56

0 6

25

15 26

B2

24 12

31 55 40 30 0

The results of studies made earlier of the annealing of irradiation-causedchanges in the properties of graphite show that the change in resistivity isthe "hardest to anneal." This parameter relaxed to the least extent also inthis case, which made it possible to perform additional isochronous annealingof specimens with in-between measurement of their resistivity. The resultsare presented in the g. aph in fig 4. In combination with the data of table 2,they can be regarded as an indication of the fact that the deviation of thematerial's properties from the initial mean values is the consequence of anincompletely annealed effect of previous irradiation. The active return ofthe resistivity to the initial value in specimens of block No 2 (specimens Nos1, 12, 4 and 15) is possibly explained by the fact that this fragment was ofthe smallest size and cooled off most qt.ickly.

Taking as a hypothesis the graphite's I igh temperature after ejection andtaking into account the results obtained, it can be estimated for the majorityof fragments as exceeding 1000 'C.

The probable high temperature of the graphite blocks at the instant ofejection could have resulted in the mass transfer into the material ofradioactive fission isotopes from fuel particles and aerosols in the air thatwere deposited on surfaces. In view of this, studies were made of thedistribution of fission fragments over the cross section of availablefragments of blocks .

~ - Pap Pe

2.IO J.J.J.J.J.J.J.

J.J.J.J.J6—

IO 3(~ m a ie m sr ag

I

( Blocks 7, 3, 5

~Blocks G, 7

-=~@~Blocks 4, 6

I

500I

IOOO

I

I

I500

Figure 4. Relaxation of Resistivity of Graphite of Blocks of DamagedReactor With Isochronous Annealing: Annealing time —20minutes. Ranges of value; of the resistivity of thenon-irradiated material are marked on the graph.

The initial diffusion specimen was in the form of a cylinder 10 mm in diameterbored out of a piece of graphite by means of a tubular cutter. The boringaxis was chosen so that the end faces of the specimen would correspond togeometrical surfaces of the original undamaged block and so that models of anunbounded plate, cylinder and tubular cylinder could be used in themathematical description of the process. After it was bored out, the initialspecimen was sawn perpendicular to its axis into 20 to 35 disks 1 to 2 mm

thick, which were sent for gamma spectrometry. Some distributions, obtainedas a result, of the specific activity of isotopes along the axis of theinitial diffusion specimens are presented in the graphs in figs 5 to 9.

An analysis of the distributions obtained shows that they do not form auniform picture. The main repeating distinctive feature of the results is thefundamental difference between the concentration profiles of. isotopes ofcesium and the remaining detected fission isotopes.

It is also possible tc note a group of activation products of natura]impurities in the graphite: Sc-46, Co-60 'e-59, Cr-51 and gf-181.

Examining the group of fission isotopes not containing cesium, the followingrelationships can be singled out,

10

10

10

Ql

o 1O

10 ~s

10 tt—

0.00I t I I i I

10.00 20.00 30.00 40.001Tlm

50.00

5'0L4J

C

60.00 70.00

Figure 5 Distribution of Concentration of FissionSection of Block No 7 (Block of Reflector

Isotopes Over Cross

10 o —Ru-103

10

10

C.a10~

U10

10 tt

10 lt

0.00

0

r0 +J~4

I I I I I I

10.00 20.00 30.00, 40.00x, 'mm

4J

O

50.00 60.00 70.00

Figure 6. Distribution of Concentration of Fission Isotopes OverCross Section of Block No 5 (Block of Reflector)

10

10»

U 10'1m

fU

V)

10

10

p pA Cs-134b O

7 .000.00 10.00 20,00 30.00 40.00 50.00 60.00 0x, mm

Figure 7. Distribution of Concentration of Isotopes of Cesium OverCross Section of Block No 10 (Core Masonry Block>

10

10

10

Ql

10D

10 ~-

10 c

h~kA,~

Cs-134 I'uAJ

D

10 ~ i ~ s I > t > < s s I s > s I s I ~

0.00 10.00 20.00 30.00 40.0D 50.00x, mm

60.00 70.00

Figure 8 Distribution of Concentration of Isotopes of Cesium Ov'er

Cross Section of Block No 7 (Block of Reflector)

10

IIII

II00 (

10Ql

Q

V 10 '1

10

II <II ~ 4( I I I »III>I II ~ I

0.00 10.00 20.00 30.00 40.00 50 00X, I71n1

60.00 70.00 80.00 90 00

Figure 9. Distribution of Concentration of isotopes of Cesium OverRaciius of Reflector Plug

The center portion of the distributions, most remote from the surfaces of theblock, is in the form of a more or less pronounced plateau whose existence canbe explained by the fission of impurity uranium in the graphite, for which alevel of its concentration of 10 t % (at.) is sufficient. At the same time,it is possible to distinguish regions of distributions having a length of 10to 20 mm from the surface into the graphite's interior, in which an impurityconcentration gradient is realized (cf. figs 5 and 6).

The di'stributions of the specific activity of the isotopes Cs-137 and Cs-134are of a totally singular character. Two groups of results can be singled outhere: 1) those that fit into the solution of the Fick equation under theappropriate boundary conditions (cf. the graph in figs 7 and 8): 2) thoserepresenting a diffusion-spread square wave of penetration of an impurity fromthe surface into the interior of the graphite (cf. the graph in fig 9).Within the'ramework of the hypothesis concer1'ling the high initial temperatureof the graphite blocks, let us examine the probable process of mass transferaccompanying the cooling-off. The source of the diffusant was, in part,microparticles of fuel deposited on the inside surface of the blocks at thetime when the fuel elements ruptured, and mainly two types of aerosols —of the

'uel and cesium —that saturated the air in the vicinity of the damaged unit.Here it is tc be expected that the fuel aerosols represented a suspension ofmicroparticles depleted by volatile fission fragments. Cesium was probablyfound in elemental form or in the form of a hydroxide or oxide.

When the graphite cooled, the gas contained in its pores volatilized. heingreplaced by air, which caused the transpiration of sols of both ty(>es into thebody. At the same time, the fuel sols were effectively filtered in the near-surface layers of the graphite. Oxide compounds of cesium decovr>ose when theycome into contact with graphite heated above 500 'C. According to theoreticalanalyses, a cesium atom behaves as a free particle near the surface of thebasal plane of a single crystal of graphite but can b> sorbed whenpolyvacancies are present in it. Events of sorption or t'.e formation of weakchemical bonds on the discontinuity surfaces of a sf igl~ crystal are alsopossible. It is also known that the rate of the desor.".>tion of cesium ingraphite has a broad peak with a maximum at T;900 'C (as published], An

additional desorption peak appears at 320 'C in the presence of oxygen. Inintervals between acts of sorption the cesium must be transported by thetranspiration flow of gas in the pore. It is not ruled out that the"transpiration" concentration profile of the cesium will be "frozen" inregions of the graphite having at this time a temperature agreeing with theneighborhood of sorption maxima. This will result in the appearance of a wavyconcentration distribution.

At the same time, diffusion, initiated by transpiration flow into dead-endpores, must take place in the graphite over the body and boundaries ofcrystallites. Thus, the actual concentration distribution profile, with arelatively low temperature and short process time, when cross flow betweenmigration channels has not yet smoothed the picture, will approximate the sum

of the individual distributions.

The cooling blocks in some cases lay with one surface on molten asphalt, whichruled out transpiration. The concentration distributions near this surfacewere produced entirely by solid-state diffusion (cf. the graphs in figs 7 and8, the right branches of the distributions).

For a calculated estimate of the graphite's temperature at the time of theexplosion it is necessary to know the parameters of the diffusion of cesium inRBMK graphite. Appropriate laboratory studies were conducted in view of this.A drawing of the specimen-can used in the experiments is presented in fig 10.The inside surface of the specimen's removable bottom was covered with agastight layer of molybdenum carbide. In preparation for diffusion annealinga weighed portion of fuel (collected at the site of the ChAES) crushed in amortar was put in an even layer on the molybdenum-coated bottom of the can andwas secured with a thin layer of cement. The can was put in the assembledstate into the muffle of an electric furnace, and after evacuation themuffle's space was filled with argon. Diffusion annealing consisted inrapidly (10 to 20 minutes) raising the can's temperature to the planned level,shutting off the heat and cooling down the diffusion assembly. The can'stemperature was constantly monitored and recorded in the process, Assemblieswere annealed under vacuum according to this procedure with maximum

temperatures of from 1240 C to 2050 'C and in argon with Tmax ~ 900 to1400 'C.

I

2

Figure 10. Specimen-Can; 1-thermocouple channel; 2--irradiatedfuel; 3—layer of molybdenum carbide

When the annealing had ended the bottom of the can was removed and the initialdiffusion specimen was made from the center portion of the specimen by axialboring witn a tubular cutter. The subsequent procedure for studying thedistribution of the diffusant is described above.

The purpose of the experiments was not to work out thoroughly the mechanism ofthe mass transfer of cesium in the graphite, but to obtain the effectivediffusion coefficients under conditions simulating possibly more closely themain process being examined, ln .particular, the thermal-shocked fuel of thedamaged reactor was used as the source of the diffusant for this reason.

A mathematical analysis was made within the framework of a model of diffusionfrom a steady source into a semi-bounded space. Here it was assumed that thefrequency factor and activation energy were constant over the entiretemperature range covered in the experiments, and that the temperature wasconstant over the specimen's volume and varied over time in accordance withthe experimentally observed relationships. The experimental material wasoptimized from the viewpoint of obtaining the best values of the frequencyfactor and activation energy. The following values were gotten: D0

0.3 cm'/s and E ~ 19,140 cal/mol

The following probable mechanisms for the formation of the distributions offission isotopes in the graphite were examined in making the final calculatedestimates: 1) thermally activated diffusion; 2) transpiration: 3) penetrationof pores from the explosion wave.

The following approximations and assumptions were made in the calculation:

At the instant of e)ection the block fragments were heated to a certaintemperature constant over their cross section.

The cooling of the graphite occurred on account of heat radiation and naturalconvection.

Block fragments approximating a plate in terms of their geometry cooled offwith one of their surfaces lying on a layer of asphalt 25 mm thick under whichthere was a layer of concrete l00 mm thick . In one calculation version suchfragments cooled off with both surfaces radiating heat.

The temperature dependence of the material's thermophysical parameters, takenfor the case of unirradiated graphite. was taken into account in calculation.This approximation results in some overstatement of the reference initialtemperature.

A model of diffusion and transpiration from steady sources was used incalculation.

The majority of the simplifications used could easily have been bypassedwithin the framework of the computational methods employed, but this wouldhave resulted in a cascade-like growth in the tree of optimization calculationoptions.

The heat conduction equation is written in dimensionless form for the casesexamined:

where v = 0 or 1 for a plane and axisymmetric cylinder, respectively;C's

the heat capacity of the graphite; C = C'/C0 (here and below, physicalparameters having a zero subscript correspond to the initial temperature ofthe graphite); r's the running coordinate; r r'/R0 'R0 is thethickness of the plate or the radius of the cylinder; 2's the thermalconductivity of the graphite; A 8'/A0 , T's the temperature; T

T'/T0 'v is time; t t/t0 , t0 C0R6p/A0 . and p is the graphite'sdensity.

The initial condition is T(r, 0) T0 .

The boundary conditions for a plate are:

where s is the graphite's emissivity factor; Sk is the Stark number;Sk TgeR0/A0 , c is the Stefan-Boltzmann constant; Av's the thermalconductivity of air at temperature Tg , Av Av'/A0 . Tg's theair temperature remote from a block; Tg T'/T0 ; Nu is the Nusselt number;

0.25Hu=0.47 Gr

if

and Gr is the Grashof number at temperature Tg

Por a cylinders 2 x R0 is used as the characteristic dimension in theNusselt number, and the maximum linear dimension of the actual body, for aplane.

In the case of a cylinder, the condition of symmetry,

BT/Br 0 with r 0 ,

arises in the boundary conditions.

The problem was solved numerically using a three-point implicit scheme with an'order of approximation of 0(r + h~) , where i is the time step and h isthe coordinate step. Values of t 0.001 to 0.01 and h 0.02 were usedin calculation. By virtue of the nonlinearity of the boundary conditions,iterations were performed at each z step in order to achieve accuracy of10 '; three or four iterations were sufficient.

The calculation of the diffusion distribution in an assigned temperature fieldwas based on the boundary-value problem

(2)

In the case of a cylinder, aC/ar 0 (with r 0 ).Here C's the impurity concentration in the graphite, C C'/C0., and D

is the diffusion coefficient.

The problem was solved numerically together with the temperature fieldproblem.

By virtue of the equivalence of equations (1) and (2), the same implicitscheme was used as in solving the temperature field problem. The calculationwas performed in the following order:

Determination of the diffusion profile

Determination of the temperature field and proceeding to the next time 'layer.

The optimization search for the initial temperature of a given fragment of ablock was conducted according to the following scheme:

The one-by-one trial of a number of values of the initial temperature, T0with a calculation of the final (corresponding to. total cooling) distributionof the diffusant and the chi-squared test corresponding to it.

Search for the best value of T0 of the number tried.

Optimization of the value of TO by the gradient descent method.

The values of the concentration of the diffusant source and of its backgroundconcentration in the material were optimized simultaneously. Thetranspiration concentration pedestal was regarded as the background in certaininstances, which, as indicated above, is justified in view of the slight crossflow between diffusion channels.

In cases not relating to fragments of reflector plugs, optimization wasperformed for various heat transfer alternatives with cooling (variouspositions of the block on the underlying foundation). The alternative matchedby the hest chi-squared test was chosen as the optimal one. The results arepresented in table 4,

Table 4.

Number256789101113141512»

of Block Calculated Initial Temperature, 'C1240990780920110012208808701500800 to 1100900 to 11402200

«This block lay at the center of the area of a fire, possibly also caused byit. The history of its cooling is not characteristic in this case. Theresult is presented as an illustration.

As is obvious from table 4, not all graphite blocks had cesium concentrationdistributions allowing a description from the viewpoint of the solid-statediffusion mechanism. In part of the cases the profiles were of a stepcharacter. As mentioned above, this can be explained by transpiration —thedrawing of aerosols into the graphite's pores by the flow of air at the timewhen the block was cooling.

It is evident from the simplest reasoning that the depth of the transpirationconcentration profile. in the case of a cylinder, bears a square-lawrelationship to the maximum temperature from which the graphite (and the gasin the pores) cools. But, then, estimates show that in the mo:t clearlypronounced case (block No 8) the graphite must have been heated to such a hightemperature that the distribution could not be flattened on account of othermass transfer mechanisms. Obviously, processes of the nonlinear (with respectto temperature) sorption of cesium in the graphite play an important, if notdecisive, role in the process of transpiration formation of the concentration

profile. Some calculated estimates can be made. The transpiration process isdescribed by equations of continuity and sorption:

t U P

——u —[ r u c —Pc + yeadc 1 0 rat r ar

where p's the density of air; p p'/pO, u's the linear velocity ofthe air at a giv n point, determined from the difference analogue of thecontinuity equation, u ~ u'/UO , uO KRO/tO ', a is porosity; a's thesorbed impurity concentration per unit volume of the graphite, a a'/pO ,c's the impurity concentration in the gas per unit volume of the graphite,c c'/pO , p's the mass transfer constant, p p'/tO ,'7 is theHenry law constant; P is the gas's pressure at a given point, P P'/PO ,

PO ppRgTO ', and Rg is the gas constant of air.

The temperature field is assigned here from solving thermal problem (2). Someresults of the numerical solution of problem (3) for various initialtemperatures of the graphite and sorption constants are presented in thegraphs in figs ll and 12. It is obvious that the solution is extremelysensitive to the values of sorption parameters, and an estimate of thetemperature of blocks according to transpiration concentration distributionsmust be based on the results of studies of the kinetics of this process.Simplified estimates result in overstatement of the sought temperature.

An alternative version of the mechanism of the origin of the concentrationdistributions discovered for fission isotopes in the graphite is theirpenetration into pores with the explosion wave when a TK [fuel assembly] wasfractured. The process of the penetration of a shock wave into a porous bodywas described by the following equation (on the assumption that Darcy's linearlaw is fulfilled):

1 BPu p Or

where p's the coefficient of viscosity of the gas in the pore, p p'/pOt is dimensionless time, t = t'/tO , tO mR)pO/POk ; k is the graphite'spermeability; and u is the dimensionless velocity of the gas, whereu u'/uO and uO kPO/pORO

0,7

0,6

0,5

0,4~40'4

0,3

T ~ .",200'C

$ ~ ".0 3 IP370) s

002

G,I

0 O,I 0,2 0,3 0,4 Oe5 Oo6 Oo7 0,8 0,9'K/R,, rel. un.

Figure 11. Distribution of Concentration of Sorbed Diffusant inTranspiration Wave; Cylindrical Geometry

0,6

0,5

T a 1200'C

p~ IO I/s

e 0,4

cJ0,3

0,2

O,I

0 0 I 0 2 0 3 0)4 0 5 0 6 0 7 0 8 Oy9 I 0'I/R,, rel. un.

Figure 12. Distribution of Concentration of Sorbed Diffusant inTranspiration Wave; Cy3 indrical Geometry

It is assumed that the temperature of a porous body is constant over its crosssection and does nnt change during the development of the process, and thatthe gas in the pores is in thermodynamic equilibrium with the body.

In this case it is extremely complicated to make calculated estimates becauseof the following facts. Nct only the value but even the order of magnitude ofthe time interval between the fracture of a fuel assembly and containmentfailure of the reactor space are unknown. It is difficult to estimate thevalue of the pressure in a shock wave.

Nevertheless, it is possible to attempt to single out some boundary values ofexplosion parameters the crossing of which would result in noticeabledegradation of the graphite's structure.

The results of a numerical solution of problem (4), performed for the case ofthe following explosion characteristics, are presented in the graph in fig 13:pressure drop in explosion wave--50 atm; time prior to containment failure ofreactor space--2 ms. An analysis of the calculated distribution of thegas pressure in pores over the radius of a graphite cylinder shows that thepressure difference that originates in open and closed pores leads to theorigin of stresses that can cause fracture of the graphite.

E25

020

I5

~i ~ I,V.IO'

/////Ylh'l=.

~i'i W4)~W//// ~ t Xhhht

YE/iV)'/ / /M~~lP2J'J.l J J

0 O,I 0,2 0,3 0,4 0,5 0,6 0,7 0,8 0,9 I,O'Z/R re].. un.

Figure 13. Propagation Into Porous Cylinder of Gas-Compression ShockWave: Pressure in Explosion Wave--50 atm. Time Prior toPressure Drop--1.7 ~ 10 ~ s

The time dependence of the velocity field of the gas in the graphite's poresis presented in the graph in fig 14. It is obvious that tensile stresses mustdevelop in the near-surface regions of the material. This can result in theappe'arance of a type of blistering. Cavities were observed on the insidesurfaces of blocks in some cases. However, an effect like this would haveresulted in a sharp drop in Young's modulus and in a growth in the graphite'sresistivity, which was not observed.

2K

IASS

I,25

0,62 w8W

-0,62

E

-I,25IJ

e -I,SS

+i ~ ] e7.10 s

m

-2,50

-4,IS0,3 0,4 0$ 0,6 0,7 0,8 0,9 I,O

rel. un.

Figure 14. Distributions of Velocity of Gas in Pores Over Radius ofGraphite Cylinder Under Efiect of Shock Wave: Pressure inExplosion Wave--50 gauge atm.; time prior to pressure drop—1.'7 ~ 10 3 s

Results of Studies of Fuel Elements, Channel Tube and Zirconium — StainlessSteel Transition Piece From Damaged Vnit No 4 of Chernobyl Nuclear Power Plant

A fragment of a fuel channel (the lower part of a channel tube with the lowertransition piece) with part of a fuel assembly (total length of approximately830 mm) was found on the roof of the gas cylinder SAOR [emergency core coolingsystem] building. The number Kh446 was marked on the surface of thetransition piece, and the number of the channel--Zh887--was determined fromthis number at ChMZ [Chelyabinsk or Cherepovets metallurgical plant) (tube No

B4461, melt No 1546-9-1-4, made 24 Jan 80). This fuel channel was in cellNo 25-17 of the core of unit No 4 of the ChAES.

Deep dents and cracks were observed on the channel tube's surface. Thetransition piece was cut off of the channel tube in order to remove the fuelelements from the channel. In the process a large quantity of fragmentedfuel, with which the space between fuel elements was filled, poured out of thechannel tube, The appearance of the fuel elements inside the zirconium tubeand transition piece (into which they apparently had been driven at the timeof the explosion) is shown in the photographs in figs 15 and 16. Photographsof fuel elements removed from the channel tube and transition piece arepresented in figs 17 to 21. It is obvious that fuel elements of theperipheral row of a fuel assembly were preserved, and fuel elements of theinside row are absent. Their place in the fuel assembly is occupied by grainsof fuel and cladding materia). Peripheral fuel elements were baked togetherat their cladding to such an extent that it was not possible to separate themfrom one another after they had been removed from the tube and transitionpiece. Individual fuel elements had extensive cladding fractures, Thecladding material was very brittle. It was possible to remove from some fuelelements fuel pellets that were extremely brittle: The pellets crumbledduring an attempt to take hold of them with the forceps of a manipulator.Gamma spectrometry and alpha spectrometry of an extract of alpha-radioactiveisotopes in a sorption column were used in order to estimate the fuel burnup.Both methods gave with satisfactory agreement a burnup of 10 + 2 MW ~

~ 24-hours/kg V.

The working time of the fuel assembly was estimated according to the ratio ofthe concentration of isotopes Cs-134 and Cs-137 in the fuel. The fast-neutronfluence (E W 0.5 MeU), which equaled (2.8 + 0.4) 10" cm ', was estimatedaccording to the buildup of Mn-54 in the fuel assembly's spacer grid. Henceit follows that the assembly was in use from the startup of unit No 4.

Ten speci mens were cut out from various sections of the fuel element claddingfor the purpose of study. The fuel element cladding material has a distinctlypronounced alphaed structure--wide and long plates of the alpha phase can beseen (cf. figs 22 and 23). The formation of such a structure is caused by theheating of the cladding to a high temperature and by saturation with oxygenand nitrogen.

Figur e 15. Appearance of ChannelTube With Fragments ofFuel Assembly

Figure 16. Appearance of ChannelTube With Fragmentsof Fuel Assembly

Photographs of the microstructure of the areas of contact of two and threeadjacent "baked together" fuel elements are presented in figs 24 and 25. Itis obvious that the claddings form a single whole with an alphaed structure,which was caused by an intense force and heat effect.

Measurements of the microhardness of the cladding material gave a high valueof 360 to 440 kg/mm~, which is associated with the elevated oxygen content inthe alloy.

Model experiments with intermittent heating of non-irradiated material in airwere performed for the purpose of determining the possible temperature of thefuel element cladding at the time of the accident. T,t was shown that analphaed structure forms in the alloy with temperatures above 1200 'C. Thus,it can be assumed that the temperature of the fuel element cladding at thetime of the accident was in the range of from 1200 'C to the melting point(1850 'C).

Figure l7. Fuel Elements Removed From Channel Tube

Studies of the microstructure of the channel tube:s material revealed anincrease in the alloy's grain size to 160 pm near the inside surface and to80 pm near the outside surface. A comparison with the results of experimentson the annealing of specimens indicates that the channel tube's temperaturereached l000 'C.

A photograph of the microstructure af the area of the inner joint of thetransition piece is presented in fig 26. Failure of the welded joint isobserved along the first four joints. Hydrides having an axial orientationcan be seen in the zirconium part of the transition piece. Traces of anintense reaction between the zirconium alloy and stainless steel are also

observed (fig 27). The thickness of the reaction zone reaches 130 pm (with anormal value of approximately 10 pm). The grain size was 100 to 200 pm, witha normal value of 80 to 150 pm, in the structure of the stainless steel of thetransition piece. The changes observed in the structure of the materialsindicate overheating to a temperature of > 1050 'C (the austenitizingtemperature of the stainless steel at the time of the transition piece'smanufacture).

Pigure 18. [Caption missing]

» powder of the fuel was x-ray photographed in a I:abye chamber. X-rayi»terference ~as not observed, which indicates either c..mplet» «>llorphi ..'..'.ii;:iof the fuel or an extremely small specific surface (less than 1000 A).

'.:onclusions

:"tudies were made of distributions of the concentration of fissior. is rope;,over the cross section of elements of the graphite masonry of the damagedreactor. Various mechanisms of the origin of these distributions w=.reexamined. An estimate was made of the heating of the reai.tcr's masonry at '<time of the explosion, within the framework of a model of the heat-aitivat.diffusion of cesium in the graphite. It was demonstrated that tb: ~.emp.:.",.oui »of fragments of the masonry at the instant of ejection was in the range of800 to 1500 'C.

It was demonstrated that the temperature of the cladding of the fuel elements

studied was in the range of 1200 to 1850 'C at the time of the explosion, and

the temperature of a channel tube and transition piece exceeded 1000 'C.

Figure 19. Fuel Elements Removed From Channel Tube

Figure 20. Fuel Elements Removed Prom Channel Tube

Figure 21. Fuel Elements Removed From Channel Tube

Figure 22. Microstructure of Fuel Element Cladding Material

Figure'23. Nicrostructure of Area of'ontact of Fuel Element Claddings

Figure 24. Microstructure ofArea of Contact ofFuel EleaentCladdings

Figure 25. Microstructure of Areaof Contact of FuelElement Claddings

Figure 26. Area of Transition Piece Joint

Figure 27. Mi crostructure of Area of Trar:sition Piece Joint

The 1st International Workshop on Severe Accidents in

Nuclear Power Plants

EXTENDED ANALYSIS OF FIRST PHASE OF UNIT IV ACCIDENT

AT CHERNOBYL NUCLEAR PO'HER PLANT

E.0.Adamov

Yu.M.Cherkashov

L.N.Podlazov

A.I.Ionov

Dagomys, Sotchy, USSR

30 oct. - 3 nov. 1989

:.T~IDED ANALYSIS 0-''INST

. ACCIDENT AT C~NOBYL ~HJCLEAR PO'.Vr~ 'PL».'

Adamov E.O, Ionov A.I.,Nikitin Yu.M., Podls~ov L.N.,Cherkashov Yu.M.

. The Chernobyl accident.is being presently analy"ed in

several aspects:- analysis of the first phase. including behaviour of

the reactor from the moment of operation of reactor protec-

tion system A35 to'eactor runaway;

canalysis of forces and processea resulting in destru

tion of .tho. core and reactor plant;V

- discharges and activity .spread which take place both

during destruction of the core and in subsequent periods;.

environmental. consequences of the accident.

llethodologicslly, these aspects have their specific

a~r.roach~a and calculation techniques snd advance in para-

i:lel .to': o~e. another',Z nap it e of . a'rea t.. number,o f publica t iona, on: these

,aspects;.:the problem is far from being-'exhsuit'ed.and: require:

further ', study and development in all of theabove-listed.'.res

s.,The present report. deils viith the. first phase of the

iccidert, vhich: i" covered'y a considerable. number of scien-J

t if ic-methodological.. publications -/1,2,'3/ ~, However, "of no

leis imports'nc~ than the'cientific-.methodological wo'rk ii'to.,oreser.t e re 1 .end consistent picture of the. sequence of

')

*'.

events and processes that have resulted in the reactor run-

s.;ay so as to become conf ident that nothing has been missed.

In this case, criterion of truth should be as a minimum

coordination of all knovm registered reference points and

events. Such reference source date may include:

distribution of neutron field, positions of rods of

the contrcL and safety system, registered 1.5 minute before

operation of reactor protection system A35 according to

special PRI."; 2 programme;

listing of the GREG monitoring system for a number

of main opercti.-.,: parameters;

os"i llo,":or" of some electrical parameters chsracteri-

sing behaviou" of the main circulating pump power sources.

"..ecsuse o their specific features, all the above-men-

tioned date media cannot be mechanically synchronized with

one another to e s"f icient accuracy which imparts some

vega ness to sequ enc e of events and necessitates specialcalculatio- ~n~ ..-.ethodologicsl research. Another source of'.-.i.efinit~.".e.""s i.=. o~eration of t'e controlled circulation

s> stem is a ret ion ar beyond the permissible 1'mits (low

power level, high coolant flo.v rates, low underheating of

coolant at entrance to the core, main circulating pumps

approach in/ csvi (ation conditions) where thermodynamic

behaviour of:'.. controlled circulation system parameters

hss not b e". s, l'icie..tlv studied.

It should be born in mind, that in the regions of coo-

lant para.. et sr . el'.:es close to saturat ion consid ersble

effect may bo. ~ 'artsd by dynamic end static measurement

er"ore. such as pressure in the separator. feed ":ster flo-i

rates, which determine the cavi'ation mar,"in an<'. ",...'er.,ea-

ting at entrance to the core.

All the considerations listed above have made us return

to analysis of the first phase of the accident, now in view

of the results of the mothodological research contained in

publication /1/.

Briefly, the main conclusions of this work are confined

to that for catastrophic runs~'ray and steaming of the reactor

channels .i.thin the time interval after operation of reactor

protection ™!rstemA35, estimated at 4-5 s, apart from posi-

tive ru;.-out of reactivity caused by expelling of water co-

lumns at the bottom of the core, additional positive reac-

tivi'.„. is needed which may result either from disconnection

of four pumps fed from the running do.rn turbogenerator or

from ingress of steam to entrance to the core from the ca ri-

tating main circulating pumps or boiling of slightly under-

heated ~+ter at throttle control valves. Since the phenome ~ a

:nish take .>lace in the controlled circulation system ."it'.:

cavi'.sting main circulating pumps cannot be interpreted

..sfinitely at present and require additional investigati07

wore detailed co.sideration and analysis were given to disc-

repancy of the data relating to possible time of disconnec-

t ion of four r ~nning down main circulating pumps. Chronolo-„;

of the registero.d events is very important for restoring

- turn 0 'elo",.ment of t'e acc ident f irst pha se. Ob vious-

1„', to restore the ser <ence of real events by the register~d

i'.ate, it is ."aces.=e .. to take into account:

I

'E3/

C

4

Dart ic.clara of regi st ering sv at em func t i or i.w;

i."..ami" "ha rect erist ics o i "..~:a~.;r i.".=, -"ha."=.~l.-.~

- dynamics of functioning of the main circulating pump

protection logic circuits.As pointed out above, the main sources of information

for restoring chronology of events vrere.'

oscillogram carrying some electrical parameters;

- DR"G svstem operating with the SKALA computer.

"-in";,.e the oscillograph in use did not provide absolute

time referencing, such referencing had to be done on the

basis of the ";.R"-G svstem data a..d, in particular, for the

moment,jhen the diesel-generator put in operation reached

..ominal voltage. Ho'ever, specific featureaof DR:-G system

functioning (scanning cycle. aperture) may cause an error of

2-3 s in referencing the time moment of this event on the

oscillogram. Other factors distorting the real sequence of

events is lag of sensor and electronic converters aa well as

-..on-standard conditions of sensor operation in the pre-acci-

dent perior!.,: "alitati re analysis of the above-mentioned

fa" tora rakes it possible to believe that two running do;v-..

mein circulating pumps at each aide were disconnected at the

43 =-, rather. than at the 46 one. Fee should bear11d rd th

in mind that according to the DREG system data, by the 36th

end 39 s, low ra t e through t h e running dove main circ la-

ting pumps drop:..e:! belo.v the miniirium permissible level o

-".'00 m Ih -hereat their emergency protection systems

operate.

:.".vie:; o indefiniteness of absolute referencing of

time en i sequ~"..ce of the registered events, it is quite pos-

.",'''."1e that. our running do-i:.."..ai... ciroulat i."g ~u-...ps discon-

nected close to the t ime of depressio:: o- the ':.~ro .. 'o

system button, which means that the development of the firstphase of the accident at the Chernobyl Nuclear Power Plant

(ChNPP) coincides with the data presented in publication /2/

and shown in .".igs 1,2.Another vital criterion to be met by the results of

simulation of the first phase of the accident at CAPP issubstantially simultaneous generation of the power protec-

tion (power rise above 10',~o nomi.nal) and speed protection

(signal of power rise period less than 10 s) signals at allperipheral lateral ioni. ation chambers between the 2 andnd

3 seconcs after operation of the A35 protection system.rd

As is seen from ."'ig.2, at operation of the A35 protec-

tion system without additional disturbances, variations of

neutron fields within the reactor volume during first se-

conds are vividly heterogeneous, while the variations per

se cannot cause simultaneous generation of the power and

speed protection signals in all the chambers, i.e. when t:.A35 protection system operated, neutron power rise yes o"

e global and intensive nature which could not stem only

from insertion of the rods of the control and safety system

at the initial volume field measured 1.5 min before opere-

tion of the,",35 protection system.

It is noteworthy, that distribution, of the volume

ield assumed as the initial one may be regarded conserva-

tive, since, as mentioned above, it was measured not earlicthar. 1.5 min before operation of the A35 protection systeL..

rb

C';:ing to an increase of steam content at the reactor top

caused at this period by a drop in flo',v rate of the r ."...ing

down main circulating pumps and a decrease in delivery of

feed water, the maximum of neutron field also displaced to

the reaotor top, which obviously diminished the effect of

expelling of water columns from the lower portions of the

control and safety system channels at operation of the A35

system upon reactivity. To analyze this phenomenon, special

methodological research is needed, vhich will be carried out

in the course of implementation of the subsequent programme

of research of the accident at the Ct8lPP. Since the P>LZ

high-power channel-type'eactors are reactors of great/,

physical dimensions sa regards combination of characteris-

ties and properties, to solve such problems it is necessa-

r„" to accomplish integrated research of spatially distribu-

ted neutron and thermohydraulic processes in view of opera-

tion of the reactor monitorir~, control snd protection sys-

t ems.

A package of three-dimensional dynamic programmes of

kinetics, tyPe T~.Z'QA is expected to be the main tool for

such a research. The programmes incorporate dynamics ofinstantaneous and delayed neutrons, heterogeneity of'hysi-cal prop crt i es and t hermohy draulic sta t e of t he core l4/.

uch ar approach to the analysis of the accident at

the Chl'PP is also dictated by the use of three-dimensional

cynamic prog r am:es by for e ign researchers ( STABZL, Cans da;

:.ST''~~.,:"',:,AQvP, Italy). The recent technical meeting

of the:."=. worki-.~~ -roup on codes of programmes used in

jIQ-,~-:= . ';T.~:-20.i ..X=.GCTi ii P~.-,~;.GC,ttTFiI: 4'4,TiFiI ~~;GiT<1 A3-5

0

6

2

2a.

lg6 iD

J~ICP. jtetnOPtlalgIR POJIOit 9 OOI,Cl.IC JCal(TOjI:I:Ign IIQn'1TII:I lnlOllnu AJ-~

.G

p

s6er ;eaz".~:oc"."oT sz.ecaeam sogs res

::::..:e.'.NC.CTJ,.F '::eJIOB i~i ~~ OTcp'TcTBEE

oo-a-.~ax cs~seN.

ansly a is o f power reac to r sa ety ("'erl n, .pri.l 1'7-21. 1'38> )

er:.phasi.".ed the need .or usin=, i.-..provi.-.g a.-.. te=-t': ."ee-

dimensional models of neutron kinetics. However, it may be

supposed that the main conclusions of the accomplished work.

will remain true.

From the above it transpires that a consistent picture

of the first phase of the unit ZV accident st the ChIlPc- is

represented by the concitions when four main circulating

pur.ps =.."..'."~ r.'own with the turbogenerator disconnected

within the time interval of ~ s after the moment of opera-

tion o he .'.=5 system, which resulted in intensive runaws.

o the reactor intensified by positive run-out of reactivitcaused by expelling of water rota the lower part of the

control arrl sa-ety system channels (."ig. ~).l'nder such con«:itions, power rise in moat energy-

stressed ".ones -...ight exceed not only the melting thre shold

of uran.'um oxide ". t also its dispersion point which shoul".

cau . intr:"...",'. enerstion of steam, destruction of fuel

elements ".".d cc.-.:.i'ercble umber of process channel pipes,."olio:<ed b„" destruct io.. of the reactor.

. ote, t.-.et the .-.ext lo:ic "tep in analyzing the acci-'.«nt should be .etailing the pro" ess of reactor destructio.".

itsel . 'ski.-.:. into account, that this destruction took

p1sce pluri;".," "-.octor "unawsy, it is planned to .viden the

"rog ro-,'«: . „-«6 .or the research to the full-scale ''I;

on= and to o:-.~1~.-..ant it .lith relations describing processo.;

of destruct '..o;::el elements and process channels.

:-'.eferences

1~ Adamov ".0., Varinger V.V., et ai, ""-"stimation of

Qualitative "'ffects of Possible Disturbances during

Accident at ChernobyliG'P'.

Adamov 7.0~, Vasilevslqr V.P.~ et al, "Analysis of Firsthase of Unit IV .'cc ident at C hernobyl '.nuclear Po".Ier

Plant;", Atomic ""nergy, 1988, v.64, issue I, pp 24-28

.',smolov l.G., 2orovoy A.A., et al, "Accident at Cherno-

byl .'PP: a year after", Atomic "-nergy, 1988, v.64, issue

4 ~ L onov .'. ~ ~, Pan in V.i I ~, et al, "Simulation of Dvnamic

Conditions of Reactors;vith Great Physical Dimensions",

i".:-A;Iorking group technical meeting, Berlin, April

17-21, 1989.

12

"ig.l. Behaviour of Neutron Power and Reactivity at.Depression of A35 System Button

Fig.2. Deformation of Pields within Reactor Volume at

Depression of A35 System Button

1- t =Os; 2- t=7 s; 3- t =11s

.- ig..". Run-Out of Reactivity Caused by Bxpelling of "/ater

from Lower Part of Control and Safety System Channels

i.. Absence of Feedback

The 1st International Workshop on Severe Accidents in

Nuclear Power Plants

ESTIMATION OF QUALITATIVE EFFECTS OF POSSJBLE

DISTURBANCES IN CHERNOBYL-4

E.O.A~~ov

V.V.Vazinger

V.P.Vasi'=vs+

Yu.V.M'":.n::v

Yu.M. ¹k'.'nL.N.Podiazov

Yu.M.Oherkashov

D~gys Sotchy, USSR

30 oct. — 3 nov. 1989

ESTIMATION OF QUALITATIVE WFECTS OF

POSSIBLE DISTURBANCES IN CHERNOBYL-4

E.O.Adamov, V.V.Vazinger, V.P.Vasilevsky,

Yu.V.Mironov, Yu.M.Nikitin, L.N.Podlazov,

Yu.M.Cherkashov

On 26.04.86 an "experiment" on checking the reliabilityof auxiliaries and power supply for the emergency core cooling

system (ECCS) provided by a slow down turbogenerator at the

Chernobyl NPP unit-4 was scheduled to take place during the

reactor shutdown.

At 01:00 a.m., 26.04.86, the reactor was operating at the

200 ZW„ capacity uith the slow down turbogenerator No.8 (TG-8)

The operational reactivity margin was 7.2 control rods (while

the minimum perm's"'ble reactivity margin according to the

"Regulations" is 15 rods). In each pumping station 3 main cir-culation pumps (MCP) .rere operating, providing the total flow

rate of 21000 m /h "..'""ough the half of reactor. The feedwater

flow rate in the drum separator (DS) of each loop of the

circulatior. circuit (' was 200-250 t/h. The pressure in'the

drum separator was about 6.9 MPa.

At 01:03 a.m. .wo pumps LICP-12 and, IJCP-22 were turned on

in the left and r'.;;".t loops of the CC, respectively. For carryin—

out the "experime.-.:" .';.e iiCP-13 and MCP-23 were fed from the

transformer SA, a;.d -.;-.e pumps MCP-14 and MCP-24 - from the

transformer SB cor=.ec .ed to the TG-8. The rest MCP were stil1soperating on the r~i.".s. The ECCS pumps were also connected to

the turbogenerator "'~-8. 'ince it was expected, that the ECC

would be turned on automatically in the cours of the "experiment"

it was cut off from the reactor circuit by manually operatedI

gate valves to prevent cold water entering the circuit.And at last the emergency protection signal (EP-5) from

the closed TG-8 cut-off valves was blocked.

There are the fo"lowing data on the basic technological

parameters of the reactor facility just before the accident.

First of all, these are the real-time data recorded by the

"DRZG" program for feedwater flow rates - FY(1 and W2 /kg/h/,

drum separator pressures - P>S and PS>2 /kg/cm / and, levels2

in them — HDS11, HDS12, HDS21 and H>S22 /mm/, as well as RCP

flow rates - GBI11'hI12'M14, GM21, G1K22'M23 and GM24

/m /h 10 / (see Table 1 and Pigs.1-6), the data are given in

the real time scale starting from 01:18:49a.m., 26.04.86.There were no signals on the mimic panels of c»~~els

indicating the excess and decrease of water flow rate in the

reactor fuel channels, that is obviously caused by turning-off

the LICP with the excess of flow rate through the reactor (whensfour hICP were turned on}, therefore, through the pressure tubessup to 35 m /h.

There is no information concerning the appearance of the

signal "1HCP suction pressure" on the panel, which is generated

by the "Skala" central monitoring system, when the positive

suction head decreased below the permissible level defined. by

the cavitation process. There are also no data on the functioning

of the hICP'uxiliary systems, suction temperature and pressure

in the LICP header.

The behaviourof the reactor power during the preaccident

period has been restored {Fig.7).The humidity in different reactor sections before the

accident was normal.

At 01:23:04 a.m.*) after the "design-basis accident" button

included in an additional scheme to perform the experiment had

been pushed, the TG-8 cut-off valves were closed, and. as a result

the turbogenerator went "running out" with subsequent decelera-

tion. The voltage drop across the TG«8 terminals resulted in

decreasing the rotational speed of the MCP-13, 14, 23, 24 and.

their capacity, respectively. The change in flow rate of, groups

of the pumps fed from the mains and TG-8 in the reactor halves

and pressure variation in the SD starting from 01:23:00 a.m.

are given in Fig.8 (according to the DREG data).At 01:23:40 a.m. the emergency protection signal (EP-5)

appeared, which may be caused by some technological reasons

(LfCP turning-off) or the reactor shut down button pushed by the

senior engineer responsible for reactor control.

At 01:23:43 a.m. the runaway" rate and power limit scram

signals appeared in the listing, indicating inefficiency of the

scram system during tl;e first seconds of motion of the rods due

to a considerable d=-crea e of the reactivity margin below the

maximum permissible level.

~)Since there was no s.rict correlation between the time referencpoints of dif e e".. computer media (DRZG, oscillograms), thetime scale of onl„" one medium-DREG-used in the present report.

Starting from 01:23:45~*)a.m., the coolant flow rates

recorded by flowmeters in the circuit loops dropped. at least

by half, and a sharp pressure increase was observed, at the

same time (Pig.8). The pressure in the left-loop DS rose 'by

9.4 kg/cm durin 3 seconds and in the right loop DS - by2

21.6 kg/cm .2

At 01:23:49 a.m. the pressure growth inside the reactor

vessel was fixed (p 0 1.5 m H20).

According .o t!ie data of the RCS (reactor control system)

recorder, the reactor po rer rised above the upper limit of the

scale of 1000 IE/(t), the LAC (local automatic control) recorder

showed 3500 MY(t) the SPDM (system of power distribution moni-

toring) recorder — 2650 M'Y .t'hisis the basic documented infor~tion concerning the

reactor technological parameters.

The analysis of processes occurred. in accidents is usually

carried out at two levels. The first level includes calculation

of the circulation c'rcuits as a whole with detailed description

of processes dependi.-- on potentialities of the available compu-

ters to provide ru-erical realisation. The second level includes

calculations according .o the mod.els most fully reflecting phy-

sical processes '.". = „"ere.e circuit elements, first of all, in

the reactor core, . ':.".+ the "best estimated" code. Xn this case

It shou1d be no .ed that, since the MCP flow rate an DS pressur~readings are taken b" bellows differential manometers and thosedesigned. for measu=i"..- stationary and slowly varying processes,the DREG listings durir~ sharp changes (like in case of pumpswitchin -off);rill:»eve signal distortions in the measuringchannels both in rgagn+.ude and d.clay (not less than 1-2s) withrespect to real time (i.e. real processes recorded in transit-time channels have occurred earlier than moment recorded by theDREG program)

an equivalent channel, vrhere neutron-physical processes are

described by the equations of kinetics, vrhich are point equations

in most cases, is considered. as a core model.

The RE!K reactor core dimensions is much greater than tl.,e

neutron migration path (D/M is 50), and. this reactor is refered

to the class of physically large reactors. The cors is a polyla-

ttice of channels d'ffering in breeding, absorbing and. scattering

properties. Hence, there are inhomogeneities of the characteris-

tics both in radius and core heigh /1/. That is why the point

model of neutron kinetics cannot be used to describe the beha-

viour of the RBLZ core parameters.

Ilevertheless, for cualitative comparative analysis of different

disturbing factors on subsequent development of processes the use

of the point model of neutron kinetics seems to .be justified. Such

simplified neutron-physical model has been used. for comparative

qualitative estimation of time-dependent pov.er variation of the

ecuivalent core channel -'n the circulation circu't from the

discharge group header (uGH) to the DS (Pig.9).The basic assumptions used in the mathematical thermal physi-

cal model o the equivalen. channel are as follovrs;

.ne boundary-value problem is stated for equations of

thermohydraulics ( ith the given time functions G. (t) and

-'ou. '" 'out ""- a non-equilibrium t;vo-phase flow, vvhere boiling in case

of l'quid phase under.-.eati."i,„- 's assumed, 's considered. in the

char".el, steam is -ss-"..e":o 'e " .urated;— the sound veloc'.y i" assumed to be infinitely great;

— the phase velocity ~ference is determined through empirical

correlations for coefficients of sliding friction or phase driftvelocity;- the two-dimenstional nonstationary thermal conductivity prob-

lem (in radius and length) with the boundary cond.itions of the

third kind on the fuel element surface is stated for a fuel

element, while the coefficients of heat transfer are calculated

on the basis of empirical correlations; a gap between the fuel

and its cladding is assumed;

- thermal physical properties of water and. steam are determined

by the base pressure (e.g. pressure present at the channel output),

which, in general, can vary.

The equations of kinetics in the point one-group approximation

are taken in the following way:

where c is the concentration of precursory nuclei, n = N(t)/H8

is the average relative power,

1 is the mean neutron lifetime (l = 0.2),A is the effective tir-e constant of delayed neutrons () =0.237).

The initial condi.ions are as follows: at t"-0, n=1, c=1 ~

The order of poss ble external reactivity disturbances forthe equivalent core chal~-eel caused by the injection of control

rods from the "'P-5 siGnal was taken on the basis of estimations

of total reactor power variations at the same EP-5 signal in

the nodes of the three-dimensional neutron model with approxi-

mated (mainly due to the G(Z)~constant appro~imation) description

of processes of thermohydraulics in the coolant /1/. This analysis

of spatial power distribution in the reactor core area, where the

accident has occurred, after injecting the control rods from the

EP-5 signal, carried out on the basis of the information concer-

ning positions of all the rods recorded 1 minute before the acci-

dent and the data from the power monitors, has shown that the main

feature of the EP-5 mode at the prophibitively low operational

reactivity margin is considerable reshaping of the spatial neutron

field with shifting the peak towards the lower part of the core,

which within the point approximation can be treated as a result

of introduction of some equivalent reactivity, shown in Fig.10.

Por the equivalent c~»el internal feedbacks, provided by

coolant density and fuel temperature variations, were described

by the following dependence:

where

f(0

is the initial

7~v is the average

7+y is the average0

vo d fract on,

temperature along the fuel element radius,in +.al temperature along the fuel element

radl.us y

9 is the power profile along the c>~~~el height L.

So, within the frames of point kLnetics of reactivityvariations

~(~) z ~~>, g gy g oat ~T„~Res

where reduced to the effect of the control rods and factors

influncing the coolant density and fuel element temperature.

The initial coolant parameters at the core inlet before

the accident*) where estimated according to the DREG data and

results of statistical calculation of the circuit parameters

according to the GXDRA.-1000 program.

The time-dependent variations of the total water flow rate

through the reactor Q, total feedwater flow rate and pressure

in the drum separator PDS average for two loops are shown in

Pig.11. The curves provide smoothed parameter values witho ta pulsation component encountered in the DREG data. The pump

speed variation for those pumps connected to the slow down

turbogenerator is show in the same Pig.11.The pressure distribution over the PCC, when a" 8 MCPs are

operating with total capacity of 56000 m /h and. pressure in the

DS of 65.7 bar (67 atm) is shown in Pig.12. The curve is plotted

on the basis of the results calculated according to the GIDRA-100:

program supplemented by the data on pressure drops and elevation

Variations of the coolant operating parameters of the circuitbefore the response of the cut-off valves are of no interest fo=estimating the preaccident initial thermohydraulic conditions "tthe core inlet.

marks of the circuit elements.

The time of coolant transport from drum separators to the

core inlet, when 8 MCPs are operating, is about 30 s. The trans-

port times for each PCC section, when four MCPs are turned on,

providing capacity of 4x7000=28000 m /h, are given in Table 2.

Table 2

Section Sectionlength,m

Innerm/s trans'rum

separator-- MCP suctionheader

47.5+40-2= 43.85

24x293 4.81 9.12

Suction header-EICP 36.4 4x752 4.38 8 31

MCP-pressure hea-der 20.0 4x752 4.38 4.57

Pressure header-DGH 17.2 22x293 5.24 3 ~ 28

DGH-core 23.8 830x505.5 830%55

(vertical section)

4.773.94

4-991.63

Total DS-core ~ 31 ~ 9 s

So, every moment a mixture composed of the circuit water

and the feedwater, produced half a minute ago in the DS, is fed

to the core inlet. On its way to the core, this mixture acquires

an additional enthalpy increment in the pump:

8'kcp= ———= 3 kJ/kgMCPGz

where NMCp 4.3 MLrI is the capacity of one pump,

G< = 11700 kg/s is the total weight flow rate through

the reactor.'Vithin the time interval under consideration four of eight

pumps have approximately constant capacities and, flow rates,and the other four pumps have NMCP and G decreasing, when the

pumps are slowing down, although iMCP is nearly constant.This fact provides the assumption iMCP const, vrhich isused in the calculations. In this case the water enthalpy atthe core inlet is determined by the follovring relation:

where iD is the saturated vrater enthalpy at the pressure inthe drum separator,

if = 722.4 kJ/kg is the feedwater enthalpy (at '170 C).boule calculating the pressure at the core inlet, the pres-

sure drop in the sections vras c>~ging in proportion to the

dynamic head. The pressure grovrth in the MCP was determined,

according to the total characteristic of the simultaneously

operating pumps wi .h regard to the slovring down of four MCPs.

The results of calculation of enthalpy at the core inlet aregiven in Pig.13. It is easily seen from these results thatund.erheated, water is fed, to the core inlet during the vrhole

period. und.er consideration. )Vhen the pressure minimum is attained

,11

underheating is also minimal and equals 17.8 kJ/kg against the

initial value of 29.3 kJ/kg.

Zn case of so slight underheating the existence of nonequi-

librium steam produced by csvitating pumps is not excluded.

The time of coolant transport from the IJCP to the core

inlet is about 14 s (Table 2).From these positions possible cavitation regimes of the

EICPs were estima.ed.

As it i" shown fror .he DREG data (Table 3), during the

preaccident period a considerable pressure variation in the

DS and feedwater low variations are observed, when the total

flow rate through the circulation circuit increases

These sre basic 'actors in determination of h, the pump

suction head margin be ore cavitation csn occur under stationary

conditions, which minimum value is usually calculated hy the

formula.

)2; -2 )v'-"n.

7IllZl g+ ----- [ m ]in

2g

The static pressure at the pump inlet (p.4 in Fig.14) is

equal to:

p p „- p Z Mcp )2 I fl/nl)G

tn DS dh (

where PD is the :. .'"-.:.in the drum separator

H i" t'.e;-= .- c'-.)n height - n the downtake header

(30.5 --),

P,t 's —.;:e =": .=""'rop in the downtake tube, equal .o"281200 n/m at t}:e ratea;vater flow rate of 24500 m /n.

12

The saturated steam pressure P at the temperature of pumped-sover liquid is calculated according to the water enthalpy at the

MCR inlet

(ZGMCP

— G fw) 1'fw 1

fm

Z GMCp

Moreover, under transient conditions and at pressure varia«

tions the parameters in the saturation line st the MCR inletcan be attained under the following conditions:

p 3 i'dt bp

mugwhere dz is the pressure margin variation prior to cavi-tation, expressed through the enthalpy J/kg s.

For the time interval of coolant transport from the steam

drum to the MCP inlet 6 t~~ (s), the margin defore cavitationis 6 i = g Ah„.„ and the maximum permissible pressure dropmarg ..anrate in the SD without cavitation equals

dp

dt('")

The results of calculation of these cavitation characteris-ties according to the total flow rate of each reactor half, ssshown in Fig.14., u-'ng the data on the operating parameters va-

riation de.ermined according to the DRZG system, sre given inTable (dH, dp/dt ) ' "gs.15, 16.

Although the " essure drop rate does not exceed, the permis-

sible value (Fig.16), the minimum permissible cavitation margin

equal to approximately 23 m H 0, is attained during

13

s certain period of time, slightly different for both reactor

loops (Fig.15).The boundary cond,itions for the snalyxed c~~el model must

also includ.e heat flow through the channel from the graphite

construction The hest flow values, as well ss graphite tempera-

ture, calculated for the given variations of the reactor power

and, hest transfer coefficient from the graphite block (K=78

Yl/m'C) to the saturated coolant, T=280 C, are given in Fig.7.It is clesz t'nat, when the accident has occurred, the heat

flow through one fuel c'nannel with regard to the povrer peak

factor of 1.4 is eaual to approximately 35 kW.

So, according to the estimations made the ~allowing initial

preaccident conditions cr the eauivalent core channel can be

assumed

- initial power is 133 k'8;

— heat flovr rate thzough the channel from the graphite

construction is 35 k;"I;

— flovr rate at tne channel inlet is 6.24 kg/s;

- base pressure is 6.6 EIPa;

- minimum coolant underheating at the inlet is (-0.0141).Because of the lack and uncertainty of the source informa-

tion concerning the causes of the reactor surge, qualitative

effects of the bas' po. sible disturbances vrere estimated in

the "pure state".These could be, o instance, introduction of some positive

reactivity by the con. ol rods after pushing theEP-5 button,

switching-off of our '..'CPs, when the turbogenerator slovr down

14

or temporary MCP cavitation.

The power behaviour on introduction of the equivalent

reactivity, characteristic for injection of the control rods

from the EP-5 signal (Fig. 10), and unchanged flow rate atthe core inlet is shown in Fig.17. Zt is easily seen that

although the relative power N/N is increased, by the eighth-

ninth second up to 4.5, no local surge is observed later on,

and maximum fuel temperature does not exceed 3874C.

".Ihen the EP-5 button is pushed at different initial void

fractions at the inlet up to $0- 0.256 (just as a coolant

containing nonequilibrium steam is fed to the core inlet)first, the maximum local power value in the course of surge

is increased, and then it drops (Fig.18). The maximum power

surge occurs at f, —0.01 and equals 32 (at the Gi ~ const

boundary condition). The fuel temperature'- under these condi-

tions does not exceed the values of about 79('.C, and. the

cladding temperature is about 650'C.

Zf the positive reactivity introduced by the control rods

is increased up to 0.8 p as compared to the previous case

(Fig.10), the power surge grows up to (N/N) ~ 880('C-7.5~8sec),

when the underheated coolant is fed to the fuel assembly inlet.The maximum fuel temperature in this case exceed. its melting

point.

The maximum power surge (N/N ) ) 1000 is observed, when

the void fraction is increased. to p, =0.01. Zn this case the

fuel and cladding will melt down.

15

The results of calculation for the case of constant pressure

difference detween the discharge header and separator (PDGH—

74 bar; PDS—64 bar) and introduction of the equivalent reac-

tivity according to Pig.10 and, Xi ~ -0.0141 are in good

agreement with the results for the operating conditions show

in Fig.17. The variation of void fraction at the inlet increases

the maximum power surge at f, ~ 0.01 up to N/H —64, and the

maximum fuel temperature momentarily attains 990 C.

The given results show that the limits of thermal state

disturbances of fuel elements caused, by only the effect of RCS

rods on the reactivity (Pig.10) in a possible void fraction

range at the fuel assembly inlet, are restricked by momentary

fuel temperatures up to 1000 C and cladding temperatures up to

810 C ~

When the KCP flow rate is decreased, by 50/o per second at

the unchanged reactivity provided, by the control rods (just

as four MCPs are turned off, when each pump capacity decreases

below 5000 m /h during the turbogenerator slow down), the

maximum power N/N attains approximately 430 ( i ~ 12 sec),

and the fuel temperature exceeds the melting point ( i = 13 s),i.e. the local surge following such a scenario is likely to

happen, when the cu.-off valves are closed 10-12 sec after a

sharp flow rate d"op a . the core inlet.The power surge (:;/N > 1000) and melting-down of the fuel

0

and cladding are also observed in case of similar disturbances

within 5-6 secorde a='.er the introduction of the equivalent

reactivity by the control rods (1"ig.19). Such scenarios are

16

likely to be realized, if it is assumed. that according to the

DREG data the flow rate drop (with regard to transit-time effectsin the channels measuring flow rate taking - 1.0-2 sec) afterthe EP-5 (1:23:40a.m.) is caused by turning-off of the MCPs

(1:23:45 a.m.) fed from the TG-8 in the course of slow down atthe minimum capacity (5000 m /h) (Set table 1 or Pig.8). Later3

on the flow rate disturbances do not cause elevation of the

temperature above the melting points of fuel and. cladding. Itshould be noted that more intensive surge occurs in case ofaxial deformations of power distribution towards the core

bottom (Pig.20).The melting points of fuel and cladding also rise under

operating conditions, provided by void fraction disturbances

at the core inlet up to P —0.01 per 1 sec, when the control

rods do not introduce positive reactivity of within 4-5 second

after the introduction of the equivalent reactivity. The mel-

ting points are reached during approximately 6-7 second afterthe occurrence of void fraction disturbance at the core inlet.

The given estimations out to the reactor was operating

under extremely unstable conditions, when the flow rate distur-bances or those provided by the void fraction at the core inletcould result in uncontrolled surge.

17

REFEREHCZ'-S

1. E.O.Adamov, V.P.Vasilevskij et al ~ Analysis of the firstphase of accident development at the Chernobyl HPP unit-4

Atomnaja energia, 1988, t.64, vyp.1, s.24-28.

2. Yu.Y.Mironov, N.S.Razina, T.I.Pomicheva. Mathematica model

and method of numerical solution of nonstationary problems in

the TRANS system, Voprosy atomnoJ nauki i te>»~ki, Ser.

Phisika i Te~b~ika jadernykh reaktorov, 1987, vyp.3, s.3.

et

'I 4'ttt St«st tieijne) „

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5F'. 72 F.77F.'5 r.02

e. ~ 7 6.87r.ne ?.00

rrnr. IleC22a tat

PCC2 HEC 1ti'O'Soaat as

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a st - taoosaaaast t . 'Isa at t

100

nieiajn~~ «o

f.2000 izrud 1.2Z 00 f.D6'0T (h,min,seo

f.Z4M

Pig.1. I'eedwater flow

PDS 'gf/cm 2

70

nr prod

a2Pznt

65

60/ (8 4 0 / /9 08 1.2000 /2r.DD''.22////

hg

CxaR.

T (h,mia,sec/2//ae

,mma) Pr eseure in DH

nz/.zsrrright loo

f/2//'252f

f. 1S.VO f. /9.//0

f/f//2 j2/

f/OQ55f

l!N4// L2)PO /ZZOr7

left loop,

I

xzsau T (h,min,seo)

fig.p. b) 1evels in DS

g . m/hMCP

7000(

I

I g

etui22I MCp12

grans former suPP17

t.'ZJDP

8'()

iV4P

Qqgp m /h3

<<G~2>~ MGP13rI I

gr

g/

I

Hfg f241

~c

6I~D0

TC-8 supply

MB8'

gaoo I

~~e4e XXSoo jzD,'H'Zit5'"

ig.4. MQl'.1'low (lef i loop)

~f76 <2u Mcp21

i

YNb'"(!

VO

/

I(I

fgI (I

I

( 4 e

(%r (

I l

6T trans former supply

poli//BOLD /. f9.'OO P ZDOO f. 2fO(t('72 DO

~MCP3

g2C g23S MCP23 I

\

I\

IJ

HZC s24s/

MCP24

Is I

g/JOB i

5000

TG-8 r upplp

om T/:8

40Ã Ls. sa.yo s shoo szu00 1.2r.uo s22 DD

N (~w)

'C

SAD

3MV

f000

Igloo Ego

25.04.ISB6=)

l

Olf

4'6.04. L9B6

~IP25'i

T (h,min,seo)

pig.7. Thermal power

G x10 3m /h1r,~ 00

P kg/cm

Tmains supp)y rG x10 ~rn~/h

t irne f i nish

ah,„„<ah'I5

timB finis

j5I ~4QH %+4 ~I

left looP

6023.00 23 (0 73.20 2330 23.90 23 0$

EP-5

~ 5Z3$0 (h,min)

Pig.8. Plow in coolant. lo ps and pressure in DS

f DGE/

3 p6~

02'ig.9.

Ca1cu1ated scheme oz p.-ima~ cir uit

S 9 lO

fig.10 .Equ valent r .activity variation after insertion

oi.'ontrol rods from W-5 simal, calculated

according to the three-dimensional dinamic

program TRIADA at G (z)=const

G x10 3rn3A

60

50

Gfw, 4'h

r2oo8OO

P. kcmSS

65'7,

IPm/000

Soo

f.2<.30 r. ~a.oi ].22.30

T (h,min,

1.23.M / 83,88

Fig.11,Va."iations of total floor through the reaotor;

'f'eedwater flow, average pressure in DS and

b1GP speed., fed 'from the turbogenerator

P kg/cm

MGP Pfl CRP core

Fig.12. PressUre distribution in primary i;ir.'.wit

01220D 722,00 4+3 00 f29~ T (h,min,=-ec;

Fig." 3. 'Zime-dependent va= ation o" coo~art

~ader heat ing at the .o in't

~ ~II

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I I ~ ILP.L>t,t I,« .>< I.«..>.f P,t,l I'',l' I l,l,l 14>'' Cut I ~ Iuen4 I'

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~ CCC GC C CCCC CC'-CC GSCC GGPGGGCGGSSSGGGGPn II.u r, . 4 0 Ci r<O O C<a,u 1 P.l.i-u,a. iiP. Plu< P.V nt.c>0<a.~u<MMP)P)PIP)ne

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r un ~

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I~ ~~ ~

~ oi M t ra M >.i r 0. a r 5 v v o': e 5 Mlp r 5 01 r 5'p. Io n t. ~ tv n 0 t.i 5 M 5 Osn I ~ r1 I. » i «0 a e 0 C 'i u> ~ 0 i- >< 5 PV N T a IO M h) s< SO N <a) P. r< t r >1 e Pl r.

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~ \t ~

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nn 'ur Vlf'V Wt.nn ~ 0 0 O. 0 0 ~ O.Wuv>10 I'>P'P Iut'.>VS<un>'<P-f<P

» u' I- <. I- p- >i 4 4 f v> 0' e r ..i . i p I"> c«s >, I I > M M <1I''«'<e r . r e n cv'i <i 44'4<0444444 i u ' . >s.4 i>''a.'u>i44>u'Ososas>u>u4s<O>a'>P»n

r>V ~»P In<.> ~g ~ .. Tqv a>ut>iu.y Mr<et.d ~r><.5~410 Vt-c ~>. ev c ~rv> v u v i >, u> u c c c c uvuve~uvpvthc<I'>i>»I>eea I

n v C c i C c ~ ~~NP,>, <,n> >pl i.p'. p><'>l<p)p<t>M>v>,p<M>s>>»'I~ < n p>ns>vn»:' >t cs<'v>vci»:,, ': 'hrhnvr><>rara ~ ~ I'<NNP<«c<pvrvc<Nn r,

steam to the turbine

h afeedwater

3(I

suction headei QCP

~@Iso

670P pressure.

header

7RY gate form~intenanc~

fu~lchanne j

pig.~4. Hydragljc scheme of the RBMK reactor csrculatxon cxrcuxt

dH, m

30

minimom o*vjtztjpn

2042000

LigI

/23na /2V 47@'<'», min, seo )

H> right loopii> left loop

l"i .15. Ave .6- - raged head variations MGPs in cf the reactorright and left loops

permissible pl es'.I lredrop rate

Z 2/op f. 22,N /duo VZ4'00

Pig.16. Pressura drop rate in DSd'P 6 L

c/5

(h,min, seo

Nwo

Io

Fig.17. Relative power variation for

G -- const and~in < acS

accorging to Fig.10

0,< 0,2 o~ <

Relative power variation after insertion

of equivalent reactivity l.™ig.10)and

different void fraction in the core inlet

0 — boundary condition

X - boUndary cond'". on

Gin = const

P~,.>= const (74 bar)

P~s = const (54 bar)

O,S

the loops by half

can to result in

a rapid power

surge and a

melting of fuel

I

I

I

I

I

B aso

60

5,0

2,01

I

<,D r,s

Fig.20. Axial distr buttons oI heat generation

P. k.

Simulation of reactivity and neutron fields changes

at reactor shutdown

N. N ~ Andriushchenko, L ~ V.Bidoo 1ya, A ~ C ~ Ka lug hin,4 ~ N.Kuzmin Kurchatov Atomic energy institute

lt is known[1],that one of the factors,wich caused severityof Chernobyl accident, was the insertion of positivereactivity at shutdown of reactor.

/I

We must remind, that possibility of that strange phenome,son

is due to absorber rod constructionswhen absorbing part ofthe rod is withdrown from the core, inside it is nonabsorbergraphite part, replacing the water and below it the watercolumn with length of ho=i ~ 25m.At the shutdown absorbersinsert in upper part of the core(;negative reactivity3 but inlower part the water is replaced by graphite Epositive reac-tivity) ~ The summary reactivity change p is determined by theinitial form of axial field 4(z) and its deformation C~(z)at moving of the rods

y. (,g y'(~) a: „y „>„~4.b)0"()da-

( q. (~ ) P"c.>»where H is the height of the core.

As any difference effect, this value is badly calculatableand therefore different methods gave for p the meaning from

zero to 1.5!3C1)~We have made experiments to simulate shutdown effect and to

receive data for testing of calculation methods.

The experiments were realized on graphite assambly with

height of 7m and diameter of 7 ~ 5m.We investigated four typesof assamblies with dif ferent initial axial disteibutions. The

cores of all as amblies contained 660 channels with naturaluranium fuel elements,36 channels to form axial distributionand some absorbers to provide criticality of the assambly(10-22rods) .The special group of 20 channels had in lowerpar'f the ~ater columns 1.5m height.Af ter criticality the absorberrods inserted in these 20 channels at d' ferent depth and at thesame value decreased the water level below. Then .»e measured thereactivity and neutron flux distribution on each water level.

