Phebus Fission Product Project: Presentation of the experimental programme and test facility (Publication No. Eur 13520 En of the Commission of the Europe)

THE PHEBUS FISSION PRODUCT PROJECT

Presentation of the Experimental Programme and Test Facility

Proceedings of the Seminar of the Phebus-FP (Fission Product) Project, presentation of the experimental programme and test facility, organised by the “Commissariat à l’Energie Atomique (CEA), Institut de Protection et de Sûreté Nucléaire (IPSN), Centre d’Etudes Nucléaires de Cadarache (CEN) and the Commission of the European Communities (CEC), Directorate General Science, Research and Development, Joint Research Centre (JRC) Ispra, Safety Technology Institute (STI) and held at the Château Cadarache, St. Paul-Lez-Durance, France, 5–7 June 1991. Programme Committee Members:

Co-ordinators:

C.LECOMTE—CEA/CEN Fontenay-aux-Roses M.GOMOLINSKI—CEA/CEN Fontenay-aux-Roses A.MEYER-HEINE—CEA/CEN Cadarache P.VON DER HARDT—CEC/JRC Ispra A.G.MARKOVINA—CEC/JRC Ispra M.C.RUBINSTEIN—CEA/CEN Cadarache W.KRISCHER—CEA/JRC Ispra

Local organisation: M.C.RUBINSTEIN Publication arrangements: D.NICOLAY—CEC Luxembourg Scientific secretary: W.KRISCHER—CEC/JRC Ispra

THE PHEBUS FISSION PRODUCT PROJECT Presentation of the Experimental Programme and Test Facility

Edited by W.KRISCHER CEC Joint Research Centre, Ispra Site, Ispra (VA), Italy and M.C.RUBINSTEIN CEA, IPSN, Centre d’Etudes Nucléaires de Cadarache, St Paul-Lez-Durance, France

ELSEVIER APPLIED SCIENCE LONDON and NEW YORK

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ISBN 0-203-27038-X (Adobe eReader Format) ISBN 1-85166-765-2 (Print Edition) Library of Congress CIP data applied for Publication No. EUR 13520 EN of the Commission of the European Communities, Dissemination of Scientific and Technical Knowledge Unit, Directorate-General Telecommunications, Information Industries and Innovation, Luxembourg. LEGAL NOTICE Neither the Commission of the European Communities nor any person acting on behalf of the Commission is responsible for the use which might be made of the following information. 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 of 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

Severe, hypothetical accidents in light water reactors may lead to the release of radioactive material, mainly fission products, from the containment barriers. The knowledge of that release and of its evolution with time and in function of certain depletion phenomena is a major point for an efficient accident management and emergency planning, as it allows to estimate the Source Term, i.e. quantity and physical-chemical quality of radioactive releases to the environment. Computer codes have been developed and several experimental programmes performed in an attempt to understand and quantify the complex sequences of events involved in core degradation and fission product transport. Based upon the scientific and technical information made available internationally by experiments and computations, the French Commissariat a l’Energie Atomique proposed in 1985 a series of in-pile integral experiments to study the release and transport of fission products under the most representative conditions: the principle was to study phenomena taking place in core, primary circuit, containment building, and the release to the environment in an experimental facility where the coupling between phenomena could be also investigated. Consequently, the Phebus Fission Product (FP) programme has been settled with the main objective to improve the understanding of FP physical and chemical behaviour, i.e. their emission, transport and evolution in the primary circuit and containment. The experiments are designed to study also the degradation of high burn-up fuel, particularly during the later phases of the transient. The Commissariat à l’Energie Atomique and the Commission of the European Communities agreed in 1988 to carry out the Phebus-FP project as a joint effort, and to open it to international collaboration. A new facility has been designed over the past two years using the existing Phebus test reactor at Cadarache, France. Fission products from overheated high burn-up fuel will be swept, by hot steam and hydrogen, through simulated reactor cooling system components into a tank simulating the reactor containment. Kinetics and results of FP transport, chemical reactions, and depletion will be measured by appropriate instrumentation and by post-test analysis. The first in-pile test is planned for autumn 1992, and five subsequent tests will follow in yearly intervals. The seminar on the Phebus-FP project was organized by the French Commissariat à l’Energie Atomique and the Commission of the European Communities. The objective of the seminar was to present the state of the art in LWR (Light Water Reactor) Source Term evaluation, the needs for safety analysis and the lessons learned from relevant projects. The presentations outlined the contributions and limitations of the Phebus-FP in solving identified problems and in providing a qualified database for the validation of code systems for the Source Term estimation during a postulated severe accident. Essential project features such as the instrumentation of tests and analytical support were illustrated. The seminar was closed by a panel discussion where a summary of the main points under discussion during the different sessions was presented and the speakers gave their recommendations on the Phebus-FP programme and, in general, severe accident research. A visit of the Phebus plant was included. P.Fasoli-Stella

