The Chernobyl Accident and its Implications for the United Kingdom: Watt Committee: report no 19 (Watt Committee Report No 19)

The Chernobyl Accident and its Implications for the United Kingdom Watt Committee Report Number 19

Members of The Watt Committee on Energy Working Group on the Chernobyl Accident and its Implications for the United Kingdom This report has been compiled by the Working Group on the Chernobyl Accident and its Implications for the United Kingdom. The members of the Working Group were: N.G.Worley (Chairman) Dr F.R.Allen F.J.L.Bindon R.Bulloch Prof. A.Charlesby Dr D.R.Cope Prof. P.M.S.Jones Dr J.D.Lewins

G.Lewis Dr. G.K.C.Pardoe P.D.Potter Dr F.B.Smith Prof. G.N.Walton J.G.Mordue (Secretary) G.F.Oliver (Information Officer)

The Chernobyl Accident and its Implications for the United Kingdom Edited by

NORMAN WORLEY C Eng, BSc (Eng) (Chem Eng) Lond, ACGI, M Inst of Energy Deputy Chairman of The Watt Committee on Energy Chairman of the Working Group on the Chernobyl Accident and its Implications for the United Kingdom appointed by The Watt Committee on Energy and

JEFFERY LEWINS MA, MSc, PhD, DSc (Lond), C Eng, PPINucE, F Am Nuc S Lecturer in Nuclear Engineering at the University of Cambridge

Report Number 19

Published on behalf of THE WATT COMMITTEE ON ENERGY by ELSEVIER APPLIED SCIENCE PUBLISHERS LONDON and NEW YORK

ELSEVIER SCIENCE PUBLISHERS LTD Crown House, Linton Road, Barking, Essex IG11 8JU, England This edition published in the Taylor & Francis e-Library, 2003. Sole Distributor in the USA and Canada ELSEVIER SCIENCE PUBLISHING CO., INC. 52 Vanderbilt Avenue, New York, NY 10017, USA WITH 14 TABLES AND 35 ILLUSTRATIONS © 1988 THE WATT COMMITTEE ON ENERGY Savoy Hill House, Savoy Hill, London WC2R 0BU British Library Cataloguing in Publication Data The Chernobyl accident and its implications for the United Kingdom.—(Watt Committee report; no. 19). 1. Ukraine. Chernobyl. Nuclear power stations. Accidents, 1986 I. Worley, Norman G. II. Lewins, Jeffery D. 1930– III. Watt Committee on Energy IV. Series 363.1’79 ISBN 0-203-21644-X Master e-book ISBN

ISBN 0-203-27264-1 (Adobe eReader Format) ISBN 1-85166-219-7 (Print Edition) Library of Congress Cataloging-in-Publication DatA The Chernobyl accident and its implications for the United Kingdom/ edited by Norman G.Worley and Jeffery D.Lewins. p. cm.—(Watt Committee report; no. 19) “This report has been compiled by the Working Group on the Chernobyl Accident and its Implications for the United Kingdom”— P. “Published on behalf of the Watt Committee on Energy by Elsevier Applied Science Publishers.” ISBN 1-85166-219-7 1. Nuclear power plants—Government policy—Great Britain. 2. Nuclear power plants—Accidents—Environmental aspects. 3. Chernobyl Nuclear Accident. Chernobyl. Ukraine, 1986. I. Worley, Norman G. II. Lewins, Jeffery D. III. Working Group on the Chernobyl Accident and its Implications for the United Kingdom. IV. Series. HD9698.G72C47 1988 363.1’79–dc19 88–7286 CIP The views expressed in this Report are those of the authors of the papers and contributors to the discussion individually and not necessarily those of their institutions or companies or of The Watt Committee on Energy. No responsibility is assumed by the Publisher for any injury and/or damage to persons or property as a matter of products liability, negligence or otherwise, or from any use or operation of any methods, products, instructions or ideas contained in the material herein. Special regulations for readers in the USA This publication has been registered with the Copyright Clearance Center Inc. (CCC), Salem, Massachusetts. Information can be obtained from the CCC about conditions under which photocopies of parts of this publication may be made in the USA. All other copyright questions, including photocopying outside the USA, should be referred to the publisher. All rights reserved. No part of this publication may be reproduced, stored in a retrieval system, or transmitted in any form or by any means, electronic, mechanical, photocopying, recording, or otherwise, without the prior written permission of the publisher.

Foreword There can have been few occasions when an accident to a major work of engineering of any kind was of such immediate concern to governments and peoples throughout the world as was the nuclear power-plant accident at Chernobyl. The collapse of a bridge or a fire in a coal mine, for example, may win headlines in many countries and cause more immediate casualties, but neither civil accidents like these nor natural disasters like earthquakes and famines are likely to be major stories everywhere for so long. For the ordinary man and woman, nuclear power is incomprehensible and horrifying, even if it is possible to distinguish it from the nuclear weapons issue; even for professionally qualified readers, some half million of whom, in the United Kingdom, are represented by the professional institutions constituting The Watt Committee on Energy, understanding is often difficult and opinions range right through the spectrum from enthusiasm for nuclear power to outright opposition. The Watt Committee Executive, when it met soon after the accident occurred in April 1986, saw at once that the implications of what had happened, both immediately and over a period of perhaps years to come, would be far-reaching: perhaps more so because the possible benefits and dangers of nuclear power were already a matter of fierce public controversy, in the United Kingdom as in many other advanced countries, and it was appreciated that important national decisions were being made and could be affected. The importance of these decisions was not limited to the few thousands who would construct a nuclear power station, work in it and live near it; they would have an impact on the technological base, and therefore on the economic prosperity, of the whole country, and on the consumers of energy—specifically of electric power which, in the United Kingdom, means virtually everybody; and these effects will be with us for as long as anyone can foresee. The attitude of The Watt Committee on Energy to the civil nuclear power question was stated in a Report entitled Nuclear Energy: a Professional Assessment, published in March 1984. The then Chairman of the Watt Committee, Dr Jack Chesters, in his foreword to

that Report, explained how I had been made Chairman of the Working Group that produced it. Briefly paraphrased, our hope was that we had succinctly described the technology and stated the main issues in such a way that the reader, whether he supported or opposed nuclear power when he began to read, would have a better understanding when he finished. We were therefore in a good position, two years later, to take part in the reassessment of nuclear power that was bound to result from the Chernobyl accident. This did not mean, of course, that we wished to start from scratch in reconsidering every aspect of civil nuclear power, even if our resources had been sufficient. We appointed a new working group, including some members of the previous group, and our instructions to it, narrow in scope but requiring thorough investigation, were to consider the implications of the Chernobyl accident for the United Kingdom. Two years later, a huge volume of information has been made available (some vital, some marginally relevant) and the arguments have been exhaustively ventilated. Our qualification for adding to the mountains of paper (not to mention other media coverage) is that the members of our group, and of the Watt Committee as a whole, include both those who are expert in the many relevant specialist disciplines and those who, having no previous connections with the nuclear industry, are familiar with the needs of public debate. The principal objective of the Watt Committee is to promote the discussion of questions concerning energy for the benefit of the public at large, bringing together, in an impartial forum, those with professional knowledge from a wide variety of backgrounds; and the Chernobyl accident and its implications, considered as a subject for study by one of our specialist groups, presents the need for this approach particularly acutely. The members of the Working Group are listed elsewhere in this Report. On behalf of The Watt Committee on Energy, I am grateful to them all, and can only mention here the outstanding importance of Norman Worley’s role as its Chairman. The study could not have been undertaken (or not on anything like this v

vi

scale) without the financial support of the Central Electricity Generating Board, the South of Scotland Electricity Board, British Nuclear Fuels plc, the United Kingdom Atomic Energy Authority and the National Nuclear Corporation, which was given on the condition, clearly understood, that the Watt Committee was free to conduct the study as it wished and reach whatever conclusions seemed appropriate. I must also acknowledge the willingness of these and other organisations to provide us with any information that we asked for, including the opportunity of visiting nuclear establishments. Our thanks are due to the many individuals in these organisations who helped us.