The initial field forms provided by loading in 36 channels

some fuel elemer<ts consistinq of al 1oy of alvmir<i«m

enr i ched urar<ium and orle II<eter I i»t s of at sort er !n 4

different sequences.Figs-1-4 give measured a::ial distr it utior<s of r<e«tl or< flu:: fc-

a I I f o <j I a s s a m ti I i e s ~ I n i t i a I d i s t r i t u t i o r< i n the f i r s t a s s e m t.: .

is symmetr ical ~ The flu:: peal ir< ~'r<d 4 assambl ies is shif ted t=lower and upper part of the core, respectively ~ At theseassambli es r<eutron flu:: di str ibut I on i s stable and f e-<

def or]i>Ates at the rods inser t inn ~ C>r< tt<e other hand, di strib« t i o

in third assamblyI initial form is double peat ed) is quitunstahl e ~

Tt<at el:plair<s the resul ts of measured react ivi ty ct<anges, gi ve -,

at f ig-5.you may see, '. t<at positive reactivity changes ar =

observed at assamblies 2 and:." The peal., positive reactivity is~ (>- I,':(<:=0-0(>F4),that is -4/ from the ful 1 ef f i ciency of allrods ~ It is seen also,tt<at ir<sertinn of r<egative reactivit.begins ol'<ly after abso< ter rods doscer<d at the depth ~' 5-:m ~

Tt<e ir<ter est ing ef feet is observed at curve 4 at f i g ~ .:afte -.

rods inser tior< to the depth 6m reactivity r ise begir<s ~ Thi I

pher<omenon is due to length of the rod is letter than that c =

the core-Ot viusly, such s i tuat i on shouldr<' admi t at tJPP reactorThe e::peri<>ental data v>ere used f or testino one of calculati c-

scheme, basing on s!ni s —sour ces method [2] - Tt<e cal cul et i ons we

made at t>>o-gr oops, th ee —<imer<s i or<al ap pro:: imat i or<(1> I ayer s

a>:ial direction) ~ Each homoger<ious part of each char<»el

des cr-it ed wi th I> 'lt: edo matr i I at thi s method ~ Tt<e < atr i ce I

coef f i ci er<ts calculated'ith 1 ><ow>< 1JINS program ~

The accuracy of the scheme i I lustr ates the table,l>he<- =

e::per imer<tal "r<d cal culat<: I valuir<vest i gated e= ambi i es at upper

TABLE

es of I'.4H are giver< for alI Od pos i t iol

As samb I yC><p

~ftt ~Bc

elfC>n may

1 ~ oooo

1.(>(llgI .(>0(>(> 1 .Cj(>(>(>

(l. yg4B (>.ggBF.

see that error at Vo value is ]atter

1.<><>IB

(> qrv~+4

than (> ~ 5/ ~ Tt =

el I'ol's at nc vtl on f I u:: di st r ibut i on are about I(>—1 /

Figs-6,7 give calculate1 and e::per imental a::ial distr!but!orf o r a s s amb I I e s 'r< d: - F i 4 s .'.=:,9 give reactivity changes at tt'=

rods ) n«ei 1> nri )ri t tie cores of 11«si..or i':s 'Amer 1 le ~ 1 t

eeri, that calculat) nr . sl sght ly over e. t sn 1 c peal poss-reactivity ~

The same calculation scheme was used for the estimatiori

pos itive reactivity Insertion at the irii t ~al stage of Che~ ri": . 1

accident ~ There was appr o>:imate simulat cori of reactor: ..1

channel s supposed contained the fuel wi th equal a rradiati ori ~ -. 'ie

initial rods position also was tal.eri appro.:xmat ly so - . crt

summary reactivity margiri was 8 rods-The calculations gave the r cacti vary r ise after 'sec of

movement, equal 1 ~ 2(~(!":=0~ 005) ~ The est imat i ori is conser vative e 5

follows from testirig of calculation schemeCsee above) ~ The mo = =- rn

calculations[33,hased ori dyriam>c codes arid correct simulztio- ofreactor condi tiori,gi ve peal.. posi ti ve react ivi ty equal 0 ~ 75fi-

We want to remind that now the construction of the rod 15

changed so,that at upper position there xs no water column u-:ergfaphi te par t f see f ]g.10) That measure e:rludes posx - . e

reactivity insertiori at the reactor stiutdow i

F,eferences1 'he accident at the Chernotyl NPP:one year ari ~ lAEA-CN-08%c .

2 ~ S ~ S ~ C'or odl:os Comparison of net algorythms of calculatiori ofrieutron f ields iri big reactors ~ "Atomnaya Energhiya",V6 C ~),1p1 -.4-190.:-Analysi s of the first phase of C'hernobyl accident ththree-dimen«ional tlodel ~ The report or> this worl shop ~

Axia8 c(is(i'6 lions f(z> ~ ~ ciss~

2 (n ) el ispo n re. (1o~ bl>a golf>m ~) copy>I p pea t Im sf vj (o?e

).0

6,o

poSit,'oWo] 2> fogSiW Coqa

LI f> pt1 ei ~ p (~) cl j 5'tc ncaa /to~ t he fo1'f>~ot 'H c. c.o Re 0) &he coper ,,I

(OS'e'tcON

Oj 2> gok5ah@ CO7g

Ll.o

vvAt e t Pe vcr~ g 2o cha~wrCg

0

2

3

p(g p

Axial dis<tifNkio~g ~3a4 t sse~CPy NS

ss ppc l 8estgQ p g+4< cc va

7.o

yll I/

I

poSit>ozo) 2otod5

in f4 Co~C

.I;Shore j.oo, Rl,o Cotto o, o) 8r: rotC

vv te. 8evcCC 'f 2~ c 's g sswg gj

2.0

Axis& clipt'I CIAi05c, f(v'J o+cycler

p py

w ppet 8i~,gThc, cong

~~iS<c~ce (~a~ the gogo~ c,$ )gp - p~

G.o

w leo (eve(ll% 1+ckA'HH f. pg 20

1,0

I

ciCpth of ih5C~H ia n ( h) ~) POC gSa't CPS ~ogg in Hh

(Q Pw~;f5

> f - D.ao &~~ )o.5 .

C~ ~

.-).oJ-—

aspic ~ P Py

j o,gg~~C8y gp 5 c„c.~ Hy

g,gga~5Cy

F;~ s

qo

a l.I

~ ~ s ~,g I i ~ i ' ' a i ~ ' < E. ~ ~ L ~ < L i ) 1, ~ i

up~~,48'7: ~oct> w'<lid.'? >urn (<)Voefg ibsC;Hect (2)

P (~) cl's9c ~cp f~o ~Aha'at'fo~ g +s? co')~

gggoz4e.w

I 'I

wc4ez1.Z'S~

2.,$ w

hSvntelic

~

L) f 0 n g) u Ps+26~ >oJ ln 3~6 C ~veops

gegye accialawk (c)> agfez. c <cree~% (8')

0.2

I

-9he .depth o) ~oRs i~se~Hio< ii ale co>e,.nsg~v gCy x p

p ()::},,'-a.ooyw)

0

.—).o

. 2eO

-E.o

Lg ~

«p(z) a9 assa~ 4 Cy hl 6: loJ5, wi'fl el~o ~> (< )~ Se>dec( (z)

p (~) e(i S1o n(( )po ~9~e 6aH~m oj 9) e Co 2 e

L) g

The I st. Int.ernaticnal Werkshop cn ~~vere Acct dent.s i n

Nucl ear P "~er Pl ant.s

ANALYSIS OF THE MAIN CIRCULATING PUMPS (MCP) BEHAVIOR

IN THE PREACCIDENT PERIOD AND THE FIRST PHASE

OF THE CHERNOBYL ACCIDENT

E. V. BURLAKOV

V. A. BABI N

L. P.. KEVORKOV

E. -. NAVI NSKI I

V. B. PPOKLOV

A. A. ZHPAKOV~KI I

Dagomys, Patchy, U~

ANALYSIS OF THE MAIN CIF.<.ULATING PUHPS (MCP) l3EHAVIOR

IN THE PREACCI DENT I-''ERIOD AND THE F14''H~EOF THE 'EPIJOBYL ACCI DEhlT

E. V. BURLAKOY, L. R. KEVORKOV, V. B. PROKLOY. A. A. SHPAKOVSKI II. V. KURCHATOV INSTITUTE OF ATOMIC ENERGY, MOSCOW

V. A. SABIN, E. G. NAVINSKII

ECIAL MACHINE BUILDING DESI "N BUREAU. GORKII

Uni on oi '=ov; et Soc]a1 i st. R~publ 1 c

>t LI$1RACl

ANALYSI S OF Tk<L'IAI kl CI R UI A Ii 'B! i"l)l'. ':,C Ihi'P'I L>S SiAVS. t;Si

IN THE PREACCI DENT PERI OI) ',iili ".II I,II'. I's ';..'>ll >.'":.l.'i TIIE

CIIF.Rkss.:!I'; I ..

Il'se

parameter

>.:;;:s>.='.=.rs;> s;>c" <>.- t .:-, s i s s >:>f . 'r»''.Vsl-B> . s I'nd t he

R>BMK-1OOC> reactor circu!at. s ",r: <..ir: s!1 . a>'i. I.:;"-i.-;ns>.d >rd analyzed

t,o estimate t.he effect, of pump cava s.at.ic>s~ ors t.he onset. and develo-pment. of t.he acci dent. at. uni t -.I. of t.he Cher nol~yl I IPP. The MCP be-havior in t.he pre.iccid~:nt;;.r i -:I .::.s>"I I,, t.he fir st. phasss of t.he

;-;ccid.:-nt i s di scs ssed. I t. >."-. -:.!..osr> t.h;> ..;.I 1 MCPs had b en provi-ding a st.able wat.c:.r del I v.=ry t.i 1 1 dest.r uct.ion of t.he ."eactor. No

convincing evidence !>'.r. 10-1'~.';.:.>~u'ti on i st I.lse sos,a1 coolant.

f1 owrat.e t.hr ough t.he core was ca>'=,!td pr i sss~.rely by cavit.at.ion has

been reveald. This flowrat.e r -duct.ion is acs ounted for by slowing-<Jowl'> of MCPs connected t.o t I se r unni ng-down t sr>- hi >;e ger er at. i on.

1 . INTRODUCTION

It. follows fr om t he material s .on t he Chernobyl accident. t hat.

t.he react. or runaway was, caused by r api d l nser t.ion of t.he posi t.i ve

react ivit.y in particular due t.o coolant. boiling at a high posit.ivevoid r eact,i vi t.y coef f icient, and t.o unscheduled st,ate of the reac—toi (11. It. is evident. t hat t he coolant. boiling rate in the first.phase of t.he accident depended both on a change in t.he coolant. flow

through the core and on a change in the reactor power generation.Since in the preaccident period it was little changed it is naturalto ask if there was a significant decrease in the oolant flow

through the react or ci rcul ati on cx rcui t. for exarnpl e, due to "cavi ta-tion breakdown" of the MCP or if the steam produced in the working

channels of MCPs because of cavitation boiling could entrain themore and give "pulse" to rise of the reactivity the more so as in'e preaccxdent period the react. or had been operating at. low power

with a little delivery of the feed water and,hence, with a small ma-

r gx n to cavitation C NPSH ).The e questions were raised and then explained in the analysis

of the accident in 1986. However they have not been appropriatelypresented in the publications and this seems to the cause of appea-r ance of some di sputabl e comments abr oad [ 2) .

The pre ent paper gives a detailed analysis of all availabledata on the 'dCP parameters and circulation circuit of RBMK-1000 re-act or i.n the preaccident and initial pericds of the accident inorder to determine the role of the pump cavitation in initiatingthe reactor runaway at the Chernobyl NPP-4 on April 26,1986.

The compar i son of the data with the pr essure and cavitationcharacteristics of the MCPs make it possible to answer the above

questions and to de'. rmine more accurately the coolant par ameter

at the core inlet in the xnitial period of the react. or runaway.

2. THE CHARACTERI STI CS OF THE CNPP-4 REACTOR MCPS

The circulation circuit of the RBMK-1000 reactor consists oft wo parallel loops ( one of them is seen in Fig. 1),each loop inclu-des two steam separ at. or s and four MCPs. Under nor mal condi ti ons

three MCPs ar e switched on and is in "hot" reserve. Before the acci-dental experiment all eight MCPs were in operation.

The CNPP-4 was provided with the modernized pumps CVN-8 I3]having the following characteristics:

mot. or power cons umpt. i on

capaci t.y

head

4800 kW

8000 m .<hr3

200 m wat.er column

1000 rpm

The pump aggregat.e CYN-8 pumps from 5000 m lhr t.o 10000 m ihr3 3

of wat.er at. a t.emperat.ure of 270 C and a pressure on the suet.ion0

sade of 7. 2 MPa. It. includes t.he mot.or and the proper pump which areI

clast.>ca.'ly coupled. On ~ he MCP rotor t.he impeller 1 inst.alled,which in> rea es t.he rotat.i ve moment. of t.he rot.or up t,o 1'5000 kg. m,

t.husprclongx ng ~ he t.i me of t.he pump rundown ( t,he MCP rundown-1const.ant.i 20 s '0 and permit.t.ing t he MCP capicity ' be maintained

xn the ini t.ial period of the accident. when t.he power supply is lost.or cut. of f .

known for xt.s t able operat icn t.he centrifugal pump must

ha.e h;drost.at.i c. head of its x nlet for t.he gi ven pressure of sat.u-rated steam of the I;quand pumped. This hydrost.atic head ( margin tocavi t.at ion -NP~H ~ equal t.o:

P — Pinh

r equi red C m pumped liquid column'3

wher e P 1nP s

is t.he total pressure at 1 he pump inlet. Pa

is the pressure of'aturat.ed steam of t.he l iquid

pumped, Pa

is the

i s t.he

pumped 1 i qui d densit.y, kg,>m 3.2free fal l acceleration. m/s

NPSH x s r equx r ed for prevent.i on of st.earn gener at.i c n due todynamic preassure drop in t.he pumped liquid flow during it.s rundow-

ning in the pump working channels. In Fig. 2 t.he required MPH isplotted as a. function of the pump delivery (curve 1) using the dat.a

obt.ai ned in cavi tat.ion t.esses wi th "cold" C 20-70 C ) wat.er in t.he0

CVN-7 pump who e work '.ng channel is t.he same as in CVN-8

It. also is known [ 45 ] t hat. with rise in t he t emperat ure

of the wat.er pumped t.he cavit.ation characterist,ics of the pumps

"improve". Fig. 3 shows t.he CVN-7 .avit.at,ion characterist.ics t.aken at.

275 C ,and Fig. 2 - t.he depedence of the permissible NPSH at. t.his0

temperature on the pump deliver y C curve 2This ".impruvement." of the cavi t.ation charact.eristics is accou-

nt.ed for by change in the thermophysical propert.ies of t.he wat,er

with rise in t,he temperat.ure and pressure C as a result. of decreasein t.he volume void fract.ion in t.he flow of sat.urat,ed wat.er when t.he

pressure in t.he flow fall below t.he aturat.ion pressure).Therefore when t.he ultimate NPSH are reached in operation wit.h

"cold water" an instant.aneous cavi t.at.i on breakdown in the pump

capicit,y may cccur. Unlike this the RBMK circulat.ion circuit, at.0a t.emperat.ure exceeding 250,C per nu ts. a short. — t.ime pump operat—

ion ander condi t.ions of the i nit.ial cavi t.ation with a small fall inthe pressure in t.he wor king channel of the pump below t.he sat.ura-t.ed st.earn pressure, which is support.ed by t.he data of Fig. 2 and by

t.he result.s of more careful inve tigation of t.he cavit.at.ion chara-cterist.ics of t.he pumps depending on t.he temperature of t.he water

pumped [ 5 3 .

The analysis of t.he results obtained from CVN-8 test s shows

t.hat. curve 2 of Fig. 2 is rat.her conservaki ve sinse during t.he t.est.sr cadi ngs of t,he manometr i c indi cator of the NPSH were cor r ect.edallowance for t.he air released f rom t.he wat.er inside t.he manomet.er

working vessel but. wit.hout. t.aking i nt.o account t.hat. this air alsoreleased in the working channels from the underaicrated wat.er of

. /,/

t.he experimental loop Dependence 3!'I'f Fig. 2, obt.ai ned by

t.ion of the cavi tat.i on char aeter i +tie for "col d water" by

r ecalcula-t.he me-

t.hod descr i bed in f 5j should=-=be considered as a less conser vat i ve

estimat e of t.he permissible NPSH for the,wat.er at 285 C in CYN-8.0

The act.ual dependence of the permissible C short-time 0 NPSH appe-ars to be bet.ween curves 2 and 3.

The available NPSH in t he RBMK circulat ion circuit is beingprovided wi t.h t.he hydrost.at.i c head C MCPs ar e inst.al led by 30. 5 m

lower than the water level in the separators ) and also with cooling0on the pump suction as a result of the separ ated water C 285 C 0 an

the colder feed water C 165 C ).0

Urder the condition of the constant pressure in the RBMK cir-culation circuit the above mentioned hydro tatic head is sufficientfor cavitationless operation of MCP up to 6500-7000 m )hr delivery,which cannot be exceeded because of the reactor operation scheduleunder conditions of zero or minimum power -hen of the feed water

supply to the MCP 1nlet is .mall, if any ( Fig. 9 ).3

When the t ~ ~ P delivery exceeds 650<. -7000 m Wr at the zero feedwater flow rate it may,Le t he initial st age of "perm1ssible" cavita-tati on. The cavitation br eakdown of CVN-8 is possible only at a pump

delivery more then 8000 m Phr.

3. ANALYSIS OF THE MCP PAPAMETERS IN THE INITIAL STACE OF THE

CHERNCBYL ACCI .)ENT

Using the data of the diagnost1c data recording code ( DDRC 0

the 1 ns'. abi l 1 ti es of water suppl y r ecor der for each MCP has been

anal; ed C in the ~DRC 0 prints the dev1ation from the average value+is — 15 iQ. For this purpose the average flow through the circul ation

loops CFig. 4 3 wer e det»rmined at a t'e from 1 hour 19 min to 1 ho-

ur 50 min wher: t he circuit was at a constant pressure and had largeactual NPSH C Fig. 4 ) . It 1 . seen that the average water flow ineach circulat1on loop ( with four MCP operting in

deviate more than by 4 % of 28000 m ehr in the "riof 28500 m 1 hr in the "left-hand" one. The ins~ abi3

each loop3 did not

ght-hand" l o ~p and

lity of the DDRC

flow rate data is accounted for by an unstable output signal in thisflow mea uring system.

In the earlier period the MCP supply stability was not affectedby reduction in the water level in the eparators until the ceiling

200 mm " is reached, which is indicative of the lack of steamcapture in the circulation circuit bottom.

One to three minutes before the accident and at a feed water

flow rate of 400-800 tonne l hr the margin to cavitation was appr e-ciabl y higher than the permissable one ( even that required for"cold water", ee Fig. 2.8 and 10 ) 1n spite of some pressure drop

C from 6. 95 MPa t o 6. 3 MPa ) in the circuit.The time of the turbogenerator rundown experiment and the ini-

tial pha e of the accident 1 very important for understanding thecauses of runaway of the reactor which had been operating at 200 MWt

before the experiment ~'and had had only few absorber-rods insertedinto the cor e.

At 1.23'. 04" the experiment began. the electric voltage on thepower supply buses of two of the four MCPs on e-ch half-circuit ofthe reactor began '. o decrea e and . hence. from that time the rota-tion frequency of the "running-down" MCPs began to reduce C Fig. 5Two other MCPs in each half -reactor maintained the nominal rotationspeed ( 980 r pm ) .

For the analysis of the MCP r egimes it is important to have thedata on variations in the rotation speed of the "running-down" MCP

in 'he preac ident per iod.During the expe. 'ent the oscillograms of the rotation speed of

one r unning-down MCP were taken in each half-react. or C DDRC recordeddeliver y fr om all MCPs ).However the stub which recorded the signalsfrom one of ~ he MCPs was faully and,therefore, we have the data on

variations of the rotation speed of only one of four running-downMCPs. Taking into account that the electrical par ameters C voltagepower ) changed equal l y for a "l r unni ng-down MCPs ther e 1 s everyreason to suppose that the MCP rundown in each half-reactor occuredpr ac t 1cal l y s ync hr onous 1 y.

Al'nment of the osci1 logram of MCP rundown in time with thepar meter s r ecor ded by DREG was made by the mar k of vol tage 1 nduced

on the stand —by ection of reliable power supply.Fig. 5—7 how change in the capabilities of all MCPs in each loopof the ci rcul ati on ci rcui t as wel 1 as t he r ot ation speed of the run-

The First International Workshop on Severe Accidentsand Their Consequences

Sochi, U .S.S.R., 30 October through 3 November, 1989

HUMAN FACTORS ANALYSES OF SERIOUS ACCIDENTS:

CHANGING FOCUS FROM HARDWARE TO MANAGEMENT

Philip E. Berghausen, Jr., Ph.D.Rohrer, Hibler, and Replogle, Inc.

1601 Old Bayshore Highway, Suite 260Burlingame, California 94010

UPS.A.(Tel.: 415-692-0652)

Abstract. This paper describes the areas that arereviewed in a Human Factors investigation of asevere accident. Hardware is examined for designflaws that may have contributed to causing humaneto operate or maintain it unreliably. Controlsand safety systems are scrutinized. It is noted,however, that recent analyses, of human failuresimplicated in serious accidents, suggest thatmanagement is an even more important considerationthan is hardware. Thus, a thorough Human Factorsinvestigation of a serious accident must alsoexamine such management factors as organizationalstructure, mission, talent (including howpersonnel are screened, selected, monitored forbehavioral reliability, trained, and promoted),and internal processes (including what behaviorsare rewarded, how mistakes are handled, and howdecisions are made).

Human Factors, as a discipline, studies the interfacebetween humans and machines. The mechanization of the industrialrevolution, followed some 150 years later by automation in thechemical industry, did much to prompt human factorsinvestigations. Of primary concern in these investigations wereanthropometry, physiological work climate, and sensory-motorabilities [1).

Human Factors accepted human characteristics a;-. "givens" andcaused hardware to be designed to accommodate these givens. Itappears that virtually no attention was devoted to the

possibility of altering the givens. Similarly, relatively littlesystematic attention was devoted to designing manufacturingorganizations —as opposed to their hardware —to fit the givens.

Human Factors has expanded considerably in scope in the pastdecades, but it continues to attend to the issues that launchedit. In the wake of serious accidents, hardware is subjected toclose scrutiny. Hardware specialists look for the causes ofhardware failure. Human Factors experts look for hardware designflaws that may have contributed to causing humans to operate ormaintain it unreliably.

Hardware Factors

Following accidents in the nuclear power generationindustry, Human Factors experts examine a variety of hardwarefactors. For example, they look at reactor responsecharacteristics to see whether these characteristics allowoperators sufficient time to take corrective actions„ underabnormal as well as normal conditions. Are there safety systemsto protect against all foreseeable contingencies'? Are thesesafety systems capable of being disarmed? Are there safeprocedures for testing and maintaining the safety systems?(Note: Many safety systems are designed to protect during normaloperation, not during the abnormal conditions that can occurduring their own testing and maintenance.)

Control room design is another major hardware factorexamined by Human Factors experts after an accident at a nuclearpower plant. This is the so-called "knobs and dials"investigation. Placement of controls, readability of gauges,adequacy of alarms, etc., are reviewed as part of control roominvestigations. Inadequacies, and their potential relationshipto the, accident under study, are identified.

It has been argued recently, however, that this hardwarereview, in many respects, is (or should be) only the "tip of theiceberg" when it comes to Human Factors reviews of seriousaccidents (1,2] . Professor James Reason, in his taxonomy ofunsafe acts, identifies two basic types: errors and violations.Eri;ors involve such processes as attentional failures and memoryfai'ures. Violations involve deviations from procedures deemednecessary to maintain the safe operation of a potentiallyhazardous system. Professor Reason makes two very important'assertions with regard to errors and violations: (1) In recentserious accidents, violations have played a much larger role thanerrors; (2) Violations are more a function of managementpractices than they are of hardware design [2].

If Professor Reason's assertions are accepted, any thoroughHuman Factors review of a serious accident will require an

investigation of management practices. To explain what thisreview should entail, it may be helpful to discuss management, ingeneral, and bureaucratic management, in particular.

Manaaement ":actors

Management determines what work is to be done and how it isorganized, staffed, supervised, evaluated, and rewarded.Bureaucratic management —the type of management that runs suchlarge operations as power plants —can be superior to other formsof management in terms of economies of scale, reliability, andconsistency. But, these advantages too often can come at a highcost: maintenance of control syst ms (budgets, plans, etc.) cansap a disproportionate amount of staff time; decision-makingprocesses can be centralized and too far removed from importantdata sources; personnel can become much better at followingpol'cies than at thinking for themselves. The potential negativeconsecuences are inefficiency, poor (even dangerous) decisions,and inability to resoond to new situat'ons [see 3,4].

The thorough Human ractors review of a severe accident willbe especiallv alert to the presence oi these disadvantageousaspe "s of bureaucracv. It will proceed systematically throughthe four aspects of management represented bv. the followingdiagram.

KI1K)55@II'QPllf

OoaaOCeeClnel

AplCIN

Sita~

~lLtNfCavoMernwI%itOnnanCe

01OJNtZATIO 1SI~QloSuncuosILAW

~severeSlaneaea

Organizational Structure

The overall organization or organizational structure will beexamined. Is it "tall" or "wide?" Are there many layers ofmanagement? Is there a route for information to make its way tothe top? Are there provisions for monitoring quality and forconducting short-, mid-, and long-term planning? Is authoritydiffuse or focused? Are reporting relationships clear or arethey complicated or undefined? Are there separate business unitsor profit centers? Are all these important areas coordinated?Is there a matrix system for ancillary services? Does it provideservices equal in cost and quality to those that could beobtained outside the organization? In sum, does the structurework for or against the organization?

Mission

The organizational mission or purpose will be studied. Isthere a clear purpose? Is it consistent with the capabilities ofthe organization? Are all personnel aware of it? Are thenecessary planning mechanisms, etc., in place to implement it?Has the purpose been translated into short-, mid-, and long-termobjectives? Do the objectives contain process provisions, aswell as those for content? (Are they concerned with how thingsare done, not simply with what is done?) What other "balancing"factors are there--i.e., factors which keep one aspect, of themission from being taken to an undesirable extreme?

Talent

Talent refers to the personnel resources in theorganization. A Human Factors review of talent will examine howpersonnel are screened, selected, monitored for behavioralreliability, trained, and promoted.

Screening. Screening refers to eliminating those who areclearly unfit for a given job. Screening variables includeintelligence, psychopathology, history of antisocial behavior,and substance abuse [5] . Thorough screening should eliminatethose who are likely to display argumentative hostility towardauthority, irresponsibility, defensive incompetence, adversereactions to stress, and emotional or personal inadaptability.Persons who display these problems deviate from the norm too muchfor the organization's built-in compensatory systems toaccommodate them.

Selection. Selection refers to hiring "the best of thebest" (whereas, screening focuses on eliminating from furtherconsideration the "worst of the worst") . Selection requires thatrequisite technical skills (or potential to learn them) anddesired management style, for any given job, be fully established

before the selection is made. P~rsonalities of superiors, peers,and subordinates are also considered. Candidates for a positionare compared to these descriptions. Those representing the bestoverall fit with the needs of the team are chosen. Selectioncriteria are reviewed periodically for currency and relevance.

Behavioral Reliability. Behavioral reliability programs arepredicated on the fact that people change. Someone who passedscreening and met selection criteria can deteriorate in abilityto function. Psychopathology often waxes and wanes. Substanceabuse can begin after years of moderate use or even abstinance.Antisocial behavior can arise after an unfortunate encounter withan undesirable influence. New diseases can be contracted.Physical trauma can impair brain functioning. Improperlydesigned shift rotations can lead to cognitive deficits andemotional lability.

A well designed and executed behavioral reliability program .

(roughly equivalent to a fitness-for-duty program) detectschanges in work habits, social interactions, and health.Corrective action is taken promptly, thus assuring that personnelcontinue to meet high performance standards and will be likely toperform as predicted in emergency situations.

Training. Training provides personnel with skills that theyneed to do their jobs. Generally, training should be skill-based, rule-based, and knowledge-based [see 1] . "Skills" refersto a practiced ability to execute a task. "Rules" refers toknowing when to execute the task. "Knowledge" refers to anunderstanding of why the rules exist. (Knowledge-based trainingis particularly valuable in emergency situations in whichcreative solutions to problems may have to be devised.)

Training materials must be relevant (plant specific),thorough, and unambiguous. They should be updatedsystematically. They need to be reviewed regularly by thetrainees ~ Skills need regular practice.

Promotion. Programs for promotion should be systematic.When well executed, they permit succession without decrements inquality, reliability, etc. The best candidates for promotion (asdefined by competence and potential —not favoritism) areidentified in advance of need and are developed, through trainingand graded increases in responsibility, for the positions theymay assume.

Internal Processes

Internal organizational processes, or organizationalculture, will be examined. These include, among others, whatbehaviors are rewarded, how mistakes are handled, and how

decisions are made.

Rewarding behaviors. Three important principles ofmanagement and behavioral psychology are: (1) People do thatwhich they are rewarded for doing; (2 People do that whichthey are rewarded for doing in the easiest way possible (thatstill will result in their receiving the reward); and (3) Rewardscontrol behaviors much more effectively than do punishments. Inexamining the rewarding of behaviors, several important questionsneed to be asked:

--Where is safety in the rewards hierarchy? (If it is notat the top, it will be compromised no matter how muchviolations are punished.)

—Are the rewards that management provides ones that peoplereally want? (If not, rewards provided by other sourceswill have a greater influence on behavior.)

--Are the behaviors that are rewarded consistent with theoverall mission of the organization? (If not, the missionwill not be fulfilled.)

--Are rewards available soon after the desired behavioroccurs? (If not, the reward will have much less effectthan it could have.)

--Are activities rewarded instead of results? (If so,people are likely to be very busy without getting muchdone.)

--Is it easier to make things happen or to keep them fromhappening? (Exercising power is rewarding. People willexercise power in the most available ways. When saying"no" is consistently easier than saying "yes," theorganization will be characterized by inertia and

. complacency with the status quo.)

—Are the consequences for poor performance much differentfrom those for excellent performance? (If not,performance that exceeds minimum levels will be rare.)

Mistakes. Uncorrected mistakes are "accidents waiting tohappen," or in Reason's terminology, "resident pathogens [2]."Uncorrected mistakes include maintenance errors, design errors,management decision errors, procedure specification errors,conflicting mandates, etc. Consistent with the precedingdiscussion regarding rewarded behaviors, unless the detection,repo ting, and correction of mistakes is rewarded (even if theyare one's own mistakes), mistakes will remain. An organizationis in trouble when it is better for employees to be quiet than it

is for them to be right

Decisions. Where are decisions made? Are they made at thelowest possible level in the organization? Are they made withgood access to relevant data? Are they made by consensus, ratherthan by majority vote, but without undue peer pressure? Are allconstituencies affected by the decision appropriately polled (fordata as well to promote acceptance of the decision)? Are theycommunicated promptly? Does everyone have the authority to makesome decisions on his/her own? Is the authority commensuratewith capability (no more and no less)? Affirmative answers tothese questions are positively with correlated good managementand ultimately with safety.

Conclusion

As important as hardware considerations are in a HumanFactors review of a serious accident, management considerationscan be even more important. A thorough Human Factors reviewshould, in addition to reviewing hardware, review such managementaspects as structure, mission, talent, and internal processes.

References

[1] Rasmussen, Jens. "Human factors in high-risk systems,"presented at the 1988 IEEE Fourth Conference on HumanFactors and Power Plants, Monterey, California, June 5-9,1988.

[2] Reason, James. "Errors and violations: The lessons ofChernobyl," presented at the 1988 IEEE Fourth Conferenceon Human Factors and Power Plants, Monterey, California,June 5-9, 1988.

[3] Lawler, Edward E., III. "Control systems inorganizations," in M. D. Dunnette, Ed., Handbook ofindustrial and organizational psychology, Chicago,Rand McNally, 1976.

[4] Rohrer, Hibler, and Replogle, Inc. The managerialchallenge: A psychological approach to the changingworld of management, New York, New American Library,1981.

[5] Berghausen, Philip E., Jr. "Demographic, behavioral,and psychometric characteristics of persons deniedunescorted access on the basis of psychologicalassessment," presented at the 1988 IEEE FourthConference on Human Factors and Power Plants, Monterey,California, June 5-9, 1988.

(rev ~ 11/89)

ning-down MCPs and change in t.he current. str engt.h on the buses oft.he st.and-by transformer of t.he running-bown MCPs. Fig. 6 also shows

change in t.he circuit, pressure measured in t.he st.earn separat.orIt. fol 1 ows from t. hese d a '.a:-from t.he beginnxng of t.he experiment C 1.23'04" ) t.he rot.at.ion

speed of t.he running-down MCPs began t.o decrease slowly. By 1.23'45"C beginning of rapid runaway of t.he react. or ) rot,ati ons of t.he MCP

rundown t.oge'er with t.he t.ur bine generator had reached 770 rpm

-at. 1.23'44. 7"t.he overpr essure mergency ystem in t.he electri-cal c;rcuit..witched off power supply from the mot ors of t hese MCPs

and t.he running-down cont.inued due to the inert,ia of the rot.or but.

at a somewhat higher rat.e of the r ot.at.ion frequency reduction beforet.his t.ime. However by t.he t.ime of rapid runaway of the reactor t,hi

rat.e exceeded 700 rp.".".

-for the normally operat.ing MCP the power supply from t.he

stand-by t.r ansf or mer mai nt.ai ned at. l east. t.i 1 1 t,he t,i me of the r eac-t.or dest.r uti on

-the running-down MCPs C according t.o t.he DDRC data ) maint.ai-

ned a stable supply which decreased slowly because of reduct.ion int.heir rot.ation frequency

-t.wo MCP on each side of t.he react. or which were fed from t.he

stand-by transformer maintained a st.able delivery up t.o t.he moment,

when at 1.23'45" the r eactor t.hermal power began t.o r i e. t.he t.ot.al

delivery of t.hese pumps C Fig. 8 3 tended t.c a slight. ri =.e during t.he

last. 15" befor e t.he beginning of r apid runaway of t.he react. or-t.he t.ot,al flow rate over t.he circulation loops wi t.hin t.he last.

20 sec. was redusing ..lowly and by t.he time'of t.he react. or runaway

had decreased by 10-15 / as compared with t.he initial flow rate inthi peri od

-at. the t.ime of t.he beginning of rapid runaway of the react. or

C 1.23'45" ) when a rapid rise of the reactor neut.ron power and riset.he pressure in t.he circuit were recorded the MCP capicit y began t.oreduce rapidly because of rising heat. release in the core and accom-

pariying rise in the void content in the core and the pressure on theMCP head

—at 1.23'47" the void content stopped to rise C possibly dueto decre:-ise in heat re'ase to the coolant because of the beginningof the heat transfer crisis when the void content in the fuel chenn-als had exceeded 30 r. ); the core water ceased to come to the sepa-rator through the steam water lines, which resalted in reduction ofthe hydraulic resistance,and the MCP capicity began to restore C at1.23'48" the DREG records has shown restoration of the MCP flowrate ) .though it is possible that this increase in the flow rate co-incided with beginning of the reactor destruction and the coolantingress into the reactor space

-comparison of the cavitation margin which was available inthe circuit within the 20 sec. before the accident with that requi-red for the cavitatioless operation of the MCP can be made using thedependences shown in Figs. 2,6.8. Fig. 8 presents the comparisn of theelivery of two running-down and two normally operating MCPs in thecooling loops of the reactor. It is seen from the figure that the no-rmally operating MCPs, could have been involved only in the initialstage of the "

p rmissible " cavitation. Assuming that the minimum-

warrated value of th= feed water flow rate C 88 tonneihr ) couldriot. be accurate because of i ts r el ati ve smal 1ness and tak i ng i tequal to zero i.e. supposing .that the cavitation margin of the MCP

was determined only by the depth of submerging the pumps under thewater in the steam separator, it, can be seen fr om Fig . 2 and 8that even .nder these conditions MCP could not come beyound the li-mits of the cavitation regimes, retaining a stable delivery

Therefore according to the operation conditions in the experi—ment period the non-running —down MCP seem to be in the initial phaseof cavitation which practically could not affect their characteris-tics but could cause vibration in the pipe lines adJacent to MCP

Cth.- personnel has testified that noise was heard in the pump rooms)Though the evidence. about. a strong vibration of MCP referred possi-

bly to the MCP operation in the initial pha e of the accident Caf terthe button "AZ" was pressed ), when a part of fuel channel had been

de troyed after the "first" explosion and the water began to boil inthe downcomer pi ping of the circuit.

Sinse within the last 15 ec. before the reactor runaway thetotal delivery of the pumps reduced it is important to understand

how much the amount of water C roid content) in the core then changed.

Knowing the change in the water upply '.o the core C Fig.9 ) it iseasy to a es the void ccatent. It follows from the data of Fig. 7and the heat balance calculation that void content at the core onletdid not change signi ficantly within the period from 1. 23 '04" to1.23'45" C about 0,2 — 0,3 mass l ).

In order to see if there were no stronger changes in the void

content in the core than the calculated ones the MCP characteristicswere compared C taking into account the rotation frequer."ie of therunning-down pumps) with the parameters rer orded in DDRC.

This c~mparison has been made by the semigraphical method

(Fig. 12). Using the Q-H characteristics of MCP the point of theiroperation has been found as a r esult of calculation of the thermophi

sical parameters of the circuit at the time before the beginning ofthe experiment and at the time before the rapid runaway of the r eac-tcrC when the rotation frequen ies of the rotors of the running-down

MCP were 77C rpm ), under a ummption that the coef ficient of > he

hydrauli" resistance of the circuit r emained constant. C Because ofthe low roid content in;he core the dependence of the hydraulicresistarce of the circuit on the flow rate can be assumed to be

quadratic ).4s seen from Fig. 9 during t he period fr om 1.22'20" to 1.22'40"

the reduction in the feed water supply was also observed, which re-sulted in deer ease in the temperature of the pumped water by severeddegress and led to an additional slight increase of the void cont ent.

The estimate of tcore hows that it is

he total change in the coolant density in the3equal to 0,03 g l cm and 1 adequate to inse-

rt.ion of the posit,ive reactivity of not. higher t.han 0,15 — 0.2 P.Thu,because, of reduct,ion in t.he rotation speed of t.he runni-

ng-down MCP a mall rise in t.he st.earn content in t.he cor e was pos-sible, which, however, could give the initial pulse to the reactorr un a.way.

Si~ce the act.ual tot,al capacit.y of MCP oper at,ing at, t.he nominal

rot.ation speed C Fig. 5 0 is somewhal lower t.hat. t,he predicted one itcan be assumed t.hat. t,he rise void cont.ent. C t,hough slow ) began 10--15 sec. before the time of t,he rapid runaway of t.he reactor.

4. CONCLUSI ONS

4." . The analysis of the parameters recorded by DDRC and oscillo-grams of t.he running-down process in t.urbine generat,or N8 at. ChNPP-4

as wei 1 as studies of t.he indirect dat.a, character 1st.ics and t.he MCP

and reactor circulation circuit parameters have revealed no i ndica-t.ions of a significa~t. reduction in t.he MCP delivery and pre surebecau e of developing cavi tati on at the preacci dent, t.ime. The run-

ning-down and non-running-down MCP maintained a stable delivery even

in the init.ial phase of t.he react. or runaway.

4. 2. The cont,ribution of t.he MCP to t.he react. or runaway was not, de-cisive. Reduction in the tot.al MCP delivery by 10-15 l because of re-duct,ion in t.he speed of the running-down might only result.a small increase in the void cont.ent in t,he core and,hence, t,o inse-r t.ion of a posit.ive react.ivit.y of t.he order of 0,15-0,2 P4. 3. A sharp reduction in t.he MCP flow rat.e at, a t,ime from 1.23'45"is due to rise in t.he pressure on t.he MCP head because of rapidst.earn production i n the cor e.

REF'ERENCES

1. Information about. t.he accident. at. the Chernobyl NPP and it.s con-sequeces,prepared for IAEA.At.omnaya energiya,1985.v.61,N5,pp 301-320.

10

2. Nordst.rom. Was pump cavi t.at.ion t.he key t.o Chernobyl, Nuclear Eng.

I nt.em., v 39, N406. May, p, 19.3. A. S. Agalakov, V. A. Babi n et. al ..The experience of operation of t.he

maxn circulat.ion pumps at. NPPs wit,h t.he RBMK-1000 reactors. In t.he

book: Nucl ear power st.at.i ons, Ed. L. M. Vor oni n. N3, M. Ener gi ya. 1980.4. A. N. Stepanov, Cava t.ation propert.ies of fluids. Elect.rical machines

and inst.al 1 at.ions, M. Mir. N2. 1964, p. 122.5. S.S. Rudnev. B. A. Kolchugin. L. R. Kevor kov, The ef feet. of the para-met.er s of pumped l i qui d on the cavi tat. 1 on pr oper t, 1 es of t.he

pumps, Pr oceedi ngs of G. M . Krzhi zhanovski i, Po~er Inst.i t.ute, N35.

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CANAOA USSRseminar

CHANNEL TYPE REACTOR SAFETY

RDIPE-89/04-20

RBMK FUEL CLADDING BEHAVIOUR UNDER LOCA CONDITIONS

Research and Development Institute of Power Engineenng

Soviet-Canadian Seminar on

"Channel«Type Nuclear Reactor Safety"

MOSCOW, October, 1989

Research and Development Institute ofPower Engineering

RBMK FUEL CLADDING BEHAVIOUR UIG)ER

LOCA CONDITIONS

O. Yu. Novoselsky, I.I.Kryuchkov,Y.Xa.Abramov

RBMK PUEL ELEbSNT CLADDING BEHAVIOUR UNDER

LOCA CONDITIONS

O.Yu. Novosel'sky, I.I. Kryuchkov, V. Abramov

ABSTRACT

The off-core test results of the RBMK fuel element cladding

behaviour under LOCA conditions are presentaL The experiments

were carried, out in steam atmosphere at 6-8 %Ps under the

cladding heat-up to 800-1300 'C. In the course of these experi-ments the cladding integrity failures (seal failure or totaldestruction) were revealed depending on the experiment parameters

and, gap within "fuel" column that was over the range 2-20 mm.

The criteria of the cladding brittle failure are determined. e.s

a dependence equivalent thickness of the oxidized layer of the

cladding metal on the temperature.It has shown that the cladding surface exposed to steam is

considerably oxidized, and metal is saturated with oxygen and,

hydrogen that leads to significant reduction of placticity.

INTRODUCTION

The information about the fuel element ccladding behaviour

is necessary over a wide range of the parameters such as tempe-

rature, rat» of temperature change, gas pressure under the clad-

ding, coolant pressure and dy ~m~cs of its changing by performing

the analysis of reactor accidents.

Under accident conditions the fuel element cladding integritycan be failed by the action of mechanical stress or from oxidi-

zing of the cladding materials in steam atmosphere or combined

actions of these factors. The cladding integrity failure due to

mechanical stresses most likely take place in the location of

the fuel column rupture where the gap between the fuel pelletscan reaches 2-20 mm. Under emergency cladding heat-up the latterwring out the fuel due to external pressure and further its strain

stops. However at the locations of the fuel column rupture the

cladding s0rain is being continued. down to the meeting of strained

parts. In this case the cladding strain is accompanied by wall

thinning and the probability of seal failure is considerably

increased.

Another factor that acts on the cladding integrity under

an accident is the intensive cladding material oxidation (zirco-nium alloy) in steam. As a result mechanical properties sharply

worsen, i.e. the cladding metal is being embrittled and it can

lead to cladding failure.The actions of mentioned factors reduce the fuel element

reliability under an accident to a large extent. In this case

the fuel element seal failure can occur either during an accidex.c

progress or it can fail due to stresses arising under cooling

or the fuel element assembly'ischarge from the reactor.The revealing the various factors leading to the cladding

integrity failure during the emergency cooling of the fuel

element, their interdependence as well as casual behaviour of

the cladding seal failure during heat-up requires the systema-

tization and the gazing of necessary statistics and appro~<~tion

of the thermal and mechanical loading conditions to full-scale

ones.

TEST PACILITY Ai%) TEST SECTION

The investigation was performed at the rig the schematic

diagram of which is presented in Pig.1. The main component of

this test rig is the test section where the test specimen isinstalled. Slightly superheated steam with the follcering para-

meters: pressure - 6-8 %Pa, temperature - 290-3104C and flow

rate - 0-100 kg/hr is fed at inlet of the test section. The

heat-up is performed using current passing through the specimen.

After keeping under these conditions the specimen can be cooled in

steam medium or by subcooled water that can be fed to specimen

after proper valve switching.

Schematic diagram of the test section is shown in Pig.2.The specimen represents a section of RBbK fuel cladding of

Zr 1 % Nb —alloy having the length of 780 mm, outer diameter

of 13,6 mm and wall thickness of 0,9 mm. The test specimen;I

together with the vessel section form an annular: slot of 3,5 mm

in width where superheated steam passess from below up to the

top of this slot. One end of the specimen is rigidly fized in

the lid of vessel, and the second one is free to compensate

thermal expansion. On the inside of annulus the specimen is.centred using space grid located nearby the inlet. Current

supply for the specimen is performed using the upper and lower

contacts cooled with steam. The fuel pellet simulations were

performed through the installation of the ceramic {A1203) pellets

of 11,5 mm in dia and 16 mm in height into the specimen tube.

Besides, each ceramic pellet has two openings for the thermo-

couples junctions of which were located at internal surface of

the specimen. The gap simulation in the fuel pellet column isimplemented by the spacer located for centre between the pellets,approx in the middle of the specimen. The spacer is made of steelwire of 0,8 mm in dia. The internal specimen space is open to

the atmosphere therefore axial thrust ~ 115 kg acts on the speci-

men end at steam pressure 8KPa. The discharge of the specimen

due to this force was allowed for the test section.

EXPERIMENTAL PROCEDURE AND TEST CONDITIONS

Test conditions for the RBhK fuel elements are given in

Table I'able I

Pressure, ZPa

Cladding temperature,'CTime of exposu"e,min"Fuel" pellet gap,zmSpecimen heating-uprate, ~C/s

6.0; 8.0800 900 1000 1200 1300

40-120 30«120 20-60 10-20 5-150-20

20-70

The following test procedure was used. Superheated steam

was supplied at the test section inlet. After the parameter sta-bilization electric loading 18 RW was supplied at the test spe-

cimen for 1-2s. The cladding heat-up to specified temperature

took peace for 10-20s. Further the achieved cladding temperature

was held constant during the certain time by the control of power.

Following that the specimen cooldown was performed. If fJxe spe-

cimen cooling was conducted in steam the electric load was turned

off and the process of the temperature drop occured up to the

parameters of cooling medium. If the heated specimen was cooled

by water (subcooling 150-200'C) the electric load did not turn

off, but the steam supply was stopped and water was supplied

instead of it. The cooling process was being continued until

the cladding temperature achieved 200-300 C. After full cooling

of the test section the latter was disassembled, the test spec.imen

was removed and the inspection and necessary measurements were

performed.

The experiment was stopped if the specimen sealing failed

during the heat-up process.

THE CLADDING STRAIN PATTERN

Under the specimen heating-up to the temperatures exceeding

800 C due to external pressure the cladding strain occured and

as a result the latter wring out the pellet column. Due to such

strain the specific annular traces appeared in place of the

pellet edge contact. In this case the cladding diameter decreased

For instance, the external cladding diameter vras 13,2-13,4 mm

at the pellets of 11,5 mm in diameter. The cladding pressing out

reveals the available defects in the pellets (i.e. pellet eplittizruptures between the pellets). These defects are the sites of

origin of increased. danger for the cladding seal failure under

its heating-up.

Depending on the gap size in the pellet oolumns the cladding

strain changed. The cladding strain diagram in the locations of

gap 2, 6, 10 and 20 mm is shown in Pig.3. The appearance of

strained claddings is presented in Pig.4. When the gap in the

pellet column was equal to 2 mm the strained. cladding had an

annular cross-section and in this case it maintained its stabilit„under pressure up to B MPa. Maximum thinning was observed in plac<

of the maximum flexure. Under the gaps equal to 6 and 10 mm the

strained cladding formed nearly symmetric three-beam fold that

practically converged to the centre of the fuel element simulator.

Maximum cladding thinning took place in the locations pressing

out the pellets on the butt-end.

EFFECTS OF TEMPERATURE, PRESSURE AND GAP DIMENSION IN

PELLET COLU'SN ON CLADDING INTEGRITX

Seal failure was attended by the formation of the hole in

cladding of area about 1 mm . Such a hole was formed most likely2

in locations of the simulator pellet column rupture or in the

pellet surface defects. Example of the cladding seal failure ispresented in Fig.5a.

The cladding failures is caused. by the plastic property

degradation of the cladding material due to saturation with

oxygen and hydrogen occur. In some cases under the subsequent

cooling the cladding failed for several hours due to thermal

stresses occured. Statistical treatment of the test results has

shown that the probability of this event is increasing with the

temperature elevation and gap value between the pellets of fuel

simulator ~ Pig 6 shows the diagram of preserving the tightness

of the claddings during the heating-up. The diagram shows the

boundaries of areas where the specimen tightness is maintained

(a) and failed (b). Between the boundaries a-b the area exists

where the probability of seal failure varies from 0 to 100 %.

Zt should be noted that external steam pressure over the range

6 to 6 lQ'a does not practically effect on the probability of the

cladding seal ailu e.The analysis of the data on the failed specimens due to

loss of the cladding plasticity shows that within the time

conditions given in Table 1 the clear temptrature boundary occures.

Besides, it should be noted that on the one side of this boundary

the failure practically takes place always, and on the another

s'de of it the failure does not occur. The cladding oxidation

and saturation by oxygen and, hydrogen were insignificant at

800 C and even 2 nours duration did. not result in mechanical

properties. The cladding failure practically always occures atcladding temperature 900 'C and more. Reducing the duration of

heating (up to min'~um range of values indicated for each

temperature in Table 1) did. not result in decreasirgthe number

of the cladding failure events. The similar pattern isobserved in the case when the specimens are cooled. in steam

medium after the heating. The behaviour of the test specimens

may be different from above-mentioned. one under other

conditions.

THE BEHAVIOUR GP THE CLADDING 'iYITH

illfATER COOLDOWN

A set of experiments with cooling by water subcooled

by 180«200 C was conducted to show the effect of subsequent

cooling on the behaviour of the heated claddings.

In this case the cooldown was substantially rigid. However

all specimens cooled witn water did not failed At the same

time the appearance of the cladding oxidation was similar to

that of the specimen failed after less rigid cooling in steam

medium. This fact can oe caused by microstructure changes of the

cladding metal. .""ig. 7 shows the plots of the cladding

temperature changes with water and steam cooldewn. It should

be noted that the cladding temperature increasing by 100-200 'C

occures during the ini.ial stage (10-30s) when water cooling.

Most likely this fact results from the motion of different

medium phases when moving the rewetting front from bottom to

top of the specimen. In this case the mean rate of water cooling

makes up about 13 '0/s, and steam cooling — 2,5 'C/s.

OXIDATZC:i OP CLADDING -THE

G "~E~ DESCRIPTION

Strong interact'cn of metal and steam occures under the

cladding temperat-re exceeding 800 ~CD The process of oxidation

has several stages. I.'".e examples of the oxidation stages are

presented in ."ig.8. :"izially black dense nonstoichiometric

oxide forms (™ig.8a)."his oxide forms at the beginning ofil

heat-up (during 5-10s). On being held at high temperature this

oxide converts to dioxide r 02 that has grey or white

colouring (See Pig.Bb) ~ This dioxide represents porous material

susceptible to cracking and flaking (see Pig.Bb). The depth

of the cladding oxidation depends on the temperature level

and test durdtuon at this temperature. When increasing these

parameters the thickness of the cladding oxidation increases.This following stage of oxidation is characterized by oxygen

penetration into cladding thickness. In this case ZrO> and a

layer ;,Zr (0) solid solution are formed. in time. The presence

of oxygen in this layer results in sharp degradation of the

mechanical properties, i.e. in this case the material embrittlement

takes place. Pig.Bc shows the typical rupture of the cladding

when oxygen diffused throughout the cladding thickness and

rupture due to thermal stresses took place. The limiting value

of eguivalent oxidized layer can be accepted as criterion of

cladding failure. This eguivalent layer includes the cladding

layer that converts into Zr02 as a result of zirconium reaction,and imaginable layer of ZrO> obtained through the evaluation

oxygen absorbed by the cladding metal in zirconium dioxide

formation. Metal saturated with oxygen and hydrogen has greyl'olourwithout specific brightness. The cladding oxidation depends

to a great extent on its surface state. The surface damages such

as the scratches, point hollows etc. promote the corrosion pro-

cess. Photograph p "esented in Pig.Bd confirms that.

HAT "RIAL WVESTIGATIOi~IS

The set of the ~terial investigation comprised the metal

micrographic and X-ray diffraction test and analysis of oxygen

and. hydrogen contents in the cladding material that was subjected

to rigorous tests as well. During the metal micrographic tests

of the claddings we revealed that metal structure differes

greatly in wall length and thickness and around the periphery.

This fact results from non-uniform temperature distribution

depending on the given way of heating up (i.e. direct current

transmission). The change of the cladding metal structure around

the periphery and in thickness of the wall is shown in Fig.9.Microharness of the cladding metal increased with increasing

temperature and test time. The degree of its increasing was

caused by the amount of absorbed, hydrogen and oxygen. 'Vhen chan-

ging tne values oi mic onardness a variety of pecularities may

be detected. Fi stly, the increasing of microhardness in rela-

tion to 'nitial cladding metal state occured only beginning

with test temperature 900'C and test duration above 30 min.

Secondly, the values of microhardness remained at initial level

even at temperatures 1200-1300oC and. at test durations of severs'.>

seconds.

The microhardness value was closaLyrelated to the extent

of hydrogen and oxygen adsorption that in turn depends on tem-

perature and test duration. Besides , microhardness distribution

over the cladding wall thickness depended on diffusion rate of

oxygen in metal. So nonuniform microhardness distribution over

the cladding section occured at temperatures 1200-1300~C and.

test duration 5-10 mi=..:-<ear the external and. internal surfaces

microhardness vas 1,5-2,0 times higner than in the middle sec-

tion (see Fig.10). It should be to refine the causes of rather

strong oxidation of the cladding on the inside. Increasing the

12

test duration up to 15-20 min at temperatures 1200-1300 C

resulted in smoothing the microhardness values over the wall

thickness. However these values were considerably above the

microhardness level of basic material,. i.e. most likely metal

was saturated by oxygen through tne whole thickness at rather

long-time durations.

Increasing Vickers hardness also began with temperature

900'C and test duration 30 min, and at short-time durations

(several seconds) its values remained at level close to initialone even at temperature 1200-1300 C.

When measuring hardness in the corners of indentation in

some cases the cracks was formed. It occured in the claddings

where hardness increased up to 280-300 kg s/mm and more due to2

simultaneous absorption of oxygen and hydrogen. This characteris-tic may serve as a criterion of the cladding material embrittle-

ment. Very strong absorption of oxygen resulted in increasing

the lattice parameters of 2 -zirconium. This process occured

during the test in steam medium at elevated temperatures. Increa-

sing the "a" parameter lattice of X -zirconium took place atconcentration of absorbed oxygen 0,6$ (mass). When increasing

the content of absorbed oxygen up to O,SW,9'mass) the para-

meterr

"a" tended. to more considerable increasing. The calculation

of parameter "c" for lattice using diffractogram also shows the

tendency to its growth with increasing the content of absorbed,

oxygen.

Based on the results of the X-ray diffraction analysis itcan be said with confidence that the volume of 4 -zirconium

13

lattice cell increases as the content of oxygen in the cladding

metal increases. The results of the measurement of the fuel

element cladding diameter confirm this conclusion. For instance,

measurement of the specimen after test at 10004C has shown that

the cladding diameter with Zr02 ranges from 14,2 to 14,4mmf and

the diameter after removtng friable dioxide - 13,7-13,8 mm. In

this case the thickness oxide film makes up 0,5-0,6 mm.

Beginning with the oxygen content 0, 11% (mass) the presence

of ordered solid oxygen solution like ZrO was detected in the

cladding metal. The occurrence of solution like ZrO in the

cladding metal complied with beginning of appreciable increasing

the "a" - parameter of " -zirconium lattice. Very strong satu«

ration of the cladding metal with hydrogen which achieved 1-2%

(mass) in some cases (at some sections of the cladding) took

place in time.

The X-ray photograph of test specimens have shown the

impurities corresponding to hydride of zirconium. As the concentra-

tion of hydrogen was increased, the peak value increased, and at

hydrogen concentration 0,77'mass) this value was equal to or

exceeded. the peak value of A -zirconium. It is believed that

at concentrations of hydrogen about 1'mass) and more the clad-

ding metal structure must correspond to compact hydride of zir-

conium (Zr Hx). The cladding metal structure varied according

to temperature, test dumtion and the cooling conditions after

the tests- As the test duration was increased, the grain growth

and amount of hydride phase increased, and its deposits become

more substantial. In this case the rate of the process of depo-

sition is increased with increasing temperature. Increased cool=

14

down (water cooling) promoted reducing to fragments of hydride

phase and it had an effect upon the plasticity. Evidently thisfact is the explanation that under water cooling the cladding

did not to failed while under steam cooling a part of cladding

was to be failed.The equivalent oxide film was calculated based on the data

on the cladding oxidation and oxygen absorption by the cladding

metal under different test conditions. The equivalent oxide film

was correlated with test temperature and the cladding integrityat the test conditions followed by different cooling rate. This

correlation is presented in Pig.11.Minimum value of the ecuivalent oxide film at which the

cladding failure did not occured when steam cooling over the

whole test temperature range, was at level 10fo of the cladding

thickness, and, under water cooling - at level 25 $.

CONCLUSION

Proceeding from the test results of the RBMX fuel element

cladding',behaviour in steam medium at temperatures 800-1300 C,

external" pressure 6-8 ZPa and, fuel column ruptures over the

range from 0 to 20 mm, the areas of preservation and loss oftightness have been established. Loss of tightness between these

regions occurs with probability from 0 to 100fo.

The cladding integrity depends not only the temperature

and test duration, but on cooling rate. It is caused by decreasing

the metal plasticity when oxidizing it in steam medium. Relativelyfast cooldown (water cooling) results in changing the metal struc-ture in particular hydride phase.

15

The data obtained according to the embrittlement critetion

are needed for further refinement ushag the necessary statistics.

Test facility

BH1

BH7 pQ1

8HZ

()Upy

Poz)( )(pa>

I IJ I

I BH3

aH5

PigBH1-BH7-valves; PQ1-PP- flovnneters;PP' test section

Test section

1g

2~7

Fig g1 - test specimen; 2 - vessel; 3,4 - flandres; 5,6-electricsl contscts; 7 - spacer; 8 - thevmocoupl;

9, 10 — vrires

Porms of cladding deformationon the fuel column rupture

Md~ bi

hli

I zs

p( I gq /

pig 3);

a) Length of rupture of 2 mm; b) 6 and 10 mm;

c) 20 mm

1 — cladding; 2- simulator of tuel;3 - hole in the pellet; < — pressed out cladding

Deformation of cladding on the fuel column rupture

a — length of rupture of 2 ran;

b — 6-10 ran; c — 20 ran

Two kinds'f failure

ho(e

'

Fig. 0

a — seal failureb — brittle failure

The borders of cladding failure during heat up in"length of fuel colunm break - temperature" coordinates

P s 6 - 8 MPa

lqxi J

."3'33

733I3

rig.6- 100 5 probabilit7 of failure

jg/- probability or failure - from 0 to 100 $

IM

OM

aOM

0

O0

4

Cl

Q O0 C)M

f

I

OOOO 00

oSSRP- r888OOO44444$

Q~8 4gjw

ADA

I I I

w GJ A

Stages of cladding oxidation

Fig. ~a — nonstoichiometric oxid

b — stoichiometric dioxid

c,d — embrittlement

e — development of oxidation on scratches

Structure of cladding metal

to outer

surface

a — 900 C, 60 min, strained section having maximum thinning

b — the same cond., section opposite to "a"

c — 900 C, 7 min ( x250 )

Diet.ebution of miorohedness

4 V

~w 6QQ

sj)

~ 2QQ~ ~

C

Q,IinnerSw rfacp

3,2 Q,3 0,4 0,5 Q,6 Q,7

RcCknd&:i 0 LhP Clgcr'cluny

Q,8 Q,9 I,Q <'""

CuterSdtj PC'8

Fig )o1200 ~C/9 min

1200 OC/10 min

1300 OC/5 mi- 1200 'C/10 min

1300 ~C/5 min

1200 ~C/10 min

"..'he border of cladding brittle failure in'temperature - oxidation depth" coordinates

8

y >Q

II

ll

IL

0'I

Tprnp pwca pu/"c, 'fill

pi~. )<

0 - brittle failure0 - no failure (0 - cooldown by water)

- steam cooldown- water cooldown

The 1st International Workshop on Severe Accidents in

Nuclear Power Plants

IMPROVEMENTS OF CONTROL AND SAFETY SYSTEM

FOR RBMK TYPE REACTORS

Z.O.Adamov

Yu.M.Che."kashov

V.V.Kondrat'ev

S.G.!Jill v

Dagomys, Sotchy, USSR

30 oct. —3 nov. 1989

I:.'P",'.OV~~,t"-?JTS GF CO?ASTROL Ai!D SA;ZZY "YDT~?l ~rOR

P~IX TYP= R".',CTORS

.4damov "'.0., Cherkashov Yu.ii.?'ondra t i ev V.V., Y kha rov S.G.,Ionaitis R.R.,Scientific Research and Design Insti-tute of'ower Engineering,:iascow, USB

,".fter the Chernobyl occident a number of experts have

analyzed the construction of the high-pcrver channel-type

reactors, type P IX; among other results of the analysis,

shortcomin.-.s of their control and sfety systems (CvQ) are

ind ica t ed.

..hs lo.-= sgo ss 1386, the report submitted by the

Soviet specialists in Vienna /1/ contained data on this

problem. Since the accident st the Chernobyl ?IPP is of the

international nature because of its scale, while the

P~IX units sre still in service, there is natural concern

of how the struct" ral shortcomings of the reactor are eli-i-nsted in t',".e U".'Z s;d ':hst is the stat of .vo k on CSS in

ps r t icula r.The present re.--ort outlines measures which have beer.

already take.". to improve CSQ for incr.ssing safety of the

P-.:iZ reactors as v'ell as those to be accomplished in the

near future.

l ~ URPOSE As i3 CO'LG Okl" PZS OL CONTROL AND

The control and safety system (GSS) of the P>MK typ~

reactors is intended to control the reactor and ensure ita

safe operation under all operating conditions of the power

unit including emergency ones. CSS ensures:

- monitoring and registration of the reactor neutron

power level and period of its rise under any operating con-

ditions. of the reactor from subcritial to nominal power;

start-up of a reactor from shut-down to a preset

power level;

automatic maintenance of reactor power at a given

level and variation of the preset power level;- manual (from the operator's console) regulation of

energy release distribution through the core volume and con-

trol of reactivity to compensate for burning uy, poisoni~,

etc.;— automat ic stabili-ation of radial-azimuth distribu-

tion of energy release in the r actor;

preventive protection - rapid controlled reduction of

esctor power to safe levels;

emergency protect ion at emergency variation of

parameters of the reactor or power unit aa a whole;

— automatic serviceability check of CSS circuits and

.:.nits.The control and safety system comprises:

— neutron fluz detectors with fixtures (hangers) for

installing them in i;he reactor;

reactivity control members (absorbing rods) with

CSS drive mechanisms displacing the control members in the

reactor channels~

- CSS equipment converting signals from neutron flux

detectors, console controls, process automatic system of

the power unit and generating on processirg these signals

the commands for moving CSS working members, changing the

preset power level, varying reactor operating conditions,

turning on or off signalling, snd supplying analog signals

to display snd record reactor parameters.

2. CRMSING SPEED OP ACTION 01 EMERGENCY

PROTECTION MECHANISMS

One of the gravest shortcomings of CSS of the PSIIK

type reactors is excessively slow (18 - 19 s) insertinn ofthe absorbing rods into the core in emergency.

This time is regarded too long /1/

Emergency protection of the PEMK reactor is known to

be effected by automatic insertion of sll 211 absorbing rods

into the co" from sny position heightwise of the. letter.Twenty four C:~ rods uniformly spaced over the reac-

tor sre selected from above-mentioned 211 rods in the emer-

gency protection mode by a special selection circuit in the

logic. section of CSS.

Durin"..-.o:msl operation of the reactor the emergency

protection ro-'s are in the uppermost position and theircoc ked ste t e i s eu t orna t ice lly checked snd indicated.

Speed of ection of the CSS drive mechanisms is incres-

sed in stages..';t the first stage, owing to modification

of the mechanism drive and increase of efficiency of the

absorbing rod, the time of introducing negative reactivi ty

by each CSS rod ia reduced for about 5 a.

Drive modification consists in installation of an addi-

tional intermediate switch which cuts in the motor of the

drive in the electrodynamic braking mode after the rod ia

lowered to 3 m.

":ormerly the motor operated in this mode from the very

beginning of rod lowering. Absorbing rod efficiency ia inc-

ressed by increasing the number of absorbing elements.

Such mechanisms were tested at power unit 3 of the

Smolensk IMP in summer, 1987, and at power unit 2 of the

Leningrad ~'O'P in autumn, 1987. The results obtained show that

therate of introduction of negative reactivity by the emer-

gency orotection system has increased about twofold and isebou t 2/s.

Today, sll po.ver units with the PiMK reactors are equip-

ped;.:ith such CSS dri,ve mechanisms.

i;atu ally, such s drive mechanism of emergency protec-

tio.. is onl„" s temporary measure intended to increase speed

of action of emerge"..cy protection as quickly as possible by

usinp to the advantage the exi.sting equipment. Radical solu-

tion to the problem is to ensure emergency protection opersti

on of. -"s and less, which is the task of the second and

third stages respectively.

ne.: drive mechanism of the fast emergency protection

( ""2) cls bors t ed st the second stage ia to ensure inset" t ion

of the sbsorbi-.~ rod into the core within up to 2.5 s.

it also uses the old some.~hat modified CGB drivemec'"...'ism.

he time of rod insertion i.-.to the "ore i "'d.;"':

removing water from under the rod and providing film cool-

ing of. the CSS channel.

The drive mechanism is complemented with a device

which is connected to the motor field winding aa the rod

reaches the depth of 6 m and ensures efficient braking of

the rod by the motor by the end of its travel of 6.75 m.

"our drive mechanisms of the fast emergency protectio.".

were tested at unit 1 of the Ignalina NPP in autumn, 1987~

:t "as fo'.::;r'. out that owing to removal of:eater from under

the rod, e="=iciency of the fast emergency protection rod

is 1.2 time as high as that of the standard CSS rod.

Total efficiency of the complete 24-rod set of fast

emergency protection system is about 2.4P, maximum speed

effi" ienc:. is ~ — i.3g/s. The time of fast emergency pro-

tect ion "od e..tension through full travel is c.l Q.05 a,

th"ough a tre rel of 6 m sufficient for full shut-down of

the reactor is l.'3 s.Calculations performed in our institute show that in

t:.e considered hvpothetical emergency situations, includi"..

those .'ith violations of schedule and simultaneous failures

0 eQuipment e..d safety systems of the reactor, operation

o. asi r.-.erg ..c.; protection with the above parameters

alon„..sir'.e '.ith other sa ety improving measures, ensures

shut.'o".in o= '-~e reactor and excludes an accident similar

to tha t at "..it -'f the Chernobyl NPP.

o. t'.> . complete set o f fast emergency protect io'-

:rive .-.,o.c.-.:-risms wire completed at unit 4 of the Leninpra.'

i'HAPP ..vere succ essfully accomplished on August 12, 1988 aft er

a year of the pilot-commerc ial opera t ion.