CONTENTS

FOREWORD

v

ABBREVIATIONS

ix

WELCOME ADDRESS M.LIVOLANT , Director of Research of the Protection Safety and Nuclear Institute (IPSN)

1

J.P.CONTZEN , Director General of the Joint Research Centre, Commission of the European Communities

3

SESSION I SURVEY OF LWR SEVERE ACCIDENT SOURCE TERM RESEARCH LWR SEVERE ACCIDENT SOURCE TERM RESEARCH IN EUROPE M.LIVOLANT , CEA/IPSN Fontenay-aux-Roses, France

6

LWR SEVERE ACCIDENT SOURCE TERM RESEARCH IN THE USA T.P.SPEIS , R.Y.LEE , L.SOFFER and R.O.MEYER , US Nuclear Regulatory Commission, Washington

10

SURVEY OF SEVERE ACCIDENT EXPERIMENTS AND ANALYSES IN JAPAN M.AKIYAMA , University of Tokyo, K.TAKUMI , Nuclear Power Engineering Center, Tokyo and K.SODA, JAERI , Tokai-mura, Japan

23

SUMMARY OF DISCUSSION S.FINZI , CEC Brussels

35

SESSION II STATE OF THE ART DEDUCED FROM PREVIOUS LARGE EXPERIMENTS CORE DEGRADATION AND FISSION PRODUCT RELEASE R.W.WRIGHT , US Nuclear Regulatory Commission, Washington and S.J.L.HAGEN , KfK, Karlsruhe, Germany

37

FISSION PRODUCT TRANSPORT J.O.LILJENZIN , Chalmers Tekniska Högskola, Göteborg, Sweden and W.SCHÖCK , KfK, Karlsruhe, Germany

46

FISSION PRODUCT CHEMISTRY IN SEVERE REACTOR ACCIDENTS: REVIEW OF RELEVANT INTEGRAL EXPERIMENTS A.L.NICHOLS , AEA Technology Winfrith, UK and C.HUEBER , CEA/IPSN Cadarache, France

63

vii

PHEBUS-CSD PHEBUS SEVERE FUEL DAMAGE PROGRAMME: MAIN EXPERIMENTAL RESULTS AND INSTRUMENTATION BEHAVIOUR C.GONNIER , C.REPETTO , CEA/IPSN Cadarache and G.GEOFFROY , CEA Saclay, France

79

REVIEW OF B9+ BENCHMARK RESULTS B.ADROGUER , CEA/IPSN Cadarache, France and P.VlLLALIBRE , CSN Madrid

91

SUMMARY OF DISCUSSION A.MEYER-HEINE , CEA/IPSN Cadarache, France

104

SESSION III CORE AND FP-BEHAVIOUR SAFETY ANALYSIS NEEDS AND MAIN PHENOMENA TO BE STUDIED J.GAUVAIN , CEA/IPSN Fontenay-aux-Roses, France and H.M.VAN RIJ , CEC/JRC Ispra, Italy

106

PHEBUS-FP OBJECTIVES, TEST MATRIX AND REPRESENTATIVITY OF THE PHEBUS-FP EXPERIMENTAL PROGRAMME A.ARNAUD , CEA/IPSN Cadarache, France and A.MARKOVINA , CEC/JRC Ispra, Italy

110

PHEBUS-FP TEST FACILITY PH. DELCHAMBRE , CEA/IPSN Cadarache, France and P.VON DER HARDT , CEC/JRC Ispra, Italy

121

PHEBUS-FP INSTRUMENTATION P.VON DER HARDT , CEC/JRC Ispra, Italy and G.LHIAUBET , CEA/IPSN Fontenay-aux-Roses, France

132

CEA ANALYTICAL ACTIVITIES: HEVA, PITEAS, MINI-CONTAINMENTS C.LECOMTE and G.LHIAUBET , CEA/IPSN Fontenay-aux-Roses, France