Foreword

The complexity of life in a modern trading nation is beyond the appreciation of most of its citizens, and the specific individual effects of even such a major event are small by the time they have ramified into every home and work place in the land. Nevertheless, there are comparatively small numbers of people, mainly in limited areas, for whom the direct effects of Chernobyl are still important, and the long-run indirect effects are important for almost everyone. What these effects are, and what decisions are required, is clarified, I hope, in this Report. G.K.C.PARDOE Chairman, The Watt Committee on Energy

Contents List of the Chernobyl accident Working Group members

ii

Foreword G.K.C.PARDOE

v

Background NORMAN WORLEY

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Section

1 1.1 1.2 1.3

Section

Section

Section

Section

Section

Section

Section

Section

Appendices

2

3

4

5

6

7

8

9

Introduction NORMAN WORLEY General description of a nuclear power system Fuel meltdown incidents Energy in the Soviet Union

1 1 4 6

The design of the Chernobyl Unit 4 reactor PETER POTTER

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Description of the Chernobyl accident FRANK ALLEN

19

The radioactive release from Chernobyl and its effects BARRY SMITH and ARTHUR CHARLESBY

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Accident management in the USSR and the United Kingdom GLYNNE LEWIS

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United Kingdom and USSR reactor types JEFFERY LEWINS

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Reactor operation and operator training in the United Kingdom JOHN BINDON

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International dimensions of the implications of the Chernobyl accident for the United Kingdom DAVID COPE

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Comments, recommendations and conclusions

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1. Glossary of terms 2. Units of measurement 3. Summary of significant dates and timing relevant to the Chernobyl accident vii

101 104 105

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Contents

4. Nuclear safety in the Soviet Union From ‘Nuclear Power in the Soviet Union’ by B.A.Semenov—June 1983 International Atomic Energy Agency Bulletin (Vol. 25, No. 2) 5. Chemical reaction aspects of the Chernobyl accident—Gilbert Walton, Jeffery Lewins and Norman Worley 6. The Chernobyl accident trial An article published in The Independent on 30 July 1987—Anthony Barber of Reuters 7. Continuing radiation leakage from Chernobyl An article released by Reuters, Moscow, 4 December 1987 8. International aspects— Norman Worley including: Learned societies European organisations Organisations for economic co-operation and development The International Atomic Energy Agency Recommendations for the IAEA 9. The Nuclear Installations Inspectorate 10. Energy casualties 11. Presentations to the Working Group on the Chernobyl accident A. Presentation on Inherently Safe Reactors Given on 11 February 1987 by Michael Hayns, Head of Nuclear Safety, Technology Branch, Safety and Reliability Directorate, UK Atomic Energy Authority B. Presentation on The Chernobyl Accident and its Consequences’ Given on 1 April 1987 by John Gittus, Director, Safety and Reliability Directorate, UK Atomic Energy Authority 12. Visits made by the Watt Committee Working Group on the Chernobyl accident A. Visit to Dungeness ‘A’ Magnox station —David Cope B. Visit to the CEGB’s Oldbury-on-Severn Nuclear Power Training Centre—John Bindon C. Visit to Hinkley Point ‘B’ AGR station —Glynne Lewis

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112

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116 117

121 123

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127 130

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D.

E.

Visit to the Atomic Energy Establishment Winfrith SGHWR station— Frank Allen Visit to Hunterston ‘A’ nuclear power station to attend an emergency exercise—Glynne Lewis

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135

The Watt Committee on Energy: Objectives, Historical Background and Current Programme

139

Member Institutions of The Watt Committee on Energy

141

Watt Committee Reports

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Index

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Background and UKAEA, as well as reports from the International Atomic Energy Agency (IAEA) in Vienna and the International Nuclear Agency (INA), were made available to the Group. The Watt Co mmittee would also like to express its appreciation of the efforts of the Working Group members who put a lot of hard work into the study. Members sometimes had to accept controversial decisions on the acceptability of items of material or the way it was presented and this they did with good humour. All the members of the Group have been able to comment on and, in some cases, correct the texts prepared by the others, although as Chairman of the Group I have had the final say. With so many authors, the contributions will inevitably be of varying length. The editor has not attempted to change the style of individuals’ work. Appendices cover special aspects of the accident which could not easily be fitted into the text or else would hinder the reading of the main contributions. A glossary of terms and a list of the units of measurement used have been included because some of these are confusing and difficult to interpret. There are many aspects of a disaster such as Chernobyl, including individual experiences, the way the news was released, the international implications and the impact on energy policies, as well as the technical aspects of reactor design and operation. There are many ways of approaching the subject, but this study attempts to review the facts in the light of the implications of the accident for the United Kingdom and its nuclear power. Because it has been written by independent technologists, it is hoped that this report will provide a significant contribution to the rapidly growing literature on the subject.

The disaster of the Chernobyl nuclear power reactor number 4, involving the meltdown of nuclear fuel and a major fire releasing radioactivity, was clearly an event that was likely to have a major impact on energy policy both inside and outside the Soviet Union and on the acceptability of nuclear power throughout the world. To the opponents of nuclear power, the accident was evidence that engineers and scientists were dealing with a source of energy that was just too dangerous. The supporters of nuclear power found their position more difficult to defend publicly, but as information about the accident was made available, it became clear that there were major differences between the Soviet Union and the United Kingdom in terms of reactor design, safety provision and operational safeguards. The Watt Committee on Energy can call on technically informed people from a wide range of backgrounds both within and outside the power industry. It was therefore particularly well qualified to form a Working Group to report on what occurred and make recommendations from an informed but independent position. The Working Group was assembled shortly after the accident happened and, in the early meetings, the material and format of the report were discussed and agreed. To aid their research, the team members were able to make visits to a number of important sites in Britain, and were able to engage in lively and often controversial discussions with many senior staff in the Central Electricity Generating Board, the South of Scotland Electricity Board, the National Nuclear Corporation and the United Kingdom Atomic Energy Authority. The Watt Committee would like to thank these experts both for their help and for the information, vital to the report, that they were able to provide. Internal reports from the NNC, CEGB, SSEB

NORMAN WORLEY Editor, and Chairman of the Chernobyl Working Group

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Section 1

Introduction Norman Worley Chairman of the Working Group on the Chernobyl Accident and its Implications for the United Kingdom, and Deputy Chairman of the Watt Committee on Energy

INTRODUCTION

element uranium, as it occurs naturally, consists of 99·3% of the isotope U238 which does not readily fission when it reacts with neutrons, and 0·7% of U235, which does. By a process known as enrichment, using a diffusion or centrifuge process, the proportion of U235 can be increased and this is often the case with uranium oxide fuels. During the operation of the reactor, U235 is destroyed, forming highly radioactive fission products. At the same time the U238 reacts with neutrons and, as an eventual result of this reaction, forms a new fissionable material (plutonium), mostly as the isotope Pu239. As this isotope builds up, some of this will fission too. In most reactors the amount of plutonium formed will not fully compensate for the loss of U235, and the ability of the fuel to sustain the nuclear reactions (its ‘reactivity’) falls until it is necessary to replace the fuel to keep the plant producing adequate levels of power. The fuel retains much of the radioactive fission products but is wrapped in a sealed can which, under normal operating conditions, forms a further barrier to their release. The nuclear fission reactions operate most readily with neutrons whose energies have been reduced from the high energies at their formation. The material which accomplishes this is a moderator which can be graphite or ordinary or ‘heavy’ water. The role of coolant and moderator can be combined when water is used. The assembly of fuel elements and moderator is called the reactor ‘core’. The fission reactions release a large quantity of heat, and the fuel is cooled by circulating a fluid past the fuel element, either increasing the temperature of the