~y now the fa st emergency protection system is imple-

ment ed in 10 of 15 PSMK reactor power unit a.The third stage of improvemen t of the PbMK reactor

contxol and safety system is to develop an additional amer

gency protection dr ive mec hanism basing on a different prin-

c ip 1e. A conceptual design of such a mechani sm has been al-

ready;;erked out. It operates on a principle of introducing

neutron a b sorb inC gas ( He or BF3) into the core. The3

ar.".poule;. it h r.as is accommodated inside the mechanism and

is outside the core behind t he shielding plug. The drive

"..leeha n i sm o: t he a d d it io na 1 emergency p "ot ec t ion (AEP)

is to be i~sta lied in the standard 0 SS channel cooled with

water or in the auxiliary channel.

The A"--E'rive mechanisms are direct-action, i.e. they

opera t e both by the signals from the C SS equipment and from

d i rect-ac t ion neutron flux t ra nsduc era built in the drive

mes hani sm.

h i s work is at t he stage of experi'~ent s today. An ex-

perimenta 1 me" han ism is being bench tested The AHP gas

drive merhenisr operat ion time is expected to be up to O.l-0.5 s.

The i ou r t h stage envi sages d evelopment of new const ruc-

t 'ns o=" a 11 ot h r ( non- emergency) drive mechanisms of C S"-,

i.-.clu 'ng rod g ~ f -e v construction. This mechani,sm, while

eraser ring all ".:.o:it i vs features of it s predecessor is to

Pl imi-e t e or co.-.sidere bly dec res se such shortcoming as para-

sitic absorption of neutrons with water cooling the absorb-

ing rod o

One of the types of such drive mechanisms, the so-

called rod with move-on absorber is being tested on the

reactors of units I at the Ignalina and Leningrad NPP>s.

Such a drive mechanism rules out parasite columns of water

in CSS channels.

Another type is cluster drive mechanism. This drive

mechanism minimi".es dewstering effect. Its regulation mem-

bers - small absorbing rods are inserted along special

guides in e stationary expelling member. Tests are being

carried o t on a critical assembly.

I.:PRCVr.i".ZiT OP CSS ELECTRONIC EQUIPNZHT

The ..~lish experts'eport /2/ supposes that the

P iZ reactor CSS cannot ensure shutdown of the reactor at

non-stable operation or failure of safety systems.

In our opinion the list of emergency signals of the

P..'!Z reactor aloes not differ substantially from foreign

ones.: '. ''.;cl. 'es signals o decreasing power rise perio.',

excessive ..'o"er, rise of temperature and pressure of coo-

lart, drop oi separator level, reduction of main circulsti."„

pur,p speci: s..d maiy others. nevertheless, the Chernobyl

a"cic!ent has .-.+de us complement this list with the folio".:iv:

si<~als:— r~a~.:ct o.. o;" operative reactivity margin;

"d'," t 'o. o=" rani;". circulating pump cavitation mar i-..

i-cro.:~s" o ".ressure in the reactor apace (reactor

increase of pressure in the primary circuit rooms

(separators, steam/water lines, lower -,ater 1iro-);reduction of water flow through each dispensing

group header

Protection of the reactor by these signals is alsoimplemented in stages. At the first stage individual device.are developed to obtain these signals. Such devices can

be manufactured and tested rather soon.

In particular a breadboard modelof the device has been

made to r e: crate emer--.ency signal as the operative react ivi-ty margi.-.:rops down.

This device computes the above-mentioned parameter in

view of energy release field shape data inserted by the

operator with the use of a switch. The data source is infor-at ion complex, comput ing the neutron f ield shape by the

signals fro';: the intra-zone sensors.

.'naturally, presence of the switch allows an error or

intentional actions of the operator to maintain reactorpo'.ver. I.o'.:ever, emergency set tings in this device are so

chosen that emergency protection will operate at a reacti-vity margin sufficiently exceeding that at the Chernobyl

acc ident.

The second stage is development of new moderni ed

O'S eouipme..t,;.rherein all parameters shaping the above-

li ted emer'.e.-,cy signals will be computed automatical'y;sit:out interference o.. the part of the operator. All the ignite

with the P '.~',-1OGO reactors will be equipped with moderni"ec

equi pr.".ant.

During this stage s modernir ed sy st em of physical mo-

nitoring of core energy release distribution is sl o elabo-

rated. This system will use no-lag intra-zone emission de-

tectors which are now employed with the P~>K-1500 reactors.!.'ew equipment of this system'omplementing, the existing

f

system of sutomstic stabilizati.on of radial-as4ruth energy

release distribution will ensure emergency protection opera-

tion at impermissible skewness of core energy release distribution.

i;PL~<'.ENTATION O'ASIC PRINCIPLES OF SAFETY IN CSS

Implementation of any newly developed equipment intende

to increase reactor sfety is accomplished in the US& only

i='ll tho. requirements of the State Atomic Power Supervisio:

Board sre .-..et.

The ne": CSS electronic equipment is designed for tripleor qus 3ruple redundancy, with location of individual sets i.-

differe it rooms remoted from one another.: his er.sures operability of CSS in case of fire, drops

o floors. ~ig.".al ~ rom all sets of the equipment are fed to

each of 211 individual driv r.".echanism control units,

.vhere they are "zjorized (by the 2 of 3 or 2 of 4 pattern).

:mergenc„, pro tect ion may be activated by the operator

both - "or.; the <"e-.eral control panel and from local stations

havir„"; relisbl communication i<rith the general control panel.

.'.ll the .-.er.".bors acting upon reactivity are divided

into 6 in<'.e",e-<i:nt "-roups, 30-36 rods in each, uniformly

~.:istr buted over '.he reactor, having autonomous power suppl„.

snd each br r.;; car eble individually to bring the reactor to

10-

subcr it ical stat e.blocking of individual channels to pa -sage o .:-..er

gency protection signals because of failure, adjustment or

shut-down for repair i s equivalent t o g enera t ion of emergen-

cy protection signals in these channels, which excludes

unaut hor iz ed disabling of the sa fety systems.

All the newly developed equipment, including CSS drive

mechanisms are subject to earthquake reliability tests. Ifsuch tests are impossible or impracticable, seismic resis-tance of the equipment is proved by calculations.

3 .s i= .".s of ar . newly developed safety system equipment

are care""lly enalyr ed for consequences of failures. The

resul ts of such analysis should prove either maintenance oft he reactor sa fety level in ca se o f equipment fa ilure orcont rolls b il ity and well-t imed detection and elimination of"G.J -.Grous fa il"res, i.e. impossibility of appearance of co.-.—

o=."=led da. gerou s a ilures decreasing plant safety level."'hen: =.sig.-.i."~ sa fety syst em electronic equipment,

...-.=-.,~re.". '- ."~ ".s-.:e".. to rest riot access o such equipment

811i..:;.,e"'.. r i"a1 locks. Ho;;ever, it is regarded un

reel tc '.:1 .. r li;;>'.-.ate the possibility to disable the

ty -'t ""s 5'' 1e operator ix he is a skilled saboteur dc.—

term- ~ 'ed to perform an act of sabotage or maintain reactor>o".er at "-:..". - "'e.

"he o.".'.: co",~ termea sure in this case is to impart

.r -.""'.c;- . -.„';ee tures to the reactor. Such work as G."..p-' ac'. to ... '. reactors ' also being carried out in th i .

=3;:'ry. ".;~.-. "o.:.;.'.'measures i.-.elude installation o a,.

tional neutron absorbers in the core, employment of fuel

enriched to 2.4,~ for 'nd =o™eothers. '.".o ever, ~.et;ai"

ed description of this work state is beyond the framevurk

of this report.

i? "7-?"!'lC- S

1- Peport oi the US~K Delegation at ZAEA Expert Lleeting in

Aug'st, 1966.

2. i".eport o. the United Kingdom Atomic Energy Authority

"'ccide".t at the Chernobyl 1P. and Tts Consequences"

:,.arc h, 1987.

The 1st International Workshop on Severe Accidents in

Nuclear Power Plants

OPERATION OF NPP WITH RBMK TYPE REACTORS AND MEASURES

TO IMPROVE RELIABILITY AND SAFETY

E.0.Adamov

V.P.Vasii vsky

A.A.Petrov

Y>;.M.Cherkashov

Dagomys, Sotchy, USSR

30 oct. —3 nov. 1989

. 1

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,.AIC.i .'GP ~VAT:-D ACCID=".T AT Cii='2'103 L i,PP t;.'iIT .;o ~ 4

The source of accident at povrer unit No- 4 of the Cher-

nobyl NPP was an unlikely combination of several violations

of reactor operation regulations made by the attending per-

sonnel. I n the course of development of accident processes

in the reactor plant brought into irregular state, some

structural features of the reactor manif est ed themsel res to

aggrava te the accident. Ouch structural features include:

positive steam coefficient of reactivi ty and core

d ev:at erin" =:ffeet;insu f ic ient speed of action oi emergency protection

under the conditions of impermissible reduction of react ivi-

ty ma r=in belo;v 15 rods of manual control (hIC);

i.'.suf ic ient n mb er of automatic means capable of

bringing the reactor plant to a safe state at viola t ions of

operation regulat ions;

ab senc e o. t echni cal prot ection for devices adapt ed

t 0 p~>age/d isen -age some reac to r prot ect ion systems.,

'.:::?u -PRIORITY:J 'AZ.'H."-,3 TO PB.";CLL'3E PO"-EDIBILITY

G."'RI!JGI:JG P -JZ-TYPZ 2 "ACTQi'. Q .',QiJ-<2-'Ai'j,"j".,",T's

soon as the group of experts who was finding out

the causes of the Chernobyl accident in situ worded their

first versions, specialists elaborated and nuclear power

authorities approved immediate implementation of first-prioritv organi~ational measures at all NPP with P~hX-type

reactors to preclude repe t ition of the situation occured

at po;ver unit No.4 of the Chernobyl NPP by the beginning of

the accide..t. As soon as on ilay B, 1986, directors of all".PP ..ith P.-;Z reactors received a directive comprising the

ollowing instructions."

b„.i no means allow o~eration of the reactor at any

po;ver level with reactivity margin below 15 MC rods; ifreactivity aarwin drops down to l5 rods, immediately shut-

down the reac tor;

cancel the right of NPP chief engineer to allow tem-

o:..1or- -ration or the reactor with reactivity margin be-

lo;" . 6:..C rods;

„-t.ictl; obey to the regulations as regards corres-

po.=dance o- main circulating pump capacity to feed water

flo.v rate; preclude main circulating pump delivery above

",QGO m, h et feed yater flow rate below 500 t/h in each

c ircula t ion loo.";

prob i o i t en@a ~ament of the fourth main c irculat in@

',.ump in each =uupi."g room if thermal power of the reactor

1s belo l,CU ...,i:

- permit start-up of the reactor in at least t:ro days

if the shut-down was at 50,u and more o.-'ominal power and

after a time period required for its depoisoning if theshut»down occured from a lower power level;

if thermal power decreases to 700 55V, unload thereactor subsequently;vithout delays or steps or shut itdown;vith emergency protection;

- before obtaining additional reasoning, prohibit using

:HAPP in any duties, except base-load one;— prohibit temporarily all experimental work at the

opera t i.'z'. ",.o; er '. nit s.( e l 1aj ea s il„, conclude today f rom the above instruc-

tions, that considered at that time as the most probable

cause of the accident;vas the following sequence of events:— bringin„= the reactor into a non-steady state by

decreasing operation reactivity margin to impermissible

limits (to 2:.C rods according to estimates);

ezha..;sti-.". the reserve to cavitation in main circula-

ting pump lo-'I .-.ection at c"t-in of all 8 pumps with imper-

missibly hi~~ u.-i' capacity (up to 8000 m /h) at reactor

thermal oo.jer o only 200 I'".iI and corresponding low flow

rate of fe.d .'ster (200-250 tlh);- connection of 4 main circulating pump power to sec-

tions of r '..'..- 'o'vn turbogenerator, followed by their

emergency '-'; '..ich caused additional increase of capa-

city of still o e "sting 4 main circulating pumps, which

with prese'"=. '..o: L.: separators and decrease of feed water

flo.v rate ".~-.;L;.: to their cavitation or surge;

drop in supply of coolant heated to boiling point

to the inlet of the reactor operating st pover level draeti-

cally changed volumetric hesatesm content in the core,caused positive Surge of reactivity and finally rsn suey the

reactor on instantaneous neutrons

3—

'." .CPi CAL V.; AERIES TO IIiCR.""'iS!'' B.'Z~~ZY AflD RZLI'ILITY0" iP ':iI H F~.'Z R ~'<C u.'..i .; K; 'IR~i':T.YG.''

After analyr.ing the causes of origin and development ofthe accident at power unit llo.4 of the ChN?P, main trendswere determined in development and implementation of techni-cal measures to increase reliability and safety of NPP with

P„.i( reactors, both operating and being under construction:reduction of steam coefficient of reactivity, increase ofefficiency of reactor protection, increase of reactivitymargin u-..der stationary conditions at any power level, estab-lishment o continuous monitoring of operation reactivitymargin in the course of reactor operation, prevention ofunauthori" ed cut-out of reactor emergency protection systems

at its operation et power level /2/.Calcula t ions of ra rious emergency processes revealed

that with a simultaneous increase of speed of action of emer-

gency protection, steam coefficient of reactivity should be

decreased in sll power units from 4 ~ 5 to 2 4 f. This would~efpreclude r~ newsy of the reactor at any violations of opera-

..tic.".s b.. automatic action of emergency protection.the irst stage, steam coefficient of reactivity was dec-

ree sed by

absorber=

i-.:telling in each reactor 80 rods of additional

i:..st-.s6 of fuel assemblies and by increasing mini-

~ m co'missi".ie rr.ectivity margin from 15 to 30 i'EC rods.rw ~ rp pyq ca : rebec~ ion efficiency was initially increased by

pre-insert io". o= ell control an ', safety system (CSS) rods

io" 0 ~ i.-.:o t::e core. his increased differential efficien

10—

c" o" rods inserted .rom the upper position on the one hand,

3..d reduced ( for c s) the t ime of rod insert ion into the

core on the other hand. At the same time, the number ofshort absorbing rods introduced into the core from beneath

Ylas increased f rom 2 4 to 32 for t he PSMK-1000 and to 40

for the r~»X-1500 reactors and the circuit was implemented

for automatic insertion of these rods into the core at oper<

tion of emergency protection. Owing to modernizing standard

servo drives of CSS rods at all units successively, the

time of full insertion of the rods into the core by the

emergenc.. pro'.ection signals decreased to 10-12 s.i,'n(;er sta t iona ry condit iona at any power level, reac-

tivi ty margin compensated by CSS rods was increased to 43-

;.8:,.0 rods at the P~hZ-1000 and to 53-58 rods at the PoMK-

1500 reactors which increased reactor stability and effi-ciency of emergency protection at the initial period of itsoperation o;;ing to immersion of a greater number of CSS

rods to the middl (in height) part of the core. To allow

on-line:.o.".itoring of current reactivity margin of the opera

ting reactor, a c„;clic algorithm with a period of 5 minutes

::as worked out-, for calculating reactivity margin and disp-

laying it on s digital indicator; the algorithm ia implemen-

ted on standard computers.

Zh~ power'it control panel is provided with an addi-

t ion.l a" tomcti" light panel fixing disengagement of reactor

e;emergency ". ro'.ection for each parameter. Interference of

the atiendinc personnel in operation of the light panel,

i.e. cutti .— out oi the signal is fully prevented. Access to

-.ro.ection on/off panels is strictly limited. 7.emporal dis-

aoling of some emergency protection circuits, needed ".o

increase or decrease reactor poorer is authori-ed by the

chief engineer in a written form and is effected by two

special keys by the shift personnel leaders.

12-

-.'G- '~R.",I .~"SLR 5 -OR I.'RM~L'JG R "LIABILITY A:iD

SA '-"TY 0 ?:PP 't7ITH ziZ R:-ACTORS

By the end of 1986, the so-called "5»~~ry I'Leasuresfa'ncreasingReliability and Safety of NPP with PohK Reactors»

were worked out. They vrere the second stage of modernizing

.='s with reactors of this type based on more detailed

study of causes and consequences of t he C hernobyl acc ident,

more deta iled study of transient dynamics and experimental

mpport of efficiency of organization and technical measures

i"..ampler. ented durirg the first stage. These measures require

add it 'ne 1 design and research work, bench tests, develop-

ment of instrumentation, automatics, etc. Considering the

".i~e needed '.o produce additional and modernize available

equipment and actual terms of implementation at NPP's, these

measures may be referred to as long-term.

he -..ain measures to increase safety include:

reduction of steam coefficient of reactivity to

Xf <.lory„. by ohengi"..r, over the reeotore to loading with

'.;el asser..blies of 2.4o enrichment of U instead of 2,d;235

de ro.lopme»t of fast-act ion emergency protection sys-

temp and implementation of them at NPP's;

- development of new actuating mechanisms of CSS to

drest icelly di~inish the effect of CSS channel dewatering;

develo;.ze.-.t of new and improvement of existing sys-

terr,s oi diaz:.opt ics o main equipment metal, pipelines,

c ha nr. e l.",;

develo ~ent of new reactor monitoring, control and

13-

.":a=sty systems to replace those in use at power units;

development of necessary instrumentation end automa-

tics for implementation of reactor protection systems res-ponsive to: reduction of.operation reactivity margin, boi-

'. l,

ling point margin at main circulating pump inlet, rate of

pressure drop in circulation circuit, reduction of coolant

flo v rate through each dispensing group header;

development of new construction of reactor graphite

stacking with 20-io reduction of graphite mass for the lastpower units with P~LK reactors under construction;

- d velopment and replacement of IG'Psystem equipment

end instruments vital for safety with those of increased

seismic r e"istance;

development and implementation at all NPP's under

construction of systems ensuring release of steam from the

reactor e t sir.".ul taneous breakage of several fuel channels

(7n) .deve1op.-..ent of new control computer systems, diagnos—

t ic snd operator, support systems to replace those used at

tippy jimprovement of ilPP fire safety;

elaboration of a concept for reconstructing the ope-

rating power units to bring them to the top level of safety

sccordi.=-. to present requirements and extend their service

1 ife

-14-

6. "..""- NT .- AT= 0" "lOBE FOR IiJCR".;'INC ~'iF""TY OF <JPP

The state of affairs for development and implementatior

of main systems increasing safety of NPP's with the P5MK

reactors by the middle of 1989 may be briefly characterised

as follows.

.""or further decrease of steam coefficient of reactivityafter the middle of 1987 fuel with enrichment of 2.4' 2 5

instead of 2,o is loaded into the PSIIX reactors. By now from

400 to 900 fuel assemblies of increased enrichment are loa-

ded in the reactors of operating NPP's. Steam coefficientoi reactivity;vas gradually decreased and is from 0.7 to

1.2R . today.I ez

Further increase of efficiency and speed of action of

CSS, diminishing of effects of CSS channel devatering were

carried out by providing all the reactors with actuating

mechanism of a new design with absorbing part elongated for

0.6 m and an elongated telescopic control rod. The time of

rod insert io.. into the core decreased from 17-18 to 10-12 s

QI t er impl erne.".ta t ion of delayed elec t rodynamic braking on

rod servo "rives. "/hat is of primary importance, st first3-4 m of t revel the rod speed is increased twofold and is0.8 m/s. As a result of the abave-listed measures for moder-

nizing CSS ect..sting mechanisms, speed efficiency of 0.9

f/s is at ~si.-.e~ at first four sectond of movement (abouttef8 times th. i.-i'.'.'01 one) which excludes dangerous surge of

power at ra."'i":eivatering of the reactor in case of a

design be is accident (DBA). Replacement of the actuating

mechanism." ~i 'o ex" ludes possibility of positive reactivity

15-

insertion at movement of any number of CSS rods into the

core..' the same time additional automatic protection visa

implemented at NPP's to exclude positive reactivity inser-

tion at the operating reactor in case of emergency dewater-

ing of CSS channels at breakage of CSS cooling circuit pipe-

lines. The above-listed measures were implemented at allpower units and completed by he middle of 1988.

'he "..est stage in increasing CSS efficiency was the

development of a ast-action emergency protection (PAEP) ~

the en o'387, experimental specimens of FAEP actuati~mechanisms .;ere successfully tested at unit No.l of Ignalina

:st, 1)dQ an experimental ZAEP system was tested

st power u.;it:;o.4 of the leningrad NPP. The following re-

suits:;ere obtaine.: ''.'u"-P 24 rods insert over 2 3 f negativEj efreact ivit„".into t'r.e reactor within less than 2 ~ 5 s In De-

camber, 138'=', the "-.'''P system within the full scope was im-

plemented at power unit Ho.l of the lgnalina NPP and todayf29 ~o-:er ".~its =re equipped with ":AEP. On other power units

,iith:-'::..'.actor.-: -'..;P systems will be installed this year.

"cwork i.; si~~ltaneously. !ising carried out to develop

end impler ent:.~.i .".lust er-type actuatirg mechanisms both

for " 'I er d .or: ctuating mechanisms of other applications

to replace sta.-.nerd rod-type devices. Implementation of

those mech ..i." '".'ll considerably decrease positive reacti-

vity at r.-.. -.:."."....'.vateringof CSS channels and cancel

special gas-1 i-..;id circuit of ™AEPchannels. Cluster mecha-

nisms»naergo -.,~.;.ch tests at present.

.".eec'.or '.:=ts 'vere completed and series production

16-

b .=an of new type of CSS rod actuating mechanisms with an

absorber lcw er section moving onto the dis. 3a cer. ':e.-emechanisms considerably diminish the effect of CSS channel

dewatering but they are not intended for use in the ZAEP

sy at em.

Systems of diagnostics of equipment, pipelines and in-

tra-reactor devices are being developed to solve the fol-lowing parti"ular problems:

detection cf weld and metal defects in large-diame-

ter pinelines, headers and steam separators by acoustic

emission '..c hni qu e;

service checks of metal continuity in large-diameter

pipelines and headers with remote ultrasonic systems;

detection at early stage of coolant leaks (if through

flaws appear in equipment and pipelinea) with acoustic,

fiber-optics and TV systems;

insp ec t ion o f cont inu i ty of metal, geometrical d i-

mensions of fuel channels and graphite stacking to take

into account their radiation distortion in the course of

servi"e.f.'.ost of the above-listed diagnostic systems are to be

developed i. 1390-1991. At present the service check of me-

tall

oi,PP equi pment and p ipelines is int ensif ied by incr pa-

.-,in;. the .;co".e end frequency of inspections. Today NPP's

equi".~ei "i:h more advanced universal flaw detectors

both Soviet- end 'oreign-made. An acoustic-emission system

of inspect i .;- —..etal is bein"- teated at power unit !Jo.l of

the Cher"..o';.. 1 ..':-'..'.n instrument .vss successfully tested

17—

.or i..to~rated measurement of -.eometrical dimensions of

..el channels and ~rapnite stacki.-g blocks. t is br:L.g

put in lot production to equip all '.HAPP's. A special ultra-

sonic fla.v detector vIas developed to check metal in fuel

channels and CSS channels. This instrument alloivs in-

service i..spection of channels as viell as check of channels

to prove through defects detected by the standard chan-

nel ini=:'.ri t„.; inspection system (CIIS) operating at povfer.

A 9 regs rd o ther sy stems, breadboard models are sub j ect to

laborator;.'ests., inca:-.1 ec t.omechanical, electronic and logic sections

of t;he '"i ":-::--- ..rat ion '~IX reactor CSS are morally and

,,hysically ob:o] ete, a rev monitoring, control and safety

svstem i.-, "einL ~!c .eloped for them to be a part of the reac-

tor plant =-" tome t ic process control system (APCS) ~ These

sy 9 t em::. a .-e ".o b» i..stalled vihen reconstruct ing NPP'. A t

"resent technical documents for the system are being elabo-

rated.:, „;; . ir.;:t -"..acimen is to be installed at popover unit

o.] o; r.-.~,'.-, -:-r'~'PP in 1331.

he .-~:'.oi; ~~ i;;iproving neutron-physical characteris-

tic.-: o: the co."e ~'opted at operating P~i~K reactors is far

from o~ti.;;.;::..;i:;c»it resulted in considerable decrease of

-".el bur. — - res tio.. and, accordingly, of NPP economic

.""=.p table vay, .vhich is implemented only

"onstr ct ion is to decrease mass of

,':ra ..'h i ".e

re phil

ri~tac:..

.".ore for about <0,~- A reactor with such a

..'.~ 1 have fairly good physical characte-

.;:.loading uel v(ith enrichment of 2.4,~

-.< ~5i::, steam coefficient of reactovity will Os+i/ =-2.5),C

fast poser coefficient of resctivitycLg = -5.4J

.10 /i57;

reactor dewstertng effect - 2.0 p (at nominal power); fuel

burn-up fraction - 26.0 MW.day/kg uranium. Structurally,

the amount of graphite was decreased by cutting corners ofgraphite blocks. To study neutron-physical characteristicsof the core with new stacking, a stand was erected at the

Institute of Atomic Energy where a series of neutron-physi-

cal experiments was carried out. Experimental results corre-

late fairly well with calculations. At present gra phi t e

stacking o; new construction is being made for power unit

'..o.5 of the .".ursk;.PP.

'-tequirements for seismic resistance of NPP equipment

and building structures have been recently made far more

stringent. Seismic resistance of NPP sites is checked and

in some cases increased. ItPP designers are provided with

lists of equipment and systems important for safety, whose

-ei™ic "..sistance sho" ld be ensured to withstand maximum

'. stgn earth'.,uakes at each llP- site. Equipment and building

.rtructures inclu& in the lists were either analyzed for

increased effects or tested on vibration beds. Most of the

installed equipment, pipelines and systems of NPP meet the

more stringent seismic resistance requirements. Equipment

a.-.d i.=-tr~ments .vhich do not comply with these requirements

are re~l3ced "it'a sei "mic-resistant ones, pipelines are

fitted ..ith additional supports and hangers, building

."t"uctures sr rei..forced.

corns t ion .vi th implementation of more stringent

-equirements for IMP safety and reliability, implementation

of nev systems and protection facilities, v!idenin~ o='oe

list of accident source events, change of neutron-physical

characteristics„ it .vas necessary to carry out additional

analysis of safety of H'PP's with PHD reactors. Cn the basi

of this analysis in 1987-1988 additional reports were issued

on technical substantiation of safety of both reactor planta

and HPP' with ™>!Zreactors. Power unit operating instruc-

tions and other operation documents were considerably re-

vised. 0 perating regulations and emergency instructions

specif.; actions of per sonnel in various non-standard situa-

tions i. more detail, strictly define the rules for enabling

an'isabling each reactor protection system, prohibit dis-

abling the .;hole set of emergency protection systems on the

operatir~ reactor, considerably restrict rights of NPP admi-

nistration for allowing operation of the reactor with devia-

tion from process parameters within a limited period of time

In vie of ".resent:IPP safety requirements, the listo '".on-de. i;:..:.«c™ir.'ents under consideration is consideraol.

.'!i.iened and th ir possible consequences are studied. Analysi,

o such ac" iuents according to this list is nearing its

completion. 'ore st ri!.,;:.nt requirements for systems of loca-

li~ip~ pccide. ts et .":PP"s with P>I~X reactors are set forth

in the r e;; .."-.ion o the normative document.

"esi'..-.;.=v;lopments carried out for the last power

units with . ". reactors which are under construction in G."-

poretc po.;'ioi.l '. of considerable diminishing of consequen-

. -i, .-..~ccident as simultaneous lateral bree~-

3;.e o: "everal fuel channels. Probability, for instance,

of simultaneous breakage of two fuel channels is estimated

at 3-6.10 1/reactor year. The scheme of emergency x'eleasa-10

of steam-gas mixture from the reactor space implemented at

power unit No ~ 3 of the Smolensk NPP is intended to releaseall steam generated at simultaneous lateral breakage of 5

fuel channels when using the main system and 10 channels

at joint operation of the main and stand-by steam-release

systems. The opexating power units are equipped with systems

which allow release of steam from the reactor at simultane-

ous breakage of i p to 3 channels. Designs were worked out

to moderni..e steam discharges for all operating power .unit..

i -plementation of these measures will allow bring steam dis-

charge system capacity to the parametexs of the system used

at power unit I~o.3 of the Smolensk NPP.

A considerable volume of expeximental and calculation

.vork is being caxried out to decrease fuel temperature in

fuel elements of the P~hK reactors. Being studied are fuel.

assemblies with xod-type fuel elements of different diameter"

and fuel assemblies with so-called "tight" bundles of fuel

elements. No acceptable alternative has been yet found to

cnnstruct ion of fuel assemblies employed at present. Along-

side:.ith the above, a programme is being carx'ied out for

additional,".~i: of behaviour of fuel element shells under

er..er~ency cool iver., analysis of fuel channel pipe stx'ength

in emergency caused by dewatering of the channels followed by

heatir~ o 'he ?uel assemblies. Another problem sub)ec to

add. t io~el e:!per imental research is study of radioactive

"..rocuc.'-...i~ration at accicents .rith leakage of fuel element:

caused by breakage or circulation circuit pipes. 'ithin the

framework of this programme, diffusion of radioactive pro-

ducts through various media and building structures is also

being studied. The same problems are studied by refining

mathematical models describing emergency processes. Dynamic

transient sn". emergency conditions are calculated by three-

'izensio.-. l pro rammes incorporating physics and hydraulics

o: the i..:.:;reactors in view of feedbacks. The programmes

eleborat d or -,oderniwed together with specialists of

th ..:" .::.'ic .esearch and Design Institute of Po ver "-ngin-

=erj..-,'..."titute o='tomic ='nergy and All-Union ~cientifi"

rc'.;:;.-:titute of .,'uclear Power Plants.

C.-, of the;.. i;-... asures which in fact determine 'IPP

sa fat„:, i;, i-.:".rovement of fire safety. To solve this problem

~rest ru".'.'r of new non-combust ible building, roofing end

thermal i..'".s"la t ion ma teria ls are being developed. Several

. o::—,".o~bi "t i.ble and corrosion-resistant cables

. ui i.. ".'o .ct ion..~ t present all cables are coated

'h ".eates '..-..".ibition spread of fire as a provisional

:.::-=sure u.:~ to replacement o f cables with f ire-resistant

o:-.es d,r i.-.:.."::.co:.st,.uctionof:"HAPP'...eactor and turbine

'".'.ildi -= -oo.'.'". is replaced. with roofing of non-combustible

.:i.-= ..";—.="-..eration P.=,.'2 reactors are in service

:o" pbo" t 1':,-'.rs curing .;hich:".'PP safety requirements

.v .re bees.-ii ...en" more stringent, they are to be r~:con

: t -:;ct-:., L.. '."'.:"..' t..ir . The reconst,action concept i-,

based on accomplishment of such a volume of building and

mounting .vork that would allow bringing power units to the

safety level as close to modern requirements 89 possible

within the framework of economic feasibility. Reconstruction

involves replacement of equipment with expired service life.In particular it is planned to replace fuel chanxels. Rates

of radiation deformation of fuel channel Eirconium pipes

and columns of reactor graphit e stacking used in the

design turned out to be somewhat underestimated. Design gapa

between 'uel channels and holes in graphite blocks have been

'' considerably taken ~p over 15 years of service. Therefore,

it i. regarded advisable to combine replacement of fuel

channels and sizing of holes in graphite columns to initialvalues with shut-down of power units for reconstruction.

The first group of fuel channels is being replaced at the

Leningrad !iPP for mastering and optimizing the procedures

oi mass replacement of the fuel channel- This work is being

carried out at po;:er unit i'lo ~ 1 which is under major repair

a t pr esen t.

7. GP RATIO!J O." l'Pr"'s '."ITH P„-lZ i!"-ACTORS I!I 1986-88

After the accident at power unit '.io.4 of the Chernobyl

NPP on April 26, 1986, unit No-3 located in the same main

building and closely connected technically with the emergenc:

unit but practically intact after the explosion was shut

down in about 3.5 hours. Units Nos 1 and 2 located in other

buildings adjoining the emergency one with one end wall

only;Iere shut down on April 27 at 1 h 13 min and 2 h 13

min respectively, i.e. 24 hours after," the accident /3/.Considerable radioa "tive contamination of the equipment and

rooms oi'nits "..os 1,2 and 3 resulted from ingress of ra-dioacti re matter through the ventilation system and destroyec

roof of »nit "..o.3 machine room. After cooling down, units

llos 1 and 2 vere changed over to prolonged dead period

for the time of start-up expectation, while unit No-3 was

laid up. The Govenrment commission put forward a task ofperforming decontamination of units Nos 1,2 and 3 of the

I

Ch;HAPP as the first stage of their preparation for start-up

a .d operation.

As early es in September,1986, start-up and adjustment

opera tions began at units Hos 1 and 2; these were performed

in view of primary measures for increasinj-,'safety of NPP's

with P>!Z reactors: formation of reactor charges with experi-

mental d termi'.ation of reactivity effect at core dewatering;

inspection, ei.','s(ment and modification of CSS; integrated

tests of reactor emergency cooling system; adjustment and

tests of C"" charnel cooling circuit; installation and adjust

ment oi'rotection enabling/disabling automatic light panels;

24-

prifiling of coolant flow rate through fuel channels; ins-

pection o energy release distribution monitoring system;

reactor temperature monitoring system, etc- Similar work

'.vB s ca rri ed out for all unit and plant sy s

terna

and equ ipment.

Unit No.l of the ChNPP was connected to the power system on

October 1; power unit No.2 was loaded on November 6, 1986.

After completion of repair and reconstruction work in Oc-

tober, 1987, start-up and adjustment operat ions began atunit ',o ~ ~ of the Ch'IPP and it was connected to the system

II

on December 10, 1987.

s .rith P .'K reactors produced in 1985 14.7 billion:.ii-h nore than in 1386 which stems from the following rea-

sons:

(a) '"reakdo:In of units Nos 3 and 4 and prolonged dead

perio.: o: u=.its 'los 1 and 2 at the Chernobyl NPP, inspite

of attaining nominal parameters by unit No.4 of the Kursk

,'c Py

(b) i.rple-":entation of measures for increasing safety

o ..r='s i:h r .," reactors which required additional shut-

.owns of power units and their operation at decreased power

levels for a certai.-. time;

(c} -rolonged major rep~'.~ (over 7 months) of unit No-

1 e t t h '...ole..sk .PP stemming from low-quality pipelines

of "i.""ulat io:. "irc it.:able - '.."e.,ents technical-economic indexes of NPP's

-;ith =. '; " ec-.ors in 1'386-88, i.e. after the Chernobyl

accident ."- t "~.=.= ives from these deta, in 1988 power ge=e-

'=�'o . 't '--'.' it;".:".E. reactors somewhat exceeded that

Table 2

Technical-Economic Inrlexea of PPP'a v;ith Pi>IU: in 1986-88

I!o. Btart-up

':PP

Kf'PP

C h)'?PP

1 l0002 10003 10004 10001 10002 10003 10004 10001 10002 10003 1000

02 12.85

!PP Unit %V Date of 1986G, b il- ICUP,lionk'iV- h pO

'6.42 73.3-7.75 88.47.53 86.07.69 87.86.14 'ra. 16.07 69.35.26 60.06.84 78.1

40.63.16 36 00 - 0

G ~ b il-lionkLV- h

8.117-236.02

7~ 32

7.02

7.205.466.176.416.780 34

1987ICUF,

, t/0

92.682 568.783.6

80 2

82. 2

6». 370-4 .

73.177. 4

G, bil-lionk>V-h

7.177 387.606.597.126.457.267.196.136.6»

6.95 .

1988IC1lP,

81.984. 386.975m ~

81 17) )

sj2 ~ 7

(ig. )

(4) ~ '>

7')

S7IPP 1 1000 3.28. 2 100'0 04.05.85 7.»0

I )PP 1 1500.. 9 [V3

2 1500 18.08.87

Total annual power generation

37- 58~ 2

7 g

84.8

7 456.366.65» ~ 53

)34 ~ 07

"~O 6

') l.0

7.cl7»')5 ~

7- 55

t't ~

g3 o

fl

'.7-'('3.

)

- 26-

in 1385..'is regards installed capacity, failure of unit Io.4oz the:.'h;.."P wss compensated for by start-up of ~.it:;o.'tthe I"'IPP in August, 1987. Reduction of ICUF of both power

units at the Ignslina NPP stems from temporal limitation of

their electrical capacity to 1250 15V.

By the beginning of 1989, installed capacity of IIPP's

;vith Pi:X-1000 and PEAK-1500 x'eactors was 44lo (16 GW) ofirstslled capacity of all '.IPP's in the USSR (36 G'iY). Over

'.8,> (1C.'. billion k"-h) of total",IPP power generation (215

billion k"I-h) were produced in 1988 at these NPP's. Thus,

power „'e-. ration per each kY( of installed capsc.',ty of NPP's

.vith P„-".Z reactors for 8;~ exceeded the x'espective value of;.:"P';vith ~~P-10CO reactors. At present two more power

units with P=iÃ-1000 reactoxs near completion of mounting

operations: unit;Io.3 st the Smolensk iG'P and unit No.5 at

the I<ursk .",PP. !essures for incx'easing safety of reactors of

this type ..ere fully implemented at these units. Further de-

velop...ent of channel-type reactors is a module-type boiling

po':,'el eec-,o r i '; h ~1i"t ricsl capacity of 800 1'I l increased

safety of ..hich i.s ensured by its intrinsic creatures. The

„-.oal of this 'rk is to develop a safe .nuclear reactor in-

dependent of external poser sources, capable of sutomati"ally

shutting do.vn et sny dangerous variations of parameters and

opera ti.-g ",.o.".litions and meeting modern ecological require-

—..ents.

1. Dolle."hei -..:;.,et al. '~om= - eculiaritios and .':xperience

of Operation of '..'"'-"'". >'"-'4"" -'-': Reactor, .:.i'., "ienna,

1983.c'. Adamov Z.O., et al. ":r.'~.ase of .-:aret.'f ".,P:"'s v.itn

PghK Reactors,",, Atomic:-.nervy. 1987, v.&c, issue ~, p zl9

3. Information of Accident at Chernobyl I PP and Its Conseo-

uences Prepared for IAEA, Atomic Energy, 1986, v.61,

issue 5, p 301

The 1st International workshop on Severe Accidents in

Nuclear Power Plants

R 8 D PROGRAMS FOR RELIABILITY AND SAFETY

IMPROVEMENT

O.Yu.Novosel'sky RDIPE

B.A.Gabaraev RDIPE

E.U.Burlakov Kurchatov AEI

A.K.Kalugin Kurchatov AEI

D.A.Mikhallov Kurchatov AEI

Depots, Sotchy, USSR

3G oct. — 3 nov. 1989

ABSTRACT

The brief review of main trends in neutron-physical and

thermal-hydraulic investigations implemented during the

RBbK development is presented. Particularly detail considera-

tion is given to investigations performed following the Cher-

nobyl accident in 1986 and aimed,.to reactor safety improvement.

The program+ of the experimental investigations on physics,thermal-hydraulics and thermal-mechanics that will be carriedout in the nearest future are briefly outlined.

INTRODUCTION

The Chernobyl accident forced to analyse carefully all

aspects of the RB5Z -type reactor safety once more. The

measures eliminating fast uncontrolled power growth have been

taken immediately following the accident (installation of

absorbers, updating of the control rod design etc.). And feasi-

bility study of the reactor safety performed all over again

raised a variety of the problems the solution of which requires

the carrying out a set of the important studies. Among these

is the development of detailed models of reactor dynamics desc-

ribing the accid.ent conditions. Thus, a demand was created for

more thorough study of reactor channels and fuel elements under

the loss of coolant accidents (LOCA) conditions as well as of

processes that occur in the accident localization systems when

gaseous fission products achieve these systems. Much attention,i

is being given to study of the conditions of the upper protec-:."

tive plate tearing off as well as to development of the measu-

res eliminating such accident.

The paper presents the main trends of the K3MK safety studie.

that are being carried, out at present. Summary of the investi-

gations on physics, thermal-hydraulics and thermal-mechanics

performed earlier during the RBbE development is given for the

more completeness of presentation.

NEUTRON-PHYSICAL INVESTIGATION AND SAFETY

The development of the RBMK wss based on the design and

long-term operation experience of the channel-type uranium-

graphite reactors at the First NPP, Siberia NPP and Beloyarsk

NPP. Therefore, the development of the RBMK neutron-physics

prediction procedure was based on the methpds of neutron-

physics prediction tested at the operating reactors. However

as a result of aiming to the continuous refueling

and some the RHE structural features a demand arose for the

additional methodsy and in some cases for the development of

new prediction procedures and codes.

The experiments were performed in parallel with the develop-

ment of numerical models. They were carried, out at the modified

graphite rig available. The channel lattice spacing

was equal to design one and made up 25 cm, and the num-

ber of the channels - 81. Obtained data were used to improve

the prediction procedures and evaluate the reactor performances.

In 1980 the critical test rig with 324 channels was construc-

ted in the Kurchatov's Atomic Energy Institute (Kurchatov AET).

The subcritical test rigs located, at P -1 reactor were also

used to test the cell characteristics.At the beginning of 1980s the experimental and numerical

studies aimed. to RBbK neutron-physical characteristics improve-

ment were extended. A variety of the present-day codes were

developed and, the experiments at the RZK test rig were started.

Following the Chernobyl NPP accident the works aimed at

the operating RBLK unit safety improvement were expanded.

They have an all-round nature. However we review only the

aspects of safety and neutron physics.

'iYithin the framework of the measures aimed at void reacti-

vity coefficient decreas the switch-over of the reactors to

make-up fuel with 2.4 % enrichment is being implemented. A set

of calculational-experimental investigations with respect to

grounding on nuclear safety when using the fuel of 2.4 % enrich-

ment have been carried out.

In an effort to study the transients and to analyse the

Chernobyl accident as well, the three-dimensional dyramic TRIADA

STEPAN cahs were developed. These codes provide the ability to

perform the detailed analysis of the power density fields both

in the steady-state and transient conditions including the

accident ones and the safety system work's well.

The codes ax@designed for describing power density field

dynamics under severe accidents (maximum design basis accident,

a variety of hypothetical accidents and accident with the drum-

separator ruptures). further improvement of thee codm is aimed

at more detail explaining of the thermal-hydraulics and impro-

vement of adopted constants (using estimated. data file ENDF/B).

The proposals on building the critical uranium-graphite

thermal test rig KRUG-T (full-scall in height with the control

and safety system ) are being considered for codes :.erification.

This experimental apparatus should be furnished with the systems

for water and graphite heat-up and up-to-date equipment for

data monitoring and processing.

The works on updating the methods of the RBMZreactor'ower

density field monitoring using the small-scale devices

are being carried out. At present the fast shutdown system

(FSS) is being introduced in the RBMK reactors. The time of

the rod insertion in the core makes up less than 2.5 s. Besides,

the additional shutdown system (ASS) based on fast ( < 1 s) in-sertion of gaseous absorber in the core is under development

in the Kurchatov AEI and RDIPE.

In 1988-89 the program of investigations using the changed

graphite stack s~cture was implemented. It is known that the

core voiding depends on uranium nuclei-graphite nuclei ratio.The next section of this paper presents the results of

the thermal-hydraulic tests for the fuel assemblies with 36

fuel elements having smaller diameters.

In these fuel assemblies the urdu.um charging per c>~noel

increases by ~ 15/o. However, the analyses have shown that in

this case the void reactivity coefficient decreases only

slightly. To improve the core performance it is necessary to

reduce the amount of graphite in cell (up to 20 5). For this

purpose, the lattice pitch should be decreased (it is impossible

for the operating units) or corners of the square graphite

block should be cut off (this method. can be implemented at the

units under construction). The majority of the neutron-physical

experiments using the modified grephite stack (i.e. the blocks

with cutted-off corners) have been carried out at the RBMK

test rig in Kurchatov AEI. The experiments and analyses perfor-med. have shown that when using the modified. stack the zero or

slightly negative values of void reactivity coefficient are

achieved, even in the absence of BO additional absorbers.

The program of more precise definement and provision of

design lifetime of the graphite stack and. channels have been

already planned and now we are implementing this program,

At present the investigations are being performed with

the aim of more precise definement of the volatile fission pro-

duct distribution (VPP) in compartments of the accident loca-

lization system (ALS) and. outside the power unit under the

severe accident conditions including the hypothetical accidents.

We have to study the VFP distribution between the phases in

the coolant before and after discharge through the rupture;

condensation, deposition and separation of liquid. phase under

the conditions close to the real ALS; dynamics of thermal-hyd-

rau1ic parameters in the ALS compartments; escape of steam-

water medium from the ALS due to seal failure.Within the framework of technical feasibility of the REIK-

safety under loss of coolant accidents (LOCA) conditions the

works on the investigation of "new" type of accidents, i.e.the accidents with primary system pipeline .not full-cross

section rupture (partial ruptures, PR) that can result in a

short-time full loss of cooling in a group of channels are

being performed. now. The investigations of these accidents

are implemented, using the numerical analyses. The experiments

are also carried out at the KS and KS3 test rigs in the

Kurchatov AEI.

Besides, a variety of the research programs directly con-

cerned with the RBMX safety increase is planned in addition

to above-mentioned ones. ihese program include the following:

- performing the probability safety analysis for the

RBMZ;

- improvement of the automatic system of technological

processes control, development of operator support systems

and. trainers;- improvement of CSS actuators and additional absorbers.

THERMAL-HYDRAULICS AND THKQVL-MECHANICS

This section gives the most general idea about completed

and continuing studies of thermal-hydraulic and thermal-mecha-

nical aspects of the RBMK safety. Pr"sent survey covers 20-year

period measured from the development of RBMK technical

pro)cot. Some of reviewed, studies have been completed before

the Chernobyl accident, while others were extended and deepe-

ned after this accident. Purthemore, a number of researches

was started anew. Among latter we have first of all to men-

tion the investigations of channel tubes (RBMK core pressure

tubes are implicated,) behaviour under accident conditions..

Researches in the field of thermal-hydraulics and thermal-

mechanics pursue next two main ob]ects:— obtaining closing correlations (development of mathe-

matical models for involved processes and, phenomena);

- verification of various codes.

In this connection it appears relevant to present briefly a

pack of codes available for RBMZ design basis accident3 analysis.Among the codes designed for simulating the whole primary system

it should be mentioned DIKRUS developed at NPOE. This code isbased on 100-c~»el (100-group) model of core and on one-dimen-

sional neutron kinetics. It provides the ability to calculatethermal-hydraulic parameters of RBMK primary system

non-statiorm"„'onditions

including the pipeline break. Other codes applied

for such calculations are RELAP-type codes. Temporal variations

of coolant flowrates, pressure and, enthalpy evaluated for typi-cal points of RBMK primary system represent the input informa-

tion (boundary conditions) for the codes complex RET-TRANS-

-RAPTA-PIPELIA(RTRP). This complex provides the prediction offuel element initial state (RET), non-stationary thermsl-hyd-

raulics in the channel tube and in the adjoined flow paths

( TRANS) f stressed-strained staM cb~~ge of the fuel element

cladding (RAPTA) as well as of the ch~»el tube (PIPELLA) un-

der accident conditions.. Fuel element cladding failure time

and nature are determined. Failure criteria need the more pre-

cise definition.

Thermal-hydraulics of the reactor internal cav'ty (RIG)

and the accident localization system (ALS) compartments under

conditions of channel tube break or primary system pipeline

rupture accidents is predicted. by means of such codes as

KRITIKA, LEVIS, BREAK, VSPLESK. Former provides the evaluation

of the coolant discharge through ruptures of primary system

as well as the steam flowrate through steam outlet tubes of ALS.

10

This cod e i 3 applicable to both the critical flow and the

usu!~1 hydraulic (subsonic) flow of two-phase mixture. LEVIS

and BREAK are designed for prediction of the RIC thermal-hyd-

raulics including temporal variation of pressure. VSPLESK code

provid.es the evaluation of,: thermal-hydraulic parameters (pres-

sure, flowrate and enthalpy) in ALS compartments under LOCA

conditions.

As a rule, techniques of thermal-hydraulic predictions reali-zed in above-mentioned, codes require some closing correlations

describing the transfer of heat, mass and momentum. The necessa-

ry information partly have been obtained from technical liters-ture however the main sources are special experimental investi-

gations.

Post-Dryout Hest Transfer

A great bulk of experimental information in the field of

post-dryout heat transfer intensity in heated rod bundles

simulating the K3MZ-1000 and. RBMK-1500 fuel assemblies is obta-

ined over the ranges of parameters: P = 7-8 MPa, ~W = 600-

2000 kg/(m s), ~.2-0.8, q = 0.2-1.1 KW/m . Experiments have2 = 2

revealed strongly manifested localization of heat transfer

deterio ation in rod bundle. This is especially true for rod

bundles with intensifying grids. This circumstance complicates

the dry spot (deteriorated heat transfer zone) detection on

heated rods surface thus increasing the measured heat transfer

coefficient uncertainty. As a whole, over 1500 data points

have been obtained for transition and film modes of boiling.

The intensifying grids expand (along the x-axis) the region of

transition boiling. Typical values of heat transfer in this

region exceed 10 llf/(m K).Heat transfer coefficients in the region of film boiling

fall into the range between 3000 and 10000 8/(m K). Obtained

data don't contradict to known results of other investigators.

Our experimental points for rod bundle with intensifying grids

are compared with Hsttson et. al. correlation /1/ (Pig.1).The intensifier effect reveals itself in deviation of a great

bulk of data on the right side of this figure. Compared data

have been obtained at EB NPOE and RDIPE.

Still being in progress the experiments at AHEI cover the range

of lower mass velocities, heat fluxes and. press~es. Post-dryout

heat transfer in the region of negative and small positive

values of x are predicted by means of correlation based, on

Stewart-Groeneveld data /2/ which were obtained for the tube

at pressure 9.0 KPa in wide ranges of mass velocity and inletsubcooling. These predictions are considered as conservative

estimate. Kore reliable information is to be obtained in expe-

riments with rod bundles after overcoming some corresponding

technical and methodical difficulties.Rod bundle ref looding studies at EB NPOE (stainless steel

rods) and RDIPE (zirconium alloy rods) have given the results

which agree quite enough with known data of other investigators.

In 1986 the realization of experimental investigation pro-

gram &tended to measure thermal-hydraulic characteristics of

closely spaced bundles for RHK channels began in

12

the test rig KS at Kurchatov AEI. Experiments have been carried

out with fulL-scale simulators of fuel assemblies cnree the ranges

of parameters: pressure 7-8.5 MPa, mass velocity 1600-4000

Rg/(m s) and heat. power up to 6000 RW. Puel assembly simulators

represented the 36-rod bundle each rod being 10 mm in diameter.

Total 4 modifications of fuel assembly simulators have been tes-

ted. They differ in the grid axial space and the rods arran-

gement in bundle. One of simulators was sub)ected to voiding

conditions which developed in case of feed water interruption.

Obtained results are similar to those for 18-rod bundle. Men-

tioned investigation program is the organic part of complex

study aimed to clarify the expediency of R3hK 18-rod 'bundle

replacement by 36-rod bundles. This study covers all aspects:

physics, thermal-hydraulics, fabrication technology, reliabi-

lity, exploitation, safety. Preliminary results don't reveal

azar decisive advantages of 36-rod bundles.

Parameters d~~+cs under LOCA conditions

In 1977-1980 RDIPE have performed the program of investi-

gation of RBMK primary system thermal-hydraulic parameters

behaviour under LOCA conditions'",in case of seal failure of the

printery system pipeline upper part or the steam tube. Por one-

circuit reactor largest probability of such accident is condi-

tioned by relief failure ,'not closed valve).

Experiments have been carried out on the test rig 3M which

had full-scale altitudes and hydraulic resistances which volu-

13

mes and power were scaled-down,'the scale of power was 1:300).On this test rig there were stuc.ied not only the primary system p

rameter dynamics for various rates and, depths of depressuri-

zation but also possibilities for natural circulation preser-

vation under such conditions and measures favouring development

and support of natural circulation. Experiments have been per-

formed for upper and. lower disposition of discharge group headers

It should be mentioned that the lower disposition of these hea-

ders corresponds to second series reactors of Leningrad NPP.

Accident versions with and without feed water interruption

were studied. Depressurization rate had values from 0.2 to 1.5MPa/min, the depth of depressurization was up to 0.7 MPa. Feed

water interruption achieved 3 minutes. Then fol1owed feed wat erflowrate recovery up to 10-1005 of nominal value. Provided the

feedwater flow was not interrupted the natural circulation de-

veloped even at d.epressurization rate 1.5 MPa/min. If after3-minute interruption the feedwater was restored only up to

10% of nominal value the depressurization depth below

than 3.5 MPa (or 4.0 MPa at upper disposition of discharge

group headers) causes the ceasing of natural circulation and

results in fuel element superheat. Obtained data have been

applied for emergency core cooling system (ECCS) improvement.

In the same period of time Kurchatov AEI performed

on the test rig KS investigation of RMZ full-scale fuel assembl

simulator thermal conditions in regimes with coolant natu-ml

circulation ceasing and restoring. Since the impressed voltage

of fuel assembly simulator was maintained constant the observed

14

decrease of power during the processes of voiding and heating-

up is at.ributed to increase of simulated fuel elements electricresistance due to the temperature growth. Experiments provided

the measurement of fuel assembly voiding time and reflooding

outset time for various values of heat power in the range

1600-4600 k'Ã. Simultaneously there were debennimd the values of heat

transfer coefficients during fuel assembly voiding as well as

of rewetting front velocities. Obtained data don't contradict

to known results of'ther investigators. These data were used

for development of phenomena models for RAPT.'<-RBMZ code.

Later the same test rig KS have provided the ability to

study dryout in the case of RBMK-1500 fuel assembly cooling

by means reversed flow of coolant. This situation corresponds

to LOCA conditions caused by tube rup';ure s,t the i ore inlet.I

Experiments have been carried out in next ranges ok', parameters:

pressure from 6,4 to 8.4 MPa and mass velocity from'1000 to

3500 kg/(m s). Dryout was observed for relative enthalpies x

from 0,2 to 0.5. Obtained data are in agreement with the cor-

relation for RBMK-100 fuel assembly.

Critical Plows

The extensive program of subcooled and saturated water,

steam-water mixture and steam critical flow investigations

is carried out jointly by RDIPE and EB NPOE from the middle

of seventies. Obtained data are presented in /3-10/.Agreat bulk

of data falls into the range relevant to RBMK:

— pressure from 0,3 to 8,5 MPa;

- inlet subcooling of water from 0 to 50 C;

- inlet quality from 0 to 1,0;- inlet superheat of steam from 0 to 30 "C.

Orifices and axesymmetrical straight ''ubes (from 10 to

38 mm in diameter and up to 120 in dimensionless length),

Laval nozzles with and without transversal injection (from

10 to 30 mm in diameter of throat) and full-scale control

valve of RBMK fuel channel have been tested. Study of critical

flows in branched flow paths is still in progress. Results of

mentioned experimental investigations are applied for

KRITIKA, LEVIS and, BREAK codes verification.

On the basis of obtained data bank the regulation

document /11/ have b'een elaborated. Besides, effective flowrate

limiters fabricated as Laval nozzles with t~sversal injection

have been developed. They are applied in PBMK-1500.

Afore-mentioned studies provide substantial expanding

of investigated ranges of flow parameters. In particular the

role of geometric scale appears to be unessential.

Investigations of the fuel element

claddings and channel tubes behaviour

under LOCA conditions

Since 1980 s much attention is being given to jnvestigation

of the behaviour of the RBMK fuel element claddings and the

properties of Zr-1 5 Nb alloy under the IOCA conditions.

In 1983 the RDIPE, EB NPOE and Bochvar ASRIIM completed a

first set of experiments on stressed-strained behaviour of

16

the RBhK fuel element cladd.ings at high temperatures.

The great bulk of these tests was carried out under inert

atmosphere conditions, i.e. without oxidation of zirconium insteam environment. The range of parameters: pressure

differi="".~-.. across the claddings from - 8,0 to +8,0 MPa,

cladding temperature 500-1000 C, the rate of the fuel

element simulator cladding heat-up 32-200 C/s.

Heating was accomplished by direct passage of electriccurrent. The criterion of cladding failure in swelling

(strained one) and the criterion of stability loss during

external pressing have been obtained by the results of about

200 experiments.

The analytical relations for these failure criteriahave been developed. The results obtained in cladding external

pressing experiments that had been performed on the fuel pelletscolumn simulator revealed a number of specific features of strainbehaviour. Due to the fact that such strain br.haviour of RBMZ

fue'laddings is the most probable under LOCA conditions, the

investigations oz fuel cladding strain and failure were perfor-med on the fuel pellets column defects (break in fuel column,

fuel pellets spalls, etc.).The studies of a such type were performed at the RDXPE. The

fuel element simulators with breaks in "fuel" column consisting

of a set of aluminium oxide pellets were tested. The cladding

was heated by the electric current passage through it as before.The preliminary comparative analysis of heating dynamics offull-scale fuel elements and simulators of fuel elements being

filled and heated electrically in different ways,'howed that

the d'rect electric heating under -.he ~~' accident conditi=."w

crea.es the =ost severe strain cond'tions for fuel elemen.s

under LOCA conditions zn comparison w'h m~X ones though

it results in add'.'onal uncertai ties in:uel cladding .e=-

perature. The tests were conducted in s.earn e vi onment a.6-S V:-a and at pressure inside the si~ator of 0.1 ~':-a, and

cladding temperatu e from BOO to 1300'G, the heat-up rate is

20.60 'C /s. The break in fuel column 's 0-20 mm in length,

The test duration under given conditions ' .rom 5 m' to 2

hou s, the less time corresponds to higher temperatures. T e

simulators were cooled down following the exposing at hign .em-

perature either in steam environment when the electric eating.

is switched off or by subcooled water without switching off

the heating. There are two types of fuel cladding failure:

fuel cladding break ir. "fuel" column defect location and brit'." =

failure of fuel cladding in.o several parts. The first fa'lure

type - loss of sealing with formation of a hole of ~ 1 mm in2

area occurred during heating-up either during exposing to gi-

ven conditions. The other one - brittle failure takes place

during simulators cooling down after exposing them to given

condi.ions. About 100 specimens we e tested. The main results

are: the regions of cladding sealing being intact or failured

under strain in fuel break local='on are determined. The sea-

ling failure occures between these regions with probability

of 0-1.0 ard the probability of cladding break rises with

increase in temperature and fuel column break length (rig.2).

1 ~ ,The steam pressu e ove" the g'ven ~e doesn'. effec".on these results. The c'rwracteristic of cladding =e:al eibri ~ .-lement 's p esented in the form of the equ'va'nt ox"de layer.Zi>~e 3 'llustrates the rela.ionship be.ween t:"'s value and

the test te..peratu"e. Past coo ir--down o simu" ators by =.ea.".s

o water results in some improvement of meehan'cal propertiesdue to reduction of the hydride phase particle sizes in metal.

The experiments are carried out in two directions:1) ga'n'ng of required s.atistics in tes.s w'h coolie~-down

by means of water;

) repetition of some tests but only with cladding heat-up

by means of ouilt-in heater.

The results obtained in these tests refer to unirradiated

claddings. The loop (ampoule) investigat'ns of the R3MK fuelcladding behaviour (shortened fuel element specimens) with

different burr.-out depth are planned .o accoun. for the effectof reactor environment -actors such as: radiation embrittle;.ent,iodine corrosion, interaction with fuel, etc. The experiments

will oe performed in the IVV-2 reactor loop in S3 RDIP"". The

fission products release into the coolant due to seal ailureof the overheated cladding externally pressing the fuel column

will be simulteneously studied in tests. About 30 fuel elements

will be tested, the tests will be begin in 1989.The program for investigation the RBhK fuel channel behavio"

under LOCA conditions is being implemented since the last year.Th' program includes the. experimental study of such problems a

19

a) strained behaviour of channel tubes at temperatures

up to 1000'C under e:excessive steam pressure over the range

from 8.0 to 0.1 EZa;

b) stressed-strained state dymunics of fuel channels at

nominal coolant parameters adjacent to chanr.el being failured

in the core,

The investigation of the problem (a) comprises:

- tests performed on specimens of full-scale fuel channel

tubes.

- tests performed on model (small-scale) tube specimens

of Zr-2.55 Nb alloy over the wide temperature and pressure

ranges.

The KT test facility was constructed to perform the inves-

tigations on full-scale tube specimens. Figure 4 shows the

test section of this facility. The specimens of full-scale

tubes of 80 mm in inner dia and 120 mm long with a "steel-zir-

conium" transition are being tested. The specimens were heated-

-up by means of the built-in electric heater.

Test conditions:

- initial state: saturated steam pressure of 8 MPa inside

the tube, nitrogen-helium mixture at 0.1 KPa, solid contact

rings and. graphite bl~>cks inside the tube;I

- accident simulation: tube heating-up at the given rate

(up to 15'C/s) by thI~ built-in heater up to 700+1000 C with

the simulteneous pressure decrease according to the given law.

P.s a whole, 8 specimens are planned to be tested.

The tests will be started in January, 1990.

20

To develop the criterion of fuel channel tube rupture, we

need the statistics which is not provided. by full-scale tube

specimen tests due to their difficulties and high-costs. The

tests of small-scale models should help us to overcome thesedifficulties. But there are some limitations related to zirco-nium oxidization processes in steam.

Tubes of 20 mm in dia, 1 mm thick and. ~300 mm long areused as test specimens. Test conditions are similar to the

previous ones. Test extend is about 200 specimens. Test resultsare: strength and strained characteristics including rupture

criterion (for instance, maximum strain) and metallographic

characteristics. The investigation should be completed in 1991.The TKR experimental facility is being developed at Zuev-

skaya Experimental Heating and Power Plant to perform the in-vestigations of problem (b). Fig.5 presents the test sectionof this facility —reactor graphite stack module (RGSM).

This is a full-scale fragment of REE graphite stack with

fuel'charnel tubes. As a whole, there are 25 (5 x 5) channel

tubes that are surrounded by a row of reflector columns. The

rigidity of "thrown off" stack is simulated by special stops,while the hydraulic resistance is modeled by throttle shields.RGSM is encased in protective housing and connected to the

accident localization system receiving the condensing steam.

Water is removed from the RGS5l housing through the drain pipes.The central channel is destined for arranging saturated water

discharge into the stack with flowrate up to 80 kg/s after the"sudden" rupture of the tub~wri.l.

The surrounding 24 fuel channel tuoes are equipped with

8traingauges�.

There are d is plac e ment Oct ec tors a."ound the

periphery of a stack that record the sag of outside row columr~.

More than 100 thermocouples and about 30 statistic pressure

transducers are placed in the graphite blocks o= the ..GCEI

forming "equipotent'al" surfaces. The computerized measuring

system providee the detection of the stean-water mixture

spreading out pattern over the stack and the graphite cooling-

down mode. The tests permit to resolve the main question-what

are there the possioilities and conditions of the dependent

failure occurrence (i.e. rupture of'djacent channel tubes)

under force factors of the first channel tube rupture.

The experiments are planned to begin in 1991.The paper summaries the main investigetion programs that

are either completed, being completed or being developed. 'iYe

don't touch upon such studies performed as reactor graphite

flaring up, burning and ways of 'ourning suppression, or as

oxidization of zirconium alloys in water steam etc. The in--

vestigations of the w*te hammer in the branched pipeline are

under development.

The first version of EQQFiZB Code have been generated and

the activit'es oz. developing the experimental facility for the

verification of this code are under way.

The experimental study of fine-dispersed moisture genera-

tion and transfer is being carried out in the simulator of

accident localization system compartments to define the empi-

rical cons. ants specifying tne mass transfer in VSPIiESK Code

22

more exactly.

Thus,;he thermal-hydraulic, thermal-mechanical and

neutron-physical investigations provide with necessary infor-

mation the design-basis accident analysis performed by means

of the available computer codes. The abundant experimental

data is used to form the new data banks and to fill the availabl~

ones. Among them there are data banks worthing attention on

dryout heat transfer in tubes and rod bundles, on criticalflows and. thermal-hydraulic parameters dynamics and on experi=:

ments performed with the critical assemblies that are widely

used in GAZA countries.

The further development of the RBMK safety studies in dif-ferent aspects is likely to be related to the development of

a program described briefly in the first section of the present

report and intended for investigation of severe accidents

with the subsequent core destruction and. fuel melting.

The specific character of these investigations is largelp

determined by the RED design features and calls for a sepa-

rate consideration.

23

Nomenclature

AHEI - All-Union Heat Engineering Institute

ALS - accident localization system

Bocher ASRIIM - Bochvar All-Union Scientific and ResearchInstitute of Inorganic Materials

EB NPOE - Electrogcresk Branch of NPO "Energia"

Kurchatov AEI - Kurchatov Atomic Energy Institute

LOCA - loss of coolant accident

MDBA - ~~~mum design basis accident

NPOE - NPO "Energia"

NPP - nuclear power plant

RIC - reactor internal cavity

RDIPE - Research and Development Institute of Power

Engineering

SB RDIPE - Sverdlovsk Branch of Research and Development

Institute of Power Engineering

h - heat transfer coefficient, W/(m K)2

p - pressure, MPa

q - surface heat flux, ZW/m2

x - flow equilibrium quality (relative enthalpy)

pnv - mass velocity, kg/(m s)2

Acknowledgements

The authors express special thanks to A.N.Ryabov, V.I.Zsi-kov and A.I.Emelyanov for amiably granted materials and valuable

help ir. preparing this paper.

Reference

1 ~

2 ~

3.

4.

5.

6.

7.

Mattson R.J. et al. Regression analysis of post-CHF flow

boiling data. V Int. Heat Transfe:." Conf. Tokyo, 1974,

B.3.8, pp.115-119.Stewart J.C., Gmeneveld D.C. Low-quality and Subcooled

Film Boiling of Water at Elevated, Pressure Nucl.Zng, and

Design, v.61, 1981, pp.259-272.

KedopKQB JI.P., JlyTOBNHQB C.3., TMXOHeHKO Jl."'X. BJDIBHHe MRCID-

TR6HHx lfsKToPos Ha KPHTHRecKHA PacxoP, Hac~eHHQH Bog+ H3

IIpeaz Tpy6 c OCTpoN axo~oA KpoMKOA. -Terxn03HepreTHKR, I977,5 7, C.72-76.

TrrxoHeHKO Jl.K., Kapacea 3.K., JlyTOBHHOB C,3., I'a6apaes B.A.,Tpy6KHH E,N, NccJIepOBRHMe xapaKTepHcTHK BcTRBQK orpaHHQeHMJI

pacxopa IIpH MopeJIHposaHHH asapzrrHOA pa3repMeTH3algfH KQHTypa

peaKTopa. -ATDMHRH 3Heprm, I980, T.49, sbnl.2, c.83-86.I'a6apaea B.A., Kapaces 3.K., JlyTOBHHOB C.3., HosoceJrbcKMA O.JO.