148

CEC SUPPORT ACTIVITIES: EC SHARED COST ACTIONS AND OTHERS P.FASOLI-STELLA and A.MARKOVINA , CEC/JRC, Ispra, Italy

159

PHEBUS-FP: ORGANISATION OF THE PROJECT AND INTERNATIONAL COLLABORATION A.TATTEGRAIN , CEA/IPSN Cadarache, France and P.VON DER HARDT , CEC/JRC Ispra, Italy

168

SUMMARY OF DISCUSSION P.FASOLI-STELLA , CEC/JRC Ispra, Italy

172

viii

SESSION IV ANALYTICAL ACTIVITIES SURVEY OF SOURCE TERM CODES M.R.HAYNS , AEA Technology Harwell and S.R.KINNERSLY , AEA Technology, Winfrith, UK

174

ESCADRE AND ICARE CODE SYSTEMS M.REOCREUX and J.GAUVAIN , CEA/IPSN Fontenay-aux-Roses, France

183

ESTER—A EUROPEAN SOURCE TERM EVALUATION SYSTEM A.V.JONES and I.M.SHEPHERD , CEC/JRC, Ispra, Italy

198

FPT0 TEST PRECALCULATIONS A.MAILLIAT , CEA/IPSN Cadarache, France, A.V.JONES and I.M.SHEPHERD , CEC/JRC Ispra, Italy

208

SUMMARY OF DISCUSSION A.TATTEGRAIN , CEA/IPSN Cadarache, France

238

PANEL DISCUSSION VALUE OF THE PHEBUS-FP AND RELATED SOURCE TERM STUDIES FOR THE SAFETY OF LWRs Chairman: H.F.HOLTBECKER (CEC) Members: M.AKIYAMA (JAP), R.E.VAN GEEN (B), M.PEZZILLI (I), C.LECOMTE (F), M.R.HAYNS (UK), T.P.SPEIS (USA), M.BANASCHIK (D)

240

CLOSING REMARKS

249

LIST OF PARTICIPANTS

250

INDEX OF AUTHORS

260

Abbreviations

AEA CEA CEC CSN IPSN JAERI JRC KfK NRC Univ.

Atomic Energy Authority (GB) Commissariat à l’Energie Atomique (F) Commission of the European Communities Consejo de Seguridad Nuclear (E) Institut de Protection et de Sûreté Nucléaire (F) Japan Atomic Energy Research Institute Joint Research Centre of the CEC Kernforschungszentrum Karlsruhe GmbH (D) Nuclear Regulatory Commission (USA) University

WELCOME ADDRESS M.Livolant Director of Research of the Nuclear Safety and Protection Institute (IPSN)

Ladies and Gentlemen, I have to excuse Mr. J.RASTOIN, Director of the Nuclear Safety and Protection Institute, who intended to open the seminar and welcome the participants. A last minute obligation at the CEA Headquarters made it impossible for him to come. This gives me the pleasure to welcome you in Cadarache for this first international Phebus Fission Product seminar. As you know, the Phebus-FP project was basically an enterprise between the Commission of the European Communities and the French Commissariat à l’Energie Atomique, but soon other organizations showed their interest for the project. The last Steering Committee was, for example, attended by representatives of NRC for the US, NUPEC (MITI) for Japan and Candu Owners Group (Ontario Hydro and AECL) for Canada. I have also to mention that the Korea Atomic Energy Research Institute (KAERI) very recently became a member of the project. May I take this opportunity to announce that this first seminar will be followed up periodically by meetings where the results of the tests will be presented in detail. These meetings will be open exclusively to the members of the project. You have received the programme of the seminar. Let me focus on the highlights of the meeting. After a survey of LWR severe accident source term research throughout the world, a state-of-the-art deduced from previous large experiments will be presented concerning the three main aspects of severe accident source term: - core degradation and fission products release; - fission products transport; - fission products chemistry. This will give some ideas about what is known on the subject. Then, safety analysis requirements will be discussed, to give some thoughts to what else is needed on the subject. Subsequently a description of the Phebus project will be given including supporting analytical activities on source term carried out mainly in Europe. Ideally, this should allow to show how this project satisfies safety analysis requirements. In practice, it is well known that the connection between safety analysis requirements and research activities is not easy to establish and maintain. Finally, as Director of Research of the Nuclear Safety and Protection Institute (IPSN), I would like to make some remarks about the team in charge of the preparation of the tests. As you might know, prior to the Phebus-FP project, two other projects were realized in the same installation: - the test series Phebus LOCA (Loss of Coolant Accident), and - the test series Phebus CSD (Severe Fuel Damage). The same team has been, and still is, in charge of the following experiments concerning the reactivity of cooling accidents in fast breeders: - SCARABEE, - CABRI 1 and CABRI 2. CABRI 1 has been terminated whereas the other two experiments are continuing activities. The success of these large-scale experiments gives us some confidence that, even if the Phebus-FP project is difficult from the experimental point of view, the team in charge has a profound experience in the field, which is one of the strong points concerning this project. The Phebus-FP is presently the largest in-pile experiment in the field of LWR severe accident source term research. I am glad that this experiment will take place in Cadarache and that it gives us the occasion of a wide international cooperation. I hope that the seminar will be sufficiently persuasive to convince other countries to join in and become members of the Phebus-FP project.