The Chernobyl nuclear power station disaster started in the early hours of 26 April 1986. It was claimed by the Soviets that this was the first major accident in the 30 years of operation of this type of reactor, which has only been built in the Soviet Union and which is a major component in the rapid increase in electrification of that country. The papers in this Watt Committee report describe the Chernobyl Number 4 Reactor, detail the main components of the accident and its consequences, and then explore what there is of relevance to the United Kingdom at a time when the future of nuclear power here is a major issue between the political parties. To set the scene, and to provide limited background information for the more specialised papers that follow, this introductory section briefly covers three areas: (a) general description of the principal parts of nuclear power plants; (b) an outline of some major incidents involving nuclear fuel meltdown; (c) the Soviet Union’s energy, electrical power and nuclear power situation. 1.1 GENERAL DESCRIPTION OF A NUCLEAR POWER SYSTEM The principal parts of a nuclear power plant are shown in Fig. 1.1. In operation, heat is generated in the fuel, which is usually an oxide of uranium or metallic uranium. The 1

The Chernobyl accident and its implications

2

Fig. 1.1. The principal parts of a nuclear power plant.

fluid or evaporating some of it. The heat is then used directly (if steam is formed in the reactor), or indirectly by making steam in steam generators, to generate electrical power using turbo-generators. The rate of heat generation in the fuel is high. To provide adequate cooling in most reactor systems and to achieve temperatures that give adequate cycle efficiency, the coolant is maintained under pressure so that the whole reactor coolant system, including the steam generator and the pumps or circulators that drive the coolant past the fuel, are contained in a pressure system. This may be a prestressed concrete vessel, pressure vessel and associated circuits of steel, or a series of pressure tubes which pass through the reactor core, with linked piping. With systems where a rapid increase in heat output can lead to a surge in coolant pressure, the whole of the reactor system can be surrounded by a sealed containment system. There can therefore be several barriers to prevent or restrict the escape of radioactive fission products from the fuel to the surroundings: (a) (b) (c) (d)

retention in the fuel the fuel cladding the coolant pressure circuit the containment system.

The fuel, in addition to containing and retaining radioactive fission products, will, in the core area, particularly while the reactor is operating, emit gamma radiation and elementary particles, e.g. electrons, a-particles and neutrons. Shielding surrounding the reactor ensures that these particles

(called collectively ‘radiation’) do not reach the surroundings in damaging quantities. As the fuel becomes less reactive, it requires replacement with new fuel. The equipment to do this, charge and discharge machinery, may operate while the reactor itself is operating (on-load charge and discharge) or shut down (off-load). As a result of the splitting of the ‘fissionable’ U235 or Pu239 by reactions with a low energy neutron, about two high energy neutrons are released which, unlike the other fission products, pass easily through the can. In steady operation there is a balance between neutrons causing fission and those surviving to continue the reaction. Some neutrons leak from the core, and are absorbed by the shielding. Some will be absorbed by the coolant, particularly if this is water, or the moderator, or the fission products, some of which capture a large proportion of neutrons striking them (they have a ‘high cross-section’), or in the can material, in the fuel itself or in special absorbers introduced in the core. The reactivity of the fuel will depend on how long it has been in the reactor. New fuel is highly reactive but this reactivity is lost with burn-up of fuel and with temperature changes, etc., involved in running the reactor up to power. Thus, excess reactivity in the fuel must be provided and this excess must be kept in check with control rods, capturing neutrons. To keep the heat production as uniform as possible, when a core has both new and old fuel at the same time, and to ease control of the plant, materials that can absorb some of the excess neutrons available are introduced into the fuel itself, into the coolant or as separate

Introduction

absorber rods. To start, shut down and regulate the operation of the reactor, rods that absorb neutrons rapidly, i.e. control and shutdown rods, are moved in or out of the core as required. Of the neutrons formed, most are released virtually immediately, but a small proportion which depend on the fissionable isotopic composition of the fuel (U235 0·7%, Pu239 0·4%), are released at an appreciable time after fission—up to a minute or so. It is this special property of delayed fission neutron release that enables the plant to be controlled by manual or relatively straightforward automatic operating systems with control rods that move relatively slowly. In normal conditions of operation, therefore, a nuclear power reactor system can respond to the requirements of the electrical power grid system, and the rate at which power can be increased will be dictated by the mechanical limits of the components, rather than by the reactor physics of the core. However, all reactors have systems by which shutdown rods can be—and in many cases are— automatically forced into the core to avoid circumstances that could endanger the plant (‘Scram’—said to have been formed from the term Safety Control Rod Axe Man at the first man-made reactor, Stagg Field, Chicago). Nuclear reactor power plants are complex systems with three special features which have to be allowed for in design and plant operation. If operation or design, or both, are faulty or inadequate, serious damage to the fuel, the reactor core or, in extreme cases, the reactor circuit, can result. These special features of nuclear power plant can be summarised as follows: (a) Significant amounts of energy can be released if the system is not properly controlled. (b) The fuel elements contain large quantities of highly radioactive material. To keep this from the environment requires containment. (c) After the reactor has operated, the fuel produces heat—a large quantity in operation, smaller but significant quantities when the plant is shut down and after the fuel is removed from the reactor. Cooling of the fuel has to be adequate at all times to avoid excessive temperatures. Complex plant often has problems, and engineers will modify equipment to avoid problems that have occurred in the past. There is therefore an enormous amount of experience built into new and existing nuclear plant which improves both plant reliability and safety.

3

Most reactor plant accidents do not have serious consequences. However, because much of the plant is not easily accessible and is often inside a containment where access is limited by radiation levels, reactor system repairs and modifications can be slow to accomplish and expensive. A characteristic, therefore, of reactor incidents is a long delay in getting the plant back into operation; in a few cases further plant operation is not possible. All remedial operations inside containment are extremely costly. However, with reactor incidents, direct loss of life or casualties are extremely rare. In this respect the nuclear record compares well with other energy-related industries (see Appendix 10). There have been accidents at nuclear plants, leading to injury and in some cases deaths. In these, nuclear plant is similar to all other process and power plant. The special feature of nuclear plant is the awesome potential power in the plant and its highly radioactive contents. Radiation is regarded by many people with special fear. Of course, everyone encounters radiation from many sources throughout life and there is radiation in our own bodies, from outer space and from our surroundings, with which our bodies appear to cope extremely well. The background levels vary considerably from place to place, but the records do not show any correlation between the incidents of radiation-related diseases (cancer and leukaemia) and the level of background radiation. All nuclear plants involve some release of radioactive material, but the effect on the environment corresponds to a very small fraction of the normal background, even near the plant. The effect of radiation depends on its intensity, duration and type and the age, food and living habits of individuals. With relatively low levels of additional radiation, other than that from background sources, it is generally assumed, although the technical data are inconclusive, that a proportion of people receiving this increased radiation will develop cancer over many years (probably 30–40) as a result. These are the effects described in the literature as long term casualties from radiation. These cancers cannot in fact be identified as being due to radiation. Even statistically, always a minefield of varying interpretation, radiation-linked cancers cannot be detected because of the wide scatter of the basic data. In normal operation at power stations, both the staff and the general public do not receive immediately damaging radiation levels. Even during repair and rectification operations, the radiation levels received by workers will be low. There