TMxoHeHKQ, Jl.K., Tpy6KHH E.N. PacweTHO-3KCIIepwTeHTRJrbHoe nccJre-

poBRHHe KpHTHMecKHx pacxopoa B coIIJIRX: C6,poKJIBpoa HR ceMHHap(

C3B "TeIIJIO+3HKR-82". —KapJrosbr Bapbr, QCCP, I982, T.I, c.I83-I'.BpbrJres E.C., BacaneacKHR B.II., I'a6apaes B.A., KapaceB 3,K.,JlyTQBHHQB C.3., TzxoHeHKO JI .K., Tpy6KlrH E.N. 3KCIIeprrMeHTanbHO~

zccJreposaHrre KpHTrr~ecKoro HCTeweHHJI aopg, rlapoao~oR CMecM.H

CJIR60IIeperpeToro Ilapa ~epe3 3arlopHQ-peryJrzpylouprR KJranaH peRKT<

pos Trrlla PBMK. —Borlpocbr RTOMHOA HayKH H TexHHKH, cepm: CH3z

Ka H TexHHKR RJJepHhlx peaKTnpoa Bbrll.4(26), c.33-38''a6apaeaB.A., Kapacea 3.K., JlyTQBHH0B C.3., TNxoHeHKQ JI.K,g

Tpy6KrrH E.N. KprrTz~ecKrre pacxo~ so~, IIapoaopsIHOQ cMecH H

Irapa s IlpRMbrx Tpy6ax c OCTpoA sxopgoR KpoMKON: C6.Te3zcoa

poKnBpoa Bcecolo3HOM KOH@peH~ Iro Tenno+3HKe H r~opgHRMH-

Ke rlpoq~ecco= KHlleHHrr H KOHpeHca~z. —Para, I982, T.II, c.l26

I

OV

,OO

O 4",/go

0o ~o

0

0

O

do o o

4 COO

4C i@

oo,0 0

o o

'SO' 'O o

Ma '.'7

g QQ Q

o oyO

ll 00 I) 00

6P tP @'

exp 10 'fg(m2~

Pig. 4 Comparison of the data arith the correlation

of Mattson, P ~ 7,38 Mfa, fv ~1000+2000kg/m s

E'

The borders of oladding failure during hest up in"length of fuel.colunm break - temperature" coordxnstes

P ~ 6 - 8 MPa

oro

I200

l J'33

'J'3

) rII We

Q em

pig.2 jI

- 100 5 probability 'of failure

/g/'- probebflfty of feilure - from 0 to 100 0

The border of cladding brittle failure in"temperature - oxidation depth" coordinates

~ - brittle failure0 - no failure (b - cooldown by water)

- steam cooldown-- - « water cooldown

Scheme of KT facility

HH3~ t~m

xr'I'

p

Hl Ei

HKT

pa

Pig. Q

KT — channel tube;

HKT —electrical heater;

IK —graphite rings;

ZS —graphite blocks;

HFB — electrical heater;

3K1" — casing;

CE —boiler;

PR — flowmeter;

BH —vacuum pump;

—rupture disk;

BH1 — BH3 —valves;P — thermocouple

h —pressure gauge

o — strain sensing element

A-A

IIr

II

II

f'lZ I II I I I

—5

1

p~P

Test section of the TER facility

Pig. g

L:Z~

5'Nltl II lhI ~"II

ll

ro

< / lc 1/ i S n O ~ zg~ i ~ i~,Oi r r'0i Iv<i jf l9 i'J

1 — upper plate; 2- central channel (for leak);3 — tested channels; . 4 — channels of reflector cooling;5 —graphite blocks; , 6 — casing; 7 — steam discharge

pipe; 8 — support plate; 9 —.feed water pipeline;10 — sunolv fdischage> ~inelines

The 1st International Workshop on Severe Accidents in

Nuclear Power Plants

NEUTRON-PHYSICAL. CHARACTERISTICS OF RBMK TYPE REACTORS//

BEFORE THE ACCIDENT AND AFTER IMPLEMENTATION OF MEASURES.!~'OR

INCREASING SAFETY

V.".Borshchev

A.D.2hirnov

V.D.Nikitin

'I.A.StenLok

':..M.".h=-rkashc;v

E.V.duz'lakov

RDIPE

RDIPE

RDIPE

RDIPEI

gi', PE

Kurcha'. - AZI

A.V.Krayushkin Ku"i ha.ov)

AEI—

V.S.Rc manenko Kurchatov

Yu.A.Tishkin Kurchatov AEI

Dagomys, Sotchy, USSR

30 oct. — 3 nov. 1989

N~UTRO'!-PHYMCi.~ C>~~>'C ~: TICSO;-'+5K

TYPZ REACTOR ~ B OR 'i'.; i.CGI3~)iT

A~dD A&iZ

ILAIL

'."-. '" ". i ""..AR~"B

POR INCR~ SI!.0

V.P. Borschev, Z.V.Burleko;, A.D.Rhirnov, .',.V.I:reyushkin,V.b.hikitin, V.S.Romaneri:o, J,.P.~irotkin, I.A.:tenbok,Tu.A Tiehkin, Yu.L!.Cherkaehov

Scientific Research and iesign Institute of

Power ='~incor in~

Kurchator Atomic i.'nervy Inst itute

Moscow y USSR

1989

Y(hen studying the eau."ea o: 'h='. accident at tne pov.'er

unit 4 of the Chernobyl .::-":o....,pri c6, 1966, particular

attention ass paid to positive steer. coe ficient of reactivi-

ty and dynamic characteristic. ni emergency protect ion.

The role of these 'actor in the development o. the acci;

dent is discussed. Physical nature of steer. coefficient of

reactivity is considered. uescribed a:e technical measures

taken for all the PiilK type reactors afte" the ac"ident.

These measures include, primarily, installation in the core

of 80 stationary neutron absorbers for decreasing. steam co-

efficient of reactivity. and mode"ri. ation o= absorbing rods.

The results of measuring reactor characteristics after imple-

menting the above-mentioned measures are presented. The me-

thods of calculations used in the analysis are outlined anc

snd the results of changes of the neutron-physical characte-

ristics after the accident at the "hernobyl NPP are given.

Zt is shown that the measures taken to improve safety of the

PS'ype reactors exclude the possibility of a ne:. accident

of the Chernobyl kind ~ Certain trends are discussed for

further improvement of the operating P~l'G( type reactors.

The accxae..t at po..rr ..x .. ~ otlirt;

. her..ob., 1 „..(I,

on: April: ~6, 1956, - reve::.',c '. '..iavi. shortcorii.".-.; i.-'.:tho

neutron-pgisical characteri. tic" o; the',. '„"'pQ re ctor."-))

/1/. These are. great-.posit ive. sto.am coeffi" ie..t o. resctivit.".and a defect in the er.";erC'e,".c.":;rot ection c)') ate.;:~ linko.'d '. ith

f;ythe construction of the co::trol ro-,'.s. The'~ir'."en" e.of:stercolumn uncer the: cisplscer (.e~ -'.i-..l) leads to

a,"~ossibilit.'f

implementing positive reactivitv b„' „.»..terri o" rods star-ting movement from:the upperr,.ost position. Ore. of th~ '' "st—

priority measures after:thi. ace ir~nt:.vas "re:.'osition='nr o;-.th~

roc upper limit sv(itn. es:o)s ce."fath o= l.-::;.'hich allo):ejelimination of posit ive reectivi't.'::-out but; pres:;nt;ecI;':::

'some problems in controlling .i.utral fielcs at:,po "er. Lster

the,qonstructior.; o-" the ro~s v;es modi'iec. by ezten~ing..th-.:

:":-telesc'opic:: roc .'onnectinj:, tnt:d'.,piaci'»r,v,'ith;:the'od') absor-'

bing'elemeri't. Beloiy is '"onsicore~ -, thi )chen=.i": i"..':"..eutrori-

physical" characteristics:o th~''::..'.:;t.'.pe .re" ctora a,"..:a 'ro'-,

sult: o,f: s: 'nuinber, of. nea mr.e.,: 'or;:i;;.p:roiri.".„:.t~e'ir: safety ./c j.

2. O'.I::.'HQDS OP CALCULATII',G 'I=UTRON-PHYSICAL

CHV..":C -PLASTIC

S

Design characteristics of the PoliZ type reactors arepresented below alongside with experimental

characteristi-

csca. Let us briefly consider the methods of calculation.The reactor cell is calculated with the uae of the 'VIMSD-4

orogramme /3/. Generated in these calculations is a libraryof two group macroscopic neutron sections for reactor prog-

rammes. lieutron sections for cells .vith fuel assemblies arepresentap in the form o a polynomial in burn-up fraction

1

depth, coolant density, fuel temperature, graphite tempera-

ture, Ze-135 concentration. "ihen calculating cells with absc

rbing rods and other types of non-fertile-material assemblie:the cell used is surrounded by an external annular area with

properties simulating cells 'Iith fuel assemblies. Boundary

conditions of reflection are set at the border of the saidexternal area.

The ST"-PAiJ reactor programmes use t;vo-group spproxima-

tion in three-dimensional geometrv. To solve diffusion eaus-

tiono, uae is made o=" finite difference approximation with

one grid p6int per cell. The ST="Pal programme has a versionfor static analysis of the reactor (k f problem) taking intoefaccount feedbacks of fuel and graphite temperatures, cooler.".

density and .'.~-lj5 concentration. These parameters are conv=.

tea at exter.-.al '' terat iona in 10-20 iterations. Another ver-

sion of the ST P.' programme solves non-stationary equation=.

An important feature of reactor progrsmmes for calcula-n

- 5-I)

,/j'ing

the PalIK type reactors is the presence of.a procedure

for restoring neutron field against„ the readings of tc",

intA'one neutron sensors. This procedure corrects the

neutron sections to coordinating the calculated and measure>1/

,I

neutron fluxes. This procedure noticeably improves reliabi-lity of forecast calculations. To consider xenon transient

processes, single-dimension SLOW programme (axial geometry

7/as employed. The mathemat ical model inserted kn the prog-

rsmme includes quasi-static neutron and thermal„engineering

calculations, as vrell as non-stationary equations for grs-phite temperature, concentrations of J-135 and Ze-135 isoto,pes.

3~ ROLE 0" R03".. I: ".0 '",,; T .T CHMNOBYL I1PP

" The present report i,:-: .".o; ai:.~=''ct detailed a.".clysis of

the causes of the accident. e; "esent nereinbelov only the

results of calculations: ~:"..o.;,"'..":~'.iythe role of rod cons-

truction in the accide..'..The'enon transient" ..cc . ''.'receding the accident v!as

mathematically modelled witn t...~: ei"; of the .single-'dimension

SLO'i'J'rogramme."

A certain distribution o;

„reactor, obtained: by averagir~-

at different time "st different

control rods heightwisi..~ of th.

distributions;;hich cccured

reactors operating et 100;o

power level was assumed a.; the '.'ial distribution Thus,

such a distribution may be consicered sufficiently ty'calfor st ea dy- s ta t e condi t ion

s.'3

Distribution of,.energy production with respect to heightJ

was calculated on the asumption that the reactor operates for8 long 't ime with overloading a t 100,~ pov!er 1evel;!ith ma inta i-

Q i

,,ning the selected distribution o. the„control rods.Ij

0It is assumed in th'e moael that trie operator manipulatin

the rods maintains K f = 1 sn!.'pt i-..;ally <'jy:ith respect to

mean square ..root deviation) preserves the preset, form of "neu-

trori f ield. (j

'he reactor of power unit 4 at the ChNPP operated for 8

long time at 100',o power level before po'.;er reduction according

to the graph of Fig.2 began about r 4 hours before the accident

Kith the above-listed asumptions, calculations of the

transient process yield the following results. Evolution of

operation reserveoccured as shown in Pig. 3 where four points .

A,B,C ~ D are marked for "hic."..-ig. ~ represents for.. unct iona

of neutron distributions:

A - initial f ield, t = ',;, rat on 'ese "ve 31 ro:sB - t = 3.51 h, operat io..: .-';.' '..7 ods;

C - t ~= 2c'. 5c h, opera t io:; .""..er'. e c 5.i rods;

D - t = 24.cG h, immed ia t . or~ the accident. operation

reserve 5.~ rods.

Lt is not e:rorthy that " ver„'ms' opera t ion reserve

(curves B, D), the neutro:; .'~:ld maximum moves to the upper

third of the core, where u.". er regular conditions the.e are

ma~ part islly immersei ro. s, enc e small "ield and b"r.".-out.

the -oment of co.-..;1"-:'.-. of slow evolution ( ')()

maximum neutron f ield is. i.-. '. ce .pper half of t h reactor,while the graphite tempera: re .—..axiu.m (:"ig.5) kept on beir~

J/

determined at moment (C) . "hus. nei: tron 'ld orm cannot be',

definitely found from t he graphite temperature field .cr

Operation of emergency protection iras modelled at the

final moment of „t ime (D) with the ~ ."e o= a series o= stat i"calculations for different pos itic..s oz the rods et constant

temperature distributio:. en -.".t=:r 'ensity. The reactivityinsertio ~ curve obtained b.: cele"1-:tio"..s is,,,shown in - ig. 6.Maximum positive reactivit„'s in. crt ac i.. ~.5 s and is0.8p . Positive runout lasted about 0 s. '.t the 0.6 s a

th

slight drop in react ivity was detected (-~ x 10 )..:,'i'~

The consideration presented is mainly of a aua1'i:.tat ive

nature. This stems from lack of information on the initialfield and distribution of rods, conventional simulation of8 t tending personnel act iona, employment of s'impl ifi ed single-

di.-..ension .-..odel, neglect of the role o= delayed .-eutrors and

other "actors o f emergency protect ion operation.

Calculations under the 'iRIADA programme without feedback

carried ou+ to estimate maximum positive runout of reactivity

during operation of emergency protection gave the same result

of O.BP . The previously published value of this runout /4/

equal to 0.5 jt ras ohtatned Itthout dtsconnectttut feedhact.

."ige7 illustrates the results of modelling movements of)I

rods at po;ver .".it ~ nf the Ch:iPP by-. a three-dimensional

non-stationary version of the ST-""PA!1 programme. At the initial

state ( before the beginning of rod travel), the neutron field

=orm ".as restored b„'eadings of the intra-reactor sensors.

«s is seen, at the initial time period after insertion of

small negat ive react ivity, the rod system starts inserting

positive reactivity, the value of 'vhich is approximately

equal to that obtai. ed under the abov~rnentioned programmes.

Thus 8. mor detailed three-dimensional analysis gives the

same picture of reac ti ri:y variation at introduction of rods

as ~ he ..LO;" progrsmne cal" lotions. Cur re ~ in ." ig.7 repre-

sents reactivi'u„" that .vo'.;1d be inserted at po ver unit

by the moderni~ed sy stem o~'ods. Thus, considerably negative

reacti rity i .ether rapidlv inserted in this case. Therefore,

. odernf ~ation of rod construction consisting in elongation

o the co."ecti;.„-. rod r les out the possibility of positive

runout of resctivi:;.

-9-4e VARIATION OF STEAM COEFFICIENT OF REACTIVITY AND

CORE HOT DEHYDRATION

Steam coefficient of resctivi'ty of the PBItx reactors

ta known to be 4-5P before the aootdent at power untt 4 of

the ChNPP. This parameter ie under constant supervision in

the course of operation of the PSMK type reactors. Some re-

sults of measuring cog before the accidents and in the cour=-

of taking the above-mentioned measures are given in Table l.Table 1

Results of Measurements of Steam Coefficient of

Reactivity st PEMK Type Reactors f Fhf tQcP3s~ d~ A«~1;r<

Date Power Average

unit fuel

burn-out,

ll1't't'.dsv/s

em'A

Resc-

qty tivity

margin,

MC rode

Qty of

2 ~ 4i~o

2 6 ~ 03.82 LNPP 1125

04.07.85 SNPP 1180

16.12.85 PlPP-1 1190

27.08.86 BlPP-2 740

02.11.86 C hNPP-1 1100

19.14.86 KNPP-3 1155

20.11.86 SNPP-1 900

21 11.86 Chl'O'P-2 880

08.12.86 KNPP-2 '902 4.12.86 KNPP-1 980

03.01.87 C hNPP-1

4 5

0 25.0

1 37.70 32 '113 50,0

30 47. 4

0, 46.9

54 58-0

81 43.1

35 41.0

54 34.0

50 45.0

5.0

4.5

5.10.052 '

5.30.91.02.2

2.01 7

-10-

26.02.88 SNPP-2 897

2 3

24.01.87 LNPP-1 1039

24.01.87 LNPP-2 1015

26.01 87 LNPP-3 1078

13.03 ~ 87 LNPP-2 882

30.03'7 LNPP-1 907

02 '4 '7 ChNPP-2 873

11,04.87 ZG'P»2 958

11,04.87 ZG'P-2 935

18.04.87 IZG'P»» 4 906

08.05.87 LNPP-3 880

10~ 07.87 ZG'P-2 846

30.07.87 LNPP- 4 860

23.09.87 ZG'P-2 851

2 5,09~ 87 SNPP-2 843

01.10.87 LNPP- 4 872

06.11.87 LNPP-2 898

10.11.87 ChNPP-1 875

24.11.87 SNPP-1 836

01.12.87 KNPP-1'602 5.12.87 ChIiPP-3 880

28.12.87 LNPP-1 943

I» 5

49 43~ 0

50 43~ 0

20 58 ~ 0

79 46.4

80 43.0

81 45.0

54 54. 5

68 53.3

74 44 ~ 0

80 42 ~ 4

80 45. 3

85 41.2

. 80 54.6

80 47.2

85 41.2

BG 54.1

80 49.9

81 49.2

89 42 ~ 0

81 46.7

80 54 9

80 47.0

0

47

2 5

1.7 .

2 ~ 5

0.81 ~ 0

1a2

1.71»4

1.61al

0.61.0

81

0

0

1.2

la4

0.351~ 4

100 0.8265 1~ 0

365 1.3

149 0 ~ 6

11.03.88 - SNPP-1 849

11.03.88 SNPP-2 896

16.03.88 LNPP-3 907

23 03 88 ChNPP-2 939

1.080 50.0

80 53.2

81 47.7

383 0 ~ 85

189 0.818$ 1.0

81 49.0, 87

-11-

1 2 3 4 5

08.04.88 LNPP-2 953 80 45.0

14.05.88 LNPP-4 926 85'4.8

14,07.88 LNPP-2 985

29,07 ~ 88 ChNPP-3 938

26.09.88 KNPP-3 884

80 45. 6

81 45 ~ 4

80 47. 4

28.09.88 C hNPP-1 934'0 44. 6

29 '9 F 88 KNPP-1 308 81 44 ~ 8

30,09,88 KtlPP-4 967

30.03.88 LNPP-1 1025

80 48.1

80 46 5

02 '0.88 ChNPP-2 381 81 45. 4

09~ 10.88 KNPP 2 899 82 45.9

29.12~ 88 Ch'tPP-1 943

18.01.83 S:.'PP-2 10C7

80 43 ~ 7

81 42. 5

NOT.: AA - additional absorber

hIC - manual control

FA - fuel assembly

04.11.88 glPP-1 899 81 45.8

25 ~ 11.88 &HAPP-2 983 81 46 ~ 3

6 7

0.9239 1.2~99 0.7229 G.9

c<~ leO

168 1.2167 1~ 2

463 1.0559 1.0303 0.3250 1 e4

29G 0.9459 1.3228 CD 3

~ l

As transpires from the measurement results, the value

of ~f at all the P:-lZ-1000 reactors is presently at a lev~

of 1 , varying from 0.6 to 1.4 for particular states.Consideration of the experimental results for o + determine

a trend for its decrease as the number of loaded 2 ~ 4'i'i fuel

a sseqbl ies is incr ea sing. This decree sed ba sed on comp 1 t e

transfer of'he entire core to fuel with 2.4'~ enrichment isestimated:. at (0.5+0.2)3..

t

This conclusion, as well as the measurement results,agrees fairly well with the calculated cata. Table 2 repre-

::.ents calculated results fork/ and hot dehydration obtained~'ith the aid of the ST""PA~I programme. The reactor under con-

sideration operates under steady-state conditions without

fuel recharging.

Table 2

Calculated Data for cLg and Hot Dehydration

Cvocal co~a)C o~l~te,

Dehydra- Burn-up fractior

t ion ef- K'(.day/kg

feet.l. Fuel enrichment 2;o, without

add it ional absorber s, opera-

tion reserve 30 manual

control rods

2. Puel enrichment 2;o,

4.8 a,0 20.3

80 additional absor-

bers, operation re-

serve 45 manual

control rods

3. Fuel enrichment 2.4~<

80 additional absor-

1 1 0.7 14.0

bers, operation re-

serve 45 manual

0 2 0 ~ 4control rods 20.0

Thus, alongside with substantial reduction of 0(y end

hot dehydration at installation of 80 additional ebsorbers

-13-

and increase of operation reactivity r.serve of up to 45 ma-

nual control rods, a noticeable reduction (about 25;~) in

fuel burn-up fraction is also fou..d. Change-over to fuel er;richment of 2.47o will make it'ossible to return the burn-

up fraction to the pre-accident level and additinnally dec-

rease the hot dehydration effect.Additional study was carried ot t for an accident with

break-down of the circulating pump pressure header in the

PS'-1000 and PSMK-1500 reactors which revealed that with

operation reactivity reserve der.ressed to 30 manual control

rods, maximum value of positive reactivitv being inserted

at such an accident (in view of operation"of emergency pro-

tectlcn end fuel temperature cadet tve feedback) te at 0.)$ .This makes it possible to conclude that runout at instants-

neous neutrons is impossible for design accidents.

-14-

5. ON CHANGES IN SUB-CRITICALITY OP SHUT-DO'AN

REACTOR

Alongside with reduction of the hot dehydration effect,

the measures for increasing safety resulted in a decrease i.".

sub-criticality of the shut-down reactor in a cold cepoisoned

state. To discuss the reasons of reduction of sub-critica-

lity, we present the results of its calculations for previ-

ously considered variants (See Table 2) in different states

which occured in the reactor during the shut-down. The calcu-

lation results are given in Table 3.Table 3

Characteristics State 80 AA Ditto as Ditto as 3

before + + mo + 2.4;~ enrich-

acci- 45 MC dern. ment

dent rods rods

1.Subcrit icality

at stopping of

reactor. Puel 13.0

and graphite

t emp era tur e

11~ 5 10.3 9.9

284 C

2. Subcr it icalityafter depoisoning 8.0

3. Subcriticality"

after cooling to

70 C and protec-

tion rod arming 8.7

6.7

5.1

5.5

3-9

4 3

2.8

-15-

It is seen that sub-criticality in the cold depoisoned

state (See Item 3, Table 3) decreased primarily owing to

loading of 80 additional absorbers and increase of the opera-

tion reactivity reserve (sub-criticality decreased from Ss7

to 5.1J)). This is explained by the chard(e in reactivity

effects as regaras coolant density, fuel and graphite densi-

ty owing to installation of absorbers in the core.

Change-over to modernised rods has additionally decrea-

sed sub-criticality. This stems from the modified construc-

tion of tbe rods ss a result of which a water column about

c.5 m long was formed at the modernized rod under tbe absor-

ber. Due to this in the operating state the upper region ofthe core is overlapped with a grid of absorbers - water co-

lumns which resulted in lower burn-up fraction in tbe upper

part of tbe core.

The data given in Table 3 show that change-over to

fuel wit h c' 4,~ enrichment; a dd i t iona lly d ec r ea s e s s b-c r i t i-cality for about 1 J (from 3.9 to 2.8P ) ..his is li..ted viit

additional intensification of react ivi ty density e feet v he."

filling the controlled circulation circuit with water in

the system with submerged rods and additional absorbera. l;ot

that at the pre-accident state this effect was negative ana

'.:ss (calculations) 1II ~ As a result of installation of the

additional a bsorbers, it displaced to the posit ive region

. and is 1.5-2.0Jf . 1he reactor shut-dovm cooling effect also

increased

Let us consider one important physical effect which to

large extent determines. the observed reauction of sub-criti-

-16-

cality. Messu .ments of axial neutron distributions in experi-

ments on burn-up charges of unit 1-3 reactors at the Ch!'7P

in 1986-1987 (in cold state) revealed that neutron field fee-

tures a pronounced maximum at the upper part of the core. A

characteristic view of neutron distribution is shown in

".ig.8- This effect is explained primarily bv prevailing of

rods inserted from above, which decreases burn-up o" uel in

the core upper part. Besides, at power operation, even ifsymmetrical axial distribution is maintained against the

heightwise field detectors, we obtain distribut'ion of energy

release in height somewhat displaced downwards. This displace

ment stems from decreasing sections of division of fission

isotopes st movement from the lower to upper part of the

core because of hardening of neutron spectrum as coolant den-

sity drops (owing to boiling) in working channels.

As a result, the core top underburning occurs systema-

ticallv. In a cold depoisoned state the meximum coefficient c

working channel lattice multiplication in the upper region

of the core is 4-5~~ greater than in the symmetrical lower

reg ion.

Owing to the above-mentioned measures, the axial neutro~.

fi.eld at power level had maximum at the bottom for s prolonge

period of time. This increased the effect of underburning of

the core top and decreased sub-criticality.Reliability of calculating sub-criticality with the aid

of the reactor progrsmmes is supported by the data of Table

Compared here are some results of calculations and measureme."

-17ee

of sub-criticality at unit 2 reactor of Leningrad NPP in June,

1989'e~

Table 4

Sub-Criticality. 6J

State

l. After shut-down, cooling snd

Exp er iment Ca lculs t iona

depoisoning

2. After increasing additional absor-

hers from 80 to 100 pcs

3. After dewstering of control and

safety system coolinz circuit

2 ~ 8

4,2

3.3

2 '

30 7

3.0As is seen, calculations show somewhat lower sub-criti-

cality.Let us now consider behaviour of reactor sub-criticality

in t"ime during transient period in different states.Calculated dependences are shown in Pig.9. Apart from su".

criticality in a cold depoisoned state with the coolant circ"-i/lation circuits and control and safety system cooling circuit

filled with water,theatre

ere shown snb-crtticsltties et dewe-

tered control snd safety system circuit and in a hot state))h

with water in the circuit of the control snd safety system.

Meant by the hot state is the state when sll the core material

have a temperature of 284 C,while coolant density is 0.8 g/crr.

In all cases the emergency protection rods sre armed, other

rods are submersed.

The curves of Pig.9 manifest drastic drop of sub-critica-

lity near 1600 ef.days from the'oment of reactor start-up.

18-

This is explained by the fact that by this moment the safetyimproving measures were taken {installation of additionalabsorbers, increase of cperstion reactivity margin, cbange-

over to modernized rode and feed of fuel with 2.4$ enrich-

ment) .After a jump-like drop in sub-criticality at the moment

of 1600 ef.days, it decreases more for about 400 ef.days.This stems mainly from redistribution of axial burn-up offuel snd is likely to be dictated by the fact that vari calcu-lations it was assumed that rod distribution in time with

respect to insertion depth was constant. In practice, action'f operators for equalising the heightwise field and

maintaining the operation reactivity margin should smooth

out, this effect but lead, on the other hand, to a greaterdecrease in sub-criticality at the initial moment at itsjump-like drop.

It is seen that drastic drop of sub-criticality in allstates was accompanied by change in the ratio between sub-

criticality, values in different states. If before the

measures, sub-',;criticality noticeably decreased at heating,I'

fter the: measures were taken, its value rema ins sub stant ia-lly equal in sll states. As regards the state with dewaterec

circxfit of the control snd safety system, this is explained

by moderni2ation of its rods, mainly elongation of theirabsorbing part which eliminates the region at the core top

not overlapped with absorbers. Referring to the hot state,it may be stated that the main reason of changing ratios ofsub-criticality in cold and hot states is linked with instal

19-

lstion of 80 additional abaorbera in the core. Ae s result,

the heating effect, as well as the density effect of reacti-

vity have decreased. This statement correlates

we'll

with

the observed run of the aub-criticality curves in the tran-

sient period. At the initial charging, with s great number

of additional ebsorbere, on the contrary, aub-criticalityll

in the hot state is higher than in the cold one, as the add'-

tional absorbera are unloaded, their values approximate each

other snd than sub-criticality in the hot state becomes

lower.

- cO-

6 ~ ON THE POSSIBILITY OF INCREASING SUB-CRITICALITY

A radical method of increasing sub-criticality is tochange-over the PSMK-type reactors to the rods with displa-cer 7 m long. Table 5 compares neutron and physical characte-

ristics obtained under the STEPAN programme.

Table 5

Comparison of Characteristics in Systems with Displace.

4.5 m Lopez (Modernized Rode) and Displacer 7 m Lom

Characterist ic 2.4'A, moder- 2.4g FA,

n,ised rods, 80 rods with

7-m dis-

placer, BC

AA

22 ~ 11. Burn-up fra ct ion, 5Ã. day/kg 20 ~ 0,

2. Hot dewatering effect, 0.83."~~Emergency protection «etght, f 15'4. 'Sub-criticality in cold depoisoned

4state (protection rods armed), P 2.8 6e9J iy'ote,that change-over to rods with 7-m displacer

will'llow

to increase criticality for about 4p At th<) same time,

elimination of water columns in the control and safety sys-

0 '12.0

!j'ern

channels increases burn-up fraction for 5g. However,e

because of elimination of water columns, hot dewatering ef-

what the operation reactivity margin, or to install more thar

80 additional absorbers. At present reactor tests of rods

with 7-m dispel.acer are being completed.

feet will rise for 0.4p. Therefore, if changing over to rod:with 7-m displacer it may become necessary to increase some-

- 21-5 t

CONCLUSION/(1. The measurements performed and calculai;ions carried

out show that as a result of the measures, steam coefficientof reactivity was cor:siderab'y decreased. Taking into account

modernization of the control and safety system rods which

preclude insertion of positive reactivity, we may conclude

that the neutron-phvsical characteristics of the P~hK type

reactors presently in operation exclude the possibility of

the Chernobyl type accident. The possibi'lity of runaway at

instantaneous neutrons is also excluded at all design acci-dents with lost of coolant.

2. The measures decreased sub-criticality of cold depoi-

soned state. The measurements and calculations performed

allow the analysis of the reasons for reduction of sub-criti-

cality. It is shown that the main reason stems from installs-tion of 80 additional absorbers snd increase of reactivity

margin. Change-over to fuel with 2. 4;h enrichment is not the

mein reason of sub-criticality decrease.;:,-,

3. Measures have been elaborated and are being implemen-

ted to increase sub-criticality. Change-over to rods with 7-z

displacer will allow elimination of the sub-criticality prob-

lem of the P5NK-type reactors-

The Author+ are grateful to Lavrukhin V-S. for presen-

ted materials on measurements of steam coefficient of

reactivity, to Smirnova L.A. and Tikhonova 0 ~ S. for assistanc

in shaping the report.

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Fig.l. Manual Control Rode in'$4iMK Type Reactors. Schematic

(a) Rod of old construction

(b) Rod of old construction inserted into core to eliminate

possibility of insertion of positive reactivity(c) Modernised rod

1 - rod withdrawn

- rod submerged

3 - absorbing element

4 - graphite diaplacer

5 - water column

7 - (b)8- (c)Fig.2. Power Variation

1 - t, hours

Operation Margin:Variation (in rods, 1 rod = O.OG05

l - t, hours

.ig.4.:orm-Functions of Neutron Flux Density at Time I'ome.".t..

A, B, Cy D

l-Z, cm

2 - Top

3 - Bettom ((

Fig.5. Distribution of Graphite Temperatures at Time 7Joment D

l - Z> CIN

2 - Top

3 - B'ottom

-29-

RZF~MENCES

'!I: Information of accident't the Chernobyl NPP snd itsconsequences prepared for IAFA, Atomic Energy, 1986, v.61,issue 5, pp 301-320.

2. Adamov E.O., et al «Increase of Safety of NPP with

P„-NX-Type Reactors«, Atomic Energy, 1987, v.46, issue 4,

pp 219-226.

3. Halsall H.J ~ «A summary of IYIMS-D4 input options.AEEN-M 1327~ 1980

4. Adamov Z.O., et al «Analysis of First Phase of

Unit 4 Accident at Chernobyl NPP", Atomic Energy, 1988, v.

64, issue 1, pp 24-28.

-31-

Pig.6. Reactivity Inserted by Rods at Operation of Emergency

Protect ion,

l-t, s

Pig.7. Variation of Reactivity at Movement of Rods in Unit

4 Reactor of Chernobyl NPP by A3-5 Protection System

1 - Reactivity, & ef2 - Rods of old construction

3 - Modernised rods

4- Time, s

Pig.8. Axial Neutron Distribution in Critical State snd

at Low Sub-Criticality

1 - Thermal neutron flux density, rel.units2 - Critical state

3 - Natural function in sub-critical state

4 - Sub-critical distribution in problem with source

5 -. Core bottom

6 - Core top.

7 - (cm)

Pig.9. Variation of Reactor Sub-Criticality in Time before

and After Measuresffor Increasing Safety. Time Moment1

of 1600 ef.days Corresponds jo Accomplishment of

Measures

Sub-criticality in cold state

2 - Sub-criticality with control and safety system circuit

d.ewa t er ed

3 - Sub-criticality of reactor heated to operating tempers-

tures

4 — Sub-criticality, ef

The 1st Internat1onal Workshop on Severe Accidents 1n

Nuclear Power Plants

FEASIBILITY STUOY OF SAFETY OF URANIUM-GRAPHITE

~RSMk TYPE REACTORS

Z.0.Adamov

V.P.Vasilevsky ~A.A.Petrov

u.M.Cherkasi. v

Dagomys, Sotchy, USSR

P9 oc~ 3 mes. 1989

PEASIBILITY STUDY OP SAPETY

OP UBANIUhi-GRAPHITE

R35X-TYE E REACTORS

ALaaov ~~ O.; Vasilerskij V,Po, Petrol AeAo., Cherkashor Yuomo

Research aad Developaext Ixstitute of Power Engineering

USSR> bllosoow

W

Since 1962 in the USSR "General regulations on

safety of nuclear power plants in design, construction

and operati:in (GRS-82)" (1), issued instead of those of

1973, have been in effect+ The main regulative document

takes into account experience in construction of nuclear

power plants (NPP) with different type reactors and con-

tains arrangements and technical requirements to safegua-

Ling means, At present a new edition of "General regulations

on NPP safetyf ~ ~ is prepared for publication, which isbased on a"-lysis of causes and posteffects of the Cherno-

byl accident in April, 1986 and gives a oomprehensive view

of the world, approach to solving the nuclear safety problem

at present stage cf nuclear power development with a defini-.--..'./'elynegative attitide of some people to this energy source,

NPP is considered to satisfy the safety requirements „

when its radiation effect on personnel, population and

envirovwent under normal operation and design basis.acci-dents is not above the stipulated radiation doses and

permiss "ole contamination standards, hPP safety is ensured

due to consistent realization of defence in the depth

principle based on application of a barrier system protec-

ting from ionizir~ radiation and radioactive substance

spread+ According to regulations the design should- include

technical rear." end arrangements providing for safety at.any design '"avis i""'+ial event (accident) and simultane'-

JZ

ous failure of any safety system active element or passiveI'Q

element with mov'rg rechanical components~ Besides, undetec-

ted failures cz ler;.ents uncontrolled during HPP operation

1

but in luencing the accident evolutio> should be taken

account ofe

I'he USSR channel uranium-grat. hite reactors

This type reactors have been in operation in our coun-

try for more than 550 reactor-years. From the first world

APS reactor, Siberia, "-eloyarsk, 3ilibin reactors to

high-power RBE-type reactors - this is the evolution of

.c~~el uraniu.';-graphite reactors, At present 13 power units

with the R3hK-1000 reactors, a single-unit power being 1000 ]If,and 2 power units with the RHE-1500 reactors, a single-unit

power being 1500 k%, are in operation in our country. The',I<

total capacity of the P~-type NPPs constitutes 16 GI of

36 GÃ, installed capacity of NPPs in the USSR, Prom start-up

of the Leningrad-1 reactor in December, 1973 till the present

time gower units with t;-.e rBMK reactors produced 850 milliarcts

kW z h, The Smolensk-3 reactor is near to start up, under

constructior. is the i'.ursk-NPP unit-5 with the R3hK-1000 type

reactors, The Rrsli. nuclear gower reactor is a heterogeneous

c'nannel thermal-(neutron) reactor with graphite used as a

moderator and boiling light water being a coolant 1660 fuel

channels located in vertical holes of graphite columns repre-

sent a zirconium a loy .ube 80 mm in diameter, ~ide the

channel:here is a fuei. ~ sembly with 13.6 mm diameter rod-type

fuel elements in . i=cesium alloy cans, The heat circulation

circuit is typ'ai "" cne-loop NPPs with boiling reactors

(F'g,1) ~ Forced-c ""ulat=on circuit (FCC) consists of two0

parallel loops, eac.'. " sponsible or cooling oX half the reac-

:cr x'uel channels," The coolan~ circulation 's. realized v'a

.he main circulating electropumps (MCP). "he supply cf

underheated water and withdrawal of steam-and-water mixture

f om each cha.gael is carried, out via separate pipelines ~

In horizontal tvpe vessel separators the steam-water separa-

tion under 7 KPa takes place. The saturated vapour is direc«

ted to two turbines and after heating and deaeration itscondensate returns to separators, from which its mixture with

separated. saturated water is fed via MCP to reactor inlet,~fter the Chernobyl NPP unit-4 accident'he safety of

the R3hE-type eactors underwent a close and rigorous analysis.

Though the accident was caused. by'ractically unpredicted

violation of both operating order and conditions (2,3), a need

for measures increasing reliability and safety of operating

NPPs and those under ccnstruction with the REK type reactors,

excluding reactor runaway and failure under any violation

of service regula-.ior~, has 'hrisen. Such measures include,

first of all, decrease of steam void coefficient of reacti-

vity and ircres e c= sc"am system fast act'n (4) ~ At allc"e ating .'.=Ps wi+:: the 23K +ype reactors .echnicai solutio=a

awed at improvirg x.".e reactor physical characterist'cs are

realizeds operat.'ve react.'vity rmrgin is increased; reactor

ontrol system (3CB) fast response has increased two times;

aosorbing rod :e. -xh 's increased; scram operation automatic

con.=ol is real'd. :urrently, ciu n~e to reactor nake-up

with up to 2.4% -."." ch d gael is completed; provi-icn cf -"1

operating powe" 'its with 2»2 ' s fast-response scram syste=

=s carried out; =CC .-..etal state auton~tic control is under

:erelopmen ..

5

2. Frinc 'les and criteria or safe .y of t."e ."="-'.'. "eac .cr

All technical means for safety of R3hX reactor provide

for the following functions:

a reliable monitoring anf control of ener~ distribution

in a core;

core diagnoses for timely replacement of failed structural

elements;

automatic loss of reactor power or reactor shutdown in

case of erergency;

a safe core cooling in case of different equipment failures;emergency core cooling due to PCC pipeline, steam line and,

feed pipeline breaks;

safety of reactor structures at any initial events;

withdrawal of coolant exhausts in case oi pipiline breaks

to accident localization system (ALS) y

provision for equipr.ent c~intenance during reactor plant

operation and when liauidatmg the design basis accident

pos te ffects ~

The RZE ." a ~or af:y "easures provide for::..'' =h quality of

sa�fety

system designs, subsystems anti whole

reactor design;

high quality of fabrication, mounting and maintenance of

equipment and .. i =-' ..es;raiwterance o .; =...~l operation systems and safety systems

by means o. c.'".ecks, tests and due date replacement of equip-

ment;

reactor 0 era G:1 in c 'se compliance with service forms

- 6-''

a-d re"o..ds, re.'ulations and"technical ".an1'~hen

designing the first l;:-'Es with the PM~".-1COC reac-

tor a liat of ini"'al accidents was formulated ~d .heir

most favourable outcomes were ana'sed ~ Eroceeding

from cperational experience and the I~c'E safety requirements,

.hat have become more stringent concerning the world nuclear

power, original 1"'st o initial accidental events was much

enlarged. Zt was essent.ally corrected after analysxs of causes

and af teref=e ts o." the C.-.e nobyl accident. Zhe list of design

basis ini tia ~ccic.e..ts of HZ<X "eactors nowdays incl"cesi

about 50 sit".a ..s ...a. &y be classified according to 5 main

types:

;) "ea tivit„- c.'".ange s'.nations - in„'ection and spontaneous

wi thdrawai<;:of RC= -a=ety "ods, failures of RCS equipment,./

false switch-on of reactor emergency cooling system, RCS c>~»els

de watering;

2) "cre cooling system accidents - MCF de-energization, break

of flow-act vated valve = ate or K~CF valve gate failures of

rQter v 4 - ~U

r'CC --'-el'-=,, ~-:m line and feed pipeli.-.e break accidents;

4) equ'p en'ite.'=-cf or a"'lure accidents - de-energization

of auxiliaries, -. "';c .e.".e"ator (iG), opening of main safety

valves (i'SV) with = o owing unclose "e;.-") other acci e.-.-.s — -.der reactor c".arging machine fueling,

:i e) i 'oci..c'- =- . o

lysis of ..='=.".—.and aevelopment of safety means

="or +he PU~~'ac-.-r ':ased o" the following design basis

s8~ e y cri eria:0

-) "~x=-.. " ="ed'ble accident (~.C i) is t.".c =';-. a ~z'.-."-.ter p'peline break with f ee bil~tera coolant discharge under

the reactor rated power;

b) the f.r -. desi~ lait to fuel el nent fai'ure u..der ."ornal

operat-'on, defining the coolant activity ascer:aired level allows:

1g of fuel elerents xith gas likable and 0,1$ of fuel elements

with direct con-.ac ~ of coolant and fuel;

c) the 2nd d s =n 1~it to fuel element failure under circulat'onci"c'. pipeline break stipulates:

- f e~ can teL;. ="a-. "e 's

local "-e .;:.cf:''2z "ar

no . vore than 1200 C

ozidaticni, 's z.:ot over 185 o= iri.ialwall thickness;

- reacted irconiu= ="action - rot nore .han 1g of fuel can ress='n t-e c.'~.""eis of one discharge ~roup header (about 40 charnels),

d) prevision c.'eactor core aZoading and fuel c~~el withdra-

oval af.er a".: acci= ".-. i.=eluding 1'DA.

:-=" "=a ='-.ion o all above-r;entionec safety reouir L..e..ts

the M":. re~c--cr " ~.".-. is 'uip.-..ed with the follovrirg control

/"=-'-.or con-.rc =""-::--..(.~C~), whi .; ':e ide traditional Q ~vlr o ~

incl =es -ca au:o.-..'.-.= r=„-u ators, ocal er e agency protecticr.

=rol'ps anc. as t s a

ene r t on;..o.";itor™-and control s„.sce with col: iJ uous cise :..=".-.="ir.:- oi cue '--„, " l=--e ever the core rad'

and hei-..t;— reactor e;.,er;=-™„.: -:.".= s„..st; (:-C3) for heat re~ection

fr cM co caus d "„pipeline break or

e cuizr.;en-. v 4 ) A gf toreutic =.'vi-.c";on according to 5

-8-

.=e.".'ent l=.orithr~ a "-1'ed .o definite::,~e acc'-e::-.s;- one-channel and group monitoring of "el :an leakage

fuel channels;

- fuel channel and PCS channel intact separate monitoring;

- monitoring and control of coolant flow rate in fuel c~~~els

and RCS channels;

- reactor ne-.alwor~ and graphite moderator temperature monitoring;

- a centralized monitoring system providing for collection and

control of input analog and discrete data, calculation of safe

process parameters and signalling cn controlled and desi~parameter deviations from the specified values;

- a system of stean-gas mixture relief from reactor under

emer=ency creak of reactor channels;

- reactor charging machine providing for recharging of spent

or leaking fuel assemblies without reactor shutdown;

- a brcken p'lire .ransport limitation system;

— compartments with large diameter "irculation circui') pipelines

and "evices of etc~-gas mixture relief to ~S;- accident local'tion system for coolant emergency exhaust,

s~eam condensa-.d;.-. a~a ".old-up of nonconc"nsing radioactive

-ases;

- redundanc„'z'quipment and electric devices with automatic

s-'.and-b„. chan=e-o; er ~

3. Desi:-n basis reactor safety analysis

The R=4E''acili.-„: plant desi~ includes feasibili.y study

of r actor afet„" at all above-me."tioned. initial accidents,

w~l„:si=- cf ec cn acc =en'ompri es study of dynamics of occ 'r-':=:

.".eutron-pi'ysicsl ana. thermal-hydraul'rocesses, "ete ri wticn

of reactor total parameters, study of the core state and the

most thermal-stressed channels, the ci.olant exhaust time and

inte~ral values etc+ Zor this purpose a complex of methods and

programs for calculation of HPP transient and emergency regimes

is developed, and a series of experiments on test facility stands

and reactor loops i carried out+ S-ome emergency conditions

that could not be representatively simulated on test facilitystands and reactor loops were additionally investigated, on a

full-scale at the firs: "ower units with the HK-1000 reactor

( or example, e"er'.enc„'e-energization of aumliaries and

change to the coolant natural circulation, determination of ECS

eal character.'sti".s e-.c. ) ~

On the basis ofcalculation-experimental investigations

accident symptoms are defined., scram system trip signals are

selected, o"eration algorithms and safety system performances

are optimized e.c. The accident alternative scheme with the

nost severe poste=fecta is .avwlysed,

1) Cf all initi~l cocci='e .ts resultir.-„ in reactivitvc".~r-e he most ='.,-ere'; =or the RZL"; reac.ors is ener-ency

stopping of cooling wa .er feed to RCS channels due to inopera-I)bility of the puny .'..a= exchange plant or break of a pipeline,

. /feeding wate to 4h=. c'."~nnels, RCS cooling circuit and. scram

system tli.". c-.l„-ori-..="..re desi. 1ed so th=-t at any cornbinatio:;

of cause=- of .=CS c':-::.-.' .";;aterin~ i. "et ecran sy-tern tripsignal is prod~ cec: . -..".er decrease of wa er low rate in the

RCS channels or pr =a;" c "op in the c.'"cuit disct~r~e header

o" water level Cec='e~ " .". the up er energency voter tanlc,

-10-

.<11 RCS "ods are inserted ='to :; co"e wi-.hin 10 , an- .-..channels remain filled with water for not less that 1>5 r: n

as water from the upper par. of the channel is drained o fto the lower circulation tank passi~ through the core,

Reactivity effect during dewatering of RCB channels of shut-

down reactor 's negative and constitutes -(0+03 0+05);os

3ut for reactor safety analysis at the given initialaccident calculation-experimental study of channel tube and

absorption rod -.em erature conditions and determination of

their servise-bility are necessary< Though in this case reactor

is automatically scrammed by emergency protection system,

within the first 1.5-2 min of the accident RCS channels are

vr«.ter cooled and absorber "'ooled down from 340 to 180 C,

after water drain the channe's are filled"-'vrith air and their

heating-up from graphite moderato" with initial temperature of

580 C takes place. Figure 2 shows .emperature variation in time

of graphite block, channel -"irconium tube, boron carbide absorber

and 'ts allumin''~. al'„'an in the central cross section of-.'.e R35Z-1CGG reac.or c.'"~~el~ ~fter 6,-7 m-'n the channel is hea-

'1

-.ed up to maximum ';:" .""tur of 47G C, and -.hen slow y cco"s

down trandferring '.".eat firs. to atsorbtion rcd. and. then to

„=.rephite> The absorption rod temperature rises more slowlv and

after 60 min attains a maximum of 350 C, ax~er that all RCS

cannel, el ments ~."= ;co ed for several hours up to the tem-

perat "e cf ""CC "".::="~ -.e.-.ch tests cf absorp .ion cans, -.he

rod. bayonet con.-.e..io.-.s, c.~a">el .ube in the design temperat 're

;.;ode ard simu at'".=- =.;ll-o ale stresses s. owed thei" servicea-

01Jikg y' ~ .":der -: =".e.-.:ydewateri.".g of RCS channe =- the

sa:ety'" -..;= i~~~. reactor is provi=ed fo" wit;;c=. a:;,.

pecial RCS channel emergency coolin- system,

In case of emergency sw'tch-off o one KCP the transient

is the most ur avourable ~hen: the flo -activated 'alve atthe head of svrLtched-off pump is unclosed (see Pig. 1)~ Itis due to circulation .loop that causes additional decrease

of coolant 'w rate through the chwmels of the given r ac.orhalf and in""ease of operating MCP apacity, thus dec easing

.heir r~r-in to avitstion. In response to a sigil of MCP

flow rate deer ase ..= tc slam setting of 50GO m /h or in

response to a i~~1 of hiCP motor switch-off tne reactor

emergency -"-tecticn s„"stem is cozen nded to decrease the

"eactor power u tc 60,"o'f its rated capacity'i Fig.3 shows

;ariation of the ~Mi reactor para eters under MCP svritch-off

and unclosure of its flow-activated valve, Investigations

showed t."~t the coo'=nt flow rate through ~ reactor channels

wit™f- ile" I'."..= '=-.=" 6 s decreses up to 749 and t.".en due

"'ecrease of core re~is+ance and r actor power slowly increases

C~.—..'.c .. c= e~:.'-". of .wo operatin"„ KC= "'„ eases

-r:-.,r.. COOC .-.c

'su"'::ed in ~y

c6 ' 67CC ~ /h of .."e coo).=.".:

.:="o:-'.its'-.c'sed f'w-s ctivated

valve escapir~ -.he "o"e, he pressure re~a'a+or affect the

re~ul .=;- valv==- o=- 'czh -'.'™bies reduc"." the stem rate ~

-='"~-ses after 56 s up to u6 k (=,'j..-.

te" 1G5 ",p ~oeinal g, 1ueo ' -erewl v 1 l I ar"

c"vi".ation

4 l 0f1

I c 3'DCL -!e 1eH-i'". /)Posit':vs injap

e ~ cela.~.v

= . case o. switchirg-off c:.'ne .',.'CP w"'th s'":.ul.=-necus

failure of one ..ormal operation lement (in our case it isthe J,.CP flow-activated. valve) ali reactor parameters vary

within the perm'ssible linitso

3) NPP safety to a greot extent depends on reliable opera»

tion of water feed to separators+ The most dangerous situation

arises in c-se of momentary stopping of water feeding, for

'nstance, under breakdown of all feed electropumps (PEP)

resulting from eed pipe break or pressure drop in deaerators,

In this case accord"'r'„,. to a signal of feed water flow ra+e

decrease below 50$ from the rated value reactor emergency shut-

down comes "'nto action. Behaviour of the main parameters under

stopping of water feec =g in the RZiZ-'I000 reactor is shown

in P g. 4 ~ A s.„stem o "utomatic accelerated. discharging unload"

both TG up to no-load preventing from considerable pressure

drop in a circuit - pressure in separators is not below 66 kg/cm ,2

Laxmum level deviation in separators is not over 440. mm of

-.he rated valve. Autc:-.mtic closure of all control valves

in "esponse to emergency protection si~~wl urloads operating

IiCP up to 65CG m /n flow rate, The transient process is not

followed by steaz. ~eneration during hCP suction i,e+ their

operation without cavitation is provided for, Reactor make-up

is carried out via e.-.. rgency feed. electric pumps (EFEP) or

via ZCS pumps feed='=.„- water from ALB pond., All these pumps

a"e utomatically -ut '~.o operation in response to signals

of feed water f'w rate decrease or pressure d op in PEP

pressure header.

4) In case of Ce-~nergization of 'TPP auxili"ries reactor

-13-

e'. r'"e .cy sh "tc.ow.. comes '.-.to acti on ~ '".e core ecol=' pll thin

the first 30 35 s is pe formed due to orced circulation of

the coolant by running"out KCr and further due to it::natu"al circulation ir. PCC+ In response to emergency p otec-

tion signal diesel-generators are switched.-on, that in 1-2

minutes supply electric power for the loads necessary for

reactor emer„-.ency coolir~ (~EP, RCS pumps, "-CS, ALS pump

heat exchan„- sys .em, s-fety system reliable supply transfor-

mers etc+) ~ In the f.'rst seconds of .he transient a certain

elevation c= hea-. power dip curve i" comparison with the

coolant flow rate dip curve is seen, but later on heat gower

fall advances .:.e flow ."ate decrease and in a minute they

stabilize at 8 and 206 respec'tively of rated levels, DZ3

in the most the~xi-stressed channels is positive for the wholelj

per" od. For substantiation of reliability of core cooling

in natural convection regimes a set of experiments on circu-

lation c'rcuit test facility stands ivith a full scale fuel

c.'~nnel representation and directly on the Leningrad and Kursk

II"-:- powe u.-.i . wit" -.he KB'-1000 reactors in carr'ed out+

Sa et'f reac.or -'.". tdo;;-.. coolin™in r~tural circulation egime

is ver'fied ~der e=er„=ency conditions, when all LCP acre

switched-off =-"- close to rated power levels, that took place

at ope ating ."l.Z

") Ii'.axir..um cre5 ble acci«ent (I:;CA) of :~:-~ with the i~~: reac-

toro con=-=''=-.".:-: a ':.„.;:c.hetial ...or:".en-."r-„ break o l:Ci

pressure ':=..;c. " w.'t.. t'". outer diar.".eter 1940 mm and wall

thickness cf ",0 r=.. (=ee "-='=~ 1)~ L-inly this accident h s the

aost can-ere~ effect o- the reactor core, "-s it is followed

14-

by momentary stoppinp oi cool~n-. feed -.c c.-.e half oi reac-.crc.'"~".els independing on E.C:- operating conditions, o e..elude

the reactor meltdown ZCS automatic switch-on by a special

algorithm is provided for.LCA is characterised by a short-time stopping of coi)lant

.'f

feed to the channels of accident area, that after 1-2 s resultsin dewatering of core re„".ion, Studv of reactor neutron- physi-

(

cal characteristics showed that in case of I'CA,standard RCS

provides for neutron shutdown, Due to positive steam void

coefficien- of reactivity, measured by experimental method, of

numerical calculation and under the most close tolerances

being not more than + 1.7A f, the reactor neutron power atg-ef'nitialstage increases - spike of its integral average

values occurs after 0,8 s and is 22g at reactor rated, power,

Beside pressure r.'se signal, emergency signals reflectingthe accident owe and "ate and, actuating scram rod system

are p oduced .n a compartment, As a result after ~ 0,7 s

RCS safe .y rods are in„-'ec .ed into the core, The reactor power

passing a maxim~ decreases and in 2 s .-fter 4'CA constitute

50$ of the rated;alue ~ S;v. tch-on of the "-CS .-.'ain subsystem

for acc dent are~ is carri d out by opening of quick-activev"-ives in re"ponse .o coinsidence of' signal of excess pres-sur rise up to 0.02 1s-(f)/cm .'n a compartment (identifie2;

reat coolant leakage) and any of ".i~o signals: decrease of

separa.or levels or dec ease of pr ssure drop between b~CF

pre sure header anc separators.up to the specified settings

(iaentify s lec.ion of accident area), akin- 'to account

delay to penor" t on o= ECS switch-on signal and .he time to

15-

valve .-.ate quick ac "'on,res;oration of;.Ic.-:..er feed .cthe accident area channels will take place within 3 s of

the KCA initi-tion and ir. another 2 s ECS flow rate ,will

a.tain the reauired value (approximately 20% of rated val e) ~

The main ECS subsystems for reactor water feeding from pres-

surized ballo s of from standard "EP provide. for emergency

- cooling wi.hin 2 minutes, Owing this period, assuming

simultaneous .Iith !:C~. de-energization of the unit auxiliaries,

ECS long-time coolin- subsystem based on water feeding to

both reactor area3 fro™ECS and ZZZ:" pumps connected with

reliable power supply system, comes into action. uuring de-

watering of accident area channels an abrupt decrease of heat

.ransfer irte'si.„. i- observed, t'h e over.''eat of wall layer -nd

forrraticn of steam f"'lm take place, heat-transfer coefficient

alls up to tens (..V/m ~ C) ~ Past response of scram system and

post OFZ hea: —.ransfer "re followed by 8 c (ick heat-up of fuel

lements. ;rom the be„-.innin™ of ECS water f eding a front o'

1 coo'.".= firs-. ':„: s ~per..ea,ted. steam and then by steam-and-

water mixture arises at channel inlet. The ate of fuel can

-:. mpera .ure grow-.h s slowed-doer.:, irrigation fronts occur

ax inlet : rid, whi h is ollowed by nucleate boiling and .convec ..'

hea ~ transfer to ;.rater, The fuel can temperature abruptly

falls up to preacc'dent values, ~o determine .heat transfer

coefficients '.d .-..e of fuel cooling, ez,eriments on 4600 k"i

"'l---.ca"e fuel .'.'. 'od 1 with electric heating -,ere

carried out, Res. l-.s of e, eriments were used i" developin~

rethods and pro==-= ."or rcrstationar„ temperature fields

- 16-

calcu"a .ion across radius and height of fuel elements .vade

MCA» ."i*~ 5 shows variation of fuel can ter,"erature distribu-

tion in 3000 k% channel of R3hK reactor w'h m" ximur.. ener~release at the bottom and neutron power local spike of 1 ~ 35 ~

Imximum temperature is attained after 5 s at a distance of

1.5 m from the chanrel inlet and const tutes 620 CD i.e.maximum temperature is much below the limiting value of

1200'D'Jnclosure of flo;v-activated valve in one DGH under

maxi'~ credible a c 'ent is considered as an independent

failure of one safety components In 40 charnels connected with

this DGH rev rsal of circulation v~thcut interruption ofI'uelfeed'ng takes "."ace. Channels are cooled under post. DIG, and

srLthin 4s of sep.-ra.or dump change to steam cooling regMe»

The steam cooling intensity is defined mainly by degree

of the multipurpose va ve open'n- at .i e c'~el inlet and

steam pressure in separa,tors» Mte p"essure drop up to 1 h:Pa

stean cooling cecor. s inc "f'ie"t, .;-.er io a f'l heat-up

and:".eat is removed zainl'... rad'at.:o." .o .he chaw~el tube

and then through -raphite to graph't blocks of earby cha "elscooled "y L'CS ~ If in circ lation reverse,t'..e fuel .can tempe a-ture a "ter 5 ~ 5 s f-om .he accident attains a maximum of 740 C

then ratner high ten"„e"azure of reactor moderator, that is550-600 C, is the —.sin cause of fuel overheat in the second

V

phase o=" a proces p oc d'n or several hours, when jointeffect o~ heat re .ova .o cooli=.„" -"aph'te a d decrease ofsi u+-down p owe cause a ' cf fuel —.e~perature» F"el cans

17

of =ost thermal-stressed channels ccnnec.ed with DGH '-~vin-

u"closed flow-activated valve attain maximum temperature

Qf 1 CGC-1 40 C af ter approximate ly 25 mi."., a ter which

thei'r ter perat re dec"eases fo" several hours, Though the

second design lim' .o fuel element faulure .'s not over,

- break of fuel can in DGH channels is expected+ MS makes

"'t possible .o locali"e radioactive products, prevent.'ng f om

"eir di"charge outs" de the +LB areao

Tellus ) all reac-'or safety increase measures ensure due

sa.ety level cr relatively small number of cperatinc .".,7'Zs

desi„-ned in the 6C-7Gth. 3"t further development of channel-

type " actor is a='-' --'igh safety reactor design, : spe-

cial sport (5 j 'c.".cer.ed with possible development of

a new hi:-h sa ety c'-~nrel-type eactor that could satisfy

the „-resent requiremen.s. Despite the fact that the most

"-;.e"e '..c"ice.-. -.oo'. -lace at the channel-type reactor,\

c;ver u.".it, c~~n;.el-t.„jpe "eactors having their specifich:eai''res

~conside"ed -.rcmisin„- i=..".'.;clear power

General re~wQ.ations on sa ety of nuclear porr " plan",s

in desi~, const ~ction ard operation (GRS-82).

L:oscow, " Wergoatomi"dat ", 19G4o

2 ~ ".he Chemo"yl accident and it"- posteffects, information

for I-'Za ~ atomic ener y, 1 986, Vi 61, N5, pe 301 ~

(

3 ~ .'.d~.ov "-~ C., Vasil vski~ V.F. et al. Analysis of the

LU11t-4 ~ Atomic energy,

=irst p'=--e of accident evolve.tion a ~ the C!:erno'oyl i%i.F

1988, Vo64, l:1, po24.

4. ~d~;:ov =.C ~ , Vasi evski<

of N.=Ps with .-.~Z-type

V.62, N4, p.219O

VePo et al, Increase of safety

reactor. Atomic energy, 1987,

h ~r'1 scLf et v

p ~ .;iM~ov 7~ 0 ~ ) :-:u,=.netacv S.F. et alo Development of .,

.r-n-u=-graphite eh~el-type reactor.I A--uc~ clu~o ia, co.=e ence on a, ety cf the

=e *erotic G: pcwer reac .c =- ) Si> 1, T.'SA, 1-5 L:av

1 988 ~

I ~ 4)'a v «uv cl w O 4« ~ ~ Ilk ~ov v j)e

%he r w Ovl ~

2 - ".~el o'.a."~".el;

3 - water p'pe (YlP),

4 - stea.=-water p'~e (S'i7:-);

e=arator;

6 dowL "; j.e,'

- uctioa reader;I ..«

was s ~«v t

10 areas re .'.cade

"g1 ~- - ~ '" Can q 1 ~ -~

",2

s .ea- - e ader;=+-p- C11~=v «W.

~L"- (~ccide..—u1iJ ~CL tion system);

17 3 « a>ve);

r o;e.-.e'a-.;".;

20

21

r ~,~ WW nr-- v-- ~ .0 .Q~ ~=~ )

SSS (=-=.. " —uo"-s:=

"oiride''" ue

. "err;ea t =-");u

22

J3

-20-

el ~ *4»»ve»ce4' <4ictQ

28 - SSS conden ate pump

2P - level re.~""tor

30 - r;."S;.ate acc;z'ulator

31-32

~CS pump

CICaJ l 6CLLJer

33- FCS quic'.:-~ ctin„- valve

34 - no""'e .-....=s flem ate limiter

"'.2, ;,.=e" .;:""varia-.ion of the RCS channel components

in -.';-. ~=',3'.-1GCO reactor u.".der emergency derrstering:

1 jra» ".its oc:.)

2 - B,CS ch=.".e'ube;

3 - the red c ed;

4 - abso"her

T.»eq

r r~ .-..-...i.-i c . "..'.;:. '.~ r&z ".":I';". ~. ".de 3vp ch-off 0

one c: -.' ;,=":;.izg !..C: in a ci"culation c'rcuit. ".-'.c".."e cf its flo;r-activated valve:

1 - neutron

2 - heat

9epc "etc'"

~ . - 4

- 21

Zlg ~ 4 ~ -erareters o= the K~K-",OGO reactor: zde; stopping

of water feed-'.~:

- neutron capacity

2 - ".herbal c pacitp'

- coolant flow rate

4 - =ress re in separators

5 - remiss level in separators

5 ~ Z='.-..= ; ar='ation of f el can terperature distribution

" o"- .he 300C ~% fuel cnann 1 of ZEKX-1OOG reacto

under ~~'-.:a credible accidents

PMc. I lIpMmpnMazbHaR cxewa AK c peaaTopoM PBE:I - geaKTop; 2 - TorumsmN zaHax; 3 - Tpyda BK; 4 - vpy6a fiB)

5 - cenapaTop; 6 - onyczsaa Tpyda; 7 - scacsee~N zomemTop8 - I'Lgi; 9 - 6aNnac; IO - HanopmN Eoszerrop; II - 3PK;I2 —PI'K; I3 - naposoN zomenop; I4 - napocdpocHQN lUIanaH;

I5 - cMcTeaa ac~aaMsacgM asapzl; I6 - sanac so~ CAOP;

I7 - perymTop ~sze~wa; I8 - TypdoreHepavop; I9 - Eo~eHca-Top; 2O - cenapaTop-naponeperpesaTats (CGQ);2I - Mo~eHcaTmN Hacoc; 22 - nogorpesavees; 23 - geaspaTop:24 — AII3H; 25 - IBH; 26 - cMamsamayN nogovpeNLvems;27 - Ko~eHcavoc6opHMx; 28 - zomeHcamadt Hacoe CQO;29 - pergaaTop yposHR; 30 - rzppoaxmyxympyjo~N yean CAOP;

3I - Hacoc ".ACR; 32 - MomeKTop CAOP; 33 - 6scTpogeNcTsym~NMzanaH CAGP; .;4 — orpaHMvMTeab Teus.

'3

0 I 2 3 4 5 6 7 8 20 30 40 50 i'0 70 80 90 lu0 OPesuW

11~MeHe~um ToMDepaTypz e~ieMeHTos zaHana CF3 pemTopa PM<-IUOU ape aaaputdHoM ode3BQIBa)61L1:

l — rpx,'>aroawN duroc; ' Tpyda zaHaza Cy3; 3 —Hapyzsaa odo~oma cTepma; 4 —Marepman uor~oTlfTOJIH.

0,8I

0,6

0,4I

0

P,KI'C/CM

70

68

-G, I !~u 60 80 I00 Bpeva, c

Pzc.3...ap2.'reT ~; ..:.—.X B peale QTKJIRReHER owoI'0 H3 Tpexa6OT~~~-'.x .."-.B

—rpKyzrauzoxsoN rreTze z He3aKpmzz ero

3 '5paT o L

I — .~e."-.=".:..:..;~c~mocTs;2 - TerrzoBaa:sottrHOCTs; 3 —pacxc"-.err.ic-. c;t-.e; ~; '* —zaanesae B cerrapaTopax; 5 —MaccomH

BGHL B ce a aTo~ax,

G-.H.ez.

i~ ~

G,B

0,6

(G

'30

I

80 IOQ 3pewa, c

~c.'. ".apa:;e-.=~ .=.:.'.-GOD .-cpm ~pezpameHHz rrogam rzTaTemHolRoke+

I — HeXTpc~.-:~ ..".m"..-.r; 2 —Terrace vomHocT>; 3 —pacxop, Ter.:c-HoczTeza; 4 —zeazerze s cerrapaTopax; 5 —vaccoszR yposeHs scerrapaTopaz.

700

600

500

400

300~WC

2000 I . 2 3 4 5 '- 6

PMc.5. NsMBHeaze PacaPegeaIBHHR TeicIIBPRT~H 060JIQQKII IIQ JIJIHHB TBBJI KBHBJIR MQBlHocTbs

3000 KaT Bo'peMem npII MIIA peQKTopa Pim-rnite.

THEORETICAL ANALYSIS AND NUMERICAL MODELLING OF

HEAT TRANSFER AND FUEL MIGRATION IN UNDERLYING SOILS AND

CONSTRUCTIVE ~>~VPS OF NUCLEAR PLANTS DURING AN ACCIDENT

R~>4SE FROM THE CORZ.

R.V.ARUTUN JAN, L.A.BOLSHOV, V.V.VITUKOV ', V.M. GOI OVIZNIN,

A.M. D YKEI9E,V.P .KISZLZV, S .V.KLZMENTOVA, I.Z.KRAYUSHKIN,

A. V.MOSKOVCHENKO, V.D.PISMEKII, A.G.POPKOV, S.Y.CEERNOV,

V.V.CHUDANOV,O.V.KHORUZHII and A.I.YUDIN.

FIAZ, MOSCOW, USSR

ABSTRACT

I'igrationof fuel fragments and core fission productsduring severe accidents on nuclear plants is studiedanalytically and numerically. The problems of heat transferand. migration of volume heat sources in constructionmaterials and. underlying soils are considered.

1NTRODUCTION

A study of probable consequences of severe accdents onnuclear plants with core meltdown and. release of fissionproducts into the enviroment is carried out in the frameworkof goverment nuclear safety program . A review .of works madein this direction in the USA was elaborated. by the specialgroup of Al?S[1j.