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THE PHEBUS FISSION PRODUCT PROJECT

Ladies and gentlemen, I thank you for your attention.

WELCOME ADDRESS J.P.Contzen Director General of the Joint Research Centre, Commission of the European Communities

Mr. Chairman, Ladies and Gentlemen, In these introductory remarks, I would like to stress the genuine interest of the Commission of the European Communities for the project which is the subject of our discussions during the seminar and in the name of the Commission, I wish to convey our best wishes for the success of this meeting. We should also thank the CEA/CEN Cadarache for its hospitality. I will refrain from making the usual comments or jokes about the weather and how good our hosts are at weather control, I shall limit myself to the brief comment that meeting in Cadarache is definitely more pleasant than in industrial environments such as Duisburg, Roubaix Tourcoing, Pittsburgh or downtown Kobe. The European Community interest in the Phebus-FP project stems from two types of considerations: one of organisational and one of technical origin. From the organisational point of view, this project is fairly unique according to Community practices. Indeed it is the only case where the European Community as such, i.e. its 12 Member States, participates technically and financially, through its own Research Centre, in a project of a significant size which has been initiated and largely implemented by one of its Member States. This reflects, I feel, an evolution in the approach followed by the European Community for the fulfilment of its scientific management. Decentralisation, closer coordination with Member States are key words in this respect. The rather centralised concept embedded in the Euratom Treaty—a Napoleonic view if I refer to the terms used by our own President Jacques Delors —should evolve into a concept where decentralisation and delocalisation have more space: if Member States offer substantial capabilities in terms of human resources and facilities which could be used for reaching Community goals, these capabilities should be fully utilised at a Community level using innovative schemes of cooperation. Important is that objectives responding to the needs of the entire Community are attained, more important than how they are implemented. In this sense, the Phebus project could be a model for other projects in the nuclear or in other fields of Community activities. Allow me, as Director General of the JRC, to express to our colleagues of the CEA my appreciation for the good collaboration established between our respective staff. What has been put in place here can be described as the first JRC outpost in a national research centre. This concept of scientific outposts is currently under discussion for a wider implementation and again Phebus can be quoted as a model case. The Phebus project, the approach which has been adopted for its implementation, responds to the needs dictated by the current situation in Europe: shrinking credits, a fact which constitutes a strong incentive for task sharing and a growing concern about building a consensus on nuclear safety issues, an element which leads to associate in nuclear safety research as many partners as possible—both those who pursue the nuclear option and those who have not opted for it or have abandoned it for the time being. This requirement for an as wide as possible consensus on safety issues leads naturally to seek partners beyond the borders of the European Community and we welcome the efforts which have been made to associate Korea, Japan, the United States and hopefully further partners, in the Phebus project. All the arguments I briefly evoked in favour of European cooperation equally apply at international level. From a scientific and technical point of view, the Phebus project fits very well to the general objectives of the European Community action in the nuclear safety field. Since the implementation of the EURATOM Treaty, the Commission has sought to promote, notably through research actions, a harmonised approach to nuclear safety in general and to nuclear reactor safety more particularly. During the last 20 years it consistently has made available substantial resources for improving our understanding of potential consequences of hypothetical accidents.