The Chernobyl accident and its implications

4

are international and national standards on radiation levels and in general the ambient levels are well below these. A serious radiation hazard can exist in a nuclear power plant only if some of the fuel is exposed, particularly if it melts. Even if this occurs, in most cases little or no extra radiation will be released to the environment. Up to the end of 1986 there have been about 100 incidents, some of them deliberate experiments to provide data, at nuclear plants involving some of the fuel melting. These events have been analysed and the results used to improve design and safety. The total number of immediate deaths attributable directly to these incidents over 35 or more years of nuclear reactor operation is less than 35—three at a military prototype reactor in 1981 in the USA and 31 at Chernobyl. However, three subsequent deaths have been reported at Chernobyl (see Appendix 7). Of the fuel meltdown incidents (excluding Chernobyl-4), eight relatively serious incidents have been selected and subjected to some analysis in the following subsection. It is noted that, of these fuel meltdown incidents, only one (Three Mile Island-2) was at an operating, fully developed power plant. All of the other incidents involved research reactors or developmental or prototype plant. Three relatively minor incidents are also reviewed where single channel fuel overheating occurred in graphite-moderated plant. 1.2 FUEL MELTDOWN INCIDENTS The following descriptions of the incidents are deliberately brief. More detailed descriptions are available, for example in the sources listed on p. 8. 1.2.1 NRX—Canada—December 1952 Reactor. Heavy water moderated research reactor. Cause. Inadvertent withdrawal of control rods, leading to core melting, formation of hydrogen and a chemical explosion. Initiation. Communication misunderstanding; instrumentation indication error; ‘Scram’ did not function. Casualties. None; some low level radiation to staff.

Follow-up. Reactor cleaned up and returned to operation. 1.2.2 Experimental breeder reactor (EBR1)—USA—November 1955 Reactor. First reactor to generate nuclear electricity; experimental breeder. Cause and initiation. Safety experiment requiring ‘Scram’ but slow acting rods inserted instead. Result. 50% of core melted. Casualties. None reported. 1.2.3 Windscale No. 1—UK—October 1957 Reactor. Natural uranium metal, graphite moderated, open cycle air-cooled reactor, one of two used for producing military plutonium. Cause and initiation. With graphite used as a moderator at low temperatures (less than about 300 °C), lattice damage to the graphite occurs which is not automatically repaired in service. To avoid operating problems, the graphite was periodically heated under controlled conditions to enable the carbon atoms in the lattice to return to their more normal and stable locations; this operation releases energy. The operation was being carried out but according to the instrumentation, temperatures were falling before completion. In fact, in parts of the graphite away from the thermocouples, energy release was still going on. However, the operator concluded that an extra reactivity boost was necessary to complete the energy release. When the reactor was subsequently returned to production, in one region of the reactor both uranium fuel and graphite were on fire. In a reactor which is air cooled and where the air is filtered and then released to the atmosphere, this was serious. After carbon dioxide was tried without success, the fire was extinguished by water jets. The immediate concern was the release of relatively short lived radioactive iodine and caesium. It was later learnt that polonium had been released. Effects were detectable as far away as France. One hundred and fifty of the fuel channels were affected, with considerable radioactive release from the stack, although filters retained most of the particulate material. District and

Introduction

national monitoring was not adequate enough to ascertain the extent of release or its distribution. Milk over a wide area had to be thrown away. Reactors 1 and 2 were both shut down and ‘entombed’. This, the Windscale accident, was for many years the most serious reactor accident; it had more radiological consequences than the Three Mile Island accident of 1979. Clearly the accident had the salutary effect of making designers and operators safety conscious in the UK. Casualties. None reported, although official figures suggest that up to 30 deaths from cancer may occur in due course. 1.2.4 NRU—Canada—May 195©8 Reactor. Heavy water moderated experimental reactor. Incident. Fuel fire in charge and discharge machine. Cause. Fuel fracture in machine. Casualties. None reported. 1.2.5 SL-1 (Stationary Low Power Reactor 1)—USA—January 1961 Reactor. Pressurised water reactor prototype 3MW(th). Incident. Control rods are disconnected when the vessel head is removed. On this occasion, in reconnecting one of the rods, it was withdrawn too far, causing an extreme reactivity excursion. Considerable damage to fuel, reactor vessel and pile cap. Casualties. Three killed. 1.2.6 E nrico Fermi No. 1—USA—October 1966 Reactor. Prototype/experimental sodium-cooled ‘breeder’ reactor power plant. Incident. While increasing power, erratic neutron monitoring readings eventually led to reactor ‘Scram’. There was a partial core meltdown due to intermittent blocking of reactor coolant flow by a loose baffle in the base of the vessel. Some radiation released.

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Result. Eventually cause located, plant shut down and repaired but later dismantled. Casualties. None reported. 1.2.7 Lucens—Switzerland—January 1969 Reactor. Experimental heavy water moderated, carbon dioxide cooled pressure tube reactor. Incident. Pressure tube burst causing severe fuel damage. Result. No radiation released to environment; reactor shut down and not reused. Casualties. None. 1.2.8 Three Mile Island-2—USA—March 1979 Reactor. Pressurised Water Reactor. Incident. When the steam generator feed pump failed, the turbo-alternator tripped, leading to an increase in reactor steam pressure which ‘scrammed’ the reactor. To remove the heat still being generated in the reactor fuel, emergency boiler feed pumps were activated but could not supply water because valves were closed. The incident developed to an accident when the planned automatic operation of the emergency system was interrupted by operators. Emergency cooling was automatically turned on but was subsequently shut off by operators because, by their interpretation of the situation, the reactor did not require extra cooling. Reactor pressure rose due to fission product heat and lack of cooling, causing a safety valve to lift. This did not reseat as pressure fell again, so the reactor lost water, uncovering the core, leading to core meltdown and to formation of a ‘bubble’ of hydrogen. Result. Core meltdown and considerable release of radiation to containment. A small amount was released to the environment. Reactor written off—operation of TMI-1 suspended. Clean-up and investigations have cost billions of dollars. Since then, there has been extensive reappraisal of safety requirements and no further ordering of nuclear plant by utilities in the USA. Uncertainty of situation and media reaction led to large public concern. Casualties. None.