The. problem of migration heat sources in the enviromentis very important because there always ezicts possibilityspreading fuel fragments dur~ the accident. Heating ofsignificant amount of fuel resulted from residualradioactivity can be large enough to destroy surroundingmaterials by melti~ , evaporation , chemical reactions etc.Damage of containment constructions and. underlying soils can'ead to fuel migration upwards, downwards and. in sides. Thespeed of migration and its appearance depend. on ini'.ial fuel

distribution, chemicalcomposition of the enviroment,constuction firmness, thermophysical properties ofmaterials, correlation of fuel and. enviroment densities,conditions forconvection and. radiation heat transfer etc.The main part of these data is unknown during accidentdestruction of the reactor. Lacking full informationsimulation of heat and. fuel transfer in the unit is of aprobabilishe nature.?nvestigation of low probability eventsand elaboration of reliable methotis to :ease the fuelmigration is necessary because of catastrophic consequences«f accidents.

The problem of heat transfer and. fuel migration insituations involved. isessentialy nonnlinear and.multidimensional, that's why an approximate solution can lead.to wrong conclusions. So acomplete numerical experiment isneeded, induding both analiti:al and. ".omputational effortsand based. on mathematical models deduced. from the basicphysi:al principles and. free from restrictivesimplifications.

The present work includes a description of somemathematical models of radionuclide migration in soils,developed. in the Kurchatov Institute of A.tomic Energy in1986, as well as some physical results, derived. from thesemod.els ~

INSTABILITY OF PLANZ FUEL LAYER IN UNDZRLYING GROUND.

1. The problem of migration of fuel fragments throughunderlying ground. is an actual physi:al problem. Consider amodel situation, where the fuel fragments forma plane layer.The layer pulling is a result of heating and. melting ofunderlying ground. with following filtration: of lightermelted ground. through the heavier fuel.

Assume that the fuel layer with thi:kness H and. volumeheat generation power q pulls down at a cons~ant rate VOT1".e fuel, the underlying ground. and. the melt occupy regions0<z".H, z<0 and. z>H, correspondingly,. in the <."oordinate systemmoving with the layer. Equations of heat conduction are asollows:

1 ~111':,VT=xT+a,

V,T =~,T

K)H,

0%. g <.<.

Zc0

where the prime denotes z-derivative, y , <. —. arethermal1 g 11. d.'f usivity; R — the thermal condu<."tivity of the,

fuel.'e

temperature and. heat-:onduction fluxes are :ontinuous-:.'oitgh the layer boundaries z=0 ancL z=H. The bottom layerb<:undary <."oin<.ides with meltirg isotherm . T=Tm. The mel-.~~>'at flux is neglect'ed because of small parameter Am .'<.', Tm'where )m- is the volume meltin~ heat of the ~round„ ) . System"

h'ave been easily integrated, and. th~ rate VO 's

determined. byc V H c272T

j:-exp z =~

(2)

Determine a simpler expressions for the pull down rate in two

limit oases.The first one corresponds to a sufficiently thin fuel

layer pulling down at a small rate. In . this case one canneglected. convection heat transfer due to melt filtration incomparison with layer heat conduction. It follow then

qH

oC<'J.'he

obtqjged. expression is valid. in approximation H

(3!2T /g) . Note, that characteristic time of stabilization2 m

of the rate of layer pull down is larger than the heat-Liffusion time for a scale of the order of the layerthickness H. The maximum temperature of the layer is veryclose to the grounti melting temperature Tm.

In a realistic situation the temperature of fuel melt ishigher than the temperature of ground melt, therefore thefuel is a solid. in a sufficient thin lay.In another limit case, the fuel layer is thick and. has pulldow ratesIf

then

H » 2 mT

x q

2T '

In this case the maximum layer temperaturethe ground. melt temperature. Thus the fuel:onsidered. as a solid., if

is higher, thancan be believed

/z T2 m

2. Iet's investigate the stability of lower boundary Z=(of thepulling down. fuel layer with compactivity ~ (in theundisturbed case (=0). Inhomogeneous disturbances are<:onsidered. as large-scale in comparison with the layo'rthickness. The instability mechanism is ".onnecte<i with amodulation of the layer tickness caused. by " <iisplacement offuel. filling relief depression of the underlaing ground..

This result feeding back the inhomogeneous mel<ting and.futher modulation of ground. r lief. In this dynamic model we

belive that at the initial. stage of the instability a narrowlayer of the fuel near the lower boundary, with a thickness

of the order of the relief modulation amplitude takes part inthe displacement. Such a case can be realized, when the fuellayer ad.nascent .to the underlying ground. is less compact, sothat fragments have higher mobility. Thus, the underlyingground. relief is a slip surface. Puel fragments in the depthof the layer are not displaced, because of sufficiently largeinternal friction. Por this reason the upper boundary of thelayer remains horizontaly. Thus, the temperature perturbationis caused inhomogeneous melt filtration and does not depend,of fuel displacement, staying near the melt isoterm.Heat-conduction equation for ingomogeneous temperature areas follows:

6T +V ~(6T ) = y b,(6T ), z>H,

C26TR~CRV ~az(M2) (I a)ei(T1~2~(MR) . 0<z<H (3)

6T3+Vo~z 6T3) = Xih.(6T3

A,t the lower boundary Z=(: ....,z<0.

6TR+(TR=O,

6T~w(T~=Q,I

3.'6T+INTR ) —w, (6T~ ( ~ ) = —X

%here the two first expressions reflect the fact that thelower boundary of the layer coinsides with the meltisoterm.The last expression describes continuity ofheat-conduction fluxes with respect to the melt.

At the upper boundary Z=H the perturbation of tempera'tureand heat- conduction fluxes are continuous.Solution $p equation (3) is searched. in the form

6T"'I (Z ) e ', where. f (z ) vanishes at inf inity. Inthegeneral case this results a rather cumbersome dispersionexpression. Consider tvo particular :ases of practi".alintrest.

1 . ) a81~2~'1

thenz 'i'

8 1'I&)y(0) = A—,

H.

vrhere A is the solution of the following equation

m X 2~—A + (3-s)A+ (E-s)e -2+ 8 =0-A

H2

i~Jote, if H<(Xkn/q) (very thin layer), then p g/Am.If

qH2

Zlm

thenV H0

~ p X

O

t,(0)=5-= 2X2

H

vrhere P is the solution of equation

"t(0) =(3-a)cT

m

V1ith an increase of parameter kH the incremen. p decreasesand if kH=0.5, then p equals to zero.2. ) c /C « 1. Volume heat capacity of the ground. is

1 2essentially jess than the volume heat-capacity of the fuel.lf k = 0 ,H «H , then

th(p) +R2 1

——= 0pIf X2>>X,, then

y=—(1-Sk H ).X1 22H

3. Investigation of instability of the lovrer boundaryapullizg down fuel layer demonstrates that the perturbationvrith k=0 has the maximum increment at fixed thiclmess of .he<ayer. The obtained. dependence of the increment on .he layerthickness,>2H :anbe easily interpreted.. At the region~< (Xhm/g ):. modulationof the lower boundary vrith, ampli;;.jde

."auses additional surface heat source vrith intensity '"'i'

Th's heat source gives ri e to izhomogeneous melt'ng Pm andheat conduction flux

/7x —, vT',

H~ 0,I Xk.m< GH

-alan<.e equat

7then '7T= „—' ~ 0, 'cause ~ ~ <./i<.m.

<ll

then the in<."rement is determined. fromQTm,ion

q$ = 3. y/X vt(.This gives

x J)CO

H((

Note, that in this range of H the perturbation growthrate p$ does not depend. on parameters g and cTm like thepull down velocity. This is bacause the characteristic time«f instability is less, than the time of. pull down velocitystabilization H/Vo. Another case corresponds to 'arge H)Ho.Here the increment is exponentially small and determinated. bydependance of disturbance of homogeneous pull downingvelocity vrith d.isturbance of lay thickness.

MUSICAL SlÃULATION OF THE PLANE LAYER INSTABILITY AND FUEL

CONFINEMENT IN THREE-DIMENSIONAL GEOMETRY.

The nonlinear phase of instability of ground. melting byplane fuel layer and, interaction of a single fuel fragmentwere investigated. numerically. For this specialthermalphysical codes were writ ten in vri:h the follovringprocesses vrere taken into aooount:

1. Transient heat transfer in one,tvro and three dimensionscaused by the dependences of the specific heat on temperatureand. spatial.

2. The motion of bodies pulling down through the melt,fuel intermixing in melt due to dissolving and ".onvection,thermal convection,vaporiztion of active parts while heatingthe fuel, and subsequent condensation (absorbtionj vrhileI".ooling the fuel.

In the framevrork of thermal model the motion, of pouringbodies in a melt is simulated as fallig of a thermal sourceson the melt bottom provided. by a fuel volume and povrerdencity are constant values. Process of intermixir~~ issimulated by distribution of volume thermal source in themelt.

Thermal convection caused. by high temperatures( greaterthen the melt temperature) is described by effective heattransfer coefficient. The phenomena of vaporization and. fluxmotion are described by introducing negative sources in theheat transfer equation. The influence of thenegative sourceis pronounced when the temperature is greater than boilingpoint.

'Ne used tvro tecniques for numerical solution of tvroensional heat. transfer equation. These tecniques are

d~scr bed in [2,3l. The resul ts otained vrith the help ofthese two. tecnique are close ach to ohter. or—.hree-dimensional problem the method [3] vras used..

he results of calculations show that the instabil'-:;y.p? oI:ess veI J'ui kly (for the time f '.;;,he order 3f O.1r.,'ether L is the characteristic time of 1'aj~~r displacement "to'-.he depth compared. with its thickness) comds to the nonlinear-hase. Nonlinear phase of melting dynamics is essensialy

initialand the

depended on the temperature, amplitud.e, anddisturbance scale directions of fuel fragmentmotion.

Based on numerical results consider some of the possibleways of fuel behavior in the nonlinear melting phase.

Cosine perturbation with amplitude 6=0.Z cm and wave length7=48 cm is laid upon the lower boundary of the flat lyer withthickness 7L=1cm and. energy release power 0=0.3 tÃJ'cubi . Itis assumed that heavy fuel parts fall on the melt bottom.According to the resuls of linear analysis such a layershouldbe stable. But the linear theory considers perturbation ofstady-state one-dimensional melting front. If the initialpertubation is not one-dimensional, the stabilization time<.an be equal or even greater then the evolution time oftwo-dimensional perturbations. In this caseinstability evolution is very sensitive to initial ."ondition.

'If initial conditions correspond to a stable phase of onedimensional melting the current perturbation decays withtime. But if in the beginning the melting zone is absent thenanother picture is observed. In figure 1 three d.if ferentmoments of instability evolution are presented.. Initialtemperature of the fuel and the medium is 700 Z. One can seethat melting begin in a zone of maximum

,thichness owing to

the space heterogeniety of the layer. Fuel mirgration ispossible only in the melting zone. Characteristic melts~time is comparable with the time nf „heating the remainingfuel= parts to the up melting-point. Under these conditions asingle drop is separated. from the point of maximum layerthickness and its independent motion begin.

While heating the remaining "solid" fuel part, anohtherdrop can be generated. This drop moves in a hotter medium:ompared. with the first one and. at eventually overtake tnefirst drop. Thus initially immovable perturbed. layer has aten'dency to splik, into several drops which can lateraccumulate in more complex structures.

The size of the first separated. drop practically does notdepend on the perturbation wave length if the latter meaterthan 6 cm. If the wave .lengty l, equals 6 <.m the irstdrop "consumes" the main part of „.he fuel layer. wd, as areset of thermal interaction the"'.neighbor drops accumulate.ogether and. a stationary lager,. is formed. again,. Thetime nf drop tearer~ off:is,,'<equal to 4 Cat/8 1 hour 30rtin. Atthe moment 4 chips 10 hour'8 39aiA the whole layer is restored.again. I<

The three-dimensional simulation of confluence of twodrops was carri'ed too. Fig. .'."-,:shows successive phases of".nnfluence of two "three-dw'en 'onal drops". Initial formthe fuel drops is a cube with 'the side of 10.5:m. Thedistance between:hem is Z.1: cm. Derog release -ower aequals 0.5 w/cm', initial temperature is "-.21 K. Tho tvtaiconfluence of two drops occur es 1n abou 25 ilats. Analogous:alculati< ns '~l 740-G. mensional a e ="ow, . ia ~ .he:onfluence time is 16 hours.'

RZLIABLE STOPPING SYSTZM STOPING AND SPRZADING OP LARGZ

VOLUMES OP FUZL OVER COOLING PLATZ

The two-dimensional program for solving heat transferequation with mov~ heat sources has been used to simulatefuel fall on a cooling trap plate. Figure 3 show the resultsof numerical simulation of fuel fall on the system ofparallel pipes foz spacing 30 cm.. In this case the break ofsafety system has taken place. If the piper spacing equals 26cm the front of the melting fuel is stopped at the distance 1cm from the surface of cooling yipes. It is assumed that thepipes are placed in sand. or concrete with heat transferconduction 0.02 w/cm grad.

The absence of fuel spreading through the closely placedpipe+ is related with. effective cooling the fuel "tongue"after its displacement by a distance of the pipe radius.

Thus, the fuel confinement by a system of parallel pipesessensially depends on the spacingbetween them. Even if thisstructure cantake awaythetotal heat flux from the fallinguniform fuel layez', the fuel:-melt will spread downtransforming into separate"tongues" provided. that the spacingis greater then critical value, wich depends on the pipediameter. Por the diameter of 200 mm the spacing should beless than 26 cm. 0

Then dynamics of fuel falling on the pipe cooling safetysystem with requied spacing or on a flat cooling surface wasinvestigated. After fuel melt front stops at adistance of theorder ofseveral centimeters (it depends on the thermalcharacteristics of the medium: from 1 cm in sands to 4 — 5 cmin concretes) fuel spreads along the horizontal surface ofthecooling plate due to concrete or sand. melting at the sidewalls of the fuel.

1. RZVIZPS of MODZRN PHYSICS, Vol. 57, No. 3, Part 2,July 1985.

2. V.M.Goloviznin, V.K.Korshunov et al. Preprini IPMAN USSR, Mosco~, N 58 ,19S5

3. R.V.Arutjunjan, L.A.Bolshov et al. Preprint IAZ ,Moscow, N 4121/16., 19S5

~i

Zing..e ",

wQ

IVIULTI-DIMENSIONAL CALCULATIONS

SUPPORT TV/0-BANG THEORY

Presented By

Dr. Seymour Baron

Associate Director

Brookhaven National Laboratory

at the First International Workshop

Severe Accidents and their Consequences

Sochi, USSR

October 29 to November 4, 1989

BROOKHAVEN NATIONAl lABORATORY $jgy (ASSOCIATED UNIVERSITIES, INC. flu 4

INTRODUCTION

Nuit i-dxrnensional calculations conducted at Brookhaven National

Laboratory tend to support the theory that turbo bangs occurred

at the Chernobyl-4 reactor. This xs an alternate scenario as

discussed by Milt Levenson of Bechtel in his recent letter to the

workshop attendees.

SL I DE I — THE TlrJQ BANGS

According to our calculat ions. the two bangs could have

occurred as follows. The power peaked locally as the accidentU)

proceeded. Temperatures at these hot spots became very high

leading to initial pressur e tube failu~ es, which would cause

steam b'owdown into the reacto~ cavity (the first bang>. It is

thought that as few as 5 pr essure tubes failing would be0suf, if ient to I i ft the r eactnr I id or deck. Once this occurs,

all the pressure tubes are broken, leading to ma„.or voiding,

causing a oower bur st due .o the posi+ive void coefficient and

the secon~ bang as the cor e di sassemb les.~1

THE TWO BANGS

POWER PEAKING INCREASED DURINGACCIDENT

HOT CHANNELS FAILED FIRST, BLOWINGpn, VAULT (BANG NUMBER ONE)

RESULTING DEPRESSURIZATIGN INCREASEDCORE-WIDE EXCURSION(BIGGER BANG NUMBER TWO)

BROOKHAVEN NATIONAL LABORATORY gy gy g

ASSOCIATED UNIVERSITIES, INC. (1411

Due to the reduced or positive void coefficient, the RBMK-1000 exhibits

spatial modes of power distribution instability. These radial-azimuthal spatial

flux modes were observed in the RAMONA-3B simulation of the CHERNOBYL-4 event,. i

In the figure, the RAMONA-3B calculated radial flux shape (at t - 43 seconds into

the transient) is presented* and indicates regions of "lead" assemblies that are

2000 C hotter than the core average assembly . The power in these lead

assemblies will continue to increase faster than the core average assembly, and

will ultimately result in the failure of the hot assemblies. In fact, this

highly non-uniform shape requires that the initial fuel failures occur in a

relatively small number of lead assemblies.

1*

(."

*In the RAMONA-3B simulation, a quarter-core model was employed which did notallow the half-core side-to-side spatial mode.

8ll

5 Controlled Cell

H Power > 1.3PAyg

g Power<0.7PAyg

HL",DE — AVERAGE VS. PEAK FUEL TENPERATUREqi

As discussed in the last viewg! aph, the RA@0'AA-ZB predict ions

show a s>>ostant 'l radial Power redistribut ior> durin>g the course

of .he tra>'osier>t. The resulting radial power shi f. moves the

peak power location to a group of chanriels located approximately

4;neters > rom the center. of the cor e. This spat i al power

red i str'esults in local "hot spots". Shown in this

Figur e is the calculated peak fuel tenaerature in these "hot

s-ots", versus the fuel temoerature c>. the average fuel assembly.

-.s se: n, the peak f>JP1 +e»ne! a~u! e ezxceeds the fuel melting

te>»pe) 8 ure Th»5 th>ese assef>lbl ies wc uld be expected to undergo

ea! y 'el "-ai1>are d>i> xng the accident.

='CE 4 — SCHE>VATIC QF CCRE DISASSEMBLY

Here we see a schematic of how the core may have disassembled

based on this two bang theory. The power peaking and high

temperatures shown in the pr eceding viewgraphs would lead to

jn' 'l pressure t'ube fai lures, which would cause steam blowdown

into the reactor cavity (the first bang). It has been postulated

that as few as 5 pressure tubes failing would be sufficient to

li f+ the reactor 1id or deck.1

Once this occurs, all the pressure

tubes are broken, leadino to mayor vc iding, which causes a. prompt

critical power burst due to the positive void coefficient, and

the second ban; as the core disassembles.

400IM-

3000.0-

"CHERNOBYL-4200 MWT TRANSIENT

CENTERLINE FUEL TEMPERATURE (C)

AVERAGEPEAK~'Q:: %~%%%%~%% 0

FUEL NELTING TEMPERhTURE

"-%GO.O-

MOGQ-~eeeoee essacs ee~ee—

40 40.6 4i 4& 43 485 43 - 485 44TIME (S)

Figure 9.= Core maximum and average fuel temperatures vs. time.

1

SLE< E ~ — THE RANONA-ZB CQ<"lDUTEF. ~~<,i.'E

The cai«o<<te<" code u'. i 1 i =ed f =< these calcu'.at ions is RAY<0<PA 38~

.' ansi: <"'.s -rid accider<cs.

which is a tht ee-dimerisioral r<eutronic, mult i-cnarinel thermal-

hydi aul.:c circ!e>designed f<~r 'predactir<g a w de rar<ge of system

The RANO<NA-ZB'alculatiori provides a

'eta i.'e 1 th<.ee-.dimens ior<sl r<eut t ori f 1<<x solut iori over the crit ire

core vo.vme. The <iodel ir<g a co<!r<ts foi all, f«oortant

ca«. t r'i <.<t <or«'<. '< eac'. i v i tv, iric 1 odin@ the m<oderato'< arid foe l.:/

".e',«Q« l 8 urrs., v -«C fvact 1or< arid c<-rit no 1 rod d eris i t y.

.ha 9".'O'JA-.:-~S =imillat io<<< of . the Chernobyl-4 reactor consists of an ~

<" .r:v< ser<:at or< ..-<f -.<i>e <:<<aeter of .ne care. Syl<lmetry is'ssumed

F.-»- ';he r es", «f the cc<re'. The iriit ial ci <no it ions'or. the

t<"ar s merit ca c<glatlor<s cot r esoorid to "the'eactor state as, of.

'<7

Cs ~g!.c'

C :) C:l I

"l~

L][] g~L> i>'.U)

l

RAMONA-38 COMPUTER CODE

THREE-DIMENSIONAL, TRANSIENT CODE(

COUPLED NEUTRONICS AND THERMAL-HYDRAULICS

PROVIDES DETAILED NEUTRON FLUXSOLUTION

ONE QUARTER OF CORE SIMULATED FORCHERNOBYL REPRESENTATION

STARTING POINT FOR CALCULATIONS VVAS

REACTOR STATE AT 01:23:0026 APRIL 1986

BROOKHAVEN NATIONAL lABORATORYgjgy g

ASSOCIATED UNIVERSITIES, INC. Cl ll I

AREAS OF POTENTIAL COOPERATION

COMPUTER CODES SUCH AS RAMONA-3BAND OTHERS COUI 0 BE USED

- PROVIDE INDEPENDENT ASSESSMENTOF SAFETY IMPLICATIONS FROMPROPOSED RBMK FIXES

- SERVE AS ADDITIONAL BENCHMARKCALCULATIONS FOR SOVIETCOMPUTATIONAL METHODS

POSSIBLE APPLICATIQNS

- INCREASED FUEL ENRICHIVIENT

- CHANGES TO REACTOR SCRAM LOGIC

- RAPID SHUTDOWN SYSTEM

- PARTIAL INSERTION OF CONTROL RODS

- ANALYSES OF BEYOND DESIGN BASISACCIDENTS

- EFFECT OF PASSIVE SAFETY SYSTEMSii,

BROOKHAVEN NATIONAL IABORATORYgjI gy g

ASSOCIATED UNIVERSITIES, INC. (5 ll I

SLIDE 6 — AREAS OF POTENTIAL COOPERATION

This presentation provides ai) exatnale of the types of

co<»putational raethods available in the United States.

>Be are aware that various changes to the RBYK design are be ng

proposed or are ]n fact underway. A part ial list ing of„s =me of

these coter>t ial fixes i s shown here.

As ~art of the pr ocess in showing how these changes ir»prove the

safety of the plant design ~ calc>:,lations r»ust be r»ade.

Ve would sugges+ .hat tasks su h as these would be fru]t . ul to

pursue as areas for no".er': ial c Jope> at ion.

ps is +he practice in the United States, ir»gor tant calculations

crr duc".ed by the r>.actor design organi "ation are independent ly

audi h e<1 hy a d.'; e> ent organ s zat ion. Thus, one pro pdsa 1 mi ght

be t= serve as an add'tional inderendent assess>»ent of cer tain

calcul at io>'<5 conrJ>>cted by the = r>1<>ni".at ion charged with r»aking

RBNK design changes.

An alternate pr««p<-«sal r»i ght bo to agree to certain accident

ace«arios that are ther> calculated by two or r»ore ditferent

=«t gani at io>>s. These c«-«>»ld then serve as benchr»ark calculat ions

< o c -"«'»pal 2 anC assess the co>»gut er ccdes.

i~

+I

;r ~ %./ .1 rr1CL~)

—„,a ~~%~~'ll 4xQ

~~C'.' „I::Yilllll~ t

.1(

Uk.. 'I(L$

.im er-g 4< II+

Ngg W%+gj

IVIII'~, h

g)! -4 I I

=--lr 8 III

3fi

'

~ I

I .~

it .:.

,lj~ ~

:::::gjig'5IRH j:

I

I

I

LESSONS LEARNED ABOUT PMR

BEYOND DESIGN BASIS ACCIDENT

PRESENTED BY

OR. S. 8AaON

ASSOC XATE OIRECTOR

BROOKHAVEN NATXONAL LABORATORY

AT THE FIRST INTERNATIONAL WORKSHOP

SEVERE ACCIDENTS AND THEXR CONSEQUENCES

SOCHX, USSR

OCTOBER 29 TO NOVEMBER 4g 1989

BROOKHAVEN NATIONAL IABORATORYIy gy g

ASSOCIATED UNIVERSITIES, INC. CI tl I

TECHNICAL CONCLUSIONS FROM THE ACCIDENT

~ Presence of water in the core protected fuel from damage

~ Reactor vessel did not fail

~ Negligible releases into environment

~ Containment did not fail

bnlaui

FISSION PRODUCT 5. CONTAINMEN1 Ht=ttA,V>Utt

About 1% of noble gases escaped into environment

~ 3X10-5% of iodine escaped into environment

~ iodine 8 Cs approx. 40% in RCS, 40% in containment

Reducing environment of core caused iodine to combine withother elements and form solids.

Peak pressure due to hydrogen burn (28 psig) less thandesign pressure of containment

No mechanical damage to containment shell and keyequipment

~ Thermal effect: Minor damage to non-safety gradeequipment due to H2 burn.

bnlaui

REGULATORY IMPACT OF THE ACCIDENT

~ Inspection

PRA

~ Regulations 8 Policies

Safety Goals

Equipment Qualification

Emergency, Planning

Fire Protection-

Enforcement

Operator Training

Siting Criteria

bnlaut

POST-ACCIDENT CORE CONDITION

~ Void in upper region covering approx. 1I3 of total corevolume

~ Debris bed up to 1m deep forms top of existing core

~ A layer of hard, impenetrable material at mid-coreelevation

~ 10-20% of core mass resides on RV lower head, 2 metersbelovr original core bottom

~ Bottom-most material in vessel is most likely non-fuelmaterial

Figure 1. TMI-2 Post-Accident Condition

naca ace»

1 ~

Control rod driveassembly leadscrew

OQ~j;yi'cj

Plenum assembly~ egg W

~ g

Damaged fuelassemblies

Core void—

Loose granular=,-'ebrisbed

e<

E

~Debris hangingfrom undersideof upper grid

Core conditionunknowi-I'i

gl

Slag-like debris(composition unknown)

bnlaui

DEFUELING

~ Biggest cleanup task

~ Very little information on core condition

~ Taskforce formed in June 1982 to remove vessel headand upper plenum

~ Problems

high radiation levels

high airborne radioactive contamination

~ Head removal took 56 hours and was completed onJuly 23, 1984

bnlaui

4

I .I

P

rf'

'EJ0

-L

CLEANUP TASKS

~ Defueming

~ Vforker Protection

~ Robotics

~ Fuel Shipments

DEFUELING (Continued)

~ Upper plenum removed May 1985 and stored in floodedfuel transfer canal

Dry defueling selected

~ Shielded work platform installed over RV

Long-handled tools

~ Surface crane added

~ Gripper rods for removing rods 8 hydraulic c!amshel.'-fot G Igg) Ag

Impenetrable layer was "one-shaped ceramic-like m".=s9 ft, in diameter 8 33 to 60 inches thick

bnlaui

DEFUELING (Continued)

~ Core-drilling machine used to make 400 overlappingholes, reducing it to rubble

~ Rubble loaded into canisters

Visibility problem caused by microorganisms in watersolved by using hydrogen peroxide

~ 20 tons of relocated fuel discovered within the LCSA{Lower Core Support Assembly)

~ LCSA disassembled ar d removed during 1988 using core-drilling machine and a plasma arc cutting device

~ Other activities included defueling of UCSA 8 RCS{pressurizer, OTSG, hot legs, cold legs, main pumps)

bnlaui

WORKER PROTECTION

~ Problem: high-energy beta activit~~ from strontium-90and yttrium-90

~ Solution

Specialized training

Improved instrumentation for beta detection

Protective clothing

Access controls to high radiation-level areas

Upgraded dosimetry

~ Doses to TMI-2 workers have been reasonable

bnlaui

WORKER PROTECTlON (Contiriued)

~ Problem: Heat stress to workers due to protective clothing

~ Immediate Solution:

Mfork time limits and education

Long-term Solution:

Engineering controls {ultimate soiution}

Worker protection equipment

~ Engineering Controls

Body-cooling system: frozen water vest

Chillers to lower building temperature

bnlaul

RQBQTIGS

~ Remote technology developed for TMI cleanup by EPRI 8 Carnegie-Mellon University

~ Robot Units used for:

decontaminating reactor building basement

sampling in auxiliary building

concrete floor scarification

washing contamir;ated surfaces

removing sediments

demolishing radiation sources

applying surface treatments

packaging 8 transporting material

bnl8ul

FUEL SHIPMENTS

~ Fuel & core debris shipped to INEL, Idaho

~ From July 20, 1986 to 1989

~ Shipments via rail through 10 states

~ Dedicated trains were used (Locomotive, caboose,radioactive material cars, buffer cars)

~ Local opposition and a minor accident were encountered

bn!aui

CONCLUDiNG REMARKS

~ Sevore accident caused by combination of operator error& design faults

~ This accident has provided valuable experience toimprove reactor safety

~ Improved confidence in containment safety function

~ Strengthened belief that presence of water protects fuelfrom damage 8 can cool a -.everely degraded core

bnlaui

First International Workshop on Past Severe Accidents and their Consequences

Sochi, USSR, 30. Oct. - 3. Nov. 1989

The fear evoked by radiation or radioactivityand its psycho-sociological consequences.

What can be done to improve this situation'?

by S. Pretre, Switzerland

The paper presents a tentative outline for a possible booklet on this subject

and for recommendations which could be given. It is proposed that this booklet

be developped within an international organization by a group composed on one

side by radiation protection professionals having some background in psychology

and on the other side by psychologists, sociologists and psychiatrists having

some background in radiation protection.

TENTAT I VE OUTL INE

1. INTRODUCTION

Motivation for this study. The observed facts. Typical cases: Three Mile

Island, Chernobyl, Goiania.

2. THE ORIGINALLY POSITIVE FUNCTION OF FEAR

'"ar is an emotion connected with the feeling of insecurity. Fear in the

animal world. Function of warning and protection. This fear is justifiedand usefull. Even in the modern world, if the sense of intuition is well

developped, the fear is an effective warning.

3. FEAR DUE TO CLASSICAL SUPERSTITIONS

Fear as warning for survival disappeared with increasing civilization. As

the need for a certain amount of fear seems to persist even if the causes

have disappeared, mankind developped a new fear: the fear due to super-

stitions. It is based on religion. Fear of the unknown, of monsters, ofthe ugly, of mysterious forces, of destroying phenomena, of all sorts of

symbols, etc. This fear is still connected to a feeling of insecurity,but this feeling is often not justified.

4. THE NEWEST TYPE OF FEAR

The classical superstitions disappeared in the 20th century. Rationalism

took over. But modern mankind is now developping a new fear. Reasons forfear must exist and if they dissapear~~easons must be invented. The

newest fear is due to the non-acceotance of death. The conditions ofhealth are so well known and medicine is so powerFul that death is not

accepted anymore. Death is a shame. Everything threatening health produ-

ces fear. It comes to the modern superstitions regarding aids, radiation,pollutants, technology, etc. The fear is partly the fear of such things,but is much more the fear of the symbols behind these things.

5. RADIATION AND RADIOACTIVITY ARE IDEAL SYMBOLS FOR NEW FEARS AND NEW

SUPERSTITIONS

Radioactivity cannot be felt. It is mysterious. It produces cancer and

genetic damages. It is related to the atomic bomb. It can pollute enor-

mous areas during very long times. Therefore it can represent evil and isan ideal symbol for an artful and destructive force which cannot be seen

or felt. And because of that, protection seems impossible. Science fic-tion authors have often used this symbol which became a field of new su-

perstitions. The fear of the symbol of radioactivity is much stronger

than the fear of radioactivity.

6. AMPLIFICATION AND CONTAGION OF FEAR

Fear is maintained or amplified by the attitudes of the neighbours, by

the media, by dramatizing reports exaggerating the dangers, by science

fiction movies. This is the negative evolution of fear. The situation

stays emotional; the rationalization cannot take place. This evolution is

archaic; it can become sectarian and even fanatic. The effect is conta-

gious and can evolve like a "psychological epidemy". It can reach a mass

psychosis (Radiophobia).

7. SOUND EVOLUTION OF FEAR

Rationalization takes place. Understanding the phenomena dissolves the

fear. A distinction is made between the thing and the symbol behind it.The real risk can be understood objectively. This is the ideal case: The

frightened persor, becomes able to slowly accept some new informations

coming from a person in which sh'e (he) has full confidence. The emotions

accompaniing fear are reduced. If the frightened person is not able to

rationalize, she (he) is at least able to develop trust in a knowledgeab-

le and friendly person.

8. ANALYSIS OF THE SOCIOLOGICAL "TERRAIN" IN WHICH FEAR EVOLVES

The 3 societies of Mary Douglas:

- The pioneering society described by: initiative, pragmatism, freedom,, r

individualism.

- The regulating society described by: method, structure, order, hierar-

chy.

- The'moralising society described by: health, cleanliness, purity, sec-

tarianism.

The pioneering society is a bad "terrain" for the development of fear.

But in a moralising society, fear can develop easily into a psychological

4

epidemy. Fifty years ago, our society was still predominantly a pionee-

ring society, but evolution took place: some pioneers became regulators,aid some regulators became moralists. The group of the pioneers is deple-

ting and the group of the moralists is growing. The "terrain" of the mo-

dern society becomes more and more favorable for modern superstitions and

psychological epidemics.

9. HOW TO TRANSFORM FEAR EITHER INTO UNDERSTANDING OR INTO CONFIDENCE?

The message about radiation,and radioactivity must be such that it can be

received. How to produce a message which can be received by a frightenedt

pe"..son? It is a difficult and delicate question of communication. The

message must be formulated in a psychological language which has a chance

to be accepted. Not,only the form of the message is important, but even

more important i ihe impression of trust 'emanating from the person

bringing the message.

10. THE FOUR BASIC COMMUNICATION LANGUAGES

The 4 psychological types according to C.G. Jung are the basis of the 4

communication languages:

- The thinking language ba ed on logic, analysis, scientific laws, ob-'ectivetruth.

- The intuitive language based on inspiration, imagination, theories and

models.

- The sensing lariguage based on specific details, concrete measurable

data, what is practical and down-to-earth.

- The feeling 1anguage based on subjective values, personal experience,

trust, compassion, sympathy.

The Myers-Briggs test shows that radiation protection professionals are

using mainly the thinking language. Some of them prefer the intuitivelanguage. But the population understands mainly the sensing or the fee-

ling languages. Translation in needed.

-5-

11. SCIENTIFIC EDUCATION AT SCHOOL

Modern technology develops too fast. The understanding of it cannot fol-low. The facts concerning radiation and radioactivity are not taught be-

cause they are not mentioned on the educational programmes. And they

could not be taught because the teachers would not be able to do it. More

emphasis should be given to teaching phenomena explaining modern science

and technology. The policy of education should realize that something

must be dore to avoid the appearance of modern superstitions.

Radiation and radioactivity should be placed in their natural context and

explained starting from natural radiations and natural radioactivity. The

risks due to radiatioi~ should be compared with usual and broadly accepted

risks.

12. PSYCHOLOGICAL EDUCATION AT SCHOOL

Some basic knowledge of psychology should be taught already early at-chool. What is fear? What are superstitions? What are psychological

epidemics? School should be a training center for developping the abilityto think and judge by oneself. School should teach how to be critic with

new informations. A new information should systematically be controlled

and cross-checked before it is accepted as a fact or as a belief. School

should train the children to reject any type of dogmatic thinking.

Children should develop a better ability to differentiate between fictionand reality.

13. THE ROLE OF THE FAMILY PHYSICIAN

The family physician, usually a generalist, can play an important role in

avoiding modern superstitions. Ke (she) is the person of trust; he (she)is by definition the person knowing about health. Therefore, physicians

should be more knowledgeable on radiation, radioactivity and health phy-

sics as they are now.

-6-

14. PROPHYLAXIS OF RAOIOPHOBIA

The following strategy is proposed:

1. Radiation protection professionals learn the communication languages

and get a background in psychology.

2. Radiation protection professionals teach the facts of radiation, ra-

dioactivity and radiation protection to the physicians and school-tea-

chers.

3. The governments introduce in the education syllabus (secondary

schools) the understanding of radiation and radioactivity. They also

introduce a basic teaching of psychology explaining fear, beliefs and

training a better distinction between fiction and reality.

FIRST INTERNATIONAL WGRKSHGF 'N SEVERE ACCI L ENTSAND THEIR CGNSEi <)ENCES

SGCHI URSS30/1<>/69 — 3/11/=:

SEVERE LGCA AT THE LVCENS EXFER ~.E!< A'!.C EAR< FCWER F'' i~N.

IN SWITZERLAND CiN 2'I .)ANUARY 1369

JF Buclxn EGS '-: =-aririe, =t«ber ~69

1 EiESCRI FTION GF FLANT

The LUCENS E>:per iment al Nuc . ear F ewer F.ar t was 'cated

vari

thewestern part o: Swi t zerlarid

Three adjacent unaerarouno ca verns ere cc ris t r u =ted:

R The reactor cavern, hou=.ina also the steam erierators, ws th allnuclear auriliary c).rcuits: moderator, control rod cool dna,shield cooling, intermediate circuit, emeraency circuits, etc..M The machine cavern, housina the turbsr.e and feedwater circuits,the electricity aenerator, access ccritr I arid decnrit rccms, someelectrical rooms, sewaae circuits, e.=...F The fuel cavern, with equipment icr fuel storaae and handlina,incl. fuel dzsmantlina inside the water pool

In the service buzld>na, at t."ie cutsi"e, were placed the controlroom, electrical plants, encl. emeraency d;esel's, fresh airplant for the caverns, CG2 storaae and handlina, waterdemineralisers, workshops and laboratories,

etc'�

..There were other buildinas at the outs:de: liqu>d waste treatmeritplant, tertiary circuits, cooiina towe.s, pumping station,Ventilation filterina station, stcraae rc oms, etc...The thermal power of the plant, was 3'Wth, incl. 2E< '(Wthneutron po~er and 2 MWith from the blowers. The bruttc electr>calpower was little over 8 MWe, the net,tc electrical power litt'eover 6 MWe.

The dimensions of the fuel elements <FE) were: lenath 3 m, withpressure tube diameter (PT o) 13 cm. Each FE contained 26 fuelcartridaes. The metallic uranium core «f each cartridae was 65cm iona, with a 0 of 17mm. The cartr~daes were canned with amaanesium alloy. The outside dimenszcris were 32 mm for the Macoolina fins and the lenaht 70 cm. Four cartridaes were screwedtogether and placed inside of 3 graphite columris, each with 7coolina subchannels.

The total fuel charac consisted of 73 fuel elements <FE) placedin vertical Zr-alloy pressure tubes <'FT!. The metallic uraniumwas slightly enriched to 0.96

The reactor co: '. i: a medium was C02 circulating at a pressure of60 b, The total vclume was about 9 m3 of C02 coolant, with a massof 1400 ka,

The CD2 circ '.a". r cn t tirouah the fuel elements was of theso-ca'led single reentrant type. Cold aas sweeps down between PTanc ~ he inte .-.. '. -:aptii te s? ructure tc the bottcm oi the PT,~here z. urns .;p inside the azaphite structure to thc top of theFE.

The moderator " r = ..t contained 23 t or DZ0 moderator, cooled totempera<a!:e 'en 30 and BG'C, borated at start-up of thereacto... as -'=borated as Xenon was building up.

The twoL

wer e swepaa wentrea -tor h

oppositesecond=ircui+,

0 i'.

r.,; -.h

11o i note =c'

he

t fotc'a f

st earn —cene? a1 cl!

mean= 10

ad «-. power of 1 MWe each,C. Two half circuits, "ieach cthers. ™+artina atmanyfold room 1, ttien tohot manyfcld room 1, to tfor 2, down to blcwer 2,;:irculation t;me was ". ss for the full circuit.

and pushed 60 ka/sri a fiaure of 8",

the blowez 1, thethe FE of the firsthe top of tt,earid so on for thefor cacti half

The two helicoidal, v:.rt ical, once throuah steam ceneratorsre ei;ed in'et cas at 380OC at the tcp, and expelled cold outletaas at 210': a+ bottom, where the blowers were flanaed-on. Theywere the f rst steam-generatozs of this type to be used at anuclear po er Fiant.

Ttie design pre =ure of the primary circuit was 65 b and thedesian temperature 520'C for the hot aas path, resp. 300'C fozthe cold path cf <he C02.

Tl c seccndary part of each steam-aeneratcrs had a volume of 2 m3,anc was es.'aned fcr 70 b. The settirigs cf the steam safetyvalves was suc i h 'h in order:o prevent a Dlowina out of C02 incase of a tube ri:.ure. The feedwater flow reached at the inlet6 ka/s, at 14G = a;id 25b. Ttie evapcrat icri temperature was 223'C,

the e;aporatior. poi nt was near the end of the evaporator partcf ttie ceil . Fact. aererator was f it ted wi

gati

a small steam/vaporseparator on its side, bet ween evaporator coil and superheaterceil. The steam output temperature was 370 C, supezheated ~

ThE'teamcircui . desian ratina was 500~C.

=eparated emergency cooling coils were placed at the top of eachsteami-aenerator. Each coil was connected to a conder.sina tankwith a voiume of 2 m3 placed above it, under the roof of thereactor caverri. Each of these condens'ic tanks was cooled bynatural circulat iori from a water reserve ir of 20 m3 placed at thetop of the reactor cavern. Wheri cooling, wa+ez evaporated andwas fed to the rcof by a pipe penetrating the containment. Afterdepletion of this amount of ~ater, an emezaency water reservoirwith a volume of 30 m3 placed outside cri the hill could be usedfor supp'ementina.

The material of (tie moderator tank was peraluman. The wallthickness of +he calandria tubes was 2.2 mm. The mantel piece andthe six safety rod positioris were reinforced in order to resistany foreseer. over- or underpzessure. Five burstina discs wez'efitted foz rupturing at 25 b. There were alltoaether 73 calaridriatubes for 73 pressure tubes and 14 contrci rods positions, of

The refueling was foreseen off-line, after disconnectina the headoi the PT at the bottom oi the standpipe. The PT could then beremoved from underneath, with the correspondina bottom plug, andbe transferred to a station where a watertight cap could befitted on it. The fuel could then be stored in the sealed PT,in a pond cooled by water circulation, iniside the fuel cavern.Facilit.es ior inspecting the FE were not provided on-site.

2 CONSTRUCTION ANC CiF ERAT ION

The years 1955-60 saw the end of the maior hydraulic plants to beconstructed zn Bwitzerlanid, Ecnloos =al op p==->tron hindred alscmost of the projects for construct>no fossile fuel plants. GnlyCTV, a 300MNe plant fueled ~". th heavy oal, could be erected, buthad to be built in the mountains, which caused many unnecessarycosts and negative effects.The f irst Geneva Conference ATOMS FCiR PEACE sn 1955 was the beganof a new nuclear era, xn Swi< zerland lake elsewhere. In 1959 theswiss atomic law was promulcated. Many smaller projects fl orishedat the tzme.

de ign & construction oi aUranzvm and D2G moderated, ofted for swiss national use andntries. As primary coolant,but the tecnnology was not yetSo CCi2 -as taken to start with.

The sw:ss industry was interestednuclear plant fueled with naturalthe PTR-type, meant to be best suiliable to be exported to other coulight water steam was a favorite,advanced enough to allow for it,Utilities were interested in gett ina soonest possible nuclearpower plant operation experience and facilities for personneltraining. For supply, they env>saged +o buy a VS El"R, because.."ey feared the swiss des>on might be to ong to develop, tooe. pensive, too much outs de the normal l>ne.

The autorities were interested in the application of the newlyvoted law arid procedures, Nuclear safety autorities wereappointed in 1960. The research institutions were interested xnthe use and impro:ement of their knowledge.

So LUCENS got a real, maybe even the f;rst very wilde-rangingnational effort in industrial matter. A society was rounded bythree groups, an industrial one, the utilities, and in addition amerger amoung all western industrial and economical forces.The swiss Confederation allocated 50~ in funds, half on loan,half as contribution.

Bite preparatioris star+ed in 1961. The rock excavation insandstone lasted from 1962 to 1963. Construction and errection oimechanical and electrical plant was performed between 1964 and1966, with a strong participation of the operating czew, whichstarted to be on site in the second half of 1964.

First criticality was reached in the last days of 1966. Physicstests could then be done on a reduced loading. The fuel gotunleaded before commissioning of plant started in january 1967.This year was committed to a complete cold arid hot commissioningof the plant, and to hot physics tests. The primary circuitcould be heated to 220 'C by the blowers. Detailed operatirig andtest procedures were set up by the operator during this time.Shift work started already during important parts of thecommissionnxna orooram.

E"tensive static and dynamic tests were performed with the plantduring the year 1968. This concerned mainly the steamgenerators, the first of their kind to be operated at a nuclearstations The plant was raised to increasing levels of nuclearPower, and most of the tests were repeated at each new powerlevel. The main control of evaporation inside the SG's could bePut to three different parameters: with varying or fixedevaporation point in the SG's, or with varying input temperature~o the reactors

Emergency cooling tests were parts of the program, blowing offsteam to outside, within approaved temperature limits, with theaim to keep for over 12 h. The tests were such successful that itappeared feasible to operate the plant at steady thermal po~erand to look for the equilibrium level achievable,

the normal case, with forced circulation ipumps> to drivewater from the condensing tanks through the emergency cooling«ilsi a capacity to remove 4 8 of nominal power was the limit.ls far more than needed. With natural circulation, it meansithout any AC-powered equipment working, it was still possibleevacuate nearly 3w of Fnom. A test was done, s~itching outPossibility of supplying the plant with electricity both fromnet or by the diesel generators, to induce a perturbation ont"e secondary side leading to the loss of the turbine, and

consequently of the main generator. It ~as then completeblack-o..t of the plant. Afterheat could be evacuated, without anyPumP or motor running, for many, many hours...Another way to evacuate afterheat on passive means was to letfeedwater be pushed by the slight steam pressure on the feedwater

to drive feedwater into the bottom of the SG's coils: 6anom could be evacuated by this ways

Further tests were performed, like: loss of primary and secondaryflow tests, load rejections on the electrical grid, keeping on»use-load, etc... In July 1968, the main test program on thesystems and circuits was terminated ~ From August to Cctober-968, a long -.ime run at or near full power was performed, withthe aim to only '.rradiate fuel. In November the first test fuelelement was unloaded and sent to the hotlab at Wurenlingen. This

was inspected and found in perfect shape.The program to the end of 1969 was simply to 'rradiate andi»pect 4 more test FE's. In the year 1970 it was then theintention to unload and dispose off the fuel, and then to modifythe reactor core.I UCENS turned out to be a perfect running racing car, but with amuch too weak motor. Metallic uranium and Mq-canninc melt atabout 600 C, and start burning-at 600'C. Canning temperature atnominal power was 450'C, with hot spots calculated to reachOat

The core survey instrumentation was not fully suited to thepurpose for an experimental reactor. Each FE had outputtemperature, flow and activity measurements; some instrumentedFE's had a dozen of canning temperature measuring points. Thesurvey of temperatures worked very good. But flow and activitymeasurements were unreliable.

Many kinds of fuel deterioration mechanisms were feared. Thisconcern got expres ed to the plant owner in November 1968.Further prudent operation was allowed in January 1969, mainlybecause everybody trusted the underground containment to be verysafe.

ORIGIN CF THE AC'."'"EN.

A simplified drawing o.'.: the blower ro.atin= eal can be fo ind

annexed. The origina'eal had been ;er, well tes«ted, alsodynamically, and found safe. E it we ran out of supply.Difficulty in providing spares .rom «arne subcontractor appeared'

sunderstandings were later detected at the constructor'soff ices on the status with dryness of the primary circuit. Itwas in reallity always too wet, but a wrong specification was setfor the replacement seals. A» a consequence, great waterin-leakages appeared at insertion of the new seals in November1968. The dryer units could not work when the primary was at lowtemperature. All effor.s made during weeks for purging, drainingand drying the primary circuit were no+ sufficient,

Everybody was confident that no co idensaticn would app~ar insidethe pressure -.ube«. The lowe«t temperat are was est imated to he

higher than the condensinc due point. Reversing the rotatingrings of the seals to their smooth face ha«been the solutionselected, despite the risk of their life to be shorter. Thereversed rotating rings operated well, it means with negligiblewater in-leakages, before, during and after the accident.

WHAT PROE'ABLY HAFPENED

(as discovered during dismantling 1970 — 1972)

During the shut-down periode until mid-January 1969 periphericalFE's got cooled to a lower temperature than estimated, becausethey were nearest to the cooled radia shields. Humidity couldthen condense during December 1968 inside four PT's. Inaddition, water droplets could be preferably diverted into thePT's at the western part of the reactor, because this was nearestto the CO2 inlet pipes. Evidence of water accumulation to a

height of over 2 m inside PT's 49 & 59 and to a height cf lessthan 1 m inside PT's 48 & 58 was found. strong corro»ion of theMg canning appeared at places where the water level stagnated fora time. Less corrosion happened below this water level, butsuf f iciently to block 3 out of 7 cooling si.bchannels inside thegraphite blocs of FE 59 ~ All water inside the four PT's couldevaporate during the start-up heating & drying periode atmid-January 1969. The blockage could not be detected by existing

instruments'uring

power escalation on ?1/1/69, the flow measurement on FE 59was not able to detect an obstruction, neither was the mixedoutput temperature measurement, nor the t.urst slug detectionequipment.

Melting of canning and fuel inside 3 «ubchannels of FE 59occured. This has led to a complete obstruction of thesechannels, therefore activity detection or overtemperaturedetection got even less possible.

FIRST INTERNATIONAL WORKSHOP ON SEVERE ACCIDENTSAND THEIR CONSEOVENCES

SOCHI URSS30/10/89 — 3/11/89

SEVERE LOCA AT, THE LUCENS EKPERIMENTAL NUCLEAR POWER PLANTIN SWITZERLAND ON 21 JANUARY 196

JP Buclin EOS Lausanne, October 1989

~ +))> @~F

gi t'Pili

f

Melting of canning 5 fuel inside 3 subchannels of FE 59Complete .-~struction of these channels, therefore no activitydetection, nor overtemperature detection possible

Bowing of the fuel qraphite column, toucPing the PTContact o. PT (Zr-alloy ) with molten U+MgBurning 2 holes into PT; 85 m from bottom a 3xlScm oval hole

facino center of tank 6G cm from bottom a 3.5 8 cm round holeC02 pressure (50b) expelling molten mixture towards FE 56,

burnina calandria tube of pos.56Tear of 60 x 20 cm, 2 m from bottom (facing wall of tank)A 1 safety and control rods insertedDefcrmation of all thin calandria tubes, pressed onto PT's

pressed onto control rodsStart of depressurisation of primary circuitLocal D20 steam explosion within moderator tank?Bursting of all rupture discs on moderator tank lidDisplacement of D20 to upper face of Er under moderator tankAxial shields experienced forces of 25b x 10 m2 = 2500 tUpper axial shield not lifted, but its fixation distorded

( All the acts of this last paragraph happened within 1 or 2 s!

R'Bb'gY PAF486f%NT

Normal shift attendance during tests or new start-up:1 safety engineer in control room as superviser1 reactor engineer as shift leader2 operators on =ontrol desk3 auxiliaries for mechanical, electrical & instrumentation

Operation: 17h00 to 17h15, r>s;nc power :rom 30 to 40<ready tc switch o.". automatic primary controland tc synchron'se t.rbo-ceneratcr

Acc),dent time: 17h20, sh~rt after end cf ncrmal working t>meMost main> enance, chemistry, health physicsworkers on their way home,but all ch;efs fertunately still on site.

in control room, a hundred of alarms ( 2*Origins cf fast SD ( all abou. s>multane

Primary circuit neo. pressure gradientPrimar" circuit global radioact vityPr>mary circ~it loss of fl«~ i 2 " 2~3 ~

Core neutron flu): > coo'ant flow <2Moderator over-pressure ~2A3)Contro'od cooling overpress.re > parr,

.=ast shut-down: all C afety an' contrEmergency dec'red,Fast loss of primary pressure: 50 b toPressure rise inside bioloaxcal shi eld

means 2 out of 3):"us):

< 2:. 2*3)2*3)

2)h 4)

stop > m;n flow (2*3)rods fully in!

within 10 mincontrol rod coolina cc)

S-.ar< of emergency procedure PV2, the cne .'or a case of"primary rupture inside biological shield"

Check: Cont a inment au".omaticaliy i solat ed: okEmergency COZ f eed to pr err ary au+ orna t i cally in: okCJ2 blowers stay on high speed: okTemperatures of FE fallzng (all instrumented FE were ok! )

accordinc to safety rule, nobody inside containment?(two persons could not be absc 'teiy certif ied as notbeing in: tedious investigation at their homes: ok)

Alarm: — Chief of operation, at home (sheduled for ne>:t shift)Plant superin+endant, at a meet zng as member of the swissri icle ~r saf ety aut >rities, where he could inf orm +hese.State of Vaud police center for information

Decisions taken at 17h40 on phone with p'nt superintendant:not to use hand ope.ated emerg--ncy ~ater floadingnot to us a cc.n,ainmen+ releaf valves to +he rockd ur;.p D23 to storage .ank

all t..is ) =ssible because FE cooling was okput main ventilation imachine and fuel caverns) onrecirculation over iodine f ilters, becauseactivi ty monitoring at stack started raising

Decisions taken at 18h25 on phone with plant superintendantas activity still measured at stack:

alarm organisation for survey of general radioactivityover swiss territory.stop main ventilation (machine and fuel caverns)

This decision was taken because first principle is to protect thepublic, especially if such measure had nn strong adverse effecton plant personnel (wore gas masks for a couple nf hours untilcase was cleared) nor negative effect on plant.

i~ I4 th +I-~~~ ~ ~ ~ ~ ysssssv asoos a a veeI ve ~ s soaaw ~vs ~ sg va yves s v aeeo ~ ovv

been taking air probes and made smear tests down-wind near theplant, with negative result. A responsible person took thecc'lected samples to a special>se'aboratory at Fribourguniversity, division of physics.

Gamma doserates inside reactor ca; em ) 100 r/h, instr. of f scaledoserates inside machine cavern 10 rad/h behind materialaccess door, 0. 2 rad/h near t ur? inc, and 2 mr ad /h in accesstunnel, near entrance to control room. Geiger-Muller countersindicated high readings, but ion chambers low values: sign ofcontamination by a caseous beta-emitter.

At Zlh30, plant s:)pezintendant arrives on-site.Technically there was no problem. Fuel temperatures 140'C.Some administrative measures to be initiated:reduction in number of persons on-site, o..ganisation of on-calllists, information of o~ner and head off ices, preparation of apress release, ordering of supplies from other nearby centers,etc...

23h, some supplies ordered were brought in. 0 member of .he''state police visited the auxiliary co;.~ma»d point oraanised nearthe fence. Experts from nuclear safety aut=ri'.ies arriving.

At 24h, help came in from the Nurenlinaen nuclear researchcenter in order to perform independent environmental surveys.

At Zh30 on 22/1/69, results came from Fribourg univers.'ty:Rb88 with traces of J133 (later recognized tc he Te 132 instead)and of 3135 assumed in the samples collected. The main activitywas from Kr88, decaying to Rb88 with Tl/2 = 3h, decaying tostable Sr88 with T1/2 = 0. 5 h ~

Overpressure inside containment: 1.6 m water gauge = 160 mb(COZ emergency feed had been correctly stopped near 18h)Fuel temperature nearly equal to CQ2 gas temperature about 100'C:thermal capacity of secondary circuit is now commanding thecooling down.

Lesson to be learned: not placing radiological rooms withmaterial supplies, showers, monitors, instruments, etc...too near to the reactor! Doserates there eri: still around50 mrad/h (streaming through access lock) during the night

The machine and the fuel caverns .were put again on ventilationon 23/1/69. Pressure raises slowly in reactor cavern, because ofcompressed air supply to blower seals.

On 24/1 inspection team identifies a leak (3 m3/h 4 0.2 boverpressure in containment) at a penetration for electricalcables. Leakeage stopped. Has origin of contamination of themachine cavern. *A* No aerosols on smear tests in vicinity ofleak i.ng penetr at ion **4At Bh start emptying ZOt HZO storage tank on roof of reactorcavern for receiving pumped DZO from floor of lower chamber.Preparations in order to depressurise containment over iodinefilters, with measurement of effective decontamination factor.Calculation of expected emissions and corresponding immissionsActivities measured on air samples: H3, 3133, Te132, 3131, Kr85

At 19h depressurisation of containment to 0.2 b over 40mm8 pipe

O alt<> F k~On Zb/1/b& av vn aepxe»ui iia~.~over iodine filters. Measured decont factor: 550. Calculatedtotal release corresponds to 0.04 mCi 3131, No harm to population

At 3h15 depressurisation terminated. Opening of fuel transferstation: 20 mr/h. About 30 cm D20-level on floor of lower chamberAt 12h acute phase of accident terminated. Majority of federalexperts leaving site. Main task: save

D20'osesaccumulated during the acute phase nf the accident farunder 0.1 rem, except 2 persons reaching this dose when stoppingturbine auxiliary equipment during first hours

At 16h about 5t of DZO pumped from floor to HZO storage tank onroof of reactor cavern (over 75< cf inventory finally saved)Start heating reactor cavern for evaporating D20 humidity

INVESTIGATIONS OF ACCIDENT CONSEQUENCES

On 27/1/69 three teams of 2 persons penetrate reactor cavern withair bottles (15 min working time, no escalator! 20m height)1st team: dose mapping shows 0.2 rad/h general, 3 rad/h 0 filters2d team: smear probes, inspection of plant, ~wi".ching of dryers:

no apparent damage outsite biological ~hield3rd team: smear probes, look inside lower reactor chamber

On 28/1 TV & broadcast teams assist at preparations for entry:air samplings; dose rate 0 entrance to lower chamber: 40 rad/h

29/1: Results from smear probes E air sampling indicate uniformdistribution of FP and fuel isotopes at all,placesThis, together with Kr85 levels, a sign of bulk fuelmelting or burning. U/Pu ratios point to peripherical FE?

31/1: Confirmation of above results. Repetition of reactorcavern leak test: 1966 S 1967: -30 m3/h 4 0.3 b

1969: -20 m3/h 4 0.3 b = 50'etterIn next days, a TV camera Sr ion chamber were introduced throughtransfer opening. No damage apparent, higher dose rates WESTAbout 100 rad/h on floor level in lower reactor chamber.Until April 1969 working in pressurised suits .'cause of H3contamination of cavern atmosphere,

Defective refueling machine (mechanism was flooded by D20) to berepaired. About 1000 rad/h over sump: covered with .5t of leadwithin 2 min. Scrubbing of floor in lower chamber: 10 rad/hvacuum cleaner after few minutes, Casting of shieldings forfilters on spot. Refueling machine repaired; but.. ~ NOTPOSSIBLE TO MOVE ANY FUEL ELEMENT!

Inspections from reactor top through safe';y rod and reservechannels. FE Nr 59 detected as defective, but main tear towardswall of tank. C02 feeders to FE 59 isolated, then primarycircuit leak tested: no other important damage discovered.

Drilling (only dry methods allowed) of an inspection hole 7m long8 cm 8 between wall of moderator tank and FE 59. FE 59 found

~badly distxoyed, nearly empty.

Decision taken to dismantle the reactor and unload it from thetop ~ Decision to investigate carefully causes of accident. Anyevidence related, to the accident to be preserved and analysed.

c ji <J ji m -/o - !Tables with doses, cos s, waste produced to be shown.

Inquiry commission created by federal government to investigateoriain and pxplain course of accident. All evidence had to betransmitted. Commission decided on investigations to be doneoutsi"e the nuclear po~er station, mainly at hot-lab of nuclearresearch center in Nurenlingen.

Dismantling finished in 1973. Very extended scientific workingpapers of commission condensed in a report, of few 100 pagespublished 1979 in french and aerman (original), available fromFederal Printing 0 Supplies Office, CH 3000 Berne, for S 10.00under No. 805.10d in german, 80 . 10f in french ~

Proceedinas of an Internationa'ymposium, IAEA SM 234/8ipp. 449-475, 1979: JP Buclin, "Ddclassement de la CentraleNucldaire expdrimentale de Lucens" in french. Comparaison ofdismantlina study made in order to modify the Lucens core after19~0, at higher irradiation but with no contamination, withdismantling experience made after the accident, with less.rradiation, but very strong contamination. Indications on timeschedules, materials used, costs, wastes produced, doses, etc...International Decommissionina Symposium, October 10-14, 1982Seatle, Hashington: pp. IV-106 .. IV-122: "Decommissionina ofLucens" by JP Buclin EOS. This report goes more into details ofdismantling techniques utilised, and underlines difficultiesrelated to underground construction, also when decommissioning.(See also Nuclear Safety,Vol.16, p. 434-435, July-August 1975:"Comments on Underground NPP Si t i ng" )

Nuclear Enaineering International is advertising a special issue,to be published in 1989, on Severe Reactor Accidents. Thechapter on LUCENS aoes more into det,ails than present text. NEIanalysed very carefully the repcrt from the swiss inquirycommissior. and asked the plant, operator for descriptive materialand explanations. Subscr i prior c pen now.

OFFICIAL I NQUI RIES

By law, swiss federal prosecutor opened an inquiry on faults tobe put on designer, or constructor, or operator, etc...He ".sked the scientif ic depa. tment of the poli =e of a greater8".a e within the Conf ederat i or. to collect and analyse theevidences. This work was perf -rmed with skill by specialistsused to such procedures. They had secured written evidence inthe f irst days (operating manuals, log books... ), interviewedpeople at the plant and at offices; were present on site atdifferent occasions, as evidence was expected to be discovered;have awaited the results from all other gremiums before releasinain 1979 a statement that, they didn't find any faultive behavior.

Plant owner also asked some sk il 1ed and independant universityprofessors and industry xepre en" atives (not connected to LUCENS)to investigate the way the different partners (constructors,plant operator, architect-engineers, etc...) fullf illed theirtasks and whether they could be made responsible.

4*1962*196641969*1973*1962~*3 LUCENS 17 MOST COMPLETE, QUICKEST 99 SAFEST AND CHEAPEST 6

SIMPLIFIED DRAWING O. BLOWER 8 AL HITH ." 'H PATH

1. PRIMARY GAS ENTERS BLOWER AT 58 b

2. COMPRESSED C02 OUT OF SLOWER WHEEL AT 60 b

3. CO2 FLOH THROUGH VPPER LABYRINTH

4, DEMINERALISZD SiAL-WATER AT C02-PRE:-SURE 0 2 b

5. ROTAT NG DI™COF COPPER N I r'tE. PLAT ED CW. CMIVM LA rER

6 ~ GRAPHITE RING, FIXED SPRINGLOADED

7 ~ LUBRICATING AND COOLING HA.ER FI M BETHE.;,N RING AN" DI™C

8 ~ CLEAN C02 PURGING LOH R LAB'. R NTH

(nct ~cr a' cFerat''na condi ticns)

9. MIXTURE OF =EA- HATER AND PURGE

1 0 ~ SEAL HA.'R SEPARATOR <'YC ON )

11. SEAL HATER TO LIQUID HASTE TREATME.'IT PLANT

12. PURGE C02 TO DRYER P" ANT

f'3f'- q

Wc))

SS~&nt CHL?'8

ERSON'~.

Personnes DOS88hommes ann0es hommes rett

DOmontage: 73 (63)Investigations: l3Decontamination: 37

75 (260)60

( ) = selon etude

Paper to appear in proceedings of ENS/AHS Top1cal Heeting on Probabil1sticRisk Assessment, Zurich, Switrer'land, August 30-September 4, 1987.

CHERHOBYL-4 /ACCIDENT ANALYSIS

by

Edward E. Purv1s IIIU. S. Department of Energy, Washington, D.C. USA

and

B. W. Spencer

Argonne Hational Laboratory, Argonne, IL USA

The US Department of Energy formed a team of experts in July 1986 to bet-ter understand the Chernobyl-4 accident. These experts were from ArgonneNational Laboratory, Brookhaven Hational Laboratory, US Department of Energy,Oak Ridge Hational Laboratory, and Battelle-Pacific Horthwest Laboratory. Theresults of the analyses performed by this team in 1986 are published in DOE

HE-0076, "Report oF the U.S. Department of Energy's Te~m Analyses of theChernobyl-4 Atomic Energy Station Accident Sequences." The team's analysesconfirmed the plausibility of the results of Soviet analyses presented at theChernobyl-4 Post Accident Review Heeting held the week of August 25, 1986, 1nVienna, Austria.

The Soviet analyses presented were, however, based on a point kineticanalysis. This analysis contained an exogeneous positive reactivity input of1.5 millik per second for three seconds starting at I:25:40.25. The US teamperformed limited 3D and 20 calculations. These calculations indicated thaton a global basis the maximum react1vity 1nsertion from this pos1tive scrameffect was about 10 cents. Although the positive scram is a plausible explan-ation, it is not likely.

With the reactor in the cond1tion of instability that existed, there wereseveral plausible events that could have triggered the rapid excursion. Eachof these should be dealt with as a safety problem.

The team found that in addition to the posit1ve scram there were a numberof other plausible initiators for the power excursion, f'r example, the 1n-creased coolant boiling owing to the flow reduction, pump cav1tation, a fail-ure of trans~tion joints or other fittings in the piping system, failure of arupture disk or pipe, opening additional steam exit valves, etc. Once thereactor had been placed into the highly unstable condit1on that existed, anexcursion became a significant risk.

-2-

The team's best estimate is that the accident was a loss of flow induced(by flow reduct1on) power excurs1on that took place in two stages. Th1s paperwill concentrate on the acc1dent energetics and, in particular, on some new1nformation deve loped fol low1ng publ icat1on of Ref . I. Three mechanisms forthe energetic destruction of the Chernobyl reactor have been examined.

1~ Fue'I-Coolant Thermal Interact1ons

Mhen a large energy depos1t1on occurs in fuel elements 1n a very shortlength of time, character1st1c of a severe power excursion acc1dent, the fuelis typ1cally heated to beyond the point where the U02 fuel begins to melt(about 2850C). The pressure with1n the fuel element that 1s acting upon thecladding rapidly increases owing to the fuel thermal expansion, melt trans1-t1on volume increase. and release and heating of gaseous forms of fiss1onproducts. The fuel element cladd1ng eventually bursts, and molten fuel ardunmelted fragments of fuel are forcibly ejected 1nto the surrounding coolantchannels. The very hot fuel interacts with any water flowing through thechannels at that time, rapidly producing steam, and may addit1onally contactthe channel wall causing local heatup. The total channel pressure is the sumof the expanded fission gas pressure, the steam pressurization owing to theforced fuel-coolant interm1xing and heat transfer, plus the or1ginal ambientprcssure'he

pressurization of the RBMK pressure tube owing to the fuel-coolantthermal 1nteraction (FCI) was calculated in the USDOE team analysis to )e ashigh as 140 to 150 atmospheres during the power excursion at Chernobyl. How-ever, this pressurization is not calculated to be sufficiently high to failthe pressure tubes solely by overpressurizat1on unless there was a degradingof the integrity of the tubes owing to such factors as: a) fuel impingementand local weakening of the wall, and/or b) weakening at the upper transitionweld due to the extraordinarily excessive heatup rate of this joint calculatedduring the power excursion. The loading of the pipe 1s by pressure waveeffects propagated upstream and downstream from the expanding FCI zone. Theupstream propagating pressure wave calculated for the Chernobyl accident wasnot of a sufficient magnitude to cause damage to the inlet piping at thisearly stage of the excursion. The direction of expansion of the FCI zone waspr1mari ly upward since coolant inflow inertial effects and check valves pre-vented downward expansion during the t1me scale of interest. The upwardexpansion of the high pressure zone causes acceleration of the water/fuel mix-ture upward collecting additional water above it and rapidly condens1ng steamas the upper region is compressed. The water/fuel slug was calculated to beaccelerated to about 40 t.,eters per second. The slug soon impacted on the endof the shield plug region in the channel, where the abrupt constrict1on wouldhave caused a water hammer-type pressur1zat1on. This pressurizat1on was cal-

»ted to be sufficient to load the channel walls far beyond their ultimate;i capability, and in fact the end plugs at the refueling end may have

sed even w1thout pipe failure under this load1ng. In e1ther event, theconsequence 1s that individual pressure tubes may have failed above the sealedreactor space 1n a reg1on devoid of any containment structures. The concrete

-. hield block, end cap, parts of pip1ng, and fuel bundle parts from 1nside the.; 'sure tube would 11kely have been ejected upward as miss1les through theproof of the refueling building if this was one of the actual failure modes as'~ggested by the analysis. A steam blowdown would have ensued through the

-3-

fa1led top of the channel, directly into the refueling building. This failuremode could have involved many of the channels.