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THE PHEBUS FISSION PRODUCT PROJECT

Because of the complexity of phenomena occurring during large hypothetical accidents, in-pile tests, reproducing realistic conditions and making use of real materials have been executed for some considerable time. I would like to mention in this respect Phebus-CSD, PBF/NRU and LOFT. In these facilities the phenomena of fuel degradation and melting were investigated under loss of coolant and special transient conditions. What remained to be more thoroughly investigated was the chain of events leading to fission product release to the outer containment during substantial melting of irradiated fuel: the so-called source term problem. In 1985, the Commission initiated an action with its Member States with a view to demonstrate the state-of-the-art in this problem area. In this context the discussion with the French authorities on our possible participation started around the end of 1985. The merits of the project itself, but also the preoccupations created by the Chernobyl accident in April 1986, created the conditions for removing obstacles to our participation in the CEA proposed project. I had the pleasure of signing, for the Commission, the agreement associating us with the CEA in July 1988. In spite of the drastic cuts in funds available for nuclear safety research in the Community framework programme 1990– 1994, we have the firm intention to pursue our collaboration. We feel that the technical goals set initially have kept their validity and that a comprehensive and difficult test programme such as Phebus-FP will contribute to the formulation, notably in Europe but also at a world level, of the best possible methods of analysis, by bringing together, around the experimental facility, analysts to develop and verify common tools. One problem that I would like to raise in closing these remarks is the issue of our credibility vis-à-vis our political masters, vis-à-vis those who have to provide the funding for such a project. With the final Phebus-FP report foreseen for 1999, we span at least the duration of 2 even 3 legislatures. Even if it has not the very long term character of the Fusion programme where politicians cannot hope to see a practical application during their own life, this project requires a standing effort over a long period of time. How can we ensure continued support? Meetings such as the seminar which begins today are a good mechanism to monitor at technical level the progress of work, to adjust programmes in order to achieve the most significant results, to verify the adequacy of the objectives with respect to the needs of the users, notably the regulatory authorities. All this is necessary and I welcome Mr. Livolant’s proposal to transform this seminar into a periodic exercise for those participating effectively in the project. But how to carry the message to the political level? How to demonstrate in simple terms that effective progress is made and that the money is well spent, that scientists and technicians are not indulging in an exercise of self-satisfaction? It is an essential problem, to which I have no immediate answer to offer. I can only tell you that we need an answer. In wishing you every success in your work, I beg you to devote some consideration to the question I just raised. Thank you for your attention.

SESSION I SURVEY OF LWR SEVERE ACCIDENT SOURCE TERM RESEARCH

LWR severe accident source term research in Europe M.Livolant , CEA/IPSN Fontenay-aux-Roses LWR severe accident source term research in the USA T.P.Speis , R.Y.Lee , L.Soffer , R.O.Meyer , NRC Washington Survey of severe accident experiments and analysis in Japan M.Akiyama , University of Tokyo K.Takumi , Nuclear Power Engineering Center, Tokyo K.Soda , JAERI Tokai-mura Summary of discussion S.Finzi , CEC Brussels

LWR SEVERE ACCIDENT SOURCE TERM RESEARCH IN EUROPE M.LIVOLANT Commissariat à l’Energie Atomique DRSN, CE/FAR, BP 6–92265 FONTENAY-AUX-ROSES CEDEX, France

1. NATIONAL SOURCE TERM POSITIONS AND PRACTICES The positions on source terms and the practices concerning emergency plans are different from one country to another in Europe. A presentation of some of those practices is useful before presenting the corresponding research work. 1.1. Switzerland Severe accident source terms are primarily used for the planning of emergency countermeasures, considering both the amount and the time scale of the postulated releases during severe core melt accidents. The time scale is especially important in Switzerland where nuclear power plants are often situated near highly populated areas. In the past, emergency planning has been based on adapted WASH 1400 source terms, ranging from PWR 2 (with iodine release reduced by a factor 10) for the fast alarm system to PWR 5 or 6 for countermeasures against ground contamination. A new reference source term for the purpose of emergency planning has been defined by the Swiss Federal Nuclear Safety Inspectorate on the basis of probabilistic studies for the two newest Swiss plants, the PWR Gösgenand and the BWR Leibstadt. The approach chosen to define the reference source term is to consider that emergency planning does not need to cover all imaginable accident situations but can be based on reasonably conservative best estimate assumptions. Moreover, it appears better to define one reference source term for all plants. So, excluding accident sequences with extremely low probabilities, and under realistic assumptions concerning accident event timing, containment break, engineered safety features and operator intervention, the following source term is used as a general basis for all Swiss nuclear power plants:

-

Start of release: Duration of release: Released radioactivity:

4 hours after shutdown 4 hours 100% noble gases (3×108 Ci) 1×106 Ci iodine (0.3%) 3×105 Ci aerosols, including 3×104 Ci Césium. 1.2. Sweden

In Sweden, the use of source terms is primarily considered as forming part of a complete safety analysis of the nuclear power plants. A variety of estimated source terms and corresponding environmental consequences may be deduced from the study of various types of accidents. Regulatory requirements for mitigative measures (guidelines given by the Government by decree of February 1986) are the following: -

Land contamination which could impede the use of large areas for a long period shall be prevented; Fatalities in acute radiation should not occur; Extremely improbable scenarios have not to be considered for meeting the requirements; The specified maximum release of radioactive substance shall apply to all reactors irrespective of site and power.