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1.2.9 Others In addition to these incidents, three relatively minor events involving local channel overheating with graphitemoderated reactors have occurred, at Marcoule, Saint Laurent-des-Eaux and Chapelcross. In each case there was fuel damage and some oxidation of graphite local to the overheated fuel elements. The fuel elements were removed, the affected channels were sealed and the reactors returned to operation. There were no casualties and no release of radiation to the atmosphere. 1.2.10 Discussion The number of serious core meltdown incidents involving operating power reactors is small. Most of the incidents reported involved experimental, research, development or military plant. The causes were operator error or lack of understanding of the state of the plant in most cases; about half were due to instrumentation errors or faults and about half due to equipment failure, a fuel element in a charging machine, a pressure tube failure (surely significant for the future safety of RBMK reactors) and a loose baffle. It is probable that the safety procedures, quality assurance, automatic systems, instrumentation and data presentation improvements and the higher level of operator training which have become general since the Three Mile Island accident would have avoided virtually all of the accidents. The sample is small, but in reviewing these cases, and other problems that have led to extensive plant damage (but no core meltdown), a number of observations can be made: (a) Nuclear plant, like all other complex installations, will have component and operational problems, the vast majority of which can be dealt with by bringing in standby plant. Routine maintenance of all plant, including standby equipment, is essential, as is operator training to deal with unusual situations. (b) The operator interaction with the instrumentation and control is the area which is most likely to lead to accidents. (c) Operator knowledge of the plant and its characteristics should be in depth, and a wide range of incidents should be rehearsed. Here, building a number of plants of virtually identical design enables experience and training to be most

The Chernobyl accident and its implications

effective and staff transfers to be carried out without problems, (d) Plant design to current standards is extremely robust and able to cope with a wide range of incidents without endangering the plant, the operators or the general public. It is interesting to note that the Chernobyl-4 accident would not have occurred if the operators had complied with the operating instructions. At Three Mile Island the operators did initially do what they were trained to do; but the bases of this training were at fault and they did not have satisfactory information on the state of the plant and its components. 1.3 ENERGY IN THE SOVIET UNION The Soviet Union has vast reserves of all of the main fuels, including uranium. There are no published official Soviet statistics and figures are therefore difficult to check, but it appears that about one-quarter of the world reserves of gas and oil and one-sixth of the coal and brown coal lie in Soviet territory. Although the reserves are vast, most resources are remote from centres of population and in difficult terrain. Figure 1.2 shows the main sources of fossil fuel in the Soviet Union. The coal mines in the Donets region are deep mines. As the seams are exhausted deeper seams have to be used and these are generally too thin for the successful use of automatic machinery. Not only is productivity inevitably falling, but also the coal removed from the mines contains an increasing proportion of rock. The main new sources of coal are east of the Urals and here opencast mining is generally used. In the northern areas the long cold winter hampers production, and in the regions where the coal is lignitic the cost of transport is unacceptably high. Here local power generation with long transmission lines to the West is used, but there have been problems with these, particularly in bad weather. Oil around the Caspian Sea has been traditionally the mainstay of Soviet oil production. However, these fields are becoming exhausted. The major production in the years 1950–80 was in the Urals where the deposits were in large fields and recovery was inexpensive. Currently production from this field too has begun to fall off and over half of Soviet oil now comes from the West Siberian field. Conditions here are not so favourable, as the deposits are in relatively small fields and the capital costs associated with opening new areas are extremely high.

Introduction

7

Fig. 1.2. Sketch-map to show the main sources of fossil fuel in the Soviet Union: C, coal; L, lignite; O, oil; G, gas.

The main gas fields lie well to the north of the oil fields and here again each new field costs more to exploit than did the earlier ones. The gas pipelines have been a major industrial success and there is considerable reserve capacity. On fossil fuels the Soviet Union has, contrary to predictions from the USA, been able to meet the targets set. Exploiting the deposits in these remote regions, however, is costly in equipment and capital. The infrastructure of roads, railways, pipelines and grid lines and running and maintenance costs of equipment under extreme weather conditions all contribute to the high costs of energy from these sources. It is also costly to induce the workforce to come and stay in these regions and often productivity is low. Oil and gas exports form a major source of Soviet hard currency. The need of the USSR for hard currency for import requirements is likely to increase as the drive to improve the standard of living gathers momentum. In 1985 oil exports provided 60% of Soviet hard

currency requirements, and gas a further 20%. Both of these figures have been seriously affected by the subsequent drop in world oil prices. Because exploration and capital costs are so high, it is forecast that Soviet oil exports are likely to fall during the next 15 years. Gas exports are likely to expand and there is spare capacity in the pipelines, but not enough to offset the fall in revenue from oil exports. Expansion of coal-fired power generation is very expensive. There will be expansion, but not enough to meet the anticipated surge in electricity demand. The 1986–87 winter was marked in the Soviet Union by power restrictions and staggered shifts. In addition to only half of the Chernobyl plant running, hydro-power suffered from the effects of a summer drought and several new plants were not ready for the winter load. The Soviet power system has virtually no reserve capacity, hence the return of two of the remaining Chernobyl reactors to full power in time for the winter

The Chernobyl accident and its implications

8

peaks. There is an enormous pressure to expand nuclear installations and they are sited relatively close to centres of electricity demand. The Soviet electrical installed capacity is about 300000 MW, about half that of the USA and five times that of the CEGB. The growth was 3·5% per annum, compared to 3·3% in Japan and less than 3% in the USA and the Federal Republic of Germany and 0·1% in Britain over the period 1975–83 (probably an unfortunate selection of dates for Britain, as over the past few years growth has been 3%). In the USSR, three-quarters of the power requirement is in Europe while about the same proportion of energy production is in Asia. Currently 70% of electricity is from fossil plant, mainly coal fired, the rest being shared between hydro-power and nuclear power. The first RBMK-type reactor started up at Obninsk in 1954—two years before Calder Hall. It is still operating. There are plans for 70000 MW of new nuclear plant to be commissioned between 1984 and 1994 but how that target will be affected by the decision to build no more RBMK reactors is not clear. RBMK reactors of the same physical size as Chernobyl, but with improved heat transfer surface on the fuel cans, are operating at 1500 MW(e) at Ignalis in Lithuania. This was the standard for new RBMK plants. The RBMK type of reactor supplied 70% of the nuclear electricity in 1985, the balance being supplied by the Soviet type of PWR (VVER). There have been problems in meeting the target production of heavy pressure vessels for these PWR reactors. However, serious problems have been reported both in organising supplies of equipment and in maintaining the high standards of construction necessary for the safe operation of nuclear plant. A British assessment of RBMK designs in 1976 was critical about a number of design features.

Fast reactor development, too, is rapid in the Soviet Union. There are three developmental fast reactors: a 12 MW(e) unit at Ulyanovsk, a 1000 MW(th) plant in Kazakhstan, used to produce electricity and steam for desalination, and a 600 MW(e) plant in the Urals. An 800 MW (e) plant is under construction. Two district heating low pressure PWR systems are operating, at Gorky and Voronezh. The intention for the next 20 years or so is to use nuclear heat for about one-third of the total heating load in the USSR. In addition, there are three Soviet nuclear-powered ice breakers—this is an important duty, as many ports are ice-bound in winter and oil and gas supplies are close to the Arctic Ocean. There are plans for nuclear oil tankers too. The Soviet nuclear programme is necessary because of the country’s economic and fossil fuel situation. Over half of the new installed capacity is likely to be nuclear, and the aim must be to release as much oil and gas for hard currency exports as can be achieved. Probably, bearing in mind the location of fossil fuel supplies, the total capital cost of nuclear plant, particularly the relatively low technology RBMK, compares favourably with coal-fired installations. Sources The Worst Accident in the World. The Observer, London, 1986. Fremlin, J.H., Power Production, What are the Risks? Adam Hilger, Bristol, 1985. Wyatt, A., Electric Power. The Book Press (Canada), 1986. Stern, J., Soviet Oil & Gas Exports to the West. Gower Press, Swansea, 1987. IAEA Bulletin. International Atomic Energy Agency, Vienna, Autumn 1986. Patterson, W., Nuclear Power. Penguin, Harmondsworth, Middx, UK, 1980.

Section 2

The Design of the Chernobyl Unit 4 Reactor Peter Potter Independent nuclear consultant Northwich, Cheshire

2.1 INTRODUCTION AND GENERAL DESCRIPTION

opportunities to monitor closely individual channels and, if necessary, to repair faulty ones. Until the accident at Chernobyl, another advantage claimed for the design was that the pressure tubes, with their individual connections, eliminated the possibility of a complete loss of coolant from the core.