It should be noted that the energetic FCI's described here take place 1nindividual pressure tubes, and there were 1559 such pressure tubes loaded w1thfuel at Chernobyl-4 at the t1me of the acc1dent. There is no reason to expectthat the FCI's would occur simultaneously 1n all 1559 channels; in fact, thisis very 1mprobable. An 1ndividual FCI and its consequences in a part1cularpressure tube has a very fast t1me scale, occurring in just 50 to 100 m1lli-seconds. The range of time over which fuel in the large number of channelswould reach failure condition and cause an FCI 1s thought to be much longerthan the FCI timescale itself. Hence the FCI energy release 1s noncoherent;i.e., it occurs at different times in the d1fferent channels. This 1s becausethe heating and the cooling of the fuel 1s not the same 1n all channels, whichwe term a var1ation 1n the power-to-flow rat1o among the channels. In areactor such as the RBMK, this is evidenced by,;,the var1ation during normaloperation of exit steam quality which is known'from instrument readings foreach individual channel {though not at the time of the accident). From pub-lished Soviet 1nformat1on, the exit qual1ty 1s known to vary from about O. 11for the lowest power-to-flow channels t~ about 0.22 for the highest power-to-flow channels, the average being 0.145. The team analysis estimated that thetime interval between failures of the highest vs. lowest power-to-flow chan-nels was about 0.4 s, and could have been much larger if the power rise wasirregular across the entire core under the conditions of the accident.

Two methods were used to estimate the magnitude of the FCI energeticsduring the accident. One method is based on the maximum energetics measuredin in-reactor power excursion tests, and the second method is based on theenergy release in the meqhani~tic base case calculation by the EPIC code forChernobyl channel conditions. The experiment of interest was the RI)-ST-4reactivity initiated accident test in the Power Burst Facility {PBF). Inthis test a single, unirradiated fuel element was subjected to a rapid powerexcursion. The fuel element failed when the fuel enthalpy radially averagedat the peak axial flux location was 350 calories per gram; this indicatedabout 100K local melt fraction. A transient channel pressurization of 350atmospheres peak was concluded to have been caused by an FCI. The coolantkinetic energy was about 3.4 KJ, and the overall conversion of thermal-to-mechanical energy was reported to be about 0.3 percent efficient. In general,in-reactor tests of this type do not show energy conversinn efficienciesexceeding about 0.5 percent when the thermal energy is based on the ent1refuel mass.

This "effiriency approach" 1s now appl1ed to the Chernobyl acc1dent con-ditions. Us1ng an average fuel enthalpy of 1.24 megajoule per kilogram. UO>,130 kilogram UOq per fuel channel, and 0.5 percent convers1on effic1ency, theest1mate of mechanical energy 1s 0.8 megajoule per channel. The t1mescale forenergy release of the FCI is < 10 millisecond, the timescale for mechanicaldamage effects in the piping system is < 100 millisecond. The total energyrelease from 1559 fueled channels amounts to 1.2 g1gajoule. Under normal cir-cumstances the timescale for this release is determined by the extent ofpower-to-flow noncoherence which would probably exceed 0.4 second overall.(However, as will be seen shortly, circumstances during this accident were not"normal.")

4

The kinetic energy of the water slug accelerated by expansion of the FCIinteraction zone in the base case EPIC calculation amounted to 10.5 kilojouleper channel. Hence the total calculated energy release 1f all 1559 fueledchannels experienced an FCI would amount to about 16 megajoules based on 'thismechanistic calculation, considerably less than the 1.2 g1gajoule energy re-lease based on the energy convers1on method of the scoping calculation.

The FCI energy release was suff1cient to cause fa1lures of the 1ndividualpressure tubes under the cond1tions of the acc1dent, but 1t does not accountfor the energetic destruct1on of the reactor. As descr1bed previously, calcu-lations indicate the pressure tubes could have fa1led from a water hamner-typeloading at or near their top end plugs, above the sealed reactor space, caus-1ng steam blowdown directly into the refueling building. Additionally, tubeswould have fa1led within the sealed reactor space owing to degradat1on of thetube wall by two mechanisms: a) fuel impingement and related heatup of thetube wall, and b) excessive heatup rate at the upper transit1on weld. Thesteam blowdown into the steel vessel of the sealed reactor space from progres-sively increasing numbers of failed channels has been shown by the analysisteam to cause fai lure of the vessel within about 0.1 second of inception ofthe tube failures. In addition to this, the channel voiding which resultedwhen the FCI's occurred caused an increase in reactivity 1n this particulardesign of reactor which progressively worsened the power excursion. (Thistype of neutronic behavior would not occur in Western-design boiling water andpressurized water reactors wh1ch use light water both as a coolant and moder-ator.) Hence it can be said that whereas the FCI energetics did not blow thereactor apart dii ectly, it produced two important consequences that did; i.e.,the overpressurization of the reactor vessel which held the core, plus thechannel voiding-related increase in reactor power.

2. Steam Blowdown

The USOOE team performed parametric calculations to examine the overpres-surization time scale for the sealed reactor space vs. numbers of failingchannels. The failure location was taken to be the upper transition welds;failures assumed to be located in the core region would not have changed theconclusions. The steel structure was assumed to fail at about 0.15 megapascaloverpressu~e, about twice as large as the overpressure rat1ng of this steelstructure, and sufficient to levitate the massive (1000 tonne) upper biologi-cal shield assembly upward. Accounting for the discharge flow through thevent pipes from the sealed reactor space plus the choked influx of steam fromthe failed channels, the time for upper head failure was est1mated to be 0.8second for 1.6 percent (25) of the lead tubes failing coherently and 0'.15 sec-ond for 6.4 percent (100) of the channels failing coherently. With progres-sively increasing numbers of channels failing with time, characteristic ofthis type of accident '+ was concluded that the t1mescale for failure of thevessel was about;,.i ,. The failure involved the lifting of the tophead, among other possible effects. The lift1ng would likely have broken allthe pipes 1n the sealed reactor space, and perhaps many of the water/steamp1pes above the head. Th1s failure caused a rapid escalat1on in steam blow-down rate consist1ng of abrupt release of the pressur1zed steam in the sealedreactor space plus blowdown of essent1ally all the channels, not only thosehigt power channels that may have already failed at the top ends of the pipingT's or end plugs plus those high power channels that were fa111ng inside thesealed reactor space and responsible for overpressurizing it. Channels of

lower power which were heretofore intact would now fail coherently due to thefa1lure of the reactor vessel. This would have had an autocatalyt1c effect onthe power excursion since any water remain1ng in the core would now void co-herently, causing a rapid reactivity insert1on. Hence the overpressurizationof the sealed reactor space and the mechanical failure of the vessel structureis not an end to the destruction per se in this reactor system, but 1s a mech-an1sm to prolong and worsen the power excursion accident.

The massive steam blowdown followinq failure of the reactor vessel was in1tself a source of energy release during this accident. The steam expansionwas a particularly vigorous process because of the large flow area from essen-t1ally all the pipes, through both the inlet ends and the outlet ends. Therewere about 800 tonnes of steam and water in the system at the t1me of theaccident. The largest par t of the inventory was located 1n the four steamdrums at essentially saturation temperature. The timescale for the blowdowndepends upon modeling assumptions for the two-phase flow from the reservoir inthe steam drums to the pipe break locations in the core region by the longerinlet flow path and at the top of the reactor by reverse flow through theshorter discharge path. The shortest blowdown time scale was calculated to beabout 7 seconds, but the actual blowdown probably took longer than this. Thesteam efflux velocity at the top of the reactor pit was calculated to be about170 meters per second for the fast blowdown case. This released steam accumu-lated in the refueling bay above the reactor. Ignoring leakage losses throughblowout of doors, windows, or missile-generated holes in the roof, and assum-ing that the roof structure would fail at an overpressure of - 0.03 mega-pascal, the delay time to blow off the roof by the effects of steam blowdownis estimated to be about 0.5 second.

The total work potential of the s:earn inventory blowing down from satu-rated conditions at 7 megapascal pressure to atmospheric cond1tions (0.1 mega-pascal, quality 0.35) amounts to about 160 gigajoule. This is far greaterthan the energy available from the FCI's, but this energy is released over therelatively long blowdown time period of > 7 seconds.

3. Fuel Boilinq

The power excursion calculated by th~ USOOE team using a ten-channelmodel to account for noncoherence effects produced fuel temperatures whichexceeded the normal boi ling temperature of the U02 fuel in seven of the tenchannel groups (- 1100 fuel channels in the Chernobyl reactor). Hence thepower excursion of the Chernobyl accident was sufficiently severe that in mostof the core the fuel vapor pressure became very h1gh, and the expansion of thefuel vapor as the system depressurized was a contr1butor to the forces thatdestroyed the reactor. The maximum work potential for the fuel vapor expan-sion based on the s1mplified ten channel model in Ref. 1 is calculated toamount to 32 gir,"jcu! 'r'' cele of the energy release is approximatelythe same as th: .,u i , .;«.. „ ,.ie first seven channel groups, about200 millisecond. The energy release is less than would be calculated usingthe core w1de-coherent model and double power burst of Ref. 4 and 1s also lessthan would be calculated if the coherence introduced 1n the channel void-1nduced reactivity insertion by lifting of the vessel head had been includedin the ten-channel calculation 1n Ref. 1. The analysis that led to thisresult is very simplified in that the positive reactivity effect of fuel homo-genization was ignored as was the negative reactivity effect of fuel axial

"((

-6-

dispersal out of the core; also, property values are l..urge extrapolations ofavailable data. Oespite these qualifications, the results indicate that fuelvapor probably was an important contributor to the destruction of the reactor,and may have been the dominant energetic process taking into account both themagnitudes of energy from the available sources and their timescales for re-lease.

SUMMARY

In svnmary, three sources of energetic events have been examined whichplayed varying roles in the destruction of the reactor and the reactor build-ing at Chernobyl. These energy release mechanisms and the time during whichthe energy is released are summarized in Table 1. From this table it can beseen that the greatest single source of energy is that stored in the massivecoolant inventory at high temperature and pressure. Steam overpressure prob-ably caused the early failure of the steel-plate reactor vessel which com-prised the sealed reactor space. The steam blowdown which followed was adominant process in the accident and probably contributed to overpressurefailures of the building roof and exterior walls. Although the abrupt steamrelease from the overpressurized reactor space and the ensuing blowdown prob-ably disrupted the reactor internals, it cannot account for the "volumetricdestruction" of this region. This extensive destruction can be explained,however, by the rapid expansion of fuel vapor heated to very high temperatureduring the severe power excursion. The FCI energy release amounts to far less

Table 1. Order of Magnitude Estimates of Work Potential

Process Durationof Release

Maximum WorkPotential, GJ

la. FCI, bounding estimate ofenergy conversion efficiency;0.8 MJ/channel x 1559 channels

lb. FCI, EPIC calculation of slugenergy, 10.5 kJ/ch x 1559

2. Coolant blowdown to 1 atm

3. Fuel vaporization, noncoherent10-channel model

- 0.4 sec 1

- 0.4 sec 1

> 7 sec

< 200 ms2

1.2

0.02

160

32

Notes: timescale determined by noncoherence of 10-channel model

timescale determined by noncoherence of first seven channels of10-channel model

-7-

than either the steam boiloff or the fuel boiloff. However, it was sufficientto cause failures of the pressure tubes both inside and outside the sealedreactor space, where 1n the case of failures 1nside the vessel, conditionswere made right by the impingement heatup of fuel on the channel wall arid/orrapid heatup of the trans1t1on welds. Hence the FCI's were important to theoverall sequence of the acc1dent.

Four factors pertaining to the design of the RBMK system were found to beof particular 1mportance 1n accounting for the energet1cs of this acc1dent.First, the rapid rate of heating of the upper transit1on weld during the powerexcursion, which caused the welds to 1ncrease in temperature at a rate fourorders of magnitude greater than the 10'C/hr allowed, undoubtedly degraded thepressure tube integrity, leading to multiple fa1lures ins1de the sealedreacto. space. Second, the overpressure rel1ef 11nes of the sealed reactorspace were s1zed for flow from failure of only a single tube, and herice mul-tiple tube failures rapidly overpressurized the space, causing the steel-platevessel to fail. Third, the failure by overpressurization of the steel reactorvessel caused the massive upper head to lift, and because the pressure tubesare welded to sleeves welded to the cover plate, all the channels would havebeen broken by lifting of the head ~ Fourth, because of the positive voidcoefficient of reactivity in this reactor design, the failures of all thechannel tubes would result in coherent boilup of coolant and voiding of thechannels. Hence the lift1ng of the reactor head 1s a mechanism to cause co-herent ',ntroduction of reactivity by causing channel failures which producerapid void increase, an autocatalytic effect with respect to energy releaseduring the accident.

References

1. Report of the US Department of Energy's Team Analyses of the Chernobyl-4Atomic Energy Station Accident Sequence, DOE/NE-0076, November 1986.

2. Transcript of the IAFA Experts'eeting regarding the Accident at theChernobyl Nuclear Power Plant and Its Consequences, Vienna, Austria,Aug. 25-29, 1986.

3. M. S. El-Genk, R. R. Hobbins, and P. E. MacDonald, "Molten Fuel-CoolantInteraction During a Reactivity In1tiated Acc1dent Experiment," Nucl. Eng.and Des., 66 (1981).

4. USSR State Committee on the Ut111zation of Atomic Energy, "The Acc1dent atChernobyl Nuclear Power Plant and Its Consequences," Informat1on compiledfor the IAEA Experts'eeting, Vienna, Austria, Aug. 25-29, 1986.

USI, Olt Hl',I,A'I'I V)i'. Ii ISS IC)N I'HOI)UC I'INDI.i)ASI!, T() PSTI Hh'I'Ic

Till: I Ultl. 'll'.EIVI)HATURI: Al'l'L'R CIIFBNOBYI. ACC1DENI

h. h, Rlg)sK1-KorsaKov

V. G. Khlopir) Hadiu)n lr)st I t»te, I.eningrad, USSR

ln the early stages of mar)aginp, the Chert)ol)yl ')cci-der)t Or)e Of the mai» pr OhlemS WaS the c:Ot'r'eCt eVal»ai,ianof temperate)tr. o f fuel, r emaining .In tl)e destr oyed zoneof the reactor, s Ir)ce 1 t determit)ecl the measures to be ta-Ker) to avo1cl fur tl)er accident. eleve I npment, Dir'ect measure-ment of temperatu)r. was at. the moment, not only impossible,but. also witho»t ob,)rct, because one did not Know wherethe hottest lower layer of cor e debris was situated. Thenit was decided to «valuate th1s t.emper at»re , »sing theactivity ratio o I iii tfer er)t f1ssion pr oducts .1n the gase-ous release from the core,,ince released composit1on dif-fered not.iceabl y I r orn expec ),ecl cor e 1»ver)tor'y.

her'c) so I san)p I e s wer e I a)cer) f 1 om t l)e p1»mr. o f . moKean(1 pas above ).hr. I»)t nit)g t eac I ot by CosXn)nt)ycir ome t a ir pl a-»es s ir)ce Apt I I 21)-11) [1), a»d .".1»ce tiny 0- tl) the gt o»p ofV. G. Khl op1» Haclii))n 1»s 1.1 I » t r. [2) bega» )alc1»g s»ch sampi esuSing h1r I Or Ce hei Ic:Opt er S a))d a.) tSarnpl ()ig t.eohniq»eS,

The results of gamma-spectrometric measurement.s ofactiv1ty rat.ios for several nuci.ides ( 1» relation t.o z1r-conium-95 ) der1ved fro)n [1, 2) are pteser)ted in Table1n that table the values of

R (1)A (1) ~ a (Zr)

A (Zr) ~ a (1)( I )

are listed, where

A(1) — is the observed activity of .1-th fission product,A(Zr) — is the observed 95-Zr activ1ty,a(1) - 1s tHe I-th fission product core inventory,a(Zr) — .Is the 95-Zr core 1r)ventor y.Gale»lated core inventor 1es were taker) from [3] for theKnown average Eue I b»r nup.

Fi ss1 or) ptoduc t ) e. I ease, r ate deper)dence o)) tempe,—r at.»r e for ur at)1»m clioxicle matr ix was studied exlierimen-tal ly and discuss«el 1)) [0, 5]. 1 t .Is assumed to fol low theexpression

log ( I (I) ) = — Ci(i) / T < C2(1) ( 2 )

where : F(1) — 1s the part of 1-th nuclide atoms,sec) fr om the tnatrix every second,

C I (1) — a cor) s tar) t for 1- fh rn)c 1 I de, deg.C2(1) — a Constar)t for 1-th nuclide, IogT matrix tetnperature, deg. Kelv1n.

r'e I e a-

Kelvin,( I/s ),

Values of C1 and C2 for cerium, bar fum, .iodine, ruthe-nium, cesium and zirconium can he derived from [4] ortaken from [5] and can he used to compare the predic-tions of models [4, 5) to the real source term of Cher-nohyl accident. On the other hand the curves ( 2 ) inconjunction with our data on Chernobyl release [1,2]can he used for a» i»dependant estimate of core debriscooling history after the accident.

For that purpose a computer code TEMPERh wasdeveloped, which can be used to fit ( by the least-squ-ares techniques ) the relative fission product releasefor each day to the ( 2 ) dependence with Ci and C2 co-efficients, taken from models [4, 5] or corrected by fit-ting procedure. Data from Table i., however, should hemodified before l>ei»g used in such fft, because theyrepresent not or)ly the "evaporated" fission products,hut also the fue( particles, carried out from the reac-tor shaft by the gases, smoke and flames. This cor-rection must he most noticeable for such "nonvolatileproducts, as 95-Zr and 141, 144-Ce.

To take this .into account, some quantity P shouldhe subtracted from the R(i) values, listed in Table i.,and the result divided hy (1-P). The case of P=O corres-ponds to the situation, when all the 95-Zr is releasedhy diffusion from fuel matrix, while P=i means that allits release is due to the fuel dust, carried by the gasflow. In our program the value of P for each day ls va-ried, until the best fit is achieved.

In Table 2, the sets of Ci(1) and C2(1) coeffici-ents, taken from [5] and those, found by the program arelisted, and the corresponding release rates F(i) vs. mat-rix temperature are shown in I lg, 1. We found, that Ci andC2 values from [5] differ drastically from the ones, ob-served experimentally. Same values, derived from [4] we-re nearer, hut still hetter fit could he achieved for lo-wer Ci values, shown in Table 2. as "Our value". This me-ans, that products were more "volatile" in Chernohyl ca-se, than ln experiments, on whiclr predictf I>ns [5] werebased. Since errors in 137-Cs measurements",during Apriland May 1986 were high, 137-Cs data was not used ln thefinal version of TEMPERA calculations.

The results of TEMPERA fit for time period fromApril 28-th ( i. e. Z. 5 daYs after accident ) until May8-th ( 1. e. 12. 5 days after accident ) is presented inFig. 2. The corresponding P values for each day werefitted separately for each daY and lay in the range from0. 2 to 0, 8, whith tendency to increase with time. Thiscan be understood as the result of general decrease lntemperature and the corresponding decrease of the rela-tive role, plaYed hY diffusion-evaporation process [5].Considering the measurement errors for all fission pro-duct concentrations about 15 / ( since no errors in [1]

were l1sted, and 1» our measur ements (2] 15 % was theaverage error value ), TEMPERA calculated ch1-squaredfit cr1teria, which ranged from 1 to 3. 5.

The general t.rend of temperature change in Fig. 2does not contrad1ct the evaluation, g1ven in 1906 USSRReport to IAEA [61, Our calculation represents thoughan independent estimate, based on data, quite differentfrom those, used 1» evaluation [6). 'I'heir general agre-ement should be v1rwed as a new and significant fact.

Another conclusion from this calculation is thatthe fission product. release rates for Chernobyl eventfound here can be successfully used as an instrument toreconstruct the teniperature of uranium dioxide fuel mat-rix under accident cond1t.tons. This method can be app-11ed 1n the temperature range 400 - 1600 Centigrade,particulary irt the case, when the fueL ls inaccessibleand even it.s pos1t,ion is unkr>own,

References

1. 10. h, H3Pa3llb W np., Pallwoar<'r'wnrlOe sarPnsnertwe npwpolt-rtbrz cpert n sorte anapww r>a qh3C.MeTeoponorwn w rwrrpoaorwn, N. 2, c rp. 5, 1907.

2, h. h. Pwwcr(wA-L(opcawon w rip. Mccrtenonartwe n@Gpoca arra-pwAworo pear(Topa IV Gnor<a ~LAIC.OT«eT Faltwerror o wricTwryTa ww, B. P, XJtonwna, No. 14396,1906 r.

3. B, M. I(oaoGamr(wn w rip, Fanwarrwonwwe zapar(TePwcTWr(W OG-zy«ennoro nnepnoi o ronnwr3a. Cnpano«nwrc, M., 3rtepro-warta r, 1903.Report t.o tr>e Amer 1 can Physical Soc.ie ty,NUHEG No. 6-04-03-0LI, Feb. 1905

5. C. D. Andriesse, "Interpretation of fission product r e-lease from over»eated fuel", lr> Froceedir>gs of IAEAsymposium "SOVRCL. TERM EVALUATION FOH ACCIDENT CON-DITIONS", p, 65, 1906, Vienna.

6. MN4opwartwz oG anapww na qhSC w ee »ocaelrcTBW5lX, nojlro-TOanennan rlrln MAI'hT3. ATOWrran Bneprwn, r. 61, c, 301, 1986.

4'l'ab l e

Helative abundance of d(fferent 1 lssior> products H(i)( compared to t.he expected fr om core irtventor y ) t.oabundance of 95-Zr, derived from tlata of [1,2].

Date1 il4

Ce1 Il 1

Ce140

Ba131

1

103Hu

106Ru

28. 04Z9. 0430. 04

1. 052, 053. 054. 055, 056, 057. 050. 059. 05

10. 0511.05

0 7r0. '(41, 034, 370, 76

0. 04

1. 131, 31, 491, 0

1, 07

11G. 47.1. 773. 01l. 06

1. 25

l. 20l. 5t. 631. 02

G. 60

1. 300. 99

1. '10

2. 75

0, 69O. 037. 37G. 66

2, 303. 0il35. 03'l. 3

19

4, l(1

1. 49, 1'1r.'l r,

0, 07

1. 09

0. 570. 626. 374. 610. 00

1. 27

0. 755. 271'0. 01. 02

G. 76

0. 52O. 833. 422. 50. 00

0. 08

1. 495. 3510, 012. 9

Z, 20

Tab l e 2.

Coefficients Gl (1) and 02(i) from equations ( 2 ),describing the fission product release rate depen-dence on temperature from publicat.ion [5], and tho-se found by TEMPHRh as best fit for t.he Chernobylevertt,

Huc)ideCl value,

degreesC2 value,

log 1/s

Data (5] Gur value Data (5] Our value

144-Ce141-Ge140-Ba131—I103-Ru106-Ru95-Zr

'(4400444001260017400370003700057600

16200167002050015300166001000019000

9. 949. 940. 23. 37. 57 rJ

14. Il

U. 110. 131. 733. 310, 121, 352. 02

Fig. i. Piss>qn product release rates vs. matrixtemperature for Chernobyl event, found byTEMPERA fit.

D h T E

The most probable history of core debriscooling, calculated by TFMPEHh from re-lative fission product release data.Temperature — in degrees Centigrade.

ACCIDENT PROGRESSION ANALYSIS OF PAST SEVERE ACCIDENTS

David J. McCloskeySandia National Laboratories

Albuquerque, New Mexico, 87185, USA

ABSTRACT NUREG-1150 ACCIDENT PROGRESSION ANALYSIS

Accident progression analysis methodsemployed in NUREG-1150 are reviewed, withapplication to the Surry plant. It isproposed that these methods could beuseful in the analysis of past severeaccidents, such as occurred at Chernobyl.

INTRODUCTION

The accidents at the Chernobyl, ThreeMile Island, and Windscale nuclearpower plants resulted from a variety ofunanticipated or uncontemplated eventsinvolving human error and machine mal-function. The origins and progressionof such accidents are often clouded byuncertainty. In the review of theseaccidents, we are interested in determin-ing what actually occurred, as well asidentifying important factors that couldhave prevented or changed the course ofthe accident. The overall goal, ofcourse, is to extract all lessons fromthe review to provide assurance againstfuture occurrences.

It is proposed that accident progressionanalysis methods employed in currentprobabilistic safety assessments could bea useful tool in the analysis of pastsevere accidents. In this Workshop on

the Chernobyl accident, the methods couldbe u-ed in simplified form to guide ourthinking, help systematize the diverseinformation presented, evaluate thecredibility of scenarios, and ensureconsistency of results. In a differentsetting, the methodology could beexpanded to form the basis of a morecomplete investigation o f alternativescenarios involving considerations ofplant operator actions, equipment per-formance, safety systems operability,accident phenomenology, and containmentresponse, with due consideration of theuncertainties involved.

The most extensive development and appli.cation of accident progression

methods'ere

performed in support of the SecondDraft of NUREG-1150 (Ref. 1). Figure 1

shows the major steps of the risk analy-ses: (1) plant systems analysis, (2)accident progression analysis, (3) sourceterm analysis, (4) consequence analysis,and (5) risk integration.

The relationship among the constituentanalyses is illustrated in Figure 2.To make the risk analysis tractable,interfaces are defined between the majorsteps, and the results for each stepare aggregated into a smaller number ofrepresentative states or bins. Theaccident progression analysis proceedsfrom an assigned set of plant damagestates that describe the status of theplant at the onset of core damage. The

frequency of occurrence of each plantdamage state is determined from the plantsystems analysis. An event tree logicis used to describe the accident progres-sion.

The characteristics of an accidentprogression event tree are summari=edin Figure 3. The tree tracks theprogression of accident sequences fromthe onset of core damage through failureof containment with release of fissionproducts to the environment. The eventtree provides a mechanism for integratingthe diverse and often differing models,analysis methods, and experimentalinformation that have accumulated fromthe past dozen years of severe accidentresearch within a framework of expertscientif ic judgment. The tree isdeveloped by asking a sequential seriesof questions about the progression of theaccident. The questions may have multi-ple discrete outcomes, or "branches." A

case structure is established such that

«chis work was performed at sandia National laboratories, w..ich ss operated for the Us Department. of Energy urde contract

Number DE-AC04-76DP00789.

the outcome may depend on the branchtaken or parameters defined at previousquestions. The case structure maintainsphysical consistency and allows all typesof accidents to be treated uniformlywithin a single tree with sufficientgenerality to accommodate uncertain ordiffering interpretations of physicalphenomena. The effects of operatoractions on plant systems in either miti-gating or worsening the accident may beincluded at any point in the event tree.

It should be noted that the accidentprogression event tree does not replaceexperimental information or detailed,mechanistic codes. Rather, it is aframework or "bookkeeping" system toincorporate the results of differentmechanistic codes and diverse experimen-tal information. The results of themechanistic codes enter the event treesthrough the filter of expert judgment.The event tree structure is particularlywell suited for the study of processesthat are incompletely understood andrequire expert judgment to resolve, sincethey can incorporate several differingalternati:e interpretations of systembehavior and physical phenomena.

The NUREG-1150 accident progression treesare large (71 to 145 questions) and havea complex, interlocking case structureto accommodate multiple time periods,multiple plant compartments, and differ-ing expert interpretations. It is notfeasible to make a schematic representa-tion of the entire accident progressionevent tree, although reduced trees can bedisplayed. However, the tree can bethoroughly documented by describing thebranches, case structure and associatedinformation for each question in thetree. A generalized event tree proces-sor, EVNTRE, was developed to efficientlyprocess .accident progression event trees(Ref. 2) and list and sort outcomes.

PROBABILISTIC QUANTIFICATION AND USE OFEXPERT JUDGMENT

progr ssion tree based on subjecti..einterpretation of the available data baseby expert panels and the plant analysts.Parameters about which there was littleuncertainty were represented by singlevalues.

Expert panels were convened to addressthe most important and uncertain ques-tions involving equipment performance,in-vessel accident progression, contain-ment loading, molten core-containmentinteractions, containment structuralperformance, and source terms. Theexperts used a variety of techniquesin their development of subjectiveprobability distributions, includingreliance on detailed code calculations,extrapolation of experimental data toaccident conditions, and development oflogic networks. Formalized decisionanalysis procedures were applied tominimize bias, provide consistency, anddocument the expert's reasoning.

A stratified Monte Carlo method was usedto sample probability distributions ofuncertain parameters, and the sampleobservations were propagated throughthe event 'ee structure to determineuncertainty distributions for variousrisk measures.

EXAMPLE OF ACCIDENT PROGRESSION ANALYSISFOR SURRY

The scheme used to define plant damagestates for Surry is illustrated in Figure4 for a station blackout accident. Theplant damage state definition includesinformation about the status of thereactor coolant system, emergency corecooling, electrical power, containmentheat removal, auxiliary feedwater sys-tems, and reactor coolant pump sealcooling. In this example, the reactorcoolant system is intact and the steam-driven auxiliary feedwater pump hasfailed, but the other systems are recov-erable following recovery of offsiteelectrical power.

A major objective of NUREG 1150 was toestimate the uncertainty in risk esti-mates. This required the developmentof probability distributions for theuncertain parameters in the accident

The Surry accident progression event treeconsists of 71 questions, starting withthe size and location of the reactorcoolant system break (if any) when thecore uncovers, and ending with the final

-2-

containment condition. The analysis isdivided into five time periods !Initial,Earl~.', Intermediate, Late, and Final).

typical path starting from the stationblackout damage state considered pre-viously is displayed in Figure 5. Thepath passes through all 71 questions.However, many of the questions are not ofparticular interes for this example andonly 38 questions are shown. Shown foreach question are the branches, theassigned mnemonic for the branch taken,and the branch probability {or anotherparameter), The case structure is notsholem.

Question 21 is shown in greater detailin Figure 6. This question addresseswhether offsite power is recovered duringthe Early time peri.od (one-half to twohours). The question has seven cases toaccommodate sequences starting from allplant damage states, Case 3 applies tothe station blackout plant damage statebeing considered. (The entry conditionfor Case 3 is that either Branch 2 or 3was taken at Question 10, since for thisplant damage state there is no heatremoval from the secondary side at thestart of an accident. Previous caseseliminated the situations in whichelectrical power has been available sincethe start of the accident and in whichpower is lost and is not recoverable.) A

distribution giving the probability foroffsite power recovery as a function oftime was sampled to obtain the branchingprobabilities used in this example.

The results of the accident progressionanalysis may be displayed in variousways. Figure 7 shows a reduced eventtree for all the station blackout plantdamage states. The end states are thecombined failure modes for the vessel andcontainment.

APPLICATION TO THE CHERNOBYL ACCIDENT

are given in Figure 8, and are descr.'bedbelow.

The first step is to determine the p)an:damage state that is the start of :heaccident progression. A single plan:damage state is determined by listing thestatus of plant systems and values ofimportant plant parameters at the lates:point in the accident in which the plantstate is certain (about 01:19:00 on April26, 1986). It is appropriate to specifythe plant state in greater detail thanfor a risk analysis.

Next, relevant time periods for the acci-dent are defined, for example:

1. Final Operation (01:19:00 to01:23:40)

2, Reactor Excursion (01:23:40 to01:24:00)

3. Damage Control (rema inder ofApril 26)

4. Accident Management (April 27to May 6)

The major important operator actions,plant systems response, and physicalprocesses are identified and listed foreach time period.

A sequential set of questions is thendeveloped for each time period. Repre-sentative examples of questions are:

Final Operation:

Did the remaining main coolant pumpstrip prior to the accident?

Did blockage of a control valveoccur?

Mat was the average core .oid frac-tion at reactor trip?

Ve have not developed a detailed accidentprogression event rree for an RBMK plantor for the Chernobyl accident. However,the methodology is generally applicableand should yield useful insights. Thesteps in applying the accident progres-sion analysis to the Chernobyl accident

Plant Excursion:

Can a pressure tube rupture propagateto endanger its neighbors?

At what pressure will the ~ pperbiological shield lift?

How far did the control rods inser.in the core?

Vas the reactor prompt critical priorto the first, explosion?

Damage Control:

Could flammable gases generated bythe addition of water to the reactorcavitv explode?

Could the fire in Unit 4 endangersvstems in Unit 3?

Accident Management:

I'hat was the temperature rise of thecore debris?

and analysis of the tree leads to identification of important issues and cr ticebranch poin s tha can be given moridetailed consideration.

Simple event trees can be employed iiqualitative form or with simple probability point estimates to guide physicareasoning, screen out unlikely scenariosand evaluate the consistency of plausibl~scenarios with all available informationA reduced evenr. tree can be used tcprovide a simplified representation ofresults. Computational methods developetfor NUREG-1150 could support a mor~comprehensive analysis leading to esti ~

mates of the relative likelihood oIvarious scenarios and sensitivity tcimportant parameters,

I'hat was the likelihood of'ailure ofthe core support structure?

The 1 is- of ques.ions should beres"ricted to those deemed most importantto I'.eep the e.ent tree structure tract-able. In the develop,.ent of the Iistof questions it is suggested that eachquestion be listed on a separate largecard oz sheet of paper. ihis providesa con.'enient mechanism for screening andordering the questions in the accidentprogression event tree.

REFERE:ACES

1. Severe Accident Risks: An Assessmentfor Fi .e US .':uc lear Po-er Pi -nts,Second Draft for Peer Review, US

."uclear Regulatory Commission,

.':UREG-1150, June '.989.

2. J. M. Griesmeyer, et al., A Ref-erence Manual for the EventProeression A»alvsis Code (EY;lTREI,Sar dia >'at iona 1 Laboratories,.'.UREG/CR-5174, September 1989.

For each question, branches are developedcorresponding to outcomes. Tf physicalconsistency requires that the branchesof the current question depend on thebranches or parameters at previous ques-tions, cases are de.'eloped to reflectthis dependenc;. The case structure andreferences to previous questions arelisted on the card for each question.For each case, impossible and certainbranches are identified, and a subjectiveprobability point estimate may beassigned to the remaining branches.

Finally, descriptive material is listedon each card to document in greaterdetail the technical basis for the branchdevelopment, case structure logic,, andassigned branch probabilities (or param-eters), The set of cards providescomplete documentation of the acciden-progression event tree. The construction

Figure 1

NUREG-1150 RISK ANALYSIS STEPS

PLANTSYSTEMSANALYSIS

PlantDamageStates

ACCIDENT AccidentPROGRESSION Progression

ANALYSIS Bins

SOURCETERM

ANALYSIS

SourceTerm

Groups

CONSEOUENCEANALYSIS

ConsequenceMeasures

RISK RiskINTEGRATION = Measures

Figure 2

INTERFACES EIETWEEN RISK ANALYSIS STEPS

PlantSystemsAnalyal ~

AccidentProgressionAnalysis

Accklenlprprpcccrcn~rcrll lrcc

'PPS Tg ~ I JAPb 7 I

iPOS fJrAPb 5 j-

Source TermAnalysis

ConsequenceAnalysis

xson

I ...—,ISTG I

RiskIntegration

( POS~~IAPb oe I WSTG ~

PlantDamageStates

AccidentProgression

Sins

SourceTerm

Groups

Acr: ' PPOS~ +APb~ hS7G

Figure 3

CHARACTERISTIC OF ACCIDENT PROGRESSION EVENT TREE

~ Tracks the progression of the accidentfrom the onset of core damageevaluates containment loadingpredicts structural failure modes

~ integrate differingmodelsanalysis methodsexperimental information

within a framework of scientific expert Judgment

~ Sequential series of questionsmultiple branches

~ Case structuredependency on previous questionstreats all accidents within a single tree

~ Large tree can accommodatemultiple time periodsmultiple plant compartmentsdiffering expert interpretations

Figure I

PLANT DAMAGE STATE

Defines the plant condition at the onset of core damage

Combines sequences that behave similarly in the accidentprogression analysis.

Example (Surry): TRRR-RSR (fast station blackout)

T: RCS intact at the onset of core damage

R: Emergency core cooling is recoverable

R: Containment heat removal is recoverable

R: AC Power is recoverable from offsite sources

R: Contents of refueling water storage tank can beinjected upon recovery of AC power

S: Steam turbinMriven AFWS has failed; electric motor-driven AFWS is recoverable

R: Cooling for reactor coolant pump seals is recoverable

Figure 5

PATH THROUGH SURRY ACCIDENT

PROGRESSION EVENT TREE

StartAccrocntPro0ress»nAnallsia

Qwsuon I: Qwsuon 0: Qwsuon 10'. Owstron 12: Owsuon II Owsuon 15: Owsuon 15. Qwst>on Il; Q ssuorr I~,hcs ln40ruf 04tw ol Hast hernosal cootrn0 lor latusl Ics pressrrrs pohvs sucu T.l hcp T.IsGThl

~I UTAFT Ac torrsrl uore sosl Icp sea47 conlarnrssw elITTAFT openl Seel FwwslC~T

TRRR RSR

1010SGCHR ~ I 0

~etSCtORVnSIO neE.SGTR

OwsUon 20: QwsUon 2I: Owsuon 22:Td Hor LeS AC Poser ICS Ptsssrrra

Faurrrol Eert77 al VST

Oweuon 2L Owl uon 2IConlskns»ni Tires el

Presarrre~clare VST tscnarSel

Owsuon 2h owsuon 21: Qwsuon 22:Fn Cere AraL Zr Ws4r InOalAaal Qatlurae Cerkf elus.Vml us-Vml VST

Qwsuon SL Owsuon 5s:Fr. CD(~ AIIII Corehetsassi hs4saert

al VS7 el VST

I 0 J EaACPneES& LA

P1 r 25 ps4,ea7Idrnpr

IOACDOCSS

Ti~2 ~ aaa [10-2rOs Pl * .5sa

.255I~rl

10Ht.f Ceh

owsuon 25: owsuon SL owsuon 20: owsuon 20 owsuon ~I: owsuen eE oeeeuani2.. Qwsuon i5: owsusn as: owsuon ihAtpr» ssor4 Tips ol Stre ol Ho4 Preserrre hiss Es.Vessel Contatnrssru Cerrtranrr»rn AC terser La4 Hoe alwtr

Fattrrref Vssaet Sreecrrl ln Vsassrl al VS'7 04ars faurrre fauree7 Losel Sprals'I Hr Srrn»Esp4s4nl treseroe7 sl VS1

.257

ti 505+ Il,useneEVSE

ti < TCEifsl0PT ~ JN

~72L-ACP L Sp

Owauon 50 Owsuon 51: Ouasuon SI Owsuon 52: Owsuon Si: Owsuon 5'rrssuon 02 Owsuon 00: Owsuon 71:La4 swnl centainsraal Areowl el la Dsurrs Does prersps vorT Lace . Iseernar fteal

10ntuonl pressers h»s7 Fsttrrrel cere In cell Iei ~~~7 ccl Darnel 10atp~l 5aeu.urrorrunl contarnrr»ntConrut teel

~S~APIE ~ 1.12 tl1 ~ 1004 potu

.25i .005 A2eHys L-H2hrrr + .200 / CDS ~all

laeoCCI L CC I

ssL2~I.~neLCf

.02iF.IIT

Frgure 6

Question 21: Is AC Power Available Eariy?

(Is offsite power recovered in time to restore coolant injection before vesselfailure?)

3 Branches, 7 Cases

Case 3 applies to TRRR-RSR plant damage state (case structure requires thateither Branch 2 or 3 be taken at Question 10).

Early period 0.5 to 2.0 hours

Mean value d'or power recovery in Early period 0.565

Value sampled from distribution 0.614

0.614 AC power recovered during Early period.E-ACP,

0.386 AC power not recovered during Early period,EaACP but may be recovered in the future.

0,0 AC power failed-cannot be recovered.EfACP

FI9ure 7

REDUCED EVENT TREE - SURRYSTATION BLACKOUT

BYPASSCORE

0AHAGE

ARREST

ALPHA

MOUE

CF

CFAT

VB

RCS LATE

PRESSURE CFAT VB (XHCL. BMT)

580

0.003

YES

0.997

0.601

PT~ l

0.008

YES

0.003

0.599

0.003

BYPASS

HO VB

VB, AI.PHA CF

0.399

0.992

1.0

0.013 HI <>200 PSI)

YES 0.0

Lo

0.005VB, fARLY CF

HI. PR.

0.0EARLY CF

l.o. PR.

VB VESSEL BREACHCF CONTAINHENT FAILURECL CONTAINHENT LEAKAGEBHT BASEHAT MELT-THROUGH

0.987

0.203

YES

0.797

0.079VB. LATE CL

(INEL. BMT)

0.310VB, Ho CF

Flexure

8

STEPS IN ACCIDENT PROGRESSION ANALYSIS

OF THE CHERNOBYL ACCIDENT

1. Determine plant damage state

2. Define time periods

3. For each period, identify important

operator actionspiant system responsesphysical processes

4. Develop sequential set of questions

5. For each question develop

branchescase structurebranch probabilitiesdescription

6. Analyze the tree to identify

likely scenariosimportant issuescritical branch points

7. Seek simplified representations of results

!~'COLGGICAL A!~'0 sJ'''".IC Cc 3<'QUEiVCES

O'~AN C»'IERNOSYL A PORC i'0'iTR PLANT AGCI DENT

Vt Er SOKOLOVy I o» o RYA30V ~ I o A ~ liYAMGEV

F A 'rI >uu'lU.ROV 'I ~ A ~ SlVVC'!!;>XO '< X ~ TASr>EV

Institut e oZ Evolu'.ion~ry Anima3.:.!or..!helotry and Ecology,

J.":SR Academ;,i of Sc'ences

Introduction

!"he Qhernobyl Atomic :"ower 1'lant (CAPP) is situated in the

northexn part of t;io Ukranian Soviet Socialst Hepublic ~ 'Phe

accident, rihich occurred on

t ne products of xission and

April 26. 1986.caused @bout 3i55 oiaccumulated in the Block 4 reactor

activation to be discharged into the

atmosphere ;"he iall-out of radioactive substances brought about

a .",ups'.antial envi'onmen'.al pollution beyond the production ax'ea

(I) ~

Lar'c,;:cale de-activation measu es, land improvement

and disinte,;ration oi short-live ..adionuclides have currently stg;biii "Gd t.'.'e r 'd't'n si tuation.".rra luaxly, 4!'|v'1'z polxuvants

; ave ccme to be t. ion -lived adionuclides, namely, cesium-137

and strontium-'30 (:,ith a hali-lido of about 30 years/ and in uhe

10 km ~one azound the CAPE'. »xlutonium-23),240 (with a halx-life

oi 24380 and 6537 y.a s resyeciively) ~ Zhe total area of the2zones with a pollution level of 15 Ci/km for cesium~and over is

10 thouthand km , izom 5 to 15 Ci/km over Zl thouthand km

the pollution dens. ty ~or cesium-137 up to 15 Ci/km almost :vezy-

where provides th=~ limit oJ the life irradiation duse ox q5 rem

i.ro:."!external and 'teznal xzzadxatxon uo t!lv human .Og.ulutxon of

'e acc!dent re.„'iona.

The bulk o 'he fo'ests in Che polluCed area is in f'olesye>

where Che forest cover makes up 70% of the 8 territory, 65$ of Che

forest species are accounted for by coniferous trees, the rest bg

har d'rsood s ~

On the vinole, about; a half oi the polluted territory is consti-

t uded by natural corn;;lezes, includin~ forests and srvamps, v>hile open

.lain areas axe comp>~:tely taken up by agrosystems ~

During the first weeks and months after the accident one of the

main problems v>as protection of the population fzom ext ernal irMadia-

Cion and intake of zadionculides alon with local foodstuffs As ahuman

first step, during Che first days after the accident Che~Jvopulation

and cattle ~sere evacuated izom a zone of 50 km in xadius around the

GAPP+ &ming Che subsi~iuent period Che first priority objectives mere

evaluation and prediction oJ.'he radioactive pollution of fazmia pro-

duce in Che areas adjacent to Che CAPP, "nd in e. broader context, eva-

luation and pro'nosis of zadioecolo"ical and genetic consequences

of the pollution.

3'ffect of Radioactive .Pollution og the I'loza

In radioactive 'allouts, the arboreous layer is an efiective

filter, t;appi~~ Che bulk of the radioactive particles calling out

from the atmosphere, "n Che CAPP zone of Che accident the crown of

Crees initially trapped 606 Co g0$ oi the radionuclides t hat fell out<

Hence, the arboreous layer zeceived Che bulk of dose loads and disp-

layed concurrently i;rot ective functions in relation to Che organisms

d:.~ellis inde> the canopy throggh reducing the concentration of radios

ti e aerosols in the near-ground oi the air and the dose of exteznal

irradiation+

I t is ion mn Ch "u amon''boreous plantsy the most radiosensi-

tive taxonomic grou.::..". uze Gymnospe;.mae, in;articular> Pinaoeae.

:3epei dia,„ on ".he seasonal conditions and distribution of irfadk.ation

dose in Cime, che lethal dose for the pine Pinus sylvestris varies

betmeen lg Gy in an acute irradiation during the vc.'eta''ion period

to 100 Gy in a chronic iz adiation (2,$)o ln accordance mith the

experimen-.al:i~Ca avaiulble (4) it mas assumed that gine forests in

the ine around the CAPP v:ere to perish by the ~nd of 1987 in Che

entixe area mhere ch Qosa: e in the ecnifozs muuld exoeed 50 Gy>

and by the end o'9"8, mheze this dose viould be in excess of )0 Gy.

T."iis area mould ha:e amou. Ced to 2 Co $ thouthand hect3zes+ Actually

Che area . f Che dea;.ii o" pine forests amounted to about 400 hectares

(5), and at about Che same area Che ~crest mas dam:.ged. Thus, Che le-thal effects in pine forests maniiested themselves at higher doses

of the iz adiation o" the needles, mnile Che axea of their death

in the CAPP zone pzovod to Ue substantially less than ~ias expected,

Tne study oi thi mor1iho.. enesis of Q..ne ve~etative organs demonsti

ted that the midest ran-o oi morphological deviations was the one

at dosa;„.es of 50 Co QO Gy. In 3.986 virtually all such txees exhi-

b'ed a substantial .!e;~ression of gzo;.th p rooesses> their needles

mexe reduced in size, fell off paxtially oN completely and young

shoots v.ere Cmisted.

The mortality o" youn,: shoots duzin„ the spring period of 1986

in the majority ~f »ivies mas accompanied by develop ment ~f replacing

lateral buds that ozi;ina..ed from the dormant buda of brachyoblasts ~

On the top shoots forced in i/86, Chere developed 1$ to 2o and ocoasio-

nally up Co $0 largo buds among mhich there mas no dominant bud.

In such bed the majority oi stzuotures mere -igantic. In faot, the

cross-section of the ".talk zudimen. mas 15 times that of control'

;;hile thaC of the m zzovJ p >4 times, Duzing tile sprin ~ of 198'7 theat

buds ave xise Co robust young shoots. Some of these shoots carried

only the "Crossly ov ergzo mn scales of dormant bud s mithout ne e '~ 1 s .Some individual yoU-~:; shoots mere cleft in the apical portion, fozmiq

a raceme of 5 to 6 stalks. In dam @ed Creesp the dormant buds in

apical stalks moke u,.~ to compensate for the ones that died ox

irradiation.

A . a dos;~e of 25 Gy, in 198'7 tneze o. cuxred a considerable

'nczease in the umv~a, of needles on Ciie shootsp The length of

needsles in 1987 and 1988 in this case reached 100 t o 120 mmp

while a mass of 100 needles increased 4 to 6 and over times compared

mith 1986. In 1988 in these trees the avera e length oi needles somemh

declined, while its total mass on Che shoot increased 2 to 4 himes.

This indicates Nn ntensive increase in the zecovexy pzocesses

at sublethal doses cf i.radiation.The hardihood xo'ts in the Ch,PP zone mainly made up of Che biz(

(Betula -endula), " .,ion (Po.:-ulus tremula), alder (Ainus glutinosa)

and oak (q@xercus zcbur) pzovod by faz more tolerant oi zadiactive

pollution compazed .~ith coniiezs. The radiation dao:~e ox the czo.ax

of hazdmood sr ecies .:'s only manifested in the immediate vicinity ox

the mz5cked reacto at levels of radioactive pollution an ozdex

hi hez tnan similar one in relation to the de„ree of the damage

of conifers. In bizcl es, .;here Che dosage reached 5GO Gyp youngthe

apical shoots ztially died offp mhile by'id-August 1986, the fo-mostly

liape turned yegllom and fell off+ I3y autumn, necnosis of

some individual br~ches;;as recorded. In the spring of 1987p in

many birches at CJ;esi sites abundant flomezi~ mas recorded, e

some of the enezaI.ive oz ana of the birch v~ere anomalous in

structure< Some male and "emale catkina branched, vere twisted in

shap > and +qmo of the nthczs vrere neczotized , gy mid-summer

in auoh areas the majority oi birches acquired a peculiar coloration

the central,".art of t:;e 2,elf lamina remaininp 'zeen, v;hile the „'.ezi-

phezal pzotion bzight yello:,>,In theupper part of the crovIIn there de-

veloped some very luz .e dark-preen leaves. '.n 1988 the birches regal

ned their normal folio,„e.In a wide zan'e cf dose loads {ug to 100 Gy) in t!re one of

the acoident, radioac'ive r.ollution exerted no subarant ial eifect on

the communitiea of herbaceous plants.

The action of zalioactive pollution on t.'~e phytocenoaes was

oc mpazatively brief. A year after tire radioactive fallout the

damaged fozesta bepan to ze„'onezate. The crocus badly Damages by iridiation <vere zegenezrrtirr'. -'lome f avourable conditions for natural

'l

and artificial zecove "y of coniferous !forests v~eze ozeated.

Effect of Radioactive 'ollution on Terrestrial

Vert ebzates

No evidence oi mcztality oz migration of vertebrates under

the direct effect of inonizin; radiation ,iaa recorded early after t

accident. Still, on accc~~it of t he doses received by t.ie moat;

heavily poilu.ed areas neaz the station plant > there aze grounds

assume that some animals uhoae home ranges mere in the immediate v

nity of t'e CAPP duzir.; the accident', may have died.

Prior to the acc'ent, the area azound the Chernobyl A4P ,uas densely populated by h~mans, ploughed ug and intensively used

J.or recreatl one ''vacua lii on of t !ic human populat ion in a radius oi

$0 km from the plant i ~a garou'ht about some "protected" conditions

for .~ild vertebrates..".:o harassment fact:ora> the most importunt

amc~ other anthropo,„.ui c factors aif ecting the numbers and dispersa

of mammals and birds ideas no ion;er operative thouphout Che entire

area,

In addition, cess"Cion of Che til3.1~ of thesoil, the 5tandi~

crops that zemained in tact in i/86, Che evacuation and zemoval

of domestic animals consider..bly improved the fora in@ and protection

conditions for mild '~erbivorous mammals and birds>

'these "actors promoted t;ie survival of animals under t.he condi-

t iona of the unusually .:iarah un3 snoa-abundant winter of1986-1987'ccording

to aire"aft; censuses, aa of a~rid„'98'p the $0-lan zone

supported about 1)0 moose (Aloes aloes), )>0 to 000 roe deer (Capzeol>

capzeolus), uI; t;o 4~ deer ("ezvus elaphus)> 2)0-)00 wild boars (Sus

aciofa), about 150 foxea (Vulpes vulpes), 7g" to 800 hazes (Lepus

eurcpaeua) and a few i;olf families (Canis lupus) were recozded.

ln 1986 the abvve Jacuoza pzomoted the growth oi tne populations

of virtually all the „.arne ma:iunala and birds. The increase in populatioi

of 4eOicQeetpla'aried v~ith speciea. By thespring of 1988, foz exam

pie, the populations of moose and roe deer increased to a small

extent, while the numbers o~ v>ild boar rose roughly

eightfold compazed saith 1986. Similar Czenda were recozded fozcarnivozes ~ noticeable

The species composition of wild ve.:Cebrates shoed no changes'1

colony of b<;aver s (Castor f iber) remained atof the CAPP, and storko ze'urned to their nesting grounds todeserted built-up areas to form colonies at some sites.

The moat widely d'atributed and numerous mamHLalo in

me Muridae represented by several speciea. Hence, in the populations

>f these an~~la the primary ecological and radiobiologioal effectsiaused by changes in environmental conditions and appearance of the

.adiation factor were evaluated.

It should be noted that the radioactive pollution of the habi;

bats is highly mosaic and henoe, theix dwellers reoeive difierent

eadiation doses.

A study of the disrersal of murids undex natural oonditions

Chat the animal population ia primarily associated with habitat shoived

proyerties without de.;..idio on the level of rodioaotive pollution.

ln September 1986 at all the ylota irresyective of the pollution

level, the rodent ~vere breeding intensively, mature underyealings

being mostly involved. Animals from the polluted plots shomed an

increased number of ovulatin ova , c orreaponding to the number of

corpoxa lutea and determinin'hepotential number of embryos+

This increase Ngs accompanied by a considerable embrygnio

mortalhty, maich in Octobor 1986 was on the average ~$ (control6'5) for the bank vole (Clethrionoagrs glareolus) ~ In the ayxing of

1986, dec~eased litters as caused by embryonic mortality at pollutedof

plots averaged Igy the control and directly depended; gn the pollu-

tion level. Nevertheless , the population of rodents in spring

began to rapidly increase at all plots.

Accordin to long-term. evidence, at one of thee experimental

plots, the highest population density of rodents averaged $6

individuals/ hectare, v~hile in the August of 1987 it reached 600

individuals/hectare. /his is most likely associated with the factQd$ j Hlk

that standing crops were left~and that predator populations

Ref autm~»s and birds increased.

The overpopulation o" the habitats by rodents depleted the fora

ging resources and the lasting Chaw during the winter oi 198'7-1988,

which caused a numbez o~ Uu'rows to Ue flooded and a subsequent

spell of cold weather r brou, h about a sharp decline of animal popula-tionss

in Che regions. 3y t ".e autumn of 1988 the numbers of rodents

began Co gradually increase, but they only reached normal by mid-

-summer of 1989.

The sex structure of the populations under study o4er the entire

observatiov period showed no deviations and a rough male:female ratio

of 1:1remained.

The study or thef: eri."heral blood of the rodents dwelling in

polluted azea (dose rate oi 60 to 100 mR/hr) revealed a twofold drop

od "lycogen in the leucocytes compared with the control. Anot;her devi-

ation oi eh/peripheral blood was lpmphoortosis {an inoreased oontent o3

lymphocytes) ~

Investigation oi ti« bone mazzow in irradiated voles demonstra-

ted that the number of nucleus-containiag cell per volume unit was

significantly higher co;a,".ared with control animals, which indicates

an activation of hemopie is and supports the action of inonizing zadia-

Cion as a stzessoz.

A macroscopic examination of the liver revelled that in 68.0$of immature and 8$ .$% mature individuals the liver is flabby. Such

a condition occurs under profound degen.orative changes of parencchyma

cells and those of th{:or 'an's vessels.the

The above evidence ndicates that despite+va:eious deviations

in the or;;ans and tissue tnaC may lead some of the animals dw~

in radionuolide-polluted "reas t'o premature mortality, theQoeaQ-

~vnal .arameters vary;»t fain t!ie normal ran.e.

1.'valuation of th'., u" i e o 'fvdzobionts

The aquatic environ.::=nc h~s a s,iecific role to play in the re-distribution oi radionucl'es, since the ,";rocess of migration and

accumulation of radionucl'eo in the >cater aze by far moxe intensive

than in te"restrial ecosy=t;e.t.a. Hydzobionts are ex,>osed to external

irradiation from the uav,er "nd bottom sedimentsc j'.n addition, many

organisms are extern'"ll; ir: ablated from aruatic plants and fiouling.

According to model e.",ti;"..tes in the CA"P cooler pond the dosage

o2 external irxadiot'~ "o"i,i hydroblonts from the:;;ater reached

maximal values of 2 to "'Gy/day during the fizst days after the acci-that is o thoutii".-.3 ti.-.es hipher than the natural level

den

zadionuc:lides and radio: c' vo 1lecay of shozt-lived radionuclides causec

the level of external ir «'lation fzom w;itez to decrease 200-fold.

The natural level o" i."radlaiion of hydzobionts was roughly

O.OOOl zad/dayi Zn the c "oler, ond the ex.erna1 irradiation oi hydro-

bionts from bottom, ssdi.;~.r.ts at t:ie end of April 1986 vlas

1O to 2~ rad/day) i.e. 200 thouthanj times highez. Thus) tGking into

account the slow decline oi 'dioactive pollution, the bottom sediments

are an important factor cf dense load on the hgdrobionts.

A 1:azticularly s"'. o-.; i': act of external irradiation v>as suf--"e"-Q by benthic anim."-i.-;, .!..;..r.'.cula", molluscs. ter Uivalvesy

t .'te

Gy

extaznal~rradiatio~ ~c" .:. bottom s dissents is hi her thun fromuscan organs.HKzc~acco-di;i:;o 'ice ev'ence oZ 1888 van only 6.10t is notev>ort>i,"; th;.t the distributi,on of the i1radiation

Qogegde,~endin~ on incozpo e ".ed radionuclides in the August 1987

10

1'Ds hetero 'Gnvtl0 ~ '.0'n;.v'1'1co, A11 4,G 1'1.;9 i .;~s ~ "1 '~../ lD. ~

in ui1e .:,.lls and in the ~tie 0 ~ 10 ud/day.

~ 'alnst t"e bacL~;zo .nd od these dose loads u detei:loA" tc'.l conditii

GA the color>ies of bivalves of the Unionidae family, naczeshell, ~'i~s revealed

and Anodonta cy„"nea, a.d also the mollusc Qroissena»olymoz.'.Eal

At the,.resent time the population oA'olluscs Gf the Au:1 ly

tinionidae have overcome the de.'ressive "actor GA radiation mid are

recoveri1g their populati.ns anJ 1,ro;, uc t ively. Dy contr" st, «.1e .'oyu.

tion of Dzeissena as of i/8> continues to be de,:ressed, .;,hich is man

fested in Iec.:used corn.'iunities, decet."ated ~zoath, etc.'ll ze are Dcout p fish s~,=c 's in t."e CAPP cooler .,;ond. Alon..

"Pith Q,'OAe8 churacter105ic Gf t;Le .!'LJGA'1 ti;/ L'>A ".i)IO, oa!l rivers Quell

as the roach (l<utilus rutilus)»'ozcn ( -'~rca iluvialitilis),pi'se-ye'tuciperca

lucioperca), bream (Abramis Drama) catAish (".iluzus

Q lani 8) f etc ~ some Iar-Eastern f 'he s mere re 1eased i nto the coo lez

pond, n m ly t':1 silver car: (Hyyoghtjal."1ichthys molitrix) and big-

head (Aristichthys nobilis), QAiver caAp (1fyooy'hthalm. chthys molitzi~ rass carp (Ct eno>hry~odon idella) and "'meAican pond 'Ashes, i.e.t:1e channel catAish (lCtalurus puctatus) and buffalo (ictiobus sp ~ ) ~

'.~e ove all dosa",,"es ze closely associated:~Ath .i1e ec io;ical;>eculiarties of;he fishes concerned,; o'r '1 Gancey I.'rv Jo;"c rute

oA external i "radiation in fishes d1~ellin„ in the thici.cts Gf g

aquatic ve,„e".ation, stAch as tne ~.erch, i1 e (1GGx lucius) c'-'"1 be no-

ticeably 'n excess of 0ise internal irradiution. ine hi„i1est 9"ses are

characteristic of be11thophages of the 1.'er erations v'g"'~ and l987,199G -l)3$ t~e accumul~t ed dcsa„e for

IPNh@4M~ brw%4@opg )p~p ~

for the silver bream (Hlicca bjorkna), 11 tu 16 Gy for the bream

)eveloped ';;eil and by t.'s,.:id-t'une ly.') at ~~4'L':4 ~ (U JUys Qvme

individuals zeac~ied 8 len:;t.'i of '7 cm.

2he studies oonduc..ed in 1987 1989 in".!ic t.e t;.'iat; "he dynamic

ba>ance between the yroceaaea oi removul ~n3 accumulation of rudio-

nuclidea in iiahes in,the cooler yond of tie: Q~P~-'ua not oeen eatab-

l,ished ln ~act, in the 1~1<togha„'ea ( the silver cary) tii: content

o2 radionuclidea auostantielly declined and durini the summer 1989

the concen,,ration oi cesium-15~ - lp'7 v~' $ ,10 Ci/hg, In tjenthopha-

'ea the ra3ionculcide concentration also declined, uhou„"h in a

i.,"entler ~iattern, 'n the predatory iiahea (; lee-oercii) the situation

ia somei')hut mole complicated 4iuh peak noncentral'iona OJ. radioncu-

lides bein" recorded durir '; the atria„-a~~mer 1987. i)iiria,.: t.he

aum"er period of 198S-19S9, Oh', examination o" the ii.ueriiul or':.aua

-n difiereni;-ai~e individuals revealed a t.endenoy f~r deoiine.

Burin,; the. Inner 1989 the oon eniratio'i ( cesium-ljOy-1)'i'Jaa (W) ~

~ 10 Ci/l~ 8;-;;Bin'.lt 5olO Cia fOr li .LQ kiatural'iUQJQround i ~ e ~-ll

100 thoug h"nd times aa high ~

~.i'eneticStudios of the Conaeouences ox the CAPP Accident

for Animals and lants

The inveaii;;ation of,,enetic effects o1 inonizin'; radia4ions

.-'ormed alon„; the "oilov>in.. liness

ens2.tive to;.iv Gw',U JE;cta 3.OZ

is Pv

Qgs lac Dt J-OQ oJ 1cli30rdtvwy rsdko G

biolo" ical dosimetry oi the: environment;

-. evaluation ox tiiegrima-y radiation-,:;enantio effects of ioni"in ~

'.one anl beyond it.r:-diationa on t:ie i~opulationa oi planta arid animals. in the jG klD

-in~esti~ation in a number oi ~enerationa of t Qe enevic conae-

~yuences OJ t;ie effect of inonizin radigtiona on th.'virun.,".Cn~al

o Ujeciia ~

ii analyze the mutut ion;.rocess in QQtirul:yo,"vulutions u~,>lanta

and animals in the )(.'-km zone and around tne CAPP 2j ex,erimencul

..lots more set aside fo1 co ductin multi-year inte "r-'ted eco-;ene-

tic studies..a n these Zealots experiment.al ..iuterial i" collected 'or

9 to 25 "..cies of;iiid .lants and QQi;mls '..e cui.'tured, . i eaaril;,,gl fur'AE'~

rodents&and 3;osoihilas..in addition, in Q nu~~ber od.'«o ylots

rulio-sensitive labor:iiiory subjects are .1'iced, namely the s,, idev'ort

oi Clone 02 and inbred mice.

An txum>'le oi im;.lei;enuaiion of the firsr ai.„.roach is found

in ex, t: rimeQ, s 51th s ji deb'iol t QJ. cloQe v'2y i convoni.ent model sub

ject, "hic i is Qi„",'.ily radio-sensitive. '.i'.tis clone is a coloration-het e

ro y-'uus crusa> t:~t is revroduced,e~etg'.1vely. Phu esi;impute oi

sum~.'ic mut.'~rions is mide in the filaments of the s.a;ren filu-.ieuts of this iiont.

in 1337 .;;>ide:>ort;.lants v>ere,placed in, lot's '.~ith duse levels

o ) 0 f 100 iii 3 2~)0 ',R/Q1' e".de;.':oiximu i number o 'ut ut ions 1Q t .iis" c, er 'ieQ'i "i 1e'3ched 2)%..'un ut.'oiicluded "...;ii 'i','Qi J.icunt

;.Qci e ..se in t.»e mutaGion iumber as induced iy radii'.il ion o curs 't~t a dose rate of g mR/hr, i.e. about'.1 rad/day..i'Qe resultsoo.,a ned:;ave ooniiened data on toePai:„:!s sadiosensitit.'ty oJ Ines

,.liint, and rI.=" s.-.idemort can thus be reco:. iended lor .he Uioind~ca-

4 ion 0 radiation J'ollution 01~ hg envii'on'Qent,

'e l.vel o" mutation process in natural r;o.uiat;iona ol the

yerennia( .'~lant Arabido;sis thaliana common in i.h'ccident ~one v~'~s

gfhdit:aL pfAhLdopSih c4 S G1858LO;euec.ic moceL, and is notsdaya

midely used both in;„.en ticSand; dlt ic .ianC ea inee'in,;.:,'lo 'nve-

aCi.;aCion o~ aluNcion proess in natu'Ql q.o ul.~ciona Gi t;lia,,lunC

normally involves an esCimAed oi "embryonic leChuls"> i.e. un

ox r ~Jiat ion- illed y1;eCa in Che yoda uni 'leCerminaC .on

o~ Che ';-ropoztion oJ such nods miCh Denies seeds in x;e3.vCion Co,che

cocai nu.:bex nJ .';le,-ious under acudys

it haa been reVealed CnaC in .;1OCS :sit a ili;i; iniCial 1OVel

G~s ralioacCive,.olluCion Che l~t|'al muCaCion rate over 'hree years

'o.a~ins iai rly hi„sh '~:ll even inci eases,k,::a,":iCe Che,;e.ie. -1 Isolineoi.'!>o

levs1 o"'ndioaotivitp '> sevosnl d orrs ol t'rise ~

toor in.,'Cdance >ox 1'ocul" Cion 1 Che Bove rGCe iQ L~)66 'ci ';ar'128 'rradia

Cion iaa 2'H/tu., in 1987 2,5 mR/hr> in 19'38 2,0 m:V'lr. !".le rico

Gf .'ll ".lanca '",iCh GJfspri! 'e;re;~Otic't E'Qr el.".Jryonic lech'Q.a v.'buried(

mi"4 ears Z,:a eCC Velyt12<2 - )+1'.o> 27m~ Qp~. and Ols0- 4.9's~ ..'l Cn'..lCrn

'c,:.:uJ.aciona op: o >'. oi leChal muCatione are r co.~:led..n C.le .oilulu-

'.uJhonucliles (about'G ming/hr an2 lower in l)9'" h>0 level of muca-

C'on oad in.au|'secguenC;ell,raCions Did nc:C i3'l'cz si- n fi.c:~lt,ly Jrnm

-tel~ level GJ mu'aa„'eneais in conCxol yo;;ul~"iona.

M C'.t ~ -t;a~Ce nx,;;en Cxc,a~~cote n m.rods rvv~:iles !,.~.. one

r~"e,us C;in;ananmal aber..atinna in pullu:ae.k plOCa i. a';.vn'i.'io~nCi~

ai,,ae".. oom:;need sit!a r:a. ooiatxol, !ae r!rodrto'eta'ied anal; oi.; o:i i.'re

.v ~ . ~

,'t'".-i. 6 tc t xi ecba " .-nvuJved 'lkidae QLt 00 g..jta 'illlanuJ.tla) E 0;.'"1'b.'03 la';.'.iree "lnCa:ki|~erent;;arith ~t.a;.ect -Co C!le dnae 'aCe ~i,anm-iri.aQia-tinny. 'lnt' ~3sl CO <.'sip aR/'nrs r lnt 1" 1 Co 2 m</hr> '..'lnt gzy~4 Co:

ciona .„:- o~>i:~ in .;loca mich a ~air.1y lo"; level of .nit iol yollucivn bJ

>e shiae3.8 shovels-no 4i,;ria Gf r:,3iusion aicLuoso.