LWR SEVERE ACCIDENT SOURCE TERM RESEARCH IN EUROPE

7

Practically, the corresponding releases to fulfill the requirements on land contamination are estimated as at most of 0.1 % of the inventory of the Cesium isotopes 134 and 137. However, the off-site emergency planning applied in practice, based on source terms at higher levels defined precedently, has not changed, because most of the various aspects of emergency preparedness, like information to the public, plans for measurements, means for communication, etc, are rather insensitive to the level of source term. 1.3. France The rationale of the source term estimation in France is as follows: Three levels of source terms, named S1, S2, S3, have been defined. S1 would correspond to a containment rupture in a short time, and of fission products release without filtration, S2 to a delayed rupture with no filtration, and S3 with a delayed rupture with filtration. The order of magnitude of release is some ten percent for S1, some percent for S2 and some per thousand for S3. The accident sequences conducting to a S1 term are generally of explosive nature, and considered very improbable. All accident sequences which may terminate by a S2 level term are studied in detail and mitigation consequences are taken, including eventually some new equipment, like sand bed filters, so as to reduce the corresponding source term level to S3. So practically, S3 is the source term used for emergency planning. As already stated, it corresponds to a delayed release through a filtered channel. The main S3 characteristics are as follows: Noble gases: I131 total: Cs137:

75% 0.9% 0.35%

The main part of the release is supposed to begin approximately 20 hours after core melting. 1.4. Other European countries The other european countries have more or less similar views, but, in general, source terms for emergency planning are not stated on a national base, but estimated on a case by case evaluation of accident sequence studies made in the frame of probabilistic safety analysis. The extreme case is Italy where there are no more nuclear reactors in exploitation. For the future plants, it is foreseen to achieve a reduction of the source terms, evaluated in a realistic manner, to such values that there is no need of preplanned evacuation plan and extensive land decontamination. 1.5. General trends The general trend, specially for the future reactors, is to impose a reduction of the level of the source term considered as realistic and used in the emergency planning. One of the main reason for that is certainly in connexion with the fact that the consequences of a nuclear accident with a source term of the order of magnitude of those actually in use, like S3 in France, will not be limited to the strict health and economical direct effects, but will be largely amplified by mediatic effects. So, there is a general agreement to consider that presently used source terms have to decrease in the future. When one considers how source terms are established, it appears that there are still many incertitudes in their evaluation. In the perspective of reduced source terms, a strong research program is necessary to avoid excessive and costly conservatism. Such a program is in progress in Europe, under the financing of national organisations, with a support from the European Communities. 2. SOURCE TERM RESEARCH IN EUROPE The research work on source term in Europe is made partly in national laboratories, partly in European Research Centers (mainly Ispra). The funding is also a mixed one. The ECC plays a role in this field by its very important participation to the PHEBUS FP program, by direct research work in Ispra JRC, and by its contribution to the shared cost actions.

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THE PHEBUS FISSION PRODUCT PROJECT