The Soviet thermal nuclear reactor power programme involves two reactor types which, until now, have been planned in roughly equal numbers. The first is the PWR for which a dedicated manufacturing facility has been built in the south of the Soviet Union at Volgadonsk. Problems in bringing this plant into production have reduced the rate of construction of the planned PWRs and resulted in a greater proportion of the other reactor type, the RBMK (Reactor Bolshoi Moschnosti Kipyashiy—Large Power Boiling Reactor). This is currently being built in two sizes, 1000 and 1500 MW respectively. The RBMK-1000 is a direct cycle, forced circulation, boiling water cooled, graphite moderated reactor fuelled with low-enriched uranium dioxide clad in a zirconium alloy. It consists of a large block of graphite moderator pierced vertically by 1872 ducts fitted with pressure tubes of which 1661 contain fuel elements and 211 are for control rods. The coolant circuit is divided into two loops, each being fed from half the reactor. The steam generated is separated from water in steam drums and is fed directly to a pair of turbinegenerators which generate electricity. Components for this design are relatively small. They thus require no large unique facilities, can be made in a number of small factories, and pose no special problems for transport and installation. As in the case of the British gas-cooled reactors, on-load refuelling enables a high burnup to be obtained from low-enriched fuel. The channel tube design provides

2.2 STATION LAYOUT At the time of the accident, the nuclear power station at Chernobyl comprised four operating units and two under construction. Each unit is made up of one reactor of the RBMK-1000 type and two turbine-generators. The two units 3 and 4 are accommodated in one block as shown in Fig. 2.1. The two reactors are separated by a compartment housing common services. Alongside is the turbine hall with the four turbines in line. The blocks accommodating units 1, 2, 3 and 4 are adjacent (so that all eight turbines are in line). The block for units 5 and 6, not now to be completed, is sited 1.5 km to the south-east. The RBMK-1000 reactors in units 3 and 4, the latter being involved in the accident, are the second generation of this type with design improvements over the earlier units. The information in this section refers specifically to this second generation. An elevation of reactor 4 is shown in Fig. 2.2. 2.3 REACTOR CORE The arrangement of the reactor core toget.her with its biological shield is shown in Fig. 2.3. The active core 9

10

The Chernobyl accident and its implications

Fig. 2.1. Plan of the block housing Chernobyl Units 3 and 4 (from Ref. 2, by permission of the IAEA). Dimensions in metres.

Fig. 2.2. Sectional elevation of Chernobyl Unit 4 (from Ref. 2, by permission of the IAEA). Dimensions in metres.

Design of Chernobyl unit 4 reactor

Fig. 2.3. Sectional elevation through core and biological shield (from Ref. 2, by permission of the IAEA).

of the reactor is roughly octagonal and is composed of stacked, square graphite blocks (250 mm×250 mm×600 mm high) containing pressure tubes in central, axial ducts (114 mm in diameter) on a 250 mm square lattice pitch. It is approximately 12 m in diameter and 7 m high. Reflector blocks are arranged to produce a cylindrical configuration 14m in diameter and 8·0 m high. The core is surrounded by a cylindrical shroud; a bottom support structure consisting of a welded metal base resting on a 2 m thick concrete neutron shield; and an upper metal structure resting on the annular tank of the biological shield which is filled with water. These components together form a light but leaktight cavity. A further annular tank between the inner neutron shield and the outer one of concrete, contains sand. The upper neutron shield consists of 3 m of concrete. Upper and lower metal structures contain ducts for the fuel and control rod channels. The complete reactor structure is situated in a concrete vault with dimensions 21·6 m×21·6 m×25·5 m. Some 5% of the heat from the fission process is released in the moderator. To improve heat transfer

11

from the graphite to the channel tubes and in order to prevent oxidation of the graphite, the reactor space is filled with a mixture of helium and nitrogen. Slow circulation of this permits monitoring of its temperature and moisture content, giving an indication of the integrity of the pressure tubes. The space outside is filled with nitrogen at a pressure 0·5–1·0 MPa greater than that of the helium/nitrogen mixture. If a channel tube should rupture, a pressure suppression system is provided to condense the steam released. Channel tubes (88 mm o.d. and 4 mm thick) are of welded design and contain fuel assemblies which are cooled by boiling light water. The upper and lower parts of the channel are made of stainless steel and the central part, located in the active zone, is made from a zirconium/2½ niobium alloy. The central part is joined to the upper and lower parts by vacuum diffusionwelded stainless steel/zirconium transition joints. The channel tube is attached to the upper duct by a welded joint, and to the lower one by a compensator unit, which is necessary to compensate for the difference in thermal expansion of the channels and ducts without destroying the leak-tightness of the reactor cavity. This type of joint makes it possible to replace a channel during reactor shutdown. Water coolant is fed to each fuel channel from below and the steam-water mixture is removed from the top. In order to compensate for variations in power distribution, the coolant supply to individual channels is regulated by isolating and regulating valves which are installed in the channel feed pipes. Removal of moderator heat from the graphite blocks to the fuel channels is through specially designed sleeves. These sleeves are resilient slotted rings of graphite, 20 mm high, which are disposed along the height of the channel (in the zone of maximum thermal loading) and packed against one another in such a way that alternate ones lie against the channel wall and the others against the surface of the graphite block. This ensures the necessary conditions of heat transfer and compensates for shrinkage of the gap between the block and the channel during the life of the plant. The design maximum temperature in the graphite is 750°C.

2.4 PRIMARY COOLANT CIRCUIT The coolant circuit is shown schematically in Fig. 2.4. It comprises two parallel loops each with two stainless steel-clad carbon steel steam drums. Each steam drum

The Chernobyl accident and its implications

12

Fig. 2.4. Schematic arrangement of main coolant and steam circuits (from Ref. 2, by permission of the IAEA).

has 482 risers (76 mm diameter× 4 mm thick) and 12 downcomers (325 mm× 16 mm). The downcomers connect to a common pump inlet header (752 mm i.d.). Each loop has four electrically driven pumps, with three operating and one on reserve. There is an intermediate system of 22 distribution headers (325 mm×15 mm) per loop between the common pump outlet header (900 mm i.d.) and the channel feed pipes. There is only a rudimentary system of pipe restraints. At full load, water at 270 °C enters the bottom of each fuel channel through individual 57 mm i.d. channel feed pipes. A two-phase mixture of water and 14.5% steam (by weight), generated as the mixture flows over the fuel, passes to the horizontal cylindrical steam separator drums via 76 mm i.d. steam-water lines. Saturated steam at 284°C and 6.9 MPa, with a moisture content of less than 0·1%, passes from the drum through 14 steam discharge pipes (325 mm×19 mm) to two steam headers (426 mm i.d.×24 mm) which connect to a common steam collector (630 mm×25 mm) supplied by all eight steam mains from the four steam drums. This collector feeds the two turbines. Condensate from the turbines is deaerated and returned to the drum separators by electric pumps. A feedwater collector is located at the bottom of each drum separator. From these, the water which has been separated from the steam-water mixture is mixed with feedwater and is taken via the twelve 312 mm downcomers to a 1020 mm pump inlet header. This creates a cooling effect below saturation temperature to give the required cavitation margin at the inlet of

the main circulating pumps. Each pair of steam drums is interconnected on both water and steam sides so that the drum levels remain identical. A common ion exchange unit is provided for the two coolant loops to purify the cooling water, the flow through it amounting to 4% of the boiler system capacity. 2.5 FUEL One fuel element (Fig. 2.5) consisting of two fuel assemblies is installed in a fuel channel. The element is attached to a special suspension, provided with a locking plug, which is set in the neck of the upper duct. This locking plug hermetically seals the duct cavity by means of a ball-valve with a packing gasket. Each fuel assembly consists of 18 fuel pins. A fuel pin is a tube of zirconium/2½ niobium alloy with an external diameter of 13·6 mm and a minimum thickness of 0·825 mm, filled with pellets of uranium dioxide. The fuel pellets are 11·52 mm in diameter and 15 mm high with cavities in the end faces. The inner space of the fuel pin is filled with an argon/ helium gas mixture. Top and bottom terminal grids hold the fuel pins and are positioned above and below the level of the core. Two fuel assemblies are combined in the element giving an active length of 7 m. The uranium feed enrichment is 2·0%. The reactor can be refuelled on load at full power. A refuelling machine is supported on a gantry running along the length of the refuelling hall. As well as