'Lli tiie.'.~ales, were orossed to laboratory exiles. ihe ..iveru„'c si~eof. t;:ie litter: diJ not; v..ay wit;h «-'lots and was 8.5 youn- .or female.A: t.otal of 2000 descendant were ob::;ined. -"he study of t.'ie rare ofdomiriant -le."hhls in males revealed no dil'~erences between animals

from different »lots ..Analysis . of the rate of anomalous s;~e.'m heads

"'-in" a~iles did: not reveal'ny substantial dif~erences between tive

j;rouj~s eit;lier, A study ol rhe rate, of reciprocal tf.anslocationsl.

demonstrated tiiat in the mioe captured in ~ lot XI and .'.1ot Ill, the

transloc~tion rate is increased:(0+)'7 and Oi43Pi ~ainst 0,17 in»,I"lot.l)»

,,Conclusion

A'tl~~.,;result time: only the moa';,era..raI an «lysis oi ucolo„'.ical-.chanc„'es in t««e irr:idiuted ecosystema in t!ie'e,,:ion. ol the CAP,

c~ .be -„"iven. Xt,-should bo noted tii8tdoses in the accident-ri ion .

are by far:in excess o 'he. iintural radiation levels.,«v..".h«c3 .;:jaime I'i«ae,t.~ere is 8:.~;roapect f.'or re;,'eriratioa.".og, cviliie-,'-: .

1',':.'--, r.,".us-trees:-': rue'~oaliy't)uou»hout the ent'~e.'.one 'o~ rudio~cti.ve,',; o1'ution,'fround tlirt~.CAPP.

equi.'i': ' -: .. Bcolo,'',"i..'Gl-.-":,'.":::-,-,';,.''-'.'-,<"ntjie. w~iole -'t'«Le+"0Gle'«if rie«,'at've conse~„uenees- oi . the i.: .adi'a-,'-',: 'ti.cin'I'Z cunifc"'.>us foiests. '; ~if.'eoted 2 t, g'km i«rci.. tiie

'„«

.s oui'oe,:,:=3c«sjii te:.the': hupe scale:.o~.:I!ic.."cciderst ~ '.ucti u« ..exi'ent

o~':iore™t gaum:.i' is f.'ai'rly- common in: t'e v'cint.ity oZ lax~„e in-:.,', 3u.;t;ri.al .. c..m;,;.lexes exploited. i.'o~ many,'years even Iiit bout

"a'ec'i-der>ts.'„

«"0411days some. 'Unprec9delited'roblems inhere, posed - t o: ~o estry:, ~orkers involved in; t.'ie: liquidation of -one -CAPP "ccident-,consequonoes '" .

, b;.'::a' o~nado 'iiic;i;if).'ected.'. tiie ..:cnst~r««. g«art of'..'t ne .'0...km:- "one,

IIII hicksf ll and broke lar.'e nuibers of trees. 0learin» the roads

'an1 outtiaps in the forest mould require of radioecolo„ists elabora-

tion oi nev> recommenda;,ioni for " emoval of obstruct;ions under coondi-

,tions of incxeased radiation background.~ ~ y, (, ~

.luch concern is caused<mushroom 14ckirg in',. t ie ~one ox

,"a~ioactive pollution by (des;.'ite the ban), «;art iularly by those <people.,

who r.:turned '~ithout g~ernission to evacuated sett lements (over

1000 j~eople). l'he cont ent .N raddionuclides vari the mushrooms as oj.

1988 c.:ian.. cd consiQerably curn@ared vrith 198/, o~ t'heisoto,res, u y

G'esium- isotopes - bein~ i~resent ~:1'he overall activit> of the muohrooms o:

all s:ieoies 'ncre'~sed b„" lOX. The inters ecific differences reached

--50 tii'es in 197'7, v~hile in 1988>~'timey ';~ei.» tvo orders.

Oa t.'i.e:>hole analysis ol', thc data vbtairied indicates "iat 'n

,-'. t'"." bulk o~ t'x~ area the levels of radioactive pollutions of all muah-

.,'Noom5 .-:--do not fi" into: the approved st'andards wit,. <~oilution den-

sity oi 5 Ci/km and over, md for m8n,"g''mushroom .;,.:cies t;ic cont.en-

~'r'ation:o'1'.-cesium-lg'7..:is in excess'f .t;rie /to.:idard even. at . a pollution

I,,'.:density oi', over l'Gi/k .!

I

", '".'ie bulb.- of te..re:-;,rial vertebrates sho~s no, degreio'n;. ex"ec't o»

,-'ioni'-.kir,,'adiations ~ ,"resu~.ably',:: in th,'':. dead'.foreat ':area, some ani-

:;„,mals:- uncayable oi rapid movements and ieith sirall individual clots

;"-died of iae ioni~inp rodi'ations, but this. is on j,a hyj.otl.eels. >n

..:som 'reas shiit:"~, a higii radiation level 'disturb=.inces in t!ie hemogxetic

', system and:r~~zoiuction isI:re recorded iii Iluridae. Still:,. t.:ie .po,';uxa-

'".tions, of, mu ids',in these aieas var'y.sync ir'onously vii.t;i t;ie'orot;rol-.

"he irradiation doses of: a~juatic.or>ani'sos. even.ip., ti~e cooler

ond o'the CaPP" are not beyond the dose ''ange ~bere the „"opulatiol

t; - of 'he 'dy.-iamios of . accumuiatiori oi riidionuc~Je'rishesi;. 'lhe ''''0>'c3.'.ci;j' 'i ' c

lides n fisnes as a inaction o~ t;.' ec~lo;ical:rou;:s;."s re-

ve leds l.-')re.rator~ fishes tne concentration o~ cesium in t.ie tissues

and or 'ans is an order laic„her than in:~lancet-,hoes.

In the cooler yond of th/CAPi'urin- tne early st'„e abater

the aocident deterioration of the condition o" trre ooiorries of

some olluscaa sr.acies @as revealed. ito;.ever, t;o Gate, t hese ani-

~als hav.: be en able to coq.'e vri';h the 3e,.ression.

Desyite the hip.'i overal doses of ir.r~3iation the iisrres in ti]e

cooler yond breed successfully and „ron ~

"rior ~u tire Chemo 1 accident it '.",as revealed tha; at a dose.and over (<<,~g QOr dIP)

rate o 4.1 rad/day Ci onronic az.~3iar,i~n ce.tain „'en *tie effectsa e reco. ied such as lioturb.'noes in. t;~e nu.aber and struct;ure of

;:.e c!;romosczes, different sirecies mutations oX yiants and anirrrals,

distu"dances associated cite chan'es in fccunditp, etc.) but even

such levels bric bout no sub;..tant ial .-„:eiretie cc.rrsec,uirrces "or~

trr~ ir,:adiated natural communities. 'he> ecolo~~ic Gi c'.lan~

,es as~cciated i~it'h t;he removal of sensitive s~iecies and r«-a:.r~~e-

ment oi the "t;ruoture of the irradiated communit es be„in at hi„.her

dose rates of chronic i."radiation, namely, 1 rad/de and over.

ti'e . resent time ti;e studies oi t.: e ex."edition <;~ t, e t're '.iSBi'<

Academy ~f Sciences become 'ncreasin lp experimental. '.i-'neir objec-

tive i., t r g'"n an '.nsi~nt int< r~dic;colo;ical, r,;diobiolo'i'ai

und;ene.ic c~ .segu noes oi t."re C;:ernob;gi ace~den., on t.re ilora and

'arUra.."..re lorr;-teim ecolo;.ical and genetic stuiies in t;he ~e ion

c t t:e Cirernor,,"il CAPP can, accordirr ~ to the leaders of t;he uc;;e.'di-

tion, serve a basis for irrt ernat i~nal coo,.er:it ion of si ecialistsoJ many countries develoyin,; nuclear ower en,'ineering.

For a de 'ai led familiarization with the evidence aud conclusions

oX the researoh of Soviet biologists in thi. remi on oX the Chernoby 1

A PP ~ ki scus sion und 1>1annin: oi future stadies the l irst 1nterna

t ional Conference on "Biological Aspects of the Consequences of

t;!e Qhernobyl APP Conference in September l999 in the settlemeat of

,.ov'er engineers Zeleny Mys ~

3 f.".E'T. R "!Cl':S

1..'."rael Yu. A ~, -'etrov V. A ~, Avdyushin 8 ~ 3 ~ et; al . '";adioact ive

gaol iuti .n o" t 'ie natural environment in the zone oi the uoci-

dent of the i"hernob .1 Atomio Power !'lant ~ ileteorolopi ya ihydroiho iya, 1987, 2 > 5-18 .in Russ ion) ~

2 . 'likhomirov ~ A ~ Ef2cot of ioni zing radiations on coo o1 op ical sys-

tems . ioscoll y Atc mizdat 1'uv 1o 1972 (in '.<us sian) ~

Sgarro;~ A.H., Bchairer T ~ A ~, !Ãoochsel 1 A.M. Tolerance of A.nus

ri 'da trees to a ten-year exposure to chronic ~e»~."-irradia-

tion from cobalt-60 ~ Radiation Botany, Q, 1~ 1965>i'' ~

i.at.'Gb911'. 1',::.llshenkov N.N. et al ., 'the injury of fore st'rboreous

1ayer from an acute pa~~-irradiation in dif ferent, 'i xono ohases.

Dokaldy A l i SSR, 1980, ~2, $ > 766-768 (in:.)usa ian) ~

1zrae l Yu. A ~, Hoko 1

overly

V< G., Bokolo'; V..'!.:~ e.~ 1~ Zcolo'ical cons!

quenc e s oX t.':~e radioactive y o1lut ion of natural environ/ment in the

regions oi t e C..~ernobyl APP Atomnaya ene ';"iya ubl i, 1988,

64f 1f 28-40 {in lkussion) ~

1989-10-19 Clean-Up after a Nuclear Accident

By Jan Olof SnihsNational Institute of Radiation Protection

Stockholm, Sweden

Swedish experiencies after. Chernobyl

In May 1986 the farmers in most contaminated areas were recommended tocut the grass on the fields and take it away for. disposal. This methodis justified even at relatively low levels of contamination.

However, this recommendation was later on (some weeks) changed andinstead the farmers were recommended to cut the grass higher up andthen use it. This was newly formed grass and most of the contaminationwas on the lower old part of the grass. At some farms the decontamina-tion was made by ploughing.

The fallout from the Chernobyl accident occurred in Sweden just beforethe vegetation period which limited the consequences for the farmers agreat deal. If the accident had occurred a few months later largeamount of ensilage and hay had had to be destroyed and disposed of.

Measures have also been made to reduce the uptake of caesium e.g.fertilizing with potassium. In some extreme cases also deep ploughinghas been made.

Swedish policy for clean-up after an accident

The purpose of all decontamination and cleaning up after an accidentis to make areas, constructions and equipment available for theirnormal use and to prevent radioactive materials to be spread anddispered in the environment.

Accident risks to consider are not only those from nuclear powerplants but also from nuclear powered satellites, aircraft andsubmarine loaded with nuclear weapons and transport with spent reactorfuel. A reactor accident outside one's own country is also moreprobable only considering the number of reactors being larger than inone's own country.

The situations at which a clean-up procedure may be considered are thefollowing:

I

the radiation levels after an evacuation are so high that relocationis not possible. The decontamination must be made by speciallytrained personell.

the products from fields and forests are expected to have highercontamination than the established intervention levels'hedecontamination will be made by govermental specialists by thefarmers and owners themselves.

Criteria for decontamination

In the normal situation with the source under control, i.e. the sourceand the operation has been planned in advance, the ICRP'srecommendations on justification of the soutce, optimization of theradiation protection and the dose limits are applicable.

In and after. an accident the ICRP's recommendations on justificationand dose limits are no longer applicable. Instead the countermeasuresshall be justified to give a net positive benefit for the individualand the society and intervention levels instead of doselimits areused. ICRP has also given special recommendations for these situations(ICRP publ. 40). However they were not easily applicable after theChernobyl accident at long distances from the accident as e.g. inSweden. Di fferent values on intervention level were used in manycountries in Europe. Therefore after the accident great efforts havebeen made on international level by IAEA, NEA, MHO and others to tryto harmonize between countries. This work is still going on.

In case of a nuclear accident in Sweden or abroad the possibilities toevacuate people in parts of Sweden to other parts with acceptableexternal dose levels are considered relatively good.

As concerns food contamination it is cons'dered plausible to followICRP's recommendations in publ, 40 if the production of food in mostcontaminated areas is restricted before decontamination of land ismade and alternative methods for farming and use of land areintroduced. Increased import of food might also be necessary.

In Sweden work is going on to find appropriate criteria fordecontamination. It is quite obvious that there are many factors thatwill influence these criteria and a decision on decontamination ornot. Psychosocial and economical effects are some of these relevantfactors to consider. It is important that there is an internationalco-operation in the works on criteria in able to reach consensus andharmonization.

Criteria or intervention levels for. decontamination and othercountermeasures should also be expressed in measurable quantities orat least formula to do so should be given in advance as e.g.concentrations of radioactive nuclides or external gamma dose ratesfor plants, ground, food etc.

Priorities

In case of large areas being contaminated it is necessary to makepriorities in time and objects.

It is recommended that the first decontamination is made to stabilizethe contamination as to prevent a spread to clean areas and drinkingwater reservoirs. The timing of decontamination, however, shouldconsider the practical benefit of a delay until the shortlivedradionuclides have decayed.

The priorities in choice of objects should involve the followingconsiderations:

— the radiation level

— the population density

the use of land

— psychosocial factors

- economical consequences.

At a low radiation level the decontamination will take a shorter timeand need less resources in terms of specialists and advanced equipmentthan at a high radiation level.

Decontamination of population-high-density areas should of course begiven high priority. Evacuated areas should probably have lesspriority than non-evacuated areas where people can stay because thecontamination is not high enough to justify evacuation, Specialconsideration should be given to those groups of the society that havespecial habits of living or diets e.g. lapps, farmers and hunters.

The use of land influences the priorities in a significant way, thatis, however, difficult to predict in general terms in advance. Landfor food production and for living generally has a high priori ty ascompared to parks, gardens, recreation areas etc. Forests may have arelatively high priority because of the long biological halflives ofradionuclides in many forest ecosystems. Selective decontamination offorests close to settlements may have a high priority.

Hany possible psychosocial consequences have to be considered inconnection with decontamination. A long evacuation period is veryexhausting for people concerned and influences therefore thepriorities in choice of decontamination in time and place. Frompsychosocial point of view it may be preferred to decontaminateprivate gardens before public parks. A delayed decontamination canalso influence the employment.

The choice of priorities is also influenced by economical factors likeproduction, exports and tourist business. There are many examples ofthese effects after the Chernobyl accident in and outside the USSR.

The responsible authorities should make preliminary priorities inadvance considering some possible consequences of accidents in owncountry and abroad. The priorities should not be compulsory but beseen as possibilities and be updated after an accident has occured.

Hethods of decontamination

Firstly, a decision has to be made whethet a contaminated area shouldbe left for selfdecontamination, the contamination should bestabilized or if a decontamination should be performed.

Areas that are left for selfdecontamination should be closed which isexpensive and therefore selfdecontamination should be made only forshorter periods and for areas with low priority.

Stabilization is important to avoid a spread of the contamination.Contamination of surfaces that are difficult to clean may bestabilized by incapsulation.

As to the special methods needed for decontamination there is nogeneral overall judgement of the applicability of various methods inSweden. Therefore it is recommended to judge and develop the methodsagainst the following needs:

— methods to estimate the size of the contaminated areas, theradiation levels and the kind of radionuclides

— plans for sampling to follow the decontamination work (e.g. numbers2of sampling points per km )

— applicability for various objects

— costs and time ( time-schedule)

— needed equipment

need of personnel and specialists

influence on the object/environment of the decontamination in shortand long term

radiation protection of the personnel

— volume of the chemical and radioactive waste

The recommendations should also include alternativeland and growing of alternative crops.

methods for use of

Radiation protection of personell

The ICRP's recommendations on limits for. workers in normal operationsare not applicable a priori because the accidental situation is bydefinition an uncontrollable situation. However. after some time afterthe accident it can be argued that the situation in practical sense isunder control. Special attention should be given to the fact that somepersonnel is used that normally do not work with radiation. A lowerlimit may be used in the first phase when dosimetry and other parts ofa radiation protection system are not well organized and may bechanged to a higher value later on when the protection system isdeveloped.

Transport and disposal of waste

Chemical and radioactive waste have to be taken cared of, transportedto somewhere and disposed of. Appropr.iate lines of transport andplaces for disposal have to be found consider.ing the short and longterm aspects of protection of man and his environment. Mould thenormal conditions and requirements for transpor.t and disposal of low,intermediate and high level waste be appropriate also after anaccident?

Organisation and need of resources

A preplanned organisation of the decontamination can have largevariations in different countries because of organizational traditionsetc. In Sweden the responsibility of a decontamination may be on alocal or central authority or the government itself depending on thesize of the contaminated area. Training, education and control will bemade by "a central authority. The need of personnel and equipment willbe analyzed in advance and the possible use of ordinary laboratoriesand equipment should be examined. Systems that are already built upfor control of radiation protection like dosimetry systems, whole bodymeasurement routines, data register systems etc. should be planned tobe used also after an accident.

Rights and obligations after an accident

Rights and obligations at and after an accident is regulated by laws.However, after the Chernobyl accident it was found necessary toreconsider the laws because they do not cover all situations thatoccured in Sweden after the Chernobyl accident and that can occur inthe future. It is quite obvious that serious contamination of largeareas can happen even from long-distance accidents. Probably theindividual rights to control own property has to be abandoned againstthe interests of the society to protect people and property e.g. bydecontamination.

Need of development

Some areas of special interest have been identified where there is aneed of more knowledge of decontamination and other mitigatingmeasures:

-" methods to minimize the transfer of radionuclides from ground tofood e.'g. deep ploughing. Mhen and where is it a good method? Other

'ethods mean different ways to cut the grass or seed, or to addmaterials to ground that bind the radionuclide e.g. claymaterial tobind caesium.

— methods to mitigate the consequences of contaminated forests andtheir wild animals. Binding material for. radionuclides in supportfodder to animals can be given as well as clean supporting fodder.The seascnal variations in uptake can perhaps be used moreefficiently. After the Chernobyl accident these seasonal variationsof diet habbits of reindeer were practiced with good results.

decontamination of lakes contaminated vhile frozen may be made bytaking away the top layer of snov if any. If the lake is not frozenthe uptake in fish may be decreased by adding potassium and/orcalcium to the lake. Studies are going on and vill be continued.

studies on the problems of decontamination of cities vill continue.A special problem to consider is vhether washed-away radionuclidesfrom valls and streets by water will cause problems as radioactivevaste somevhere else e.g. in the sevage system. Also the risk ofcontamination of drinking vater system as a result of decontamina-tion a city must be considered.

THE USE OF TRADESCANTIA CLONE 02 FOR

BIOLOGICAL DOSIMETRY IN THE REGION OF

THE CHERNOBYL NUCLEAR POWER STATION

Nilova I.N.,Semov A.B.,Taskayev A.I.,Shevchenko V.A.

N. I.Vavi lov Institute of General Genetics, Moscow 117809 CSP-I

To obtain express-estimation of the biological action of

radioactxv contamination and to carry out biological dosimetry

under complicated radiation conditions, we performed experiments

with Tradescantia clone 02, a well known genetic te t-object

highly sensitive to ionising radiations. Mutationa 1 events in

Tr adescantia were recorded in stamen f i lament hair,:.> (SFH) by the

change o f ce 1 1 co lor f rom dark b lue (heterozygous state o f a gene

controlling flower coloration) to pink (homozygous tate) . In

June-August,1986, daily counts of mutations induced by r adiactive

contamination were made. An ana lysis o f the mutation process in

Tradescantia in Moscow (control) was carried out in para 1 lel. For

three experimenta 1 variants the leve ls o f gamma-ray ci-ntamination

made up '5, 5 and 0.Z mR/h, respectively. The yield of mutations

averaged for a per iod of June-July turned out to be equal to 1.1,0.6 and 0.2~/, respectively. Thus, the level of mutation process

at a dose rate of 0.3 mR/h does not exceed the contro 1 value. At.

dose rate o,' and 15 mR/h a genetic radiation ef feet is observed

signif icantly exceeding the contra 1 level.

In 1987 boxes with Tradescantia clone 02 were exposed to hig-

ger dose rates of gamma-radiation: 20, 100, 200 and 250 mR/h. The

exposure continued for three months boqining f r om May 25. About

0 000 stamen f i lament hair were analyzed. On the who le, the mu-

tation dynamics i" qualitatively similar to that ob erved in 1986

a t the highest dose —ra te — 15 mR/h. 4 c lear max imum o f the f re-

quency of mutational events xn SFH is observed 15- ~ 0 days since

the beginin of the experiment. Later on, approximately on theil

~0t~~~ day, the mutation number reaches i ts minimum a f ter wich a

gradua 1 increase in the leave 1 of induced mutations zs observed

up to the 65th day since t he start of thc experiment. The recur-

rence of a maxima 1 mutation f requency with! n two years may be ex-

plained by the substitution in the course of the experiment of

zn f lorescences f r om leading shoots by those f rom advent i thous

shoots appearing later than the former ones.

Fig.3 showes a relationship of an average (for the whole ex-

perimental period) level of induced mutational events to the do-

se rate of externa 1 gamma-irradiation. It is seen that the average

level of mutations increases up to a dose rate of 200 mR/h and

with futher increasing he dose rate a stationary leve 1 of muta-

genesis is observed. It should be noted that at acute ioni ing ir-radiations a dose-response curve is usual ly obtaned with a ma..i—

mum of mutations at ':about 100200 rad. An abso lute number of muta-

tions at a maximum per 100 FH does not usua 1 ly exceed 10-15. In

our experiments the level of mutagenesi at a dose rate of 00- 50

mR/h reached in some days ~0/ and an average level was equal to

1 /. It follows that at chronic ionizing exposure a maximal fre-

quency of induced mutations is higher than at acute exposures.

Thus, the analysis of the mutation level in plant populations from

Chernoby l Nuc lear Poaver Stat@on permi ts us to predic t higher le-

vels than those expected on the basis of acute exposure experi-

ments under laboratory conditions ~

1989-10-19 Strategy for Handling Contaminated Food.A Nordic Proposal

By Jan Olof SnihsNational Institute of Radiation Protection

Stockholm, Sweden

Some time after the Chernobyl accident a Nordic Task Group began awork on a common strategy for handling contaminated food. The work isnow in its final phase and the main principles are summarized in thefollowing.

General philosophy

As a general rule and principle attitude of authorities responsiblefor food, contamination of food shall not exist. However there areinevitable situations when this principle cannot be followed as e.g.after an accident and the realistic level of ambition is thereforethat the contamination shall be as small as reasonably achievablewhich involves considerations of health, costs, phychosocial factorsetc.

It would not be reasonable to discuss new criteria for food each timethere is a need for such criteria. Therefore the authorities shouldgive permanent intervention levels which are applicable as a lowerlevel at all times. In emergency situations, however, it is necessaryto be flexible and the intervention levels might be changed. Duringthe first days after an accident it is also necessary to releaseresources for the most urgent efforts and therefore predeterminedinterimistic intervention levels are introduced to reduce the need foracute decisions, to increase the possibilities to manage the situationand to avoid an information crisis. After some time these levels willbe reconsidered and changed, probably decreased, to other levels whichmay coincide eventually with the permanent levels.

These levels will occur at various times and conditions and thereforeit is necessary to structure the characters and phases of the accidentsituations and countermeasures.

Analysis of the accident-situation

Releases

The consequences of an accident and need of countermeasures is much

depending on the type of accident in terms of releases. A usualclassification is gases, aerosols and hot particles. Releases in airare most probable but releases in water may occur. All these releasesexcept noble gas may result in a contamination of food by directdeposition on vegetables etc. or by uptake by roots and by uptake byanimals.

Effected areas

The accident situation can also be classi.fied after the effects invarious geographical areas.

The »ear-field (zone 0) can be defined as the area up to a few tens ofkilometers from the accident. In this area there may be a very heavycontamination of the ground and various detailed regulations on foodproduction etc. are irrelevant. Instead countermeasures may benecessary to protect people from acute radiation effects (non-stochastic effects) or stochastic effects from high radiation doses.There may be a complete stop of all food production in the area.

The far-field will be areas more than 100 kilometers from the accidentand consist of areas directly contaminated (zone 1) and areasindirectly effected by import of contaminated food (zone 2).

Time phases

The accident situation can be divided in time phases. The normalsituation is the time before the accident and from a time after theaccident when all countermeasures and restrictions are terminated.

The phase of observation (phase 0) is the time before the accidentwhen there is a technically unstable situation and an accident isapprehended to occur and the time during and shortly after theaccident. No measures are taken to reduce the consequences even ifpreventive measures may be taken. The phase 0 is characterized byincreased emergency preparedness including measurements. This phase iscompleted if no contamination in air and on ground is found and thesituation passes over to normal or if contamination is found thesituation passes over to the acute phase (phase 1).The phase 1 is characterized by measures based on predetermined plans.Phase 1 is initiated a few hours after the accident and will continueduring maximum 30 days. During this time the actual situation and itsseriousness will be clarified as concern the consequences for food.These consequences will be very much depending on the season. Thephase 1 is characterized by time pressure. Rapid decisions have to betaken on uncertain basis and facts, there will be a great need ofinformation and predetermined measures, criteria, communication linesetc. are needed.

The late phase (phase 2) is defined as the time when the actualsituation is known and there are data on the environmentalconsequences. Phase 2 is initiated a sht.'rt time after the accident andbefore 30 days have elapsed at the latest. It can continue for severalyears until it passes over to the normaj'ituation. The contaminationof food is mostly by uptake by roots and in animals even though someby direct surface contamination. More accurate data will be availablecontinously and the need of information will gradually decrease. Long-

.term solutions will be decided.

Analysis of countermeasures

Heasures to reduce the contami»ation of food may concern both directand indirect contamination. By direct contamination is meantcontamination of the source of drinki»g water and surfacecontamination of vegetables, ftuits and bert ies. Indirectcontamination is caused by uptake in plants via roots and in animalsvia fodder.

The countermeasures concern the production, the marketing and theconsumption.

Production. The measures include prevention of uptake, transfer andenrichment of radionuclides in the production chain. This is made bypreventive measures or by decontamination.

Preventive measures against direct contamination are mainly ofinterest in the phase of observation (phase 0) and in the near-field{zone 0). Example of preventions are covering of vegetables etc. anddelay of planting. Preventive measures against indirect contaminationconcern cultivation of soil and plants and of lakes and regulation ofthe fodder to the animals. Examples are addition of lime, potassiumand nutriments'to soil, reploughing, use of one-year plants andcutting hay at a relatively high level. Lime and potassium can also beused for lakes. Contaminated fodder can be completely forbidden(pasture prohibition, for phase 0 and I) or mixed with other fodder.

Decontamination of directly contaminated food is made by e.g. washingvegetables, fi'tering drinking water. Decontamination of indirectlycontaminated food is relevant mainly for milk, meat etc. from animalsby additives to the fodder to bind the radionuclides or use oflowactive fodder to increase the netto biological release ofradionuclides from the animal. By preservation there will be time fordecay.

All the various methods for preventing and decreasing the activity infood should of course be used on the basis of optimization.

Marketinq. The purpose of restrictions on marketing food is thatpeopIe we'll feel safe when buying and consuming food. The main measureis prohibition of import and marketing of the food of interest. Thisis made on the basis of an intervention level and the judgement iffood should be prohibited or not is based on measurement or theprobability that the activity is lower than the intervention level.

Other measures are storage, dilution and replacement. By storage theshortlived radionuclides will decay. An example is making powderedmilk or cheese that then can be stored. Dilution is used to spread thedoses among people, individually and collectively. It is a method thatis questionable from radiation protection point of view because thetotal collective dose will not change but it may be justified onpsychosocial reasons. A condition however is that other methods havebrought the doses much below the intervention level. Replacement offood follows prohibition of food but could also compensate a decreasedproduction in a contaminated area.

Consumqtio». measures co»cerni»g consumptio» may involve regulation onor a8vxce For choice of food and methods for cooking. It may e.g. bejustified to advice people not to eat directly contaminated vegetablesa»d to drink rai» water. Advices on diet .are relevant mainly for foodproduced at home like vegetables, some meat and milk, and food fromhunting and fishing, berries and mushroom. The advices tell how muchone can consume of the various foodstuffs without exceeding the givenintervention level.

Measures to reduce the i»take by various methods of cooking are onlymotivated at relatively high activity levels. However there are somesimple methods like washing the vegetables, throwing away the waterafter boiling meat etc. There are, though, negative health effects byreduction of vitamins, minerals and salt and other negative effectslike reduction of taste of the food.

Intervention levels

Intervention levels are predetermined values of dose or activitythat are one of, the basis for decisions on countermeasures to avoidgiven levels of dose. As regards food the kind of intervention levelsare

— primary intervention level or. reference level expressed in mSv peryear

intervention levels for marketing expressed in Bq per kg.

Intervention levels for the production will depend on the methods forproduction and decontamination and will be calculated on the basis ofthe intervention levels for. marketing.

The intervention levels for the consumption will be based on theprimary intervention levels.

The calculation assumes given dosimetric factors mSv per Bq,classification of radionuclides and classification of food.

Primary intervention levels

the dose shall not exceed 5 mSv during the first thirty (30) daysafter that the fallout from an accident began

the dose should not exceed 5 mSv in the first year after that thefallout began

the dose should not exceed 1 mSv per. year during the followingyears.

The level of 5 mSv per 30 days refers to the acute phase, phase 1.

Dosimetric factors

For the calculation of intervention levels in !/q the most interestingradionuclides are grouped togethet in a simplified table according tothe proposal by Codex Alimentarius Commission. The special dosimetricfactors for small children refer to pregnant and suckling women andchildren under two years

Dosimetricfactor Sv/Bq

Typicalradionuclides

Comments

0 5Am-241Pu-239

small children

1010-'m-241PU-241

Sr-90I-131

except small children

children and adultssmall children

10 Cs-134Cs-137I-131

children and adultsII

except small children

Intervention levels for marketing

The intervention levels for marketing should be expressed as activity-1concentrations (Bq kg ) and arc called Derived Intervention Levels

(DIL). Three kind of DIL are given

— a common Nordic permanent DIL that is valid also for the normalsituation

— a common Nordic interimistic DIL that is valid for a maximum of 30days after a fallout has begun i.e. phase 1

— national special DIL based on national political, economical andother considerations.

The levels. refer to food ready for consumption and will be the basisfor the possible intervention levels for the ptoduction of food.

The common Nordic permanent DIL correspond to 5 mSv a i.e. the same-1

as for Codex Alimentarius. The assumptions for the calculations of DILare conservative and will result in real doses being much less then

-15 mSv a . The Codex Alimentarius values are for international trade.

The respective DIL refers to the sum of radionuclides in each groupand each group can be considered independent of the others. Naturalradionuclides are not included.

Permanent DIL for food except milk and child-food

Radionuclides with

dosefactor Sv Bq-1

Examples DIL Bq kg

10 Am-241Pu-239

10

10 Sr-90 100

10 Cs-134Cs-137I-131

1000

Permanent DIL for milk and child-food

Radionuclides with

dosefactor Sv Bq-1

Examples DIL Bq kg-1

10 Am-241Pu-23910-'r-90I-131

100

10 Cs-134Cs-137

1000

These levels are relevant for phase 2 and the normal situation and. forzones 1 and 2 (the directly and indirectly effected far-field).

The interministic DIL refers to phase 1 (acute phase) in zones 0 and 1and are ten times higher than above.

Intervention levels for consumption

These derived intervention levels are the basis for advice on diet and

consumption and are expressed as annual intakes of radionuclides.These intakes expressed in the same groups of radionuclides as above

-1correspond to the primary intervention levels of 5 mSv a the first

-1year and 1 mSv a the following years. They are as follows

(correspond to 5 mSv a ).-1

Maximum annual intake for the public in general

Dosimetric factorSv Bq

10

Examples

Am-241PU-239

Annual intake-1

Bq a

5000,'j/

10 Sr-90 50 000 '.

10-'s-134Cs-137I-131

500 000

Haximum annual intake for pregnant and suckling women and for childrenunder two years of age

Dosimetric factorsSv Bq

-1Examples Annual intake

Bq

10 Am-241Pu-239 500

10 Sr-90I-131 50 000

10 Cs-134C8-137 500 000

190q-10-1 General Rec»m<iiu»<iati<»is oii 1i) ti ) '»t 1<!» Levelsa»d Strategies fot M<.asu) emi »1.;

Hv J»» D]<if Si)11i:Na'tin»al 1»st)<ut<')i R 1<1)el< 1~)ll 1 <i)t<)<'1 i»1)

.it<)ckhi) 1<l), 'i cc»

Intervention levels

Predetermined intervention levels vill m )k<. 1ecisions easier in asituation after an accide»t when a lot of. i1ecisi<)»s have to be taken,often in a short time and o)) i»complete i»fo) matin» a»d data. Theinterventio<) levels indicate the level oi ),)di»tio» at which theauthorities should considet. ti)e need <)f some kind o[ cou»termeasure.The application of interve))tio» levels m»st b» ma<le with goodjudgement and they should»ot be co»side) e<i as;) 1>orderli<)e betweendangerous and not dangerous and they may also be exceeded withouttaking any countermeasure if. it appears i » the special situation thatthe consequences of the cou»termeasure a<e u») casu»able. Everyintervention level shall refer. to a given measure a»d the i»terventiondose is the extra dose people get if »o measure is taken (or the dosepeople will exceed if »o measure is taken).

It is often necessary in the application of. a» i»tervention level toderive the level from measurements and therefo) e D rived InterventionLevels (DIL) are used.

The intervention levels or more the DIL should be used in comparison,with actual situation expressed in estimated resulting total dose ordoserate or activity concentration as indicators how serious thesituation is. If t: he measured or estimated values a)e much below their.terventio'n levels or DIL no countermeasure is justified. If they onthe other hand are much higher than the intervention )evels of DIL acountermeasure is strongly justified. But not alway" strong enough forall situations e.g. to evacuate people in s»owstorm a cold winterday.

With all these reservations the following i»terve»tio)) levels havebeen recommended in Sweden. They are applicable for the first monthafter the accident.

Countermeasure Lnterzmisiir <i)terveniinn level,mSv/month for the f.irst month

Being indoors

Measures concerning food

Recommendation thatpregnant women shouldmove

Evacuation 50

Stable iodineadministration 50 ( to thyroid)

From these values it is possible to estimate DIL and the followingtentative values are given to be discussed and developed in the nearfuture.

Heasurement Interimistic DIL Countermeasure

Gammaradiationoutdoors (onemeter above ground,the value refers tothe first days)

50 pSv h-1

Recommendationthat pregnantwomen movewithin a month

500 uSv h-1

Recommendationthat pregnantwomen movewithin 24 hrs.Evacuation ofothers withina month

5 000 uSv h-1

Evacuation ofpregnant womenas soon aspossible, otherswithin 24 hrs

I-131 in grass 20 000 Bq kg-1

2 000 Bq kg-1

Grazing prohibited

Take samples andmeasure

Cs-137 in grass 2 000 Bq kg-1

Grazing prohibited

200 Bq kg-1

Take samples andmeasure

I-131 in milk 2 000 Bq 1-1 I'onsumption and

marketing prohi-bited

Strategy for measurements outside nuclear power plants in a nuclearaccident situation

The following strategies concern measurements outside the affectednuclear power plant.

Measurements before and during the release.

Measurements are made by stationary patrols that in a nuclear accidentsituation are transported to and staying at 2-3 predetermined placesin each 30 sector in the wind direction up to about 5 kilometers from0

the plant. In an accident three adjacent 30 sectors are occupied. If0

there is no specific winddirection (( 3 m/s) the patrols are locatedequally divided around the plant. The predetermined places should becarefully chosen e.g. on a fairly flat surface, overgrown by grass,and far enough from disturbing objects like forests (20 m), singletrees (10 m), large buildings (20 m) and open water (20 m). The placesshould be easily identified and found.

The instrument is placed 1 m above ground and it measures

gammaradiation by 2 GM-tubes from 0.05 pSv h to 10 Sv h . The-1 -1

average doserate during the last 5 minutes is automatically given andthe value is read off each 10 minutes and reported to a centralstation.

Besides these patrols each fire-station within.50 kilometers hasinstruments to make measurements of the doserates. There are also anumber of stationary TLD outside and around the plant to give theintegrated dose as a later verification of directly measured doses.

Measurements after the release has ceased.

By the stationary measurements the contaminated area is identified andmeasurements are then performed by mobile patrols along given routesand at given places along these routes in the sectors of interest upto about 50 kilometers from the plant. These measurements will advicethe authorities if there is any need for countermeasures likeevacuation etc. and therefore priorities are given to places wheremany people are living or staying. If mobile measurements are made the

speed should not exceed 50 km h

Sampling of air, vegetation and soil.l

Within the sectors of interest airborn activity (iodine and aerosol~)is sampled in filters on 3 given places at about 5 kilometers from thesite. This sampling should start before the release occurs and duringthe release. The filters are measured in special laboratories whichshould be located at least 50 kilometers from the plant.

Samples on vegetation and soil may also be taken in many placeswithin the same sector. The samples should be taken by well trainedstaff.

Heasurements l>y air.

During the fir st day the <<e;«;»> «.>< rrr; < rl Jfl

ill f or ma t i oil orl dr spr r '; r <;rr «I« I <i<I»I< < r r r <..';

are made from air'plarres dur irrg rh< relr.r.,< ~

contaminatior> of the «i>pl;«ru;hi< h «i11At a later stage (da,;) i t m;>; h» al>pr <>pr r af!om airplanes to mal> the gr.our><l < orrr;«> rr;<r

planning of possible co<lllr.el lrle'«. ll <rs «'>I>< elSpecial equipment is r>e«.ssr>r, l <>r .:li<~ < r r i>',

locating tire place ol rrreas«r.e<rl<'l>\.

ar e r>ecessary to give,< le;>.;< . If measureme»ts

<her< i.; also a risk of.'r «r l>< later measur'emerrts.

r <> m;rke measurementsi<>r> 'r> 1arge areas and for.<i i r<g 1 <>od pr'oduc t iolr.«!l> i ««eas<rremenr s a»d for

Heasurements by sea.

Heasurements f rom a ship ar e ><>r ..r r ol>gl'' <I><if i vrr ted but there may bespecial circumstances wherr they ar <.. llowever r lre measurements can onlyindij<ate passage uf a pl<><<>«a»d <a«<i»ly gi.<- l>u<>r qua<it itive data.

Other measur.ement. strut» i<

'incemore than 30 years ther» are 2G orat iorrary iorrizing chamberslocated in the "hole of Swede>> which give alarm irr case of abnormalincreased air.— arrd groundacrivi ty. The alt>rr: is given to the NationalIr>stitute of Radiation 1'rote< tin>> (Stater>s Stralskyddsinstitut, SSI)in Stockholm. Af ter tire Circ< r>ol>yl ac< ider>t these irr" trume»ts give nowalarm automatically to SS1 i f r.he 24 brs <lose irr< reases 10-20K abovenatural background (1 500 — 4 500 nSv per 24 hrs). The numbers ofthese instruments will in<>r<as<. l>,' 10 ir>str urr>er>ts in a nearfuture.

Furthermore there are statio»ary airsamplir>g e<luipments'at 8 places in-3

Sweden by which ai rl>or n act i vi ty (aerosol.;) above 1 r>Bqm can bedetected. The organisati«rr of the alarms;stem is sr>eh that SSI is

3alarmed if the airactivity i'igher tlran 1 mH<ld/m . The measurementsare made by the Swedislr Defense Research I»stitute (FOA).

Af ter the Chernobyl accident every mur>i< i pal arrd < ounty administrationauthority ( in total about 300) has a hand instrument to meet localrequirements on measurements arrd information but also to complementthe assessment of ground contamination. Tlrese instruments are a partof the overall emergency preparedness system to indicate increasedactivity in Sweden caused primarly by accidents outside Sweden. Theyare regularly tested to train the staff and rr> ger the vari,"-.tions ofnatural radiation.

For food measurements a number of dairies arid research laboratorieshave equipment for measurement.;., mol>i le rrrea. <rremer>ts can be made frorrrcars and airplanes are prepared furr mapping gr nund a< tivity and tolocate incoming radina< r i re i ln<><]s;

ANALYSIS O'F THE EFFICIENCY OF DNA REPAIR IN PLANTS

FOR THE GENETIC MONITORING IN THE CHERNOBYL VICINITY

Syomov,A.B., Ptit yna,S.N, and Sergeeva,S.A.

N.I.Vavilov Institute of General Genetics of

the USSR Academy of Sciences, Moscow.

Ionizing radiations prove to be an important ecological factor,

having diverse effects on the populations. When evaluating the genetic

consequences of exposure to chronic radiations, it is extremely

important to study the capacity of a population or an organism to

withstand the damaging action of radiation. As was established by a

numbers of authors, a long-term exposure of plants to chronic radiation

may increase both their radiore istance and radiosensi tivity.The strains of chlorella, isolated from the soil samples containing high

concentrations of Sr-90, were found to have an increased

radioresistance. The experimental strains were also characterized by a

more intensive mutation process. Later, it was shown that

the radioresistant strains of chlorella had a more effective system of

the single-stranded DNA breaks repair. The penomenon of radioadaptation

was also revealed in a number of other organisms. The fate of a

population exposed to chronic i rradiation depends both on the type the

organisms and on the biophysical characteristics of irradiation (RBE,

LEL, dose rates, etc.). The populations of plants growing under the

conditions of chronic irradiation exhibit the signs of adaptation to

this mutagenic factor. When exposed to mixed gamma- and alpha-

radiations (radionuclids of the uranium series), some species of

plants become more radioresistant, but others, on the contrary, — more

radiosensitive. Especially interesting are the data on the relative

radioresistance of various populations of one and the same plant

pecies but growing under different radioecological conditions. Such

re ults are obtained, for instance, in the experiments with Vicia

cracca(a perannial herbaceous leguminous plant). As follows from these

studies, the radioresistance of a population of one and the same plant

species may either increase or decrease, depending on the type of

irradiation.

At the present time, the radiosensitivity of an organism or a cell

is considered to be dependent mainly on the functioning of its DNA

repair systems. The DNA repair systems prove to be one of the most

important factors influencing the natural radioresistance of plants.

New mechanisms influencing the radi oresistance may appear at the

level of a population. Such factors as selection of most viable species,

changes in the genetic load, gene drifting, etc. are essential for

the population. At the same time our data, obtained before the Chernobyl

accident, testify to the fact that the radioresistance of natural plant

populations depends greatly on the DNA repair. Based on this parameter,

it is possible to prognosticate'urther genetic changes in the

populations. Therefore, it was very important to elucidate the

operational mechanisms of postradiation DNA repair in the populations

of plants growing in the Chernobyl vicinity.

MATERIALS AND METHODS

The experiments were carried out with the seeds of Taraxgcum

officina le, Arabidopsis, and Vicia cracca, collected on the plots

exposed to radiation of different dose rates: from 250 to 0.01 mr/hr.

The eff iciency of formation and repair of single-stranded DNA breaks

induced by acute gamma-radiation of 50 and 100 Gy wa determined in the

plants grown on photostat under laboratory conditions. The number of

DNA breaks was registered by means of denaturating gel electrophoresis

and by the method of the alkaline unwinding'f DNA with the subsequent

chromatography on hydroxyapatite (the Ridberg's method modified by us

for the experiments with plants).

RESUI TS AND DISCUSSION

When analyzing the populations of Taraxacum officinale (the samples

were collected in June, 1986), the highest yield of DNA breaks

as compared to the control populations, was registered in +~o population

exposed to the highest (of the studied) dose rate — 30 mr/hr (as of May

25, 1986). The yie'ld of DNA breaks was found tn be the same in the

populations exposed to a lower dose rate (3.7 mr/hr, as of May,25,1986).

The efficiency of a single-stranded DNA repair was identical in all the

populations and approximated 100K.

A study of the efficiency of formation and repair of DNA breaks in

the populations of Arabidopsis (the samples were collected in June,

1987) showed a significant increase in the number of'reaks per dose

unit of additional irradiation. This value constituted 5x10 " breaks

per 1 rad per 1 dalton i n the control, whereas in the populations

growing under exposure to the radiation of 10 and 60 mr/hr (as of June,

1987) it amounted to 28 and 33x10 '2, correspondingly. In this case all

the breaks were repaired both in .the control and in the irradiated

populations.

The seeds of Vicia cracca were collected in 1988 from three

different plots (control, 10, and 8 mr/hr, as of Augu-t, 1908);

the sinqle-stranded DNA breaks were registered by both of the methods

and showed "imi lar results. In the populations exposed to the additional

irradiation the process of the DNA breaks formation wa" more intensive

than in the control ( 13.10 and 8xl0 '" breaks per 1 rad per dalton).

A complete repair wa" registered both in the control and in the

population.". exposed io chronic irradiation.

Earlier we have shown a decrease in the level of the DNA repair

synthesis and an increase in the yield of DNA breaks per do e unit of

the additional irradiation in the population".. of Vicia cracca growing

under the conditions of chronic alpha-irradiation. By contrast, in

the ca e of chronic beta-irradiation we observed an increase in

the intensity of DNA repair synthesis. However, the efficiency of

the DNA breaks formation and reapir did not differ from

the corresponding processes in the control populations.

It is possible to suppose that the increase in the yield of "ingle-

stranded DNA breaks, registered the populations of Vicia cracca,

Arabidopsi-, and Taraxacum office'na)e exposed to chronic irradiation in

the Chernobyl vicini ty duri ng the whole vegetati ve period .is attributed

to the alpha-component of irradiation. Inspite of the fact that the

alpha-nuclids do not i ncrease significantly the specific radioactivi ty

of soil (by about 1X only), we cannot underestimate their role in the

observed biological changes, as their biological activity i" known to

be very high (the quality coefficient is 10-20).

The alpha-radiation is known to induce a more serious DNA damage,

than that caused by the random-ionizing gamma- or beta-radiation".

The alpha-induced DNA lesion"- ar e not so easi 1 y r ep;» red,,nid,;incor ding

to some data, they can even inhibit the repair of other 1esions, which

are normally easily repairble. Evidently, apart from .',uch Ie.',ions,

DNA breakages, modification'- to the base'tc., the alpha-radiatior>

capable of inducing much more DNA-DNA and DNA-protein bicros link , as

well as the -.:mpairments of superhelicity and DNA conformations

(transition into the Z-from). It may also change the methylation and the

ADP-ribosyi Iation. There are some literature data which confirm that

such lesions (or some other DNA lesions which have not been identified

yet) may persist in several generation. of the exposed cells and result

in their increased sensitivity to the --additional irradiation. In our

opinion, namely these lesions are responsible for the additional

radiosensitivity of plants growing under the condition" of chronic

irradiation by the alpha-nuclids of the uranium series.

The genetic processes which are seen in tHe population" of plants

growing in the Chernobyl vicinity resemble those observed in the case

of chronic irradiation by the alpha-nuclids of the uranium series.

However, we failed to obtain any data confi rming the genetic adaptation

in the case of chronic beta-irradiation. This can be ascribed to the

mixed character of i rradi ation, si nce the alpha-, beta-, and gamma-

~ nuclids are registered in the Chernobyl vicinity. Besides this,

the specificity of dose rate formation may also play its role: an

abrupt decrease in the first months after the accident, due to a

disintegration of the isotopes of short half-decay, and then,

a comparatively slow disintegration of the rest isotopes of long

ha'1 f-decay. It i s qui te probab1 e, that even i f the dose rates of chroni c

irradiation are practically constant, there will be some other genetic

What is and what is not in NUREG/CR-4? 14*Health Effects Models for Nuclear Power Plant Accident. Consequence Analysis,

J.S.Nuclear Regulatory Commission„ Hay, 1989'ielWald, H.D.

University of Pittsburgh, Pittsburgh, PA 15261, U.S.A,

The major purpose of this presentation is to give a brief generaloverview of the Nuclear Regulatory Commission's recently published report un thescient.ific bases for the hea'th effects models used in nuclear power plantoccident consequence analysis. The report is an update of the "consequences"sections of WASH-1400, the Reactor Safety Study under Norman Rasmussen thatuas published by the same agency in 1975.

The effects discussed include the acute radiation syndromes and otherearly and continuing effects; cancers and thyroid nodules; and genetic effects.In view of the scientific uncertainties involved in modeling radiation-relatedhealth risks, central, upper and lower estimates are provided for all modelparameters.

The usefulness of these models in the assessment of the Chernobylaccident health consequences should be determined. Also, the unique early,present and future biomedical data from the Chernobyl radiation-exposedindividuals and populations should be utilized to improve the models.

Omitted from the report, which is focussed on somatic health effects,is an important but difficult area for assessment, the mental healthconsequences of nuclear power plant accidents'he results of some studies ofthis health impact in people in the area of the Three Nile Island accident willbe mentioned.

EVALUATION OF GENETIC DAMAGE CAUSED TO THE POPULATION

OF CHERNOBYL VICINITY BY THE RADIATION EXPOSURE

V.A.Shevchenko.

N. I.Vavi lov Institute of General Genetics.

The evaluation of genetic risks of radiation exposure for man,

given below, is based on the reports submitted to the General Assembly

of the United Nations Scientific Committee on the Effects of Atomic

Radiation (UNSCEAR) in 1982, 1985, and 1988, As well as on the results

of a cytogenetic assay of the affected population of the townPripyat'nd

also of those people who eliminated the consequences of an accident.

This study was carried out by the researchers of the Institute of

General Genetics and several other Institutes.

Before considering the genetic effects of radiation exposure on

man, it is necessary to say several introductory words. Speaking of the

genetic consequences of radiation exposure, we mean the genetic effects

revealed by analysis of the descendants of radiation exposed people. In

other words, under the term of genetic effects we mean those effects

which could be transfered through the gametes to the successive

generations, i.e. these are various gene mutations, chromosome lesions

'nd any damage caused to the integrity of'he chromosome ensemble of

man. Similar changes in the somatic cells of the organism ( i.e. in the

cells of the liver, stomach, blood, etc.), which are not transfered to

the following generations, are known to be one of the main trigerring

factors for the so-called somatic impairments, e.g. cancer, teratogenic

anomalies, and radiation sickness. Thus, we speci fy the genet':.c and

somatic effects of radiation exposure. The genetic efects prove to be

most long-term, for they tend to be revealed in the subsequent

generations of the exposed people.

When evaluating the genetic effects of ionizing radiation on man,

the UNSCEAR experts make several fundamental assumptions. First, they

assume the linear dependence of the frequency of mutations on the dose

of ionizing radiation (the hypothesis of a linear dose-response).

Numerous experiments prove the 1 inear character of this correlation for

the moderate and high radiation doses. However, as to the low doses, the

experimental data for man and other mammals are so far scarce. In this

view, the UNSCEAR experts decided to accept the hypothesis of linear

correlation for the low doses, as well, at least until the appearance

of some other data, which may prove different dose-response effect.Second, the spectra of natural (spontaneous) and radiation-induced

mutational changes are found to be similar. This is quite natural, since

a considerable portion of natural mutations is definitely connected with

the exposure of man and other organisms to the radiation from natural

sources (e.g. the radiation from mountainous rocks, cosmic beams etc.).This similarity makes it possible to compare the natural mutational

process in man and the expe=ted mutational effects of ionizing radiation

exposure.

The thi rd importamt assumption postulates the similarity in the

spectra of mutations seen in experimental animals ( mi ce ) and man.

According to t'his assumption, both in man and in experimental animals,

various biological factors (sex, stage of the germ cell, etc.) and

physical factors (type of ionizing radiation, dose rate and others) have

the same influence on the level of genetic damage caused by radiation.

This allows using the data obtained in the experiments with animals for

evaluating the risks of radiation exposure for man.

The methodology for evaluating the genetic effects of r adiation on

man is based on a number of methods and criteria employed for the

general evaluation of the expected genetic damage to the exposed human

populations.

The main method used in the radiatson genetics consists in

the registration of various genetic lesions with respect to a dose of

some or other type of radi ation. Point mutations and impairments in the

chromosome structures are known to be the most common genetic lesions.

Estimation of mutation frequency per dose unit proves to be the main way

for evaluating the genetic risks of radiation exposure in man (i.e. the

so-called direct method for evaluating the risks). However, so far, we

are lacking the data on the mutation frequencies per dose uni t with

regard to man. Therefore, these estimations are based on the data

obtained for experimental animals. Currently, this method is employed

for evaluating the risks of hereditary anomalies of the dominant type

and various chromosome lesions registered in the descendants of the

exposed people.

Another method for evaluating the genetic risks of radiation

exposure in a doubling dose method, which infers determining the dose

causings the effect similar to that seen in the case of natural

mutagenesis. The method is based on one of the cited assumptions and

is now proved experimentally. This is an assumption of the postulated

similarity in the spectra of the natural and radiation-induced

mutations.

A question arises: what is the level of the natural variability

of man? C'reat progress has been made in this field lately. At present

more than 4000 of hereditary diseases has been studied and their number

increases every year. Over a long period of time the natural variability

of man has been evaluated on the basis of the results of an epidemiolo-

gical screening of a big human population living in British Columbia

(Canada). These studies show that 10.5X of various anomalies have been

manifested by the age of 21. A mass screening of population carried

out recently in Hungary has given new data on the natural variabilityof man. As was shown, by the age of 70, i.e. during the whole life span

of man, the number of hereditary anomalies revealed exceeds the

corresponding value registered in the population of British Columbia by

one order of magnitude, the greater part of these anomalies being

revealed in the middle-aged people and in the elderly. The main part of

hereditary diseases, coiistituting the genetic load, is represented by

the so-called multifactorial irregulary heritable diseases (such as

gout, ulcer, diabetis, asthma, etc. — in total 25 diseases). Their

absolute number equals 600 000 per 1 min people (60K). Thus, the

screening carried out in Hungary made it possible to estimate a total

genetic load for the human - populations (without taking into account

those genetic changes which lead to early embryonal death, spontaneous

abortions, and stillbirths, the mutational component of which is also

large enough).

The data obtained ir British Columbia, USA, and Hungary laid

a basis for the general estimation of the level of natural v riability

of man (according to the 1988 report of UNSCEAR, it amounts to 67.6X

~be~+) . Thi s value may even increase, i f we taI e into account of

some other types of genetic variability, which so far have not been

e. timated quantitatively, but are known to have a significantly high

mutational component {e.g. hereditary malignant tumors).

To evaluate the expected genetic effects on the first generation

after exposure (i.e. in children of the exposed parents), the following

approaches have been used. First of all we should keep in mind the fact

that the level of natural mutational process in human populations

actually repr sents a historically formed equilibrium level, which

depend.:s on the intensity of mutational process, on the one hand, and on

the intensity of selection against the arising mutations, on the other.

In the case of chronic irradiation of human populations the equilibrium

leve1 of mutagenesis resulting from radiati'3n-induced heritable changes

apoears no later than 7-10 generations .after the start of irradiation

(-F-''".~). At the doubling dose of radiation the equilibrium level of

mutagenesis caused by ionizing radiations will be equal to the

equilibrium level of the natural variability of man. According to the

estimates given by the UNSCEAR experts, a doubling dose for man equals

1 Sv. In the first generation we observe approximately 1/10 of those

mutations, which.'- 10 generations later, will constitute theI

equi1iorium level. As follows from Table 1, when evaluating the genetic

effects on the f irst generation, the UNSCEAR experts did not take into

account all the categor'ies of genetic load of man, but only some of

them, namely, dominant ance, X-1 inked diseases, recessive and chromosome

diseases. Their total expected incidence in the first generation amounts

to 17 cases per 1 min newborqs at a dose of 0.01 Sv. It is easy to

establish that the mentioned above categories of diseases, on the basis

of which the evaluation of geneti risks of radiation exposure for man

t

l

'C1 '

was carried out, constitute as little as 2.4x of all the currently

revealed genetic load of man. So far (as of 1988) there has not been any

evaluation of the risks of congenital anomalies and multi factorial

diseases, which constitute 97.6X of all the genetic load of man. Why

did not the UNSCEAR evaluate .this prevailing heritable variability of

man in 1988'? This was not dore, because it is still unclear, whether itis possible to employ a doubling dose of 1 Sv for the diseases of

complicated etiology, and, what value of the mutational component must

be used to est,imate the risks of radiation exposure. However, if we

assume that mutational components for these diseases range between 5

and 50%, and these are the figures given by the competent committees

(the BEIR 1980 report, and the NUREG 1985 report), then it would be

legal to use these values for estimating the risks of radiation exposure

for man within the whole the spectrum of human genetic variability.

Finally, at,a dose of 0.01 Sv (1 ber ), per 1 min newborns one may

expect a surpluss to the natural level of variability ( which, as was

noted, equals 67.6R, i.e ~ 676 000 per 1 min of newborns) that wi 11

constitute 50 more cases of hereditary anomalies (17+33), if the

mutational component for the multifactorial diseases equals 5X. Whereas,

if the mutational component equals 50m, the number of expected

hereditary anomalies will increase abruptly — up to 347 cases. In

principle, the expected mean value of the mutat,ional component for

multifactorial diseases may vary greatly, which, naturally, can alter

tne estimated values of risks ("- g. ). However, the most likely level

of ri sks of radi at ion exposure expected in the first generation appears

to be that of 50-347 cases of hereditary anomalies at a dose of 0.01 Sv

per 1 min newborns.

If numerous populations are exposed o chron>c rad>at>on, tne

expected equi librium level amounts to 450-3400 cases of hereditary

anomalies at a dose of 0.01 Sv per generation and per 1 min newborns.

Taking into account the hypothesi" of linear correlation, we can

determine the genetic risks for any dose of chronic irradiation. So, at

a dose of 35 ber, the expected number of hereditary anomalies in the

f;rst generation of the exposed parents wi 11 constitute 1 750- 12 100

cases per 1 min newborns. Although this level of expected hereditary

anomalies constitutes a comparatively small part of the level of natural

mutation process, it is difficult not to be concerned with the high

absolute number of expected radiation-induced genetic anomalies. We must

clear up the question of a quantitative comparison of the expected

incidence of radiation-induced hereditary diseases with the natural

level of mutations. hlormally, this comparison gives low values of

surplus to the natural level of mutation,'hich is in favor of the

'safety'f some or other level of radiation. For example, at a dose of

35 ber a surplus to the natural level of mutation constitutes 0.3-1.8%.

However, this a comparison proves to be rather conditional. Strictly

speaking, we must compare the frequencies of natural and induced

mutations, which arise during one generation, either compare the

corresponding equil'ibrium levels. Otherwise, we deliberately include

into the estimations a 10-fold difference between the frequency of

mutations in .~he first generation and the equilibr'ium level.

The mentioned above approaches to the evaluation of genetic ri ks

of radiation exposure for man can be employed for determining distant

genetic effects of ionizing radiation exposure on the population of the

Chernobyl vicinity. In this. case the main indicator to be primarily is

the effective equivalent dose. To determine the absorbed dose= we

decided to use one of the methods of the biological dosimetry, ~.e. the

evaluation of the f requncy of chromosome lesions in human peri phera 1

lymphocytes. In the May of . 1986 these studies were ini,tiated in

Chernobyl and they are still under way. At present, the bank ofcytogenetic data available at our Institute contains the results ofabout 1 000 assays. Special comparison of the estimated doses, obtained

on the basis of genetic anaiysis, with the data of physical dosimetry

has shown their in good correlation. The mean value of the absorbed

dose, determined on the basis of the examination of 102Pripyat'esidents,

was found to be 0.13+0.03.~„This sampling included the most

severely affected part of population (militia men, doctors, workers ofthe,city services, etc.). Evaluation of genetic risks of radiationexposure for the residents of Pripyat', based on the above mentioned

methodology, gives a value of 7-45 hereditary anomalies per 10 000

individuals of the first generation of exposed people.

It must be emphasized that genetic damage of ionizing radiations,even provided with a short-term exposure, is caused not only to the

first generation of exposed people. The genetic effects are exhibited

in several successive generations, gradually reducing to the natural

mutation process. On the whole, the total genetic damage caused to

several generations may be several times more than the effects expected

in the first generation.

The cited quantitative indicators of risks are refered to the

number of the expected cases of severe genetic diseases. The term

'severe genetic disease', used in the documents of the UNSCEAR, infers

poor health, leading to a in man's capacity to work, physical and

9

mental disorders or disability of genetic or~gin, which can be

manifested at any age — from the birth of man ti 1 1 his old age. Several

criteria for evaluating the genetic damage caused by ionizing radiations

have been developed, so that we can estimate the actual genetic load,

responsible for the appearance of genetic disorders, and influencing the

individual, social and communal indicators of heal th. In the UNSCEAR

1982 report the indicators of disable life (stay in hospitals, home

isolation etc. ) and shortening of life span were employed for evaluating

the genetic damage. A number of objective diff icuties were faced, when

developing and applying the criteria for evaluating the genetic damage.

Nevertheless, the UNSCEAR experts considered it reasonable to estimate

the indicator for the genetic damage, based on the incidence of genetic

disea es. Assuming an average expected longevity of life to be 70 years

(or 70 000 000 per 1 min newborns), the cited above estimates of the

risks(i.e. 50-347 cases of hereditary genetic .anomalies in the firstgeneration per 1 min newborns, i the dose of radiation equal 35 ber

(0 35 Sv)), give ihe following figures of genetic damage cau..ed to the

first generation: 39 000-247 000 years of disable life and 46 000

358 000 years of shortened life span per 1 min newhorns. In principle,

it is possible to to evaluate the genetic damage caused by radiations

in monetary values. Thus, based on the criteria and methods for

evaluating the effects of ionizing radiations exposure for human

populations, developed by the UNSCEAR, it i- possible to estimate the

potential genetic damage caused by radiation to the people of the firstand subsequent generations. Using various indicators for evaluating the

genetic damage, we can give a most general prognosis for the economic

and social damage cau..ed by the radiation-induced heritable anomalies

10

of man.

Howe''<er, this progno is will give f'airly approximate figures, for

it fails to take into account a number of important factors, which can

elevate indicators of the expected risks of radiation exposure. First

of all, the operational mechanisms of 'small'utations are stillunclear. Interacting with other mutations, 'small'utations may affect

greatly the health of man. Besides this, the problem of hereditary

oncological diseases is far from being solved. Little has been f(nown yet/'f

the genetic sensitivity of human germ cells, as well as of the early

stages of embryonal development. Elucidation of these and other problems

of human radiation genetics, obviously, requires further development of

the corresponding fundamental research in this field.

Genetic consequencies of Chernobyl accident for the population

of Gomel and;.:ogilev i",,egion-

G,.I.La"juk (i,;insk).

he research of genetic consecuencies resulted from Chernobyl

accident were carried out by Byelorussian Institute for Heredi-

tary Diseases by examining malformations in embryos and. fetuses

obtained through induced abortions in the female residents of

i;:insk, Gomel and the Southern areas of Gomel and i!ogilev Re-

gions ( able I) and by counting malformations in newborns re-gistered under Byelorussian Genetic l.onitoring Programme in

1982-1988 (Table II).I.'i%IN COFCLUSIONS

1. Since 1986 the frequency of fetal malformations found in the

residents of the South of Byelorussia has increased. from

5.64-0.3'o 9.-2.1't=$ .Q). The frequencies for IFiinsk and

Gomel have remained statistically unchanged.

2. The frequency of malformed children born in 17 districts of

Gomel and I,:ogilev Regions with the highest contamination level

has increased from 4 '7-0 '8 to 6.89-0.59 per 1000 births

(t=3.41). Ilo statistically significant variations have been

found for the same malformations in other areas of Byelorussia

Over particular areas the variation. frequencies after Chernobyl

events were the same as before the accident. So, in Goretsk

district, i!ogilev Region, the malformation frequency was

7.11-1.89per 1000 newborns in 1982-1983 and was not statisti-cally different from the increased figures in contaminated areas

of Gomel Region recorded in 1986 to 1988.

3. The study carried out during three years suggest that among

population of contaminated areas of Gomel and Llogilev Regions

the number of malformed births is tending .o increase ~

":"e '."ice frequency of cases; ith di "tur-ed intr uterine

'e'clop."..ent "an 'oe ascribed to a;,""'nt ef ect of ioni" in"- und

.;;:emica 'actors on the organism,;;nich fact:~as in" irectlye'idenced oy a nigner f eouency o f chromosomal a'oerrations found

"n oregnant;.omen and net'oorns from she contaminated ones and

'oy a larger amount of anomalies mainly of .ultifactorial etiology,

:urther studies are needed for making a definite prognosis ~

TABLE IThe frequency of inal formations (;~) in induced abortuses in several regions o f IIyeloru sia

The Regions under . i;!inskcon t1 ol

Gomel The Southernareas ofIiogilev Region

The Southern The Southernareas of Gomel areas of CornelRegion and I lo gi l ev Re

gionTo tal

The time of select- 1980 - 2nd 2nd half ofhalf o f 1 986-1987

1985 1986-1987

2nd halfo f 1986 — 1sthal f o f 1988

2nd half of 2nd half of 1961986 — 1st 1st half of 19$hal f o f 1907

The number of obser-vations 10168 2861

including informative 5732 1920 674 158 38

573

196

The number of malformed

abortuses 321 96 29 13 18

Anomaly incidence g 5.6-0.3II

Signi ficant di fferences,'omparedto those from

I PMinsk for 1980 to 1985

5 '-0 '

1 .00

)0.05

4.3-0.8

1.40

) 0.05

8.2-2.2

1 '0p0.05

1 3 ~ 2 5 ~ 5

F 00

( 0.05

9.2-2.1

2.13

( 0 '5

'"he frequency of malformed chili. en births through records

in the South districts of Gomel and;.:ogilev Regions for1982-198S (per 1000 nevmorns)

Districts 1982-1983 1984-1985 1986(the secondhalf of the year)

19S8(the firsthalf of the year)

GOI,:r'L REGION

"=ra<in, Buda-i.o she 1ev, 42 56 93'; e tka, Do brush, '1 sk,lelchitsy, Kormyansk,

Loev, iiarovlya,Khoiniki, Chechersk

3 '3+0.50 4.55-'0.61 7.66«0.79

Petrikov(control district)

2 71.24-0.88 4 '7-1.69

105.33-1.68

iiOGILZV PZGIOiV

Bykhov, Klimovichi, 23Eo stukovichi, 3.59-0.75Slavgorod, Krasncpolie,Cherikov

22

3 '8-0 '7841

5.61«0.87

Goretsk (control district) 14 87.11-1~ 89 4.23-1 ~ 49

11

4.94-''1 .48

Total for the zoneso f thorough control 65 78.