2.1. The Commission Source TERM Activities The initiation of an activity related to severe accident analysis, and in particular to the evaluation of the “source term” started in 1984. At that time, following reviews made by various institutions in the world, it was estimated that it was useful to reassess systematically the potential source terms related to severe accidents, taking into account the new information available and performing in a short time a series of well defined tests and analytical development. A Reactor Safety Shared Cost actions program was set up in 1985, focussed on Fission Products and aerosols behaviour in the reactor containment, an area where it appeared possible to make substantial progress with limited resources. Simultaneously, an Ispra team began also to work on code assessment and test analysis. At the end of 1985, the French CEA asked the Commission to participate in the funding of the in pile test program PHEBUS-FP, which was in a very preliminary stage of discussion in France. The Chernobyl accident emphasized the need for research on severe accident mitigation, and the Commission effort in the field was largely increased. In the two years 1987–1988, the JRC launched a new group of experimental and analytical SCAs in the source term area, part of them with the specific objective of preparing in collaboration with the Member States the participation of the Community to the FP project. The agreement for that participation was finally signed in July 1988, with an increase of the direct JRC effort: a team was detached to Cadarache and directly involved in the project work, the Ispra team contributing mainly to test analysis and code development validation. At the same time, the Member States sent experts to the Scientific Analysis Working Group and the Technical Group of the PHEBUS-FP program, and the ECC consultative group CGC 5 set up an adhoc working group on source term where the work performed by the Member States and the shared cost actions are periodically discussed. The main operation directly driven in the field of source term research by the JRC is the ESTER system: the main idea of such a system is to make possible the use of codes made by various organisations in the same system, in order to built in a relatively short time a code able to calculate a whole accidental sequence from modules made for partiel calculations. A data processing structure has been established, and some modules like ICARE, FIPREM, JERICHO and VICTORIA are implemented, or in course of implementation. 2.2. The French program Due to the large number of reactors in France, it was estimated that a good comprehension of the physical bases of the severe accident phenomena was necessary on a national base. So a large experimental program including analytical and in-pile experiments was defined some time ago, and is still proceeding, in parallel with the ESCADRE system for calculation of accidents. The ESCADRE System is able to calculate a complete sequence of a severe accident in a PWR up to radiological consequences. Each phenomenum is represented by models currently validated on analytical experiments carried out in France and abroad, with the needed simplification to allow sufficiently quick parametrical studies. The first part of the HEVA program has been completed. The release rates of fission products and the aerosol characteristics are determined by using pre-irradiated fuel rod sections heated in a furnace to a temperature of approximately 2000°C, in a steam jet with or without hydrogen. The additional program SOPHIE allows to examine the kinetics of the deposit and revaporation of certain selected fission products, like Cs I, Cs OH, I2 Te. It is planned to extend the HEVA program to the examination of fission product release at higher temperature (>2300°C) in a new loop called VERCORS. Experiments carried out as a part of the PITEAS aerosol physics program consist in injecting dry aerosols in a 3 m3 tank containing air and steam with variable saturation rates and monitoring the changes in them. The TUBA loop, designed to examine the retention phenomena of aerosols in small pipes representing steam generator tubes, began operation in 1988. An extension of that type of study to large pipes is being studied now (TUBA GROS TUYAU, i.e. TUBA “BIG PIPE”). Naturally, the top of the program is the PHEBUS-FP program, which will simulate a complex set of phenomena, from fission products emission to deposition in containment. 2.3. UK research program An extended research is made in UK on the nature and behaviour of fission products in the primary circuit and the containment, with a particular attention to chemical effects. Experiments are conducted in the FALCON facility, to study the interaction of real or simulated fission products emitted from fuel samples heated up to 2500°C with aerosols. The transport of the released material is followed through a complex pathway simulating the reactor core, the upper plenum, hot leg

LWR SEVERE ACCIDENT SOURCE TERM RESEARCH IN EUROPE

9

structures and the containment. An important modelling work on vapour-aerosol interaction and iodine chemistry is in progress and the corresponding models are incorporated in the INSPECT and VICTORIA codes. Another topic well studied in UK concerns pool scrubbing, for which the BUSCA code is developped. 2.4. Other European countries Some other experimental and theoretical studies made in European countries are relevant of the subject. In Germany, the VANAM tests are run in the Battelle model containment, with a multicompartiment geometry roughly modelling PWR situation. Aerosols are injected at a high level together with steam and conditions of natural circulation and stratification are established. The aerosols distribution and decay are monitored. The results are used to validate coupled thermohydraulic and aerosols calculation codes, like FIPLOC. A research program is in preparation on pool scrubbing with the possibility of tests in Spain and in Italy and the improvement and validation of calculation codes, like the UK BUSCA code, already mentioned. 3. CONCLUSION Such a review of research program on the source term in Europe may give the impression of some dispersion of efforts in various directions. In fact, for each country, the coherence of the program is estimated on a national base, but, the large extend of exchanges under bilateral agreements or under the auspices of organisations like ECC or OECD allows some specialization for the research teams, at least for the large experiments which are expensive. It is one of the interests of a program like PHEBUS-FP to give opportunity for the best experts in the field to meet and work together, which is probably the most effective way for international collaboration.