Design of Chernobyl unit 4 reactor

13

Fig. 2.5. Fuel element design (from Ref. 1). Dimensions in mm.

removing spent fuel and replacing it with fresh, the machine can also verify free passage through the fuel channel using a gauge simulating a standard assembly. The semi-automatic machine accurately locates the coordinates of the channel to be refuelled, locks on to it, adjusts its own pressure to slightly above that of the channel, removes the channel plug unit from the top of the channel, withdraws the spent fuel element, checks the channel with a gauge, loads fresh fuel, seals the channel and takes the spent fuel to the storage pond where it is unloaded. While the machine is connected, there is a slight flow of clean water from it to the channel to prevent it becoming contaminated with water from the primary circuit. The refuelling machine can carry out five charge/discharge cycles at full power in 24 h, but in normal operation only one to two channels per reactor are reloaded each day.

2.6 CONTROL AND SAFETY RODS Reactivity is controlled by ‘rods’ consisting of articulated absorber elements formed from hollow cylindrical sections of boron carbide (65 mm diameter ×7·5 mm thick) sheathed in the annulus between two aluminium alloy tubes of 70 mm×2 mm and 50 mm×2 mm respectively. They are inserted or removed from the core at a rate of 0·4 m/s (the 12 local automatic control rods are withdrawn at 0·2 m/s) by individual servomotors installed at the top of the control rod channels. With the exception of the automatic rods, all the rods are fitted with graphite followers so that, as they are withdrawn, they are not replaced by water. The square lattice of 211 control rods and 12 vertical power profile sensors has a pitch of 700 mm and is angled at 45° to the fuel lattice. The channels are made

14

of the same zirconium alloy as the fuel channels and are 88 mm in diameter and 3 mm thick. They have their own cooling water supplied from a special circuit. The control rods are divided into the following groups according to their purpose: 115 manual control (MC), 24 safety (S), 12 local autocontrol (LAC), 12 average power control (APC), 24 local safety (LS) and 24 shortened absorber (SAR) for levelling off the axial power distribution. The SAR rods are inserted from the bottom of the reactor. In order to compensate for the initial reactivity margin, auxiliary absorbers (AAR) are installed in some of the fuel channels instead of fuel assemblies. The absorbing elements in these are inserts of boron and stainless steel. By varying the ratio of the numbers of inserts, the absorption of the auxiliary absorbers can be changed. The entire active zone is divided arbitrarily into groups of cells containing 12 fuel channels with fuel elements, two channels with auxiliary absorber rods and two channels with control and safety rods. This grouping is varied at the reactor periphery by reducing the number of control and safety rods to flatten the radial power distribution. The control rod system provides for automatic control of the required reactor power level and its period; reactor startup; manual regulation of the power level and distribution to compensate for changes in reactivity due to burn-up and refuelling; automatic regulation of the radial-azimuthal power distribution; automatic rapid power reduction to predetermined levels when certain plant parameters exceed preset limits; automatic and manual emergency shutdown under accident conditions. A special unit selects 24 uniformly distributed rods from the total available in the core as safety rods. These are the first rods to be withdrawn to their upper cut-off limit when the reactor is started up. In the event of a loss of power, the control rods are disconnected from their drives and fall into the core under gravity at a speed of about 0·4 m/s, regulated by water flow resistance. The power density distribution is controlled by the 12 LAC and 24 LS rods. The average power control system is used as standby in the 20–100% power range and is switched on automatically when the LAC system malfunctions. The automatic control system holds reactor power to within ±1 % of the required output in the range 20–100% full power and to within ±3% in the range 3·5–20% full power. Removal of the LAC rods is blocked automatically when a power overshoot signal is registered in one of the channels of the local safety zone. If such a signal

The Chernobyl accident and its implications

appears in both channels of an LSZ, two LS rods are motored into this zone until at least one of the signals is corrected. When there is any malfunction of the control system, the withdrawal of more than 8–10 of the MC and S rods is automatically prevented. There is also a built-in limitation on the continuous withdrawal of the LAC rods for more than 8 s. A number of different categories of emergency situation are defined depending on their nature. The highest of these, Level 5, results in the reactor being tripped automatically and shut down completely. The reactor is also tripped at Level 5* but can be started up as soon as the emergency has passed. Lower levels result in reduction of reactor power at different speeds to safe values corresponding to the capacity of plant still in operation. The reactor is tripped automatically when any of the following occur: permissible limits exceeded in power output, water pressure in the steam drum or channel feed pipes, water level in the steam drums, pressure in the leaktight compartments or reactor cavity; permissible limits undershot for reactor period, feedwater flow, water level in the steam drums, flow in the control rod coolant circuit or level in its coolant tank; voltage loss in auxiliary electrical supply; simultaneous tripping of both turbo-generators or tripping of the only operating turbo-generator; simultaneous trip of three out of four main circulating pumps in any pump room. 2.7 CORE MONITORING Neutron flux detectors positioned by hangers are placed in fuel channels, the reflector and the annular water-filled biological shield tank. They are located as follows: (a) Water shield tank: 8 startup range ionisation chambers 16 power range ionisation chambers (b) Reflector; 4 fission chambers (during startup) (c) Core: 24 fission chambers Twenty-four ionisation chambers situated in the water shield tank are divided into groups and used to drive three banks of automatic control rods. The automatic control signal is generated by summing the relative deviations from the set power from three out of four ionisation chamber measurement channels. Twenty-four in-core fission chambers located in the fuel channels monitor neutron flux and are used to

Design of Chernobyl unit 4 reactor

15

drive the local automatic control rods. Their output is displayed in the control room and they are also used in the safety system. In addition to the control and safety systems described above, the reactor is equipped with a number of monitoring systems. The principal systems are designed to provide data as follows: radial (across 130 channels) and vertical (12 channels) power distribution; temperatures of the graphite stack and metal core structures; individual channel water flow control using ball flow meters; the state of the principal components of the main coolant circuit such as steam drums, main circulating pumps and their inlet and outlet headers; burst fuel element can detection using a short-life volatile fission product detecting photomultiplier moving sequentially across the steam—water lines at the outlet to each channel; channel tube integrity using moisture content and temperature of the gas in the reactor space.