3.39-'0.42 4.27-'0.48134

6.89-0.59

Byelorussia (vrithouttne districts mentioned 1890 1691 2274

above) 5.65-0.13 5.04-0 ~ 12 5 ~ 65-0 ~ 12

Techniques to Reduce Acute and Chronic Population Dosesafter Major Releases of Radionuclides

Werner BurkartRadiation Hygiene Division

PAUL SCHERRER INSTITUTCH-5232 Villigen PSI

Contrary to most acute disasters such as earth quakes, dam failuresand fires, there is a large and complex web of events from the release of

radioactivity to the biological effects in human populations and ihe

ecosphere. This complexity makes risk assessments a difficult task. butoffers many possibilities to modify the course of events leading to harmful

effects. Therefore, knowledge of source terms, atmospheric dispersion,

radioecology, human behaviour, and radiation biology should not only beused to quantify exposures and risks, but to reduce both acute and chronicdoses. Given the generally limited resources of time, equipement and

expertise. during the early phases of emergencies, the most effective coun-termeasures have to be recognized beforehand by sensitivity analyses. In

addition, severe radiological events should not be aggravated on tlute psy-

chological level by introducing theoretical risk concepts and estimates notunderstood by the media and the public.

A preliminary assessment shows the importaiice of widespread

radioecological expertise oii all decision levels, as well as in the agriculturaland food distrubtion syst'm. Effective and fast links with electronic media

is a crucial prerequisite'or the protection of affected populations, i.e. toelicit individual and collective responses needed to reduce doses and dosecommitments. The hazard meter, a heuristic tool, is introduced to inducemodesty in health effect prophecies.

Medical response I'.o severe nuclear accidents

'J.C. NFNOTInstitut de Protection et de Surete Auclcaire

Departement de Protection SanitaireB.P.n' - 92265 Fontenay-aux-Roses Codex, France

ABSTRACT

The medical management in case of a.severe nuclear accident which results or may

result into acute health consequences to the workforce and/or the population coversthree main chapters:

I. Handling of large groups and the sorting of patients into classes of injuryand/or diseases for the purpose of expediting clinical care. One of the main

tasks is to determine the emergency of the cases. A simple way as a first step is

to divide the virtually exposed groups into three categories, such as (l) the

overexposed individuals with combined injuries (who in most cases should beconsidered as real medical emergencies), (2) the individuals likely to have

received high doses (who will be classified later on in the classical categories ofdose levels, and for whom some basic examinations should be performedurgently) and (3) the individuals who are likely to have received only low dosesand who are free of any other injury (followed as out patients).

The parameters used for thd classification into dose levels categories should bedefined carefully; preplanning the medical response should be performed

deeply in details and its efficiency and performance tested, through real sizeexercices.

2. Diagnosis, prognosis and treatment of victims whole-body irradiated. Although

these three medical steps are now well codified, it should be reminded that an

irradiated victim cannot be directly compared with a patient irradiated formedical reasons. The assessment of the dose level is the first step and the mean

absorbed dose to tne bone marrow determines the degree of severity, but other

parameters are of prime importance, such as the distribution of the dose in

space and time. The prognosis and therefore th'e type of treatment which

should be applied depend upon the degree of heterogeneity of the exposure, as

even a few small isolated patches of marrow irradiated at doses below around

2 Gy will be sufficient for the. repopulation of heavily irradiated areas. The

evaluation fo the regeneration capability can only be based on very specifictechniques, such as stem cells cultures.

3. Perspectives and future developments for the treatment of irradiated victims. It

may seem that the general practice for treating patients with radiation induced

aplasia is nowadays sufficient; in fact, the accidents of Chernobyl and Goiania

have demonstrated that some specific developments could bring great help.

Developing techniques for biological dosimetry which would be easier, quicker

and cheaper than the current cytogenetic analysis would provide a good basis

for the early and gross diagnosis. Using hemopoietic growth factors as

stimulators of proliferation and/or differentiation might shorten the acute

period between irradiation and autologous hemopoietic regeneration and would

avoid major and dangerous therapy such as allogenic bone marrow

transplantation.

workshop on rules for exemption from regulatory control

USNRC / OECD-NEA

washington D.C. 17-19 october 1988

A COMPROMISE REGARDING THE THREE DIFFERENT

CULTURAL RESPONSES TO THE "DE MINIMIS" QUESTION

S. Pretre

Swiss Nuclear Safety Inspectorate

Division of Radiation Protection

CH-5303 Wuerenlingen, Switzerland

- 2

A Compromise Reoardinq the Three Different Cultural Resoonses to the "deminimis" Ouestion

S. PretreSwiss Nuclear Safety Inspectorate

I. IntroductionIn the documents preparing for the present NRC/NEA Workshop on Exemption Rulesthere is a stimulating list of 6 issues for Session 4. I have chosen thefollowing tnree as a starting point for my paper:

- Basis for Derivation of Individual Dose Exemption Levels.- Need for Collective Dose Lower Bounds.- Method'or Establishing Derived Exempted quantities.

Another part of the ideas developped in this paper comes from the book "Risk andCulture" by Mary Douglas and Aaron Wildavsky. In this book it is shown quiteclearly that the perception of risk is much more a social rather than a techni-cal problem. This is evident when questions such as the following are formula-ted:

- How safe is safe enough?, or- How small should a risk be to be considered as negligible?

The answer to these questions is a matter of culture or perhaps of psychologicalprofile.

2. A soectral analvsis of the societvImagine that there exists a parameter which describes the different responses ofpersons to risks. If we ask a large number of persons how they wou.ld react toseveral dangers or risks we obtain, according to sociologists andanthropologists, 3 main groups of reactions.

Response of modern society to risks

(a spectral analysis)

C0Vl

Qa0I

JDEz s I i I I I I i

purity

order

freedom

Parameter describing the type of response to risks

- 3

For the first group the key word is FREEDOM. For the second group it is ORDER

and for the third it is PURITY.

These 3 tendencies are not just of theoretical or intellectual nature; they arean observed reality which has been thoroughly studied among primitive societies.[1,2]. Anthropologists have observed the reactions of primitive people toseveral problems connected with danger or impurity. They discovered these 3tendencies, which among primitive people are more pronounced, less nuanced thanin modern society. In modern man, the 3 tendencies should to certain degrees be

simultaneodsly present in the same person.

A mixture of the 3 groups forms the whole society, and a society where these 3

groups are represented in a well balanced way and where the extreme views arenot too strong, will react soundly to dangers or risks. Let us now look closerat these 3 groups.

3. The oioneerino societvThe sociologists call it the "heroic" society but I prefer to use the label"pioneering society". The individuals of this society are initiative, optimis-tic, pragmatic; they consider freedom as being the most important thing. They

are individualists, and the extreme exponents of this group will tend towards

ruthless seekinq of power.

These people have little relation to risk and are rather stimulated by dangers;they consider dangers as being interesting challenges.

The pioneering society does not really consider our "de minimis" problem as

being a proolem. When you ask them what the "de minimis" dose, or the dose of no

signific nco could be, they answer:Of course, this "de minimis" dose must be distinctly below the lowest thresholdfor non-stochastic effects; If it seems mandatory to have a figure, choose forexample a value around natural background. But we don't see the need for settinga value.

4. The reaulatino societvThe sociologists call it the "bureaucratic" society. Because of the negativepolitical flavor of this word I prefer to use the more neutral term of"regs:iating society". The individuals of this society prefer method, structure,rules and laws. The most important thing for them is order. They have a strongtendency towards introducing hierarchy everywhere. In extreme cases this groupwill tend towards caste distinction.

In this society, dangers are used to reinforce the rules or the laws. For the"regulating society", the "de minimis" problem is an important one, and it must

be solved quantitatively for the sake of ORDER. For them it is important toclearly set a value, but which value is less important. Nevertheless, the chosenvalue must be compatible with existing laws in order not to threaten the inter-nal coherence of the structure of rules.

4

The moralisina societvThe sociologists call it the "sectarian" society. Once more, I want to avoidthis term because it is too negatively loaded, and I propose to use the label of"moralising society". The individuals of this society are strongly motivated byhealth aspects, cleanliness and protection. Their key word is purity becausepurification is most important to them. They are usually rather pessimistic, andtheir views tend to become sectarian. At the extreme, this group could tendtowards self-destruction.

In this society, the dangers will be used to reinforce the cohesion of thegroup which wants to protect itself. This society will therefore shaw thetendency to encapsulate itself.

With regard to the "de minimis" question, the moralising society reacts asfollows:Even the smallest dose can have a lethal consequence, therefore, there is inprinciple no "de minimis" dose other than zero. If a value other than zero is tobe chosen by international bodies, we insist that this value be very small, andfurthermore it should be connected with a limitation of the collective dose.

6. A sound or well balanced societvAll 3 above mentioned tendencies play useful roles in modern society, but withinevery t.endency, the extremes views are not constructive. The whole situation canbe represented ;n a triangle where it is important that the center of gravity ofour modern society remains near the center of the triangle.

regulating(bureaucratic)

society, Cy

~@I 4.

0

g\

fo O

moralising(sectarian)

society

gc /r+

3/

/

//

//

//

/initiative

pragmatism/

',(freedom) /individualism

power seeking/

//

/

pioneering(heroic)society

3 basic culturalresponses to risk

5

All three tendencies by themselveshave limited usefulness, but togetherthey correspond to necessary func-tions, like the engine, the steeringgear and the brakes of a car.

If we look back at the evolution nfour sacs.ety within the last 40 years,we observe the following trend whichis worthwhile mentioning: The earliersociety strongly believed in techno-logy ano the FREEDOM Group (Pioneeringgroup) alone represented the majority.With time a shifting occured from thepioneers to the regulators, and lateron from the regulators to the mora-lists.

Some 10 to 20 years ago, the pioneersand regulators together had such acomfortable majority that they did notlisten to the warnings expressed bythe moralists. As a consequence, themajority of that time failed.to seethe limits of growth.

d f~. L

~ steering

i r egulating

, society

c"- -7/ ter

X/ e0

We needall three

groups!

I/

freedomII

/I/

V

pioneeringsociety

moralising i

~society

Trends in our industrial society 1950~1988

:=- freedom.==-

====:= order ====:-

:==:- purity ===

pioneeringsociety

regulatingsociety

moralisingsociety

Today, tne moralists are becoming an important dffd strong group, and we are now

observing a dramatic change which has already taken place in some countries: a

new majority of regulators and moralists is appearing.

This 'new majority could possibly rule out the group of the pioneers, and wouldtherefore destroy the "engine" of our society. That would be the beginning ofdecadence.

6-

I am not trying to moralise, but whatI would like to show is, that allthree tendencies have a positive roleto play in a modern society. The otherimportant fact is that the regulatinggroup is always the winner, because ithas a steering role and because itsviews are felt to be less extreme thanthe views of both other groups.Therefore, the regulating group mightin the future receive more responsabi-lity than the other two groups.

old

majoritThe old malontyhas failed to seeIhe limits ofgrowfh,

?. What about our own tendency?We radiation protection people havethe responsability to set up standardsfor the protection of the workers andthe protection of the public. In thisway, we decide for them. Therefore,the society has the right to requirefrom us, that we protect it accordingto its views. In other words, ourprotection standards should avoidextreme tendencies, and should be acompromise not too far from thecenter of the triangle.

possiblenew

majority

pioneersI

/V

The new maioritycould destroythe engine" andbring decadence.

Thus, each one of us should know about his (her) conscious and also subconscioustendancies, coming from our education, our religion, the structure of ourlanguage, and our culture. All these influences have created a computer softwarein our head that we have to live with. This computer software might be called"psychological profile" if you prefer. And this profile has influenced thechoice of our profession. Why do we deal with protection? Why do we feel res-ponsible to protect other people? All radiation protection people probably havesome sort of common bias'owards the regulating (OROER!) or the moralising(PURITY!) society. We should be conscious of. that bias and should try to com-pensate it by avoiding too extreme positions.

Once.a group is out of balance, i.e. when a large majority of the group thinksand feels along the same lines, this tendency will tend to reinforce itself,wi 1 1 glide towards exageration and will become contagious. The minority wi 1 1 beswallowed and the collective bias will become more and more extreme. This groupdynamic process is called "psychological epidemy" by C.G. Jung [3,4] and caneven reach collective psychosis.

It seems important to me, that the group of radiation protection people willbecome conscious of its bias which is not pathologic yet and will restore thebalance by trying to reach the center of the triangle.

8. National biasesThere are also national biases:Let me recall the chaos of diverging reactions to the Chernobyl fallout. It isinteresting to observe that there were very marked cultural differences: Whichcountries reacted excessively as- a pioneering society? - a regulating society? - a moralising society?

Countries like France, Germany, Italy, Sweden, Norway and Austria reacted invery different ways, which are perhaps predictible when you study the dominatingcultural and religious background of these societies.

Thus, arriving at a harmonised radiation protection all around the world seemsto be a difficult challenge, because the differences are deep witltin the soul.Nevertheless, if you and I are participating in international bodies who pretendto set up international protection standards, we should at least be aware of thequestion: where is the real problem located? And we should at 1east try not toimpose upon others our own psychological and cultural tendencies. Internationalharmonization is perhaps a good remedy because it forces every country to leaveits national bias. Let us hope that international harmonisation will also be .

usefull in moderating our professional bias.

9. Application to our "de minimis" oroblemAt a first glance, it seems that the 3 positions to the "de minimis" question(as described above) are so far from each other, that a compromise isimpossible. But it has also been seen that the regulating group is the key gr'upfor several reasons. In addition, the positions of the pioneering andmoralising groups tend to neutralise one another. Therefore it seems to us, thatthe cn1y way out of this impossible situation is to ask a moderate regulatinggroup to propose a compromise.

In the following pages, I will try to represent this group and to pr.opose such a

compromise.

10. Facts about fatalitv rates in modern societvLet us take as a basis the overall risk of death in our society.

Fatality Rate Depending on 4ge and Cause(Switzerland, Mean ol 1979-1983)

ai

Sn

I

1Q

'0

.l10S

IS.

~g Disease/Natural Death,;

R~zS Si

~ li graf Suicide/Homicide/gin:

IIQther Accidents

~ 11Traffic Accidents

MINI!t!IWm l

lllff i@fr.."...Pte,r......S ~ I

S

g ge~ ~seunf 119lu/II III l test l IJ I, -210881S«s

,«

1 11 1 11 SS SS Ss SS 11 «) s ~ 1 S SS S S'1 SS S|S S SS SS 0 s sS S ~ dJ S SS 4 S" ~ S

I I I I I I

20 30 40 4g 60 70 . 80

During childhood it lies around 3 10-4 per year, between 20 and 40 years of ageit is approximatly 10-3 per year, and at 60 years of age it is already 10-2 peryear. Let us propose the following idea as being the basis for an internationalagreement on what could be called a negligible additional risk:

During childhood, when the overall risk of death is ~3 10 4 per year,an additional leukemia risk of 10 6 per year ',s negligible.

Otherwise, most cancers occur at ages when the overall risk of deathis > 10-2 per year. An additional cancer risk of ~10-6 per year istherefore negligible.

11. The "de minimis" doseEven if we use pessimistic risk coefficients, we see that at all ages, anadditional individual dose of less than 0,1 mSv/a (< 10 mrem/a) is to beregarded as a negligible dose.

Therefore we finally come to the following proposal

1. For any individual, an additional annual dose of < 0, 1 mSv

(< 10 mrem) is negligible.

2. The annual sum of all "de minimis" doses for each individualshould be lower than 0, I mSv ( 10 mrem).

3. The "de minimis" upper bound for any single source could be~ 1/5 of that value.

4. The "de minimis" dose from any single source could be 0,02 mSv/a(= 2 mrem/a).

This value corresponds to less than 1 % of natural background. It is alsosmaller than the yearly fluctuation of the natural background at the samelocation, or for the same person. At this trivial level of individual dose, the"regulating society" feels that collective dose considerations are not neces-sary.

References:

[1] Mary Douglas and Aaron Mildavsky "RISK AND CULTURE"

An Essay on the Selection of Technological and Environmental Dangers.University of California. Press, 1983

[2] Mary Douglas "PURITY AND DANGER"

Routledge and Kegan.Ltd., London 1967

[3] C.G. Jung "PRESENT ET AVENIR"Buchet-Chastel 1962/1970

[4] C.G,~"Jung "ASPECTS DU DRAME CQNTEMPORAIN"Bu"het-Chastel 1983

-8-

The monitirinp of seeds of herbaceous phytocenosisin 30-kilometer zone of the Chernobyl disaster

A,I ~ Taskaev, N.P. Frolova, O.N. Popova, V.A. Shevchenko

Qstitute of Biology of the Komi Scientif ic Centre of the Ural

epartment of the USSR Academy of Sciences; N.I. Vavilov Insti-ute at'eneral Genetics, USSR Academy of Sciences

This paper presents data on soiae investigation of wild

erbaceous plants growing over a period of 3 years under chronic

rradiation conditions on the territory contaminated with radio-

c .ive isotopes as a result of Chernobyl accident. The study was

ndertaken with a view to estimating genetic consequences of

rradiation for natural populations of higher plants,

The seeds were collected on 15-sites of contaminated zone

iffering from each other by the level of gamma-beta-irradiation.

'ig.1. shows the deposition of these sites around the reactor..'he nultlber of seed samples from each site ranged from 2 to 10

.nd more. Thus, we examined the essential part of natural plant

iopulations of 15-20 species of aboriginal phytocenosis The

'ollowing indices of seeds were studied: the mass of 1,000 seeds,

.he germinatic ability at laboratory and field conditions. In

'ome plant species we analyzed the chromosomal aberrations in root

ieristems of the germ>nated seeds. We tested also the reaction

>f dry seeds on gamma-irradiation with high doses. Differences in

"adiosensitivity of seedswere evaluatedin terms of the depression

~f seedlings growth and of mitotic abnormalities in roots afterthe exposure of dry seeds by gamma-rays

Examining such important population characteristics as mass

of 1,000 seeds and their germinantion we did not find any signifi-ant differences between groups of seeds of the same species

collected at different parts of contaminated zone (Fig. 2-5) ~

This is, for example, tne results of analysis of seeds of

Setaria viridis, which is an annual weed widely spread in

aboriginal phytocenosis. Pig. 2 shows that an average value

of mass of 1,000 seeds (Fig. 2, dark columns) is maintained

throughout the period of 3 years in the course of our moni-

toring. Abundance of seed samples in 1987 and especially in

5988 gives us a possibility to understand a variability of

this index in Setaria viridis on the observed territory and

demonsrate that this variability is rather small as a rule.It can be observed on all sites of radioactive zone without

any relationship with the variability of intensity of radio-

active contamination of habitats. The same is true of the mean

value of another indices - the germination of seed. (Fig 2.

dotted line). As for variability of the germinatic abilityof seeds collected at different sites of the zone it may be

explained on the assumption of different rate of germinan-

tion of seeds of this weed.

A general survey of Pig.3-5 indicates that analogous

conclusion can be made concerning 3 another plant species—Alopecurus pratensis (Fig. 3), Echinochloa crusgalli (Fig.4)and Plantago lanceolata (Fig;5) The only conclusion to be

drown is that by now chronical irradiation of natural plant

populations in Chernobyl zone does not seem to have played

any distinguishable role in the variability of this two

important population indices of herbaceous planta.

We did not find any significant deviation from some

mean value of intrapopulation variability of such important

indices as a production of seed (Pig. 6-7) of different plant

ader investigation over a period of 1987«1988.

t observe any visible shifts of reproductive

if wild herbaceous plants grown all over 30 ki-

me around the reactor.

inalysis of the frequency of occurence of aberrant

.mbryo root of germinated seeds o Dactylis glome-

eta tinctoria, hgropyron repens, hllium fistulosum

wt in 1986 did not reveal any significant differen-

individuals from different zone sites (Table 1)~ The

.ue of Plantago lanceolata seeds of which were inves-

'er a period of 't986-1988(JeK.Z),

ie course of examination of seed populations we used

i of additional gemma-irradiation of dry seeds with

3 with a view to reveal the "hidden" genetic varia-

plant populations. But, the use of this method did

my additional inf ormat ion.

',amma-exposure of seeds of P lanceolata, however,

in 1987-1988 revealed some heterogeneity of popu-

this species by this time, The seeds sampling from

iwing on the territories with relatively high level

ion appeared to be more sensitive to strong gamma-

in. It should be noted that this peculiarity could

~th in the depression of seedlings growth (Table 3)(s.~~.~3

ncrease of ab'brrant cells in root merist em'erif dry seeds by gamma-rays before their germination.

noted also that among seed samples collected fromRied

p'ly contaminated zone sit e as/were one seed sample

bs sensitivity to irradiation did not diff er from

Vield of aberrant ana-telophases in meristems of seedfrom different sites in Chernobyl

Table 1

glomerata L.

1986

Genista tinctoria L. Agropyron repens

1987

i hllium fistulosum L.

e, :Cells with abeeDose rate,:rations, % :mR/hr

:Cells with aber:Dose rate,:rations, go .mRghr

:Cells with :Dose rate,:Cells with:aberrations,g:mR/hr :aberrations,',

2,76+-0,31

2,20-0,69.

3,94 -0,384,07M,59

0,3

5 '

672-0,07

1,51-0,85

0,1-0,5

5,02,00

1,74

contpol 4,92-1,082,58-0,70

2,82-0,47

2,76+0,77

2,2-0,61 20,0

40,0

0

1,81-0,48

20,0 1,702,84&, 2

1,6A),2940,0W, 50 6,04-0, 63

seeds collected on relatively clean sites. Thi s phenomenon

was registered while using both criteria of radiosensitivi-

ty.

We suppose that all these facts of different radiosen-

sitivity of P, lanceolata seeds collected on sites with

different level of gamma-irradiation post two and three years

of Chernobyl accident may be ascribed to the effect of radia-

tion. High level of ecological plasticity of P.lanceolata

I1I, its metal-and-radiosensivity I2, 3» gives us the possi-

bility tO use this plant species as a natural test-object

which is convenient with a view to radiation monito~ ing of

the environments

References

1 Zhukova L,A. Genus Plantago// Biological phlora of

Mosoaw, i'. M: 1983, P. 188-209.

Preobrarhenskaja E.I. Seed radiosensitivity, Moscow,

1971, 232 P.

3. Pollard A.J.//iNew Phytol., 1980, 86,, 109~117.

field of chromosome aberrations in ana-telophases in Plantagoi anceolata L, (root s)

Taole

1986 1987 1988

te:Cells sco -:Abnormal, fo:red.:Dose rate,:Cells : Abnormal, 5 : Dose rate, Cells'uH/hr :scored :mR/hr :scored

:Abnormal,

739513638573

1,5-0,51,3-0,52,0M,52,4-0,8

1,8-0,28

0,020,10,10,30,32,0

1520

200

1989. 2382

65829151141:2693191629512626

0,4-0,21:,9-0,90,8W,41~3~0.4

1,8W,51,3-0,31,8-0,82,8-0;61,6+0,25

0,02 1.1170,05 12530,15 15590,7 11531,0 1642

10 85715 113915 59570 2652

3,5~,74,2-1,02,8M,62,9-0,42,9-0,41,4~,51,0-0.32,3-0,71,4M,32,5-0,4

Table 3oefficients of growth depression of Plantagoanceolata s edlings from different sites inhernobyl after f -irradiation of dry seeds

:'(0,02-0,2),

+25,4+20, 8415,7-14)9-11,4-12,4

9,2

Dose rate, mRghr

(1-3)1987

-18,5-21,9

-21,4-31,8

35 t3-33~5-56,1

(<5-20)

+3 t 7-10,2

27 p3

-34>21 1 $ 1

-42, 2

-64, 4.

(0,12-0,15)root shoot :root:shoot :root :shoot

1988

(0,6-0,8) : (8-10): (10-15)root :<hoot

-30,8-26,2-54,3

-23,0-20, 5

-46.,0

-23,4 -25,6-45,7 -43,6-67,2 -54,2

-20,9-41,9-54,8

-16,2 -56,2 -55,8-39,7 -72,5 -61,8-45,8 -68,4 -61,8

its only:fects of stimulation of seedlings growh

Table 4

Insreasing of chromosome aberrations frequency in root meristemsof P. lanceolata L. after f -irradiation of dry seeds (1988)

Dose rate,mR/hr

ControlCells :Abnormal,scored

200 GY 500 GY

Cells Abnormal, ~~>: Cells Abnormal,scored : scored : scored

0,030,050,150,71

10151570

11171 253155911531642

9571139

5952652

3i5"Ci74, 2-1,02,8-0,62,940,42,9-0,41,4-0,51,0-0,32,3-0,7'1,4-0,3

1457450

822218

1030

15,3+-3,610,5-2,9

28,5-5,749,0-13,010,9-2,4

3511881634550327869193777

19,8"3io18,8-1,731,7-4,123,0-3,828,5-7,236,0-4,252,5-6,420,4-2,6

.+', ".f l.>.; ic';. i~$ (~ ~;..g$..y eg <i (i. ClP . ~c.2. 5s>cv sip

('-' (',,g,,g l.',:.2.:..L4,'~, /4p ( 'c'-ai

j 'I ) j IB(,i(,.'z'i

y-..n.E 3j 5 '~THE ROLE OF AGHICULTURQ AND NATURAL ECOSYSTEMS

IN THE, INTERNAL DOSE FORMATION IN THE INHABITANTS

OF A CONTAMINATED AREA

M.I. Balonov, I.G. Travnikova

Research Institute of Radiation Hygiene,The RSFSR Ministry of Public Health, 197101, Leningrad,

UoS.SeR.

'he laws of the.radionuclide body intake bp the i~>obit'~~'s

~ territory exposed. to radioactive contamination are evident-

iependent upon the numerous natural and social factors, and.

> the. special features of radiation exposure as well.

~g the natural factors we pick on the soil-and-.climatic ones

the area'.s pertaining to a certain biome. The most Moor-

; social:factors're the: specialliatlon of agricultural

iuction, the structure'of population diet, and also the

.-em:of measures of. radiation protection fx'om.the internal

~diatlon..The main characteristics:of raalation 'exposure

vive'" duration, isotopic. composition and. amount of the radio-

ive:fallout, season .and climatic conditions, under which it,'.

1X'ed'.his

report concerns the radiation situation formed.in:the0

tern part of the Bryanak::region-of the: Russian Soviet Fede-

, }.ve Socialist Republic after, the Chernobyl acc1.dent,- ln

ticular, to the . consequences of soil. contamlnatlori, with

caesium radloisot'opes. This'erri.tory ls situated. between

latitude 52'-54', North. and. the longgitude 51'-0l'. -East . at

-230 kilometres'istance in 'the: north--east,:direction from'f

the Ghernobyl nuclear power p3rnt. Its climate 1 temp.rate-

continental, the relief is flat, soils are mainly podsol ond

turf-podsol with the mineral salts deficiency promoting the

migration of caesium trough a "soil-plant" chain. About a

half of the local population lives in villages and is occupied

with plant-growing and meat and milk cattle-breeding. The

rural population lives mostly on the individual farm product.;,

and uses the natural products - fish, forest mushrooms and

berries.

The radioactive fallout of a well-known isotopic compo i-tion occured in the last days of April 1986 follovkng the

accidental release at the Chernobyl nuclear power plant.

Because of a warm weather in spring 1986 the cattle had been

taken to pastures by that time. The density of caesium-1S

soil contamination exceeds 0.1 T"q/km on the are- of more

than six thousands square kilometres. Ir the so-called

controlled area of about tv<o thousands square kilometres,

where the contamination density is above 0.6 TBq/km, the

active measures of the population radiation protection have

been continuously implemented, in particular, fields plough-

ing and applying fertilizers pastures improvement, increased

nourishing the cattle with roots etc. Since May 1986 the

population has been provided with the red'.ationMree meat and

milk products imported from other regions. Zn August 1986

the individual cattle was bought up by the state agencies.

The inhabitants were informed about the necessity of refusinp

the consumption of the local animal food products, green-stuf

he present investigation gives the results of the analysis

he dynamics in caesium-137 and -134 content in the africul-1 and natural products produced or stored up in the western

s of the Bryansk region in 1986-1989 and consumed. by the

1 rural population. Information from the sanitary and epi-

ological service and from our Institute obtained by employ-

the gamma-spectrometric, beta-radiometric and chemical

ods - about twenty thousands measurements in all, has been

in it. The analytical methods are certified by the metro-

cal service of the USSR. To reveal general laws aU. the

urements data of caesium-137 content in food products

kg) from different farms were divided by the known caesium-

soil contamination density (GBg/km ) in the area, where

oroduct is cul"ivated, and analysed in common. All samples

ained both caesium-137 and caesium-1W. The Cs/ Cs

vity accounting for 0.5 ! 1 in May 1986 later on decreased

rding to the exponential curve with a period of two years.

o estimate the radiocaesium body intake with food products

he rural population and to ascertain the effectiveness of

ective measures, in September&ctober 1987 we questioned

oughly 585 adults {193males and 392 females) about theirBIld sources of supply with food products bef ore and after

Chernobyl accident. The radiocaesium whole body content

measured by the scintillation method in the same group of

>ns from 1 to 8 times.

he main materials obtained by the processing on a computer

he initial results of radiometric measurements of samples

and by the public poll are presenteu in Figures and Tables 1->

By Figure 1 we can judge of a primary monotonous reuuction

of raaiocaesium concentrations in the agricultural food pro-

ducts. The radiation level in early green-stuff ( parsley, dil

onion, sorrel) contaminated directly from air in April-May 198

decreased by 2-5 orders of magnitude during several months.

On the contrary, the radiocaesium content in potato and roots

(beet, carrot and others) was low and changed little during

three years. Data of the animal food products reflect the

process of the initial accumulation of the surface pl~t and

soil contamination by animal organism, its metabolic clearance

and subsequent consumption of feeding-stuff with different

radionuclide concentrations. Caesium-157 milk concentration

dynamics had a decreasing wavy-like character with a fluctua-

tion period of one year and is well approxim-ted by the

averaged equation

c(t) (s~=») e+(5 5 '02) C

of caesium-157 + caesium-1W with

oc9(.5;6.or) e(1 ~ <i~ 2r t)

p1 GBq/km'f caesium-137 on

soils. The curve maxima fall at a period of pasture and tne

curve minima — at that of indoor maintenance of cattle with

combined feeding with hay, roots etc.Similar season fluctuations of the radiocaesium content in

fish can be seen (Table 1), although a tendency to the time

reduction of indices is more pronounced here, reflecting a

radionuclide transfer with runnung water, its fixation in siltand deepening. The tissue specific activity in raptorial fislis higher than in other species. Strawberry was contaminated

for>i.'vii'io i 1fi 1)i a1 J'-i e ~ Lcd'ic cctlvit j;;o id..pily uo xmiuo it on o mcica (m t of rcMo»

um v:a. os o.'."zal tn m3.ctu" Wtouu) and uc'ot .~ctical-~a cu for turoc yoarG.."h~. rolo of shroom ac a uourco

ra-iocaociua ooug inta;o by tho L~bobitanto ha" rolc-iacrooaod v~ith tim.. oto that up to 9J'. of ramo-

.us ~ trcnaportM into mlution in oool-.in„unct boilinG.

.bio P. ala~a an avorago uaily oonauvption of tho min

:uJ.turci foou ~~uuctG and a ronge of conau~~ti.on of

al food aroauctu by tno c~t <~~bitanto, tm latterm-t domvdont urn t>u inuividual ta"tea anQ habita.

"ta me obt'c.ino'rom c soll of roak~t" of nine

.",o'nd az'o in ~oi corrolation v:ith tuoao of the

.iul -tati"tic". 'opo~r:,ith tno loca. foou rxemct

:.'"i.l;. t'ai~ contr„~ co="ux u lt-:: tu fidit;c fc.t;, ga t, au",ar, tLi M foou an "" on.

:Jm..in„- th« 'uurnobgl accirJont, c~ it re rovoalai

Cho uiatrict;;aa highl„conte'31uctel viith raMoactivc

;unco end, ocmciclly Gfter uuyinp up tho individual

i": liuent 19<'u> accordin„ to roco~ondationG of tho

;a"y o~ido~lo~ical service tho codou~ion of tho

L food rrvoducto mo considorGbly Qccremel (I'inuo 2 and

, 5). who co~u~ion of local Mll: vioo mat docroaooJ

.5 tinoa r.iron< in 'My, by 2-2IS tinea in uxor and by

iintor it'had fc lion to 2" of tho initial lovol. Dy may

>am:ication for duficicnc„'f local aninol nutrition the

~itcnta huvo baur. i u i&iu-'"itv tlirt ro;~lcr iwportoii

L'4 LL t.ioa 1'1'UL.'OC olsu IfM f004 'LL'o~c L ~ L.Lail;UL fIl 4i<.

cmn-umtion of frv~h tutor f~ uuJ ~~'j~~w in T)~)i.'lc>~~acta~'&o +4a frequency ~txiUu&om; of tee

1»~bitGnt'y

cno 1cntol nf .consu~&ion of thoso ftxxl ~oductc,. L.s itio soeo fear,'„~ Table 3, after the Ghorxubyl Gccidont tie <~i~~

bitants eetirsly refused to otoro uo thcao natural gvoduct"

(M ) and mmy of &~ (66,') Gto~pai ~tiurin~™ ~~oooo ot011. Ilote that hors aro peosentod the ~4xl cd;o on the

~~~bitanto of four vi>>~~a in Cho controliod aroo. i1utMuc

it'ouadsry, chan~a in the diot uw 1~ ouMtantiol

ano of mlti&ication of tie:vora;c raducoosiun

coucont in food abducts fro~ Fi~ 1 and 'ublo 1 by the

0~unto of their conmaption fro< i'oblo 2 azxl by dint of Cho

//'~lmo..uonttim into~Cion nc obtoinod c'. ~i~to of ti~J„m.~le;'1th Ch". w~crl w Y>a of dict

( s able ~') '4 a<~9'i@el v~o fQO. l l Ql314 ~t ijLQuLl'/

rc~~iocl c" coo'iu~lg ovcrag. Ml) body intcQx (Dggdhg)

ir Ci~ Gazoo aith ~ conm~.t.&n dsns1ty of 1 GDq/~

viewed tee.t the diet boa unchon~ boforc and aftor the

rr i<occtivo contminotion. So for the noturol food pcoducts,

the cstimtos of their consume;ion ranGG domndinp

on Cho <~»»@tents'ooD habits In the l~~ "2otal" Chc

lour list rolctos to a standard tarn consul@ only ~icul-tural food products on Cho ovexooo lcvol, axial tho up~ ono

- to o man, ebo ~~<tio~e~ <y conou~ 0 good doa1 of naturcl

food;~uucto! up to O.1 kg/day of fish ond ~~~i~~ ocul

nloo;~on- Cuff, and u~ to U.D> '~pity of fore.",t borrioc;.

iota fro"1 ~'QDlo 0 ~Ai~l ti& r:Gll+zioi.ili Mcl~i"".1 - .;ExxQ

Let. l nor W, tao mat ox'aAocuvuiua uAcra Chv or.-~>iam

tuo inaabltaato riith milk un'Rc,~duce - '7~&- in ~i

ia, cbo ilo noC conou~ natural foode. our'inG Cho firotitha after Cho rauioactivo follout a co~arablo uxor ovm

abator contribution to tho body intake could ~ tbo inton-

re coocuc&ion of prom-atuff and oarly borrioa. ~tor on

halo of vo~tablo ~icultural peo3ecto and o~ mac aaalL

i cacatu~ body entry uk,th ~t averosod, from 'k7.~ CoM'~"

the ~od of obocrvatlon ln the folloriln~~ ooaaons

eio of foroot mushroom arel fiah ae sn important oourco of

: ~ioiuxci.ide humn boy iaco~ration eai. rolutivvly

~ iar.. Zn ~Mco ie olxmrwod a stoady docrca'e of tl~

.iocm"tua food conta~tion (+ m> order of m,~itudo for

~i~'u) ~ co~0.ju~cc .)f Jio~ 1hc~ ch'Gi;in~ fr > .l Clk~

ilia@ t~wmrtaM.c fora. 'horoby ita conteaxt 4> tho ~ot

by W< Ci~ in ovo~ in Cho mev~ aeaJon f3(.'G

tlx. ourfacc deconta~tion, and for Cho folio~i year- it docreaaed aGain by 2 tines.

"~auroa adoptoQ. Co protect tho «>~bitanta from internal

.~ation arul, Ln the firot ploce, thc aubotitution of tho~ of eussy ~vith mat and ~ rmducta (ooe I gyxro "-

< Tablo 5) i:~~y chan~ tho lovel, dynmico areal co~"i-in of the ra'cacao"iud body intro by tho <~~~bitanta of

u'illagoe of the controlhxl aroo (Table 0). uata for

ilo 5 cicero calculatod oo for Table 4, but with re;ard for

~~~ in tho conan;<ion of local ~icultural and natura.l

food:~Jaactu. I'4e rc~iocuo iuw body iricor.mrction x,1th

foodU vicu'ctuoLly docruouod by e-5 tirna i'u~ t)tQ.inco outuna, ofter ronevi.ng tho doir~ittle fma tbo arivnt~

f~, tho total body intoho hoo docrooam by ~ t~ Go

co@pored mith routine conou~ion of local focal ~ducto.'Zhoso ries also offected tho dynaMco of Co Ozxl Ca

body cxmtent, which L" not considorod in detail horo.

Change of' rolo of difforont Groups of food oroducto in

tho r -<iocooo4n body intake by tho 4~~bitonts ottrccto our

attention. As the dairy produce mare arocticolly ontiroly

substituted for thooo froo of contmination, thoir contr ibu-

tion to this vctuo ho" docrocnod by pro than on order ofmagnitude sinco autumn '|966 Tho co~mpeion of local +oct

has boon roducod only by 2-2.5 ti~ one, according to our

c"ti'%Co, it i<~ laco~ ono of tho ~; foo'-~tuff hl

ctrocturo of tho ~iocoosium 1m'ncor.mrotion in tiecontrollod aroo. A role of citato, vogotobloo and oiMlor

ducts has sommvhat incroosed. As ~ol, tho cmunt'fcon"u~ mushroo~ snd fiah o~ent~e>>y Gffocts the lovel of

raliocaes~ hxly intake, but in case of'ring the by~ionii

limitations and gathering muahrooms as "before the occMcut"

they could bocomo the min oource of intornol AvscLiation ofooae rosidsnta. ~o a role of natural ond ~cultural oco-

oystom and their food ~otucts in the @oooo internal irra-diation under conditions fo~ in tho nanak region afterCho Chornobyl accidont is ecgunlly portent both with routino

unchain;~ diet of tbo x~1 inbabitonto and uitb the utrict

4 1'OatL ictiol!G i"t'X)UCKL ~

by Cho Jato fred kubla G one eau .,ucuAQ.'y ~ effectivo-

of tao ..xetoctivo ~aalu odo~id. to docrc~o who intor-

. irraliGtion of tho ~i<mbitanto of tie contro11e6 argo in-. r~ne ~~~lye in tho f~+ ~os tho aubotitu-

in of food ~ducto, end secoa~ t4o ~pote -~~<col arrnnGo-

;to in tbo peivato forego. For "top and tho ou~ secern of

4 tho ovorcgq doily intadce cd! reliocaoatuz bg the 4~bi-.'tB~ the ti86 into~cL1 of ~<~> iB pKoport10081 to tho

nittoci internal dose, aaG rcduco6 + ~X" on the averago

~r ~ of the o5oated protective boa"urc', end fms subunit

QGS to outu~ of 1907 it +as lecxuec'L bp ~9P 2bo

;al hl+ effoctivmooa of ~6" eau~; of the inteme1

.niic,tXon doae durin the fixate 1 5 clara after tbo Chosxabgl

:idont, oxxl conocqucnt+ cbout ~M Dovtng of ttae ~lo ~.cxncLl and, in~ornol doooo for who fr~'uitanto of tho affectsio undo'~~ conf~ tho axpodicGce of Cho cooductml grrr.

:t'ive ~urea.

Cs, tO k~//kye e [ e e e e

)e e e e

/e

Q

9I'ccrc

stuff

t0-'l~qmiCk

~~'L. miCk

l I

i

grcctl - ctuf f'eeetebtes I

PZZP/P7/iTS.0,1— potato/ footsI

, Su, A,VY,Sp,SuI

1986 '! 1 9

f j9.1. DTANDAItteZ ED (Dq/leg)/(GBG/ka ) s 10 ka /kg CLEiZUW-1S coliTsJZ

Ill hOKR LOCAL AGHICULTVRAL fOOD fBODOCT5

(3-

0,1- morat'

meat

(03- It99$ e99$

miCk

00<aonth os,, ov,, os,, ~i, I

01~,03,,os,,07,1985 0 9 8, 6 I 1 9 8 7

F<9.2. CHL!6B Zll THa', AV aRlCiB COl4UVPTZON OP THE ICAL Al<ZHAL FOOD

HlODUCTS HZ THB ADUlll NJRAL FOPULATZOH OF THB COHTROLLBU AREA

AFTER TH CZMUloun ACCZDBHT (ry'aaF)

STANDARDIZED (Bq/kg /(GBq/km ) = 10 km~/kg

CAESIUM-157 CONTENT IN THE NATURAL POOD PRODUCTS

1986 1987 1988Pood

productsMay- Autumn- Spring- Autumn- Spring-

Summer Winter Summer Winter Summer

Lake fish 22 —11 9 — 0 10 — 1 2.5 —0.4 3.5 - 0.4

River Xiah 30 —5 14 —2 16 +- 1 4.7 +- 0-7 6.5 ~ 0.8

P orestberries 5 c 40 < 10

CONSUMPTION OF THE LOCAL FOOD PRODUCTS BY THE ADULT RURAL

MRJLATION OP THE WESTERN AREAS OF THE BRXANSK REGION

BEFORE THE CHERNOBYL ACCIDENT (kg/day, one year average)

Agricultural food products

Animal V eget able

Misc and Meat and Potatodaix.y meat Eggs Bread and Vegetablesproduce produce roots

Green- Pruits,stuf f berries

M 0.78 0.20 0.07 0.59 0.67 0.30 0.01M.1 0 ~ 15

F 0.58 0.19 0.05 0.28 Oo56 0.28 0 01M. 1 0.15

SexNatural food products

Local fish Mushrooms~

Porest berries

M 0.04 OWo 1(0-0.1)OW.05

0.025(0-0.1)

OW. 1 OW.05

Tablo 3

DISTRIBUTION (fi) OF flic I~T r4DAL EORJLATIOD

OP THE CONTHOLLi2) lU(L'I BY Tlh JaLVhL OF flATURAL FOOD PRODUCING

CQRRN'fPTION BEFOkb; Ail~ AFTER THi'HL2QiOBIL ACCIDIIIT

1985 1986 19s7 1985 %86 1987

Consume moderately,in a season

0 1 46,'' 0 1

Table 4

mmuC. PF T ntuWCur.m DOUX IImXu a C~IUu-1%SX Till'DULT UU~ LUHASITAl64 UITJl BOUTINS

DIHI'986

ProductsMay- hutuma-Uumor Qintor

1987 1988 1989

opc1Dg hutlgtl b pviQg- huttCh5fntor 8unoor Ufntor

~un'Q

rmlmn1 5 2.5 2.6 1.5 1.6

0.5-17 0,2-2 0.2-1 0.2-0.4 0.1~.5

Datarol

2-5 2-5

TabLe 5

UI;~IIC., uP r „, ...„,.h.:~i~ Don error~ UI cs~luu-1%UI giu /uiU~Z ."Ji~~ I:eiiJ'I'ii82 (K: Qu'Pii~iRu m rJ~,

Gf M. D(Qil:i h i'lp.. Li'ii aI T pili'li.J i'"I 3$l'*CTI'ilw 'U~lh~

t+@Autumn Jinter ~grind mare hutusa

u.o o.5 o.o o.o o.5

hgricultumlvegetable

o 5-5

8aturol o-o.9

9-15 u.9-2 7 0~4 2 2 0 5-2 O 0 5-1 6

Toblo 6

GTAllON&Z Li) L'Oi)X IIi~ls.i. GF CAa'P< l% (10 knP/dang)',tITll DIFF~IT i)LM /AU Zii rIVaCTIVQ~ (Q GF Tie'l>gakdiAh

IHbADIATIOH M7i~IOll Gl TiL. JQ%LtZIOil IX'GIrIS30i ~

Diet'la~

iallCQLRl '«iDCQx'pcillS w~ /gag+

17-% 6-11 ~ ~ ~ 3-7

09-27 0422 05-25 0&6 0&)

47-76 75 )2 M6 77~99 k&96 GM

CYTOGLIIETIC EPPECTS OP IONIZING RAIIATION IN

CEEPIS TECTORUM POPULATIONS GROWING IN 30-KI-

XOIZL'THE ZONE OP CHERNOBYL ATOMIC POWER STATION

ACCIDENT

L.I.Grinikh, V.V.Shsvchenko

N,K.Koltzov Institute of Developmental Biology Ac.Sci.USSR,Moscow, USSR

The aim of this work was to carry out the cytogenstic investi-

gation of the effects of chronio irradiation dur5.ng a numbex of suc-

cessive generations on a wil4 plant Crcyls tsctorum growing in the

region of Chernobyl accident. C.tsctorum is a very suitable sub)act

fo. cytogenetic studies having four pairs of large and well distin»

gu:shed chromosomes. C. teotorum is a widespread plant in the 30-

kilomatrs zone of xadiation pollution caused by the Chernobyl acci-de: t. It is especially convenient to conduct fast soraen1ng of chro-

mosome aberrations in the oells of the apical rant msristam of the

gszminating seeds.

This study «as carried out with the seeds of three populations

C. tectorum collected in august-september 1986 from the plots with

tbe dose rate of gamma-irradiation 5«10 mR/hr (in august) and the

scads of the other six populations oollsctsd in June-July 1987 from

th plots where the dose~tea were from 20 to 0.02 mR/hr+(the dose

zates of beta-irradiation were about 10 times higher). Approximate«

ly 50 plants of each population were cut and put into the water inthe laboratory wheze every day during a week seeds were collectedfrom all plants of each population in one sample.

To analyze cytogenetic effeots of ionizing radiation continuous

colchicine treatment was used which made it possible to distinguishclearl» the metaphasss of the first, second and third mitosss and

thus to differentiate between cell populations with different peri-

ods of the first cell cycle. Seeds were put into Petri dishes on afilter paper with a 0.01 5 colchicine solution and then Petri dishcwere kept in a thexmostat at 25 C. The emerging seedlings with 1-2mm rootlets were selected every 3 hs and fixed 2-3 hrs after ths sa

lection in alcohol«aoetic acid (3:1).Squashed acetocaxmine prepa-rations from root-tips wars made for an analysis of chromosome aberrations. In each root tip no mora than 50 metaphasss were scored.Seed gezmination wae analyzed during 14 days.

The analysis of germinating capacity and gezmination rate reve-aled ths following> 1. in populations studied gexminating oapacityvazied from 28.9 to 86.1 5 and there was no correlation between thivalue and the dose rate1 2. there were great differences in the gs)mination rate between the populations studied (gezmination began1-3 days after imbibition and finished 4-13 days aftex it)t an4 thire was no cozrelatoin e1thez between the gemination rate and thedose rate. Taking into account these data we used the index "pexcent of gszminated seeds of the total number of gexminated seeds iithe sample" to make adequate comparison of cytogenetio effects 1ndifferent populations. All the material analyzed was divided into4 groupss early gezminating seeds (the index equaled to 1-25 5),2 groups of mass gezmination (26-50 % and 51»75 $) an4 lately ger-minating seeds ( 76-100 % )~

The data on the frequency of chromosome aberrations in diploidcells of root tip apical msristsm in C. teotorum seedlings from thpopulations chronically irradiated for 4-5 months are presented inTable 1. As assn from this table the frequency of cells with aber-rations in ths early gszminating -aads wsa considerably below theaverage, that is at first ths seeds with very small damage of rootmsristem gszminatsd. The highest frequency of damaged cells waa ob

-4-effects in the other six populations of C. tectorum. These effeotsserved in lately germinating seeds of population 1 (10.2 - 15.3 %),were low and mors than 1 5 cells with aberrations were observed onand among them 7 seedlings with more than 30 $ of damaged cells we-ly in populations 4 and 5 growing on the plots with the two higheere found. The frequency of chromosome aberrations per 100 cells was

higher than that of damaged cells, that is there were cells wi>hdose rates 20 and 1.0-1~ 3 mR/hr (1~ 4 + Oo20 and 2o2 p Oo32 $ res-pectively). In the other four populatdons the frequences of meta-multiple aberrations, especially in population 1. Almost all aber-phases with aberrations constituted only 0,3&.9 $ (the dose ratesrations were of chromosome type, there were only 3 chromatid aber-for populations 6, 7, 8 snd 9 were 0.2-0.3; 0.05-0.07; 0.10 andrations among 568 aberrations scored (altogether 10031 oells were

scored). We have shown earlier that cells of the root meristem ofthe embryo of 0, tectorum doxmant seed are synchronized at 01 phase

of the cell cycle and it is known that only aberrations of the chro

mosome type are observed after irradiation at this phase.

0.02-0.03 mR/hr1 the average per cent for all 4 groups were 0.6 +

0.191 0.9 + 0.27; 0.3 + 0.13 and 0.5 + 0.17 % respectively) ~ We ha

no control populations but irradiation level for populations 8 and

9 was very low and only 0.3-0.5 $ metaphases with chromosome aber-rations. were observed in them. This value was not higher than that

We notioed that C. tectorum seeds gexminated in such a way thatfor sponteneous aberrations for C. teotorum. As to the frequency oimmediatelv after a seed coat was broken a root 1-2 mm long with a

lot of dividing cells in its root meristem appeared. So we appliedan indirect method to score chromosome aberrations in the cellswhich entered mitosis early and passed it before a seed coat was

aberrations there were no significant differences between the fourheed groups. Multiple aberrations were very seidcm. Most aberratio)as in the previous year, were of chromosome type (121 of 131 aber+tions scored). The frequency of chromatid aberrations was 0.08 perbroken. We analyzed the first tetraploid oells which appeare4 among100 cells. '!ll in all we scored 11825 cells.the mass of diploid cells and extrapolated the data obtained on di-

ploid cells from which these tetraploid cells were formed. Such

analysis of the seeds from populations 1'nd 3 collected in thefirst year after the accident showed that there were

chromg?fag'omparing

the cytogenetic effect of radiation in the material oi

the first and the second year after the accident we can concludethe following: 1. the presence oi multiple chromosome aberrationsin root meristem cells in the first gear after the accident and theaberrations in the early dividing cells. The latter could be indu-absence of them in the second year probably suggests a certain con-ced by the effect of incorporated radioactive products at 02 and/or

S phases in the course of seed gezmination.

Unfortunately the populations of C. tectorum analyzed in thefirst year after the accident were destroyed because of the resto-ration works and so we could not follow cytogenetic effects during

tribution of densely ionizing particles to the cytogenetio effectsin the first year and that using C tectorum test system made it po;sible to reveal this effect; 2. the presence of chromatid aberratioi(which were revealed in the analysis of tetraploid oells under con-tinuous oolchicine t=eatment) in the first year an4 their absence iithe two successive years of chronic irradiation. In the second year the second year indicates that the cytogenetic effect of incorpora-after the accident we obtained the data on the level of cytogenetioted isotoas was >resent in the f'"~if ~ri~ ~ ~ " ' ' ~ i~ ' "~-

-5-test system made it possible to reveal this effect.

In the second year of chronic irrad4,ation besides the seedlings

with chromosome aberrations in some cells of the root meristem we

-6-for this structural rearrangements (karyotypic analysis of the api-

cal stem meristem could confirm ;his but in our work we dido% make

it). As to those changes of karyotype which were observed only infound seedlings in which all the cells of the root meristem (or part one seedling it is impossible to be sure th t in these ca es t-of them) had the same chromosome rearrangement1 in other words, we

found seedlings with changed karyotype. Altogether we observed 21tional events had occurred before meiosis and then had passed thro-

ugh it. They could be induced in the course of emb~nic developmen

such "viable" rearrangements which did not destroy mitotic and and then got only into the initial cells of the root (but not stem)tic cell cycles. In Table 2 all the cases of structural changes of

karyotype in the 6 populations analysed are shown. As can be seen

from the data obtained we observed not only reciprocal translocati-

ons and inversions usually passing through mitosis and meiosis but

complex rearrangements with increased or decreased amount of'enetio

material (Pig.1). Maximal frequencies oi'uch changes of karyotype

were observed for the highest dose rates (in populations 4 and 5-I6i2 and 6.6 5). 20 of these aberrations were heterosygous and obser-

meristem.

Thus the appearence of the plants with changed karyotype confirm

the presence of active microevolutzonary processes in chronically

irradiated populations. The analysis of karyotypic changes of plant

exposed to chronic irradiation also elucidates the role of the kary

otype rearrangements in the genetic adaptation oi these populations

The data on the frequency of chromosome aberrations in the popula-

tions exposed to irradiation with different dose rates allow us toved in 1, 2 or 3 seedlings and one (pericentric inversion in C-chro- recommend C. tectorum for use in biological dooi etry of the levelsmosome) was both hetero- and homosygous (in population 5 this inver-

sion «as heterosygous in 17 seedlings and homoeygous in 2 seedlings)

We registercc altogether 17 reciprocal translocations (between chro-

mosomes A, B, C and D)g 6 inversions (in chromosomes A, B and C)i

one complex rearrangement which caused the change of chromosome len-

gth and the increase of chromosome number up to 9i one rearrangement

with extra chromosome fragment in chromosome Di one deletion in chro

mosome C which was observed only ln a few eel+ cf the root meristam

(this inversion was probably induced during embryonio development).

As to the structural changes of kaxyotype which were heterosygous

and found in more then one seedling it is possible to conolude that

these chromosome rearrangements passed through meiosis and were pre-sent in micro-and macrospores (C. tectorum is a cross-pollinated

plant). Hence all the tissues of these seedlings were heterosygcus

of radioactive pollution, especially in complex combinations of di-fferent alpha-, beta- and gemma-radiations.

w7w

Table 1. The frequency of chromosome aberrations in root meristem

cells of the seedlings in three C. tectorum populations

(the first year of chronic irradiation)

)opu- group of.ation seedlings

(5 from thetotal numberof germina-ted seeds)

&WWWMWWWWWWWWMWW WWWWWW

seed- cellslings, Vo%aX vries cSromoso-number number me aberrations

number

aberrationsKola% ""per 705

"'umbercells

1-2526 - 75

76 - 100

1-2526 - 75

76 - 100

26 884

18 833

20 2 ~ 3+0,50

3 0.4+0.22

55 2267 23 1.0+0, 21

36 1193 23 1 ~ 9+4+ 40

25 1250 191 15.3+1 ~ 02

43 1921 184 9.6+0, 67

104 4364 398 9.1+Oo44WW

11 550 0 0

28

267

239

534

23

26

2.321.412.4

12.2

Oe4WW WW WW MWW

102

1-2526 - 75

76 - 100

31

18

123

1000 15 1e5+Oe38

2000 67 3.4+Oo41

400 8 2.0+0,70

3400 90 2.6+0.27 109

1.74.22 ~ 0

3.2

Table 2. Structural rearrsngements of karyotype observed in root

meristem cells of the seedlings in six C. tectorum popu-

lations (the second year of chronic irradiation).

pution

seed-lingsgnumber

195

changed karyo-tyaes~ w w wo w

number

12 6.2+1.73

description of root withksryotypicsl identicalchanges changed

kargot~e

translocation AB 1II AC 1

sD1I AD 1tt BsD 2II BD 1Il BD 1II BD 1

inversion B 1II C 3

extra fragment+Ds 1complex rearran-gement (9"chromo-somes) 2

roots with

number

16 8.2+1 ~ 96

507 7 6,6+1.10 translocation ACII AsC

AsDinversion A

lt Bff C

deletion (mo-saic) C

1111

KK17+2

23 21.5+1.82

62 0

1 2.0+2.02 inversion 2 4. 1+2.86

75 1 1.3+1.31 inversion 0 1 1 ~ 3+1 ~ 31

9 0 0

R - letters « symbols of C.tectorum chromosomes; As — short arm of chromo-some A; Bs - sho'='"t srm of chromosome B

iKK - 17 seedlings were with heterozygous inversion, 2 seedlings were withhomozygous one.

Pig.1. Structural rearrangements of karyotype in root meristem ofC. tectorum seedlings after chronic irradiation

a - normal C.tectorum karyotype (2n~B);b - complex rearrangement which caused the change of chromosome ler

and the increase of chromosome number up to 9;C - heterozygous inversd,jn in chromosome C.

'.;'TTC <: '.;.,!".1'.t'.;!CR., <;!'?«ART'CTTVP. C !lTA!!T!lATT''!«'Y!ASTB'.,T.,P"PI I IATT"!JS

Alirnm«v V..T., aednrenk? .Ll?, Shevchenl n V. A..

T< e:<timete genetic ronsequencea of radioactive contamination

(<!nnt. nn<l animal . populations, it is neceaanry to study the mu-

ion proceaa dynamics in chronically irradiated ponulationa

It i.. necessary to know the tempo of genetir. load nccumuln-'? .

< in chronically irrndinted: populations,iu generations, tn atudy

relationship <'f? the genetic load to the level of radioactive

!nminntion Adaptability of <lifferent plant nnd animal speciea

the contamination factor should be determined aa well

All these problems are the ob.'ects of our work started in

5.in the 30 km sane of the Chernobyl lluclenr Power Station (Cl!PS)

One of the ob<ecta chosen for-inveatigati<.n waa Arabidopsis

!.irma found in abundance on the contaminated territory. Tl.ia

ll annus! rosette plant from the Crucifern fnmily ia a favourite

<t ob'ert of geneticists This aperiea ia fnat-ripening. hi~hly

tile, aelf-pallinntinp nnd ia yell studied by peneticinta Tt

<ld be noted that natural populations of Arnbidopsis are roT<re-

ted by n mixture of winter nnd spring foms, whirh impede.. n

tie their 'studies under laboratory conditions C?nsidered na n go-

ic load mny be any mutations decrenaing the. fitneaa of their<'iera 'Ye 'at»<tied. the fresuency'f embryo lethal mutation., in p<'.—

<tiona nn<< tbe frequenry af those.'chlorophyll mutntinns that nre

. <. lethal in the honosy(Tote atnte

The frequency of embryo lothnia nndrh!oraphy!1 mutations wna

<,ysed iu '!<e 'iel<1 mnterinl T thin en?!, !he mnin f1ower-bonrlup.

.a were r»i off from plants prowinp an !!.eoxperineutnl aitea,

nnnlyaia, the upper nn<t ya<tnpor pads «ere examine<t with nbinocular'lens

to detect lethnl embryos nnd embryos with rhloranhyll mutations

Tree parameters of the mutation proress nrc used in studying induced

mutagenesis, by the embryo test: frequency nf t!s plnnts producing

segregating offaprings ( m )~ frequency< of fruits (pads) containing

mutations on 14> plants ( mb), and frequenry of mutants in N ( m )

In addition to analysis of planta in the field, the genetic

load in the offspring of these plants grown in n preenhouse «an stu»

died. 50 and more plants from each population nnd 3-5 pods from each

plant were analysed

Table 1 presents the results of analysis of the genetic load

in the offspring of plants that grew on the contaminated territory

in t986 The offspring wns grown under laboratrry conditions'?

As seen from Table Y, there is a relat,ionship between the level

of radioactive contamination and the level of genetic load in Arabi-

dopsis populations from contaminated regions This relationship isapproximated best of all by n linear regression equation y ~ 5,6Y+

0.29x. The correlation coefficient is 0, 932, P ~0<02

In studying the mutation process in chrrnirnlly irradiated na-

t<:rnl plnnt populations one always faces the prab1cm of control since

it ia usual]y unknown what penetic lond waa in n poaulation befor

contnmi.nation Therefore in t986 we attempted to estimate the genetic

land in an Arabidopais population from the outskirts of Moscow (Stro-

Cino) and from site 5 t0 near the CIIPS 30 kr< anne (teasel ) where the

contamination level is comparatively amnl1-0, 3 mr/h The population

fr<a Stropino gre» «n the aide of n heq<>y-trnf fir. hip)<way The level

of embryo lethal mutntiona in thin populntiona ( m ) nnde up 4, t,"

The mutation frequency in the papulntion fr<m '..1"eel ronatituted O,OOX

( n 1 << r mnnr«tt ~ alii utr< s«~<( ~ <roi n< < i <<in < rt w nnn< ~ tnt'<«<

fr<1.":.'.;«<ti<r"-"<,nr'.". -'Tt. may' pr'ob«bly,'' lie «s<s'<<ed,':"'<<ovie".er, .:-thai

iletir l<irid,iu'iiat«r«1 'Ari<bidi! pais p p«~at'irma rial.e: .i<p '1 l

n ) "-".<«."'„''ii pre11:.<i!)i<ry,.conclusion ran, be "<iade thrit" in Ara-

p p«1«tirn's witl< tl<e 1eve1 of initial- ivi li active:rontamina-

i<at: up 'tn -'O.mr/h .tl<e 1olel .of <<;enetir.. 1nad in the, first. gene-

.does snt exceed significantly .the control lete1

Tdb1'e '1

regiIuency of embryo: 1 ethalsi in Arabidnpsis .pnpulatinris

rom the 30 km sone nf:CNPS . (laboratory analysis of. L<>)'

~ 1,

'Dree -rat'e,.'Plant.l

sed as well as the genetir. load iu population .".1 where radioactive

contamination together with <ther constant environmental factors seem

to act as selective factors against plants heterosygous fnr lethal mu- ..tations Reduced UV resistanre of Arabidopsis plants heterosygous for

lethals under highland conditions was rep <rted in the works of P D Us-

<<<anny

Table 2

Prequency of embryo lethals in Arabidopsis populations

from the 30: km zone of CllPS (Pild analysis of 52 in '1987)

Xecalities . mr/h 1number: ma . 'b 1 mctDose ratetPlant :

y "Les:.. '240..: 51 .,.70;6-6,4::.40,3-+3,2 ',G3-0,02v. - .-::,.:60 .:.-: 1.54 -...45;-7-+4,0 —. 27,5+-',6 ., 6',59-0,13

ye:::. -.': .10: ...,198,:.'-.4,64-+1,5

obyl -" .. 10: - 179 '.- ''.5<0-1<63. 65: . 3,1-2,1+

1; ..::- '.: 0,-3 ':,.121;. '' "0,8-0,8'w";

. - '02.. '" .74'' ',; 4, <23:;

N3 Stechanla '. 0,02 . 50 14,0+-4,9, 6,6-1,9 '1,3- '>

N6 Chernobyl 0,5 50 6,0+-3,4 3,3-1,4 0,5-0,1ll1 Tolsty,Ims 2 ' . 55 27 '-6,0 15,9-2,7 .: 3,3-0,2

Kopachi 17,0 50, 42,0-7,0 20,9-3,2 3,8'-0,2

.. 12-0,Yanov 7,0 6G 36,4-5,9 1G,4-2,5 3,2~0,212-1 Yanov '.12<0 58 53 5 G,G 36,2-3,3 7f9«

.12-2 Yannv . 24,0 '

54 64,8-6,5 49,7-3,8 1v,5-0,4,12-3 Yanov . '130,0 ' 50 - 78,0+-5,9 44 '-3,7 11,8-0,412-4 Yanov 240,0 52 80,8-5.5 47,2-3,7 9,1-0,3

ble 2.presents:.the results'of field"analysis-<,f the same 02.ge-o:of Ara< bidopsis that in 1987 grew, on the'aris sites . Qonitoring of the mutation'load dynamics in a succession of gene-

s'relationship of '.tlie embrio lethil freuuenrj to the do..e rate rationswill permit. the regularities of accumulation or elimination of

nic irradiri'.i <<1'oreriented in Table '2 is approximated best of . the. genetic load from. Arabidopsis populations to be established

~:power- function Its form for.m,'„''. io;;respective1y,: is: '.: In addition to the mutation load ana1ysis populations from con-

q4-zO ~ 2?, !y . 1>':59<zO 2" .': v -2 255:.zO 30? ' ;:--',':: . :,taminated.regions the influence'of rhronic irradiation nn,the genetic

is: seen'rom:;Table 2 tliat: fron,,'the'i<piilations iinitia] ly take'n 't»cture,'of. the popu1atlonri "<as sti«lied'16<. nnti<ral Arnbidnpsis pnpu-

'"stipatinri":in-'1 9A'f:-::we'.,'«<a'n'aged 'o'a'nal'yse '<oh'1 v: t<v'n .-'' 1<1 s'iid, „'f,1ations .were'nalyzed - Tlie ppnpul ations. were ..rnmnnred for= eler troph«re-

Xl, 10)H, 6-PGD) The study cnntinubd for 3 years Dose rates

from o,o1 to 250 mr/h for different populations

se results of the study in 1906 showed that out of 17 loci and

sites of the ensymes 14 displayed variability and 3 were mono-

: in all populations The portion of polymorphic loci calcula-

-.h 95% criterion varied in different populations from 0 to 0,69

1, which signi,ficantly exceeds an average level (P 0, 19) deter-

for 33 species of self-pollinating plants (Hamrick et al,1979)(P«0+77)

ll as the level of polymorphismYin mountain Lrabidopsis popula-

,Pamiro-Alai) It can hardly be imagined that irradiation of

ieration in Lrabidopsis populations could increase to such an

the level of polymorphism It would be more correct to suggest

P at 955 criterionYear .'

:average : population N1 : average1

population N15

1986 0,531987 0~46

1988 0, 44

Oi57

Oe47

Oe25

1 ~ 29

1m 29

~~ 22

'e33

1,22

1 ~ 12

Tnble 3

Indices of genotypic diversity of hrabidopsis populations

growing under conditions of increased ionising radiation

bncl.ground

r -0,86PW,33

r -0,97P 0,13

1 r -0,57P( 0,001co ef P «0, 20studied

s it follows from Table 3, in all the populations the portion

dimorphic loci was higher in 1986 and then it gradually decreased

ne was observed in population N1 (Tolsty Les) studied for three

ind exposed in 1986 to ionising radiation at a high dose rate-/h. However, in the both cases the differences in P were in-

icant (Table 3)

+- the correlation coefficient was calculated from the data for 14

lociJn six populations linkage disequilibrium wns found for ~4 pairs

of loci We have noted that the number of pair nf assosiated alleles

significantly differing from the number of random combinations is lar-

ger at low dose rates < f ionising radiation It is probable that high

radiation levels disturb genetic coadaptntion nfgeneo

A direct correlation between —the intrapopulation diversity nnd

the number of pairs of associated alleles has been established (r 0,60,

P( 0,01) Thus, the larger is the number of coadaptive loci in a popu-

lation the higher is its heterogeneity Pormation of associations of

loci seems to be one of the ways to mnintain the intrnpopulntion diver-

sity in self-pnllinntors Tt can l.e nnnumed tl.nt nt high levels of ra-

dioactive contnmination crndnptive gene complexes nre de..troyed, which

n index of intrapopulation diversity (P )was calculated for po-

ohic loci ( P= (PP~ 'P< " JP.J ). The popuiations studied

ierably differed in the value of JP coefficient However, the

lence of intrnpopulation diversity on the dose rate hns not

established The indexp nlso tends to decrease in time like the

on of polymorphic loci <n the whole, the differences between

opulations for all loci are insignificant, but in popu'lntion

he p value significnnt1y decrenned in time (r «-O,r>7, PC 0, 001)

e 3). Thus, n conclusion rnn be rade that. high level., of rndio-

xch level of polymorphism was originally present in the popula- Oorr r 0 95

of polymorphic loci in krabi4opsis populationsfrom the 30 km area of Chernobyl

ns

1986 7101060

240

0.630.290.540.640.570.53

Year Dose rate, mr/h P (958 level)

ARkt

orms)orms)

)987

1988

0.020.10.32.5

7121724

130250

0.011

5050

0.650.570.120.4700.130.470.290.350.230.46

0.690.250.380.440.44

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