LWR SEVERE ACCIDENT SOURCE TERM RESEARCH IN THE USA T.P.Speis, R.Y.Lee, L.Soffer, R.O.Meyer U.S. Nuclear Regulatory Commission Office of Nuclear Regulatory Research Washington, DC 20555

SUMMARY Fission product releases to the environment, or source terms, arise as a result of a highly diverse group of phenomena involved in any particular severe accident sequence. For light water reactors (LWRs) these include core heatup, fuel element degradation and melting, pressure vessel attack and failure, possible high pressure melt ejection, interaction of core debris with concrete, retention of fission products within the reactor coolant system, effects of hydrogen burns or detonations, retention of fission products by suppression pools or ice beds, late revolatilization of fission products from surfaces, and, clearly, the effect of containment integrity or containment bypass and time and location of containment failure, if it occurs. Because of the multiplicity of accident sequences that can occur for a given plant as well as the diversity of the, as yet, imperfectly understood severe accident phenomena, it is not surprising that PRAs such as, for example, those documented in NUREG-1150 have indicated large uncertainties in source terms which represent a significant contribution to the uncertainty in the absolute value of risk. Because of the difficulty and expense involved in performing prototypic experiments, substantial reliance has been placed on the development and validation of detailed mechanistic computer codes for analyzing severe accident phenomena and the source terms associated with them. This paper discusses the extensive research and other efforts that have taken place over the last decade to address the technical issues which have a bearing on being able to describe quantitatively the source term(s) and its characteristics. It also summarizes our present state of knowledge and points out areas where additional research will add further to our understanding. In this context the paper discusses the information that could be provided by the PHEBUS-FP program and its use to assess severe accident integral evaluation codes such as VICTORIA and CONTAIN. Finally, this paper discusses the NRC’s efforts to revise the licensing source term (TID-14844) and the implications of this revision, especially for siting and design of future power plants. 1. INTRODUCTION AND BACKGROUND Radionuclide releases to the environment, that is, the type, quantity, timing and energy characteristics of the release of radioactive material from reactor accidents (“source terms”) are deeply embedded in the regulatory policy and practices of the U.S. Nuclear Regulatory Commission (NRC). For almost thirty years the NRC’s reactor site criteria (10 CFR 100) have required for licensing purposes that an accidental fission product release from the core into the containment be postulated to occur and that its radiological consequences be evaluated assuming that the containment remains intact but leaks at its maximum allowable leak rate. Evaluation of the consequences is used to assess both plant mitigation features such as fission product cleanup systems as well as the suitability of the site. The characteristics of the “source term” into the containment, which must be distinguished from a release to the environment, is contained in Regulatory Guides 1.3 and 1.4, but is derived from the 1962 report TID-14844 (Ref. 1), and consists of 100% of the core inventory of noble gases and 50% of the iodines (half of which are assumed to deposit on interior surfaces very rapidly). Regulatory Guides 1.3 and 1.4 also specify that these are instantaneously available for release, which has significantly affected containment isolation valve closure times; and also specify that the iodine is predominantly (91 percent) in elemental (I2) form. The regulatory applications of this release cover a wide range in addition to plant mitigation features and site suitability and include the basis for (1) the post-accident radiation environment for which safety-related equipment should be qualified, (2) post-accident habitability requirements for the control room, and (3) post-accident sampling systems and accessibility.

LWR SEVERE ACCIDENT SOURCE TERM RESEARCH IN THE USA

11

In contrast to a specified source term for design basis accidents, severe accident source terms first arose in probabilistic risk assessments (e.g., Reactor Safety Study, WASH-1400) in examining accident sequences which involved core melt and where containments could fail. Severe accident source terms represent mechanistically determined “best estimate” releases to the environment, including estimates of failures of containment integrity. This is very different from the combination of the nonmechanistic conservative release to containment postulated by TID-14844 coupled with the assumption of very limited containment leakage used for Part 100 siting calculations for design basis accidents. The worst severe accident source terms resulting from containment failure (especially early failures, i.e., within a few hours from onset of an accident) or containment bypass can lead to consequences that are much greater than those associated with a TID-14844 release into containment and where the containment is assumed to be leaking at its maximum leak rate for its design conditions. Indeed, some of the most severe source terms arise from some containment bypass events, such as “event V” and multiple steam generator tube ruptures. Source term estimates under severe accident conditions began to be of great interest shortly after the Three Mile Island (TMI) accident. A major NRC research effort began about 1981 and has been under way since then to obtain a better understanding of