All the data are fed to a central computer where they can be displayed and recorded. 2.8 EMERGENCY CORE COOLING SYSTEM (ECCS) The ECCS is shown schematically in Fig. 2.6. It is designed to provide a sufficient supply of water to ensure removal of stored energy in the short term and decay heat for a longer period, following interruption of the normal coolant supply. It is assumed that the emergency system has to supply only one half of the segregated cooling system, and provision is made to identify which half this is. In the event of a loss of coolant, water is first injected from two sub-systems into the distribution headers between the common pump outlet header and the individual channel feeder pipes of the failed circuit from tanks pressurised to 10 MPa which are connected by fast acting valves. At the same time, a third sub-system

Fig. 2.6. Schematic arrangement of ECCS (from Ref. 2, by permission of the IAEA).

16

The Chernobyl accident and its implications

Fig. 2.7. Schematic arrangement of part of the containment system (from Ref. 2, by permission of the IAEA): 1, reactor space; 2, compartments (downcomer pipes and main circulation pump outlet headers); 3, compartments (distribution group header and lower communication lines).

supplies water from the electric feed pump. Each subsystem is capable of supplying at least 50% of the required output. Before these supplies are exhausted, a pumped system takes over, drawing water from the suppression pool beneath the reactor. This system consists of three separate identical loops, each with electrically driven highand low-pressure pumps, and each again capable on its own of providing half the required coolant capacity. At the same time, three parallel loops with electrically driven pumps supply water to the undamaged circuit from a tank containing clean condensate. 2.9 CONTAINMENT The reactor containment system, shown in Fig. 2.7, is designed to cope with the release of steam resulting from a failure of the primary circuit. The circuit passes through a number of leaktight compartments which are designed to withstand different overpressures: 0·08 MPa for the reactor space (1 in Fig. 2.7) and the compartments housing the distribution group header (2 in Fig. 2.7), and 0·45 MPa for the strong compartments (3 in Fig. 2.7) housing the downcomer pipes and pump outlet headers (average diameter 900 mm). These compartments are separated by a system of non-return and release valves and connected by a steam distribution corridor to a two-storey suppression system. In normal operation, the leaktight compartments are all held at a slight underpressure. Should a failure of the primary circuit occur, steam and any other gases released raise the pressure in the

relevant compartment. Non-return valves in the strong compartments open to the steam distribution corridor when the pressure differential rises above 0·002 MPa. For the other leak-tight compartments, this differential has to be over 0·02 MPa to operate the valves. The steam passes into the suppression pool beneath the reactor where it is condensed. Isolation valves are provided on all primary circuit pipes penetrating the leaktight compartment walls with the exception of the risers from the channel tubes to the steam drums. Steam emitted from these can be at least partially condensed by the ventilation system. The suppression pool is supplied with heat exchangers to provide cooling in the event of prolonged operation. 2.10 EMERGENCY POWER SUPPLY The emergency power supply for circuits which are absolutely essential in the event of a reactor trip, consists of a storage battery with static inverter transformers for circuits that cannot tolerate interruptions in power supply or can tolerate interruptions up to several seconds in any regime, and automatic diesel generators for those which can tolerate interruptions of between tenths of a second and tenths of a minute in the same regimes. 2.11 WEAKNESSES IN DESIGN A team from the British nuclear industry led by the British Nuclear Forum visited the first RBMK-1000

Design of Chernobyl unit 4 reactor

reactor power station at Leningrad in 1975. As a result of analyses of the information supplied to them by Soviet engineers, members of the team reached the following conclusions concerning the RBMK reactor: — By UK standards, control rod investment was inadequate to meet possible shutdown/holddown requirements — There was no secondary shutdown system: this is mandatory on British reactors and some even have tertiary systems — The void coefficient of reactivity could be large enough to result in a positive power coefficient under some conditions — There did not appear to be an emergency core cooling system — Venting between channel tube and graphite moderator bricks was probably inadequate to allow the escape of steam from a single channel tube failure, thus facilitating fault propagation to other channels — The core restraint structure was not strong enough to withstand a channel tube failure — The operating temperature of the moderator was considerably higher than would have been accepted in the UK, allowing the possibility of fire when exposed to the atmosphere — The buildings housing the primary circuit were not capable of resisting the pressures likely to result from a LOCA (Loss of Coolant Accident) Possibly as a result of these and the criticisms of similar teams of foreign visitors, the Soviets undertook a number of measures to improve the reactor: — The number of control rod channels was increased at the expense of fewer fuel channels — An emergency core cooling system together with a suppression system was provided — The Accident Localisation System was designed around most of the primary circuit In spite of these changes, the reactor was left with the possibility of a fast positive power coefficient, with a graphite temperature that was too high, and without a secondary shutdown system. In addition, it is now clear that the speed of insertion of the control rods under emergency conditions was totally inadequate. REFERENCES 1. The accident at the Chernobyl nuclear power plant and its consequences, State Committee for the Use of Atomic Energy of the USSR, 1986.

17 2. Summary Report on the Post-accident Review Meeting on the Chernobyl Accident, International Nuclear Safety Advisory Group of the IAEA, Vienna, 1986.

APPENDIX: CHERNOBYL UNIT 4 DESIGN DATA Reactor system Gross fission heat, MW Active core diameter, m Active core height, m Diameter over reflector, m Height over reflector, m No. of graphite columns Duct diameter, mm Mass of graphite stack, t Number of fuel channels Fuel channel pitch (square lattice), mm Number of control and safety channels Control channel pitch (square lattice), mm Mass of uranium in core, t Channel tube material Channel tube outer diameter, mm Channel tube wall thickness, mm Channel tube length, m

3200 11·8 7 14 8·0 2488 114 1700 1661 250 211 700 190 Zr/2½ Nb alloy 88 4 8

Fuel Feed enrichment, % Number of sub-assemblies per fuel element Number of fuel pins per sub-assembly Pin outer diameter, mm Clad thickness (min.), mm Pin overall length, mm Fuel pellet diameter, mm Fuel pellet length, mm Fuel pellet density (min.), g/cm3 Mean mass of uranium dioxide per pin, kg Mean mass of uranium per assembly, kg Spacer/support grids per sub-assembly Active fuel length, mm Plenum length, mm Burnup, MWd/kg Refuelling Isotopic concentration of fuel at discharge, kg/t: uranium-235 uranium-236 plutonium-239 plutonium-240 plutonium-241

2·0 2 18 13·6 0·825 3644 11·5 15 10·4 3·6 114·7 11 2×3428 175 20 on load 4·5 2·4 2·6 1·8 0·5

Power Channel average power, MW

1·92

The Chernobyl accident and its implications

18 Channel maximum power, MW Channel maximum power (design), MW Average fuel rating, MW/t U Radial power factor Vertical power factor Critical power margin (min.)

3·00 3·25 16·8 1·48 1·4 1·25

Number of steam drums Drum diameter (internal), m Drum length, m Drum wall thickness (min.), mm Drum pressure (working), MPa Drum pressure (rated), MPa Weight of steam drum (dry), t

4 2·6 31 110 7·0 7·5 280

Reactivity coefficients Void coefficient at operating point, 10-6% steam Fast power coefficient, a°w, 10-6 Fuel temperature coefficient aT, 10-5/°C Graphite temperature coefficient aG, 10-5/°C Minimum weight of control and safety rods, DK, % Averaged reactivity effect of replacing spent fuel by a fresh fuel assembly, %

Main circulating pumps 2·0 –0·5 –1·2 –6 10·5 0·02

Steam Coolant flow, t/h Coolant flow in max. power channel, t/h Channel inlet temperature, °C Channel inlet pressure, MPa Channel outlet temperature, °C Channel outlet pressure, MPa Mean steam quality at channel outlet, wt % Maximum steam quality at channel outlet, wt % Moisture content of steam at separator outlet, %

37·6×103 28 270 7·8 284·5 7·0 14·5 20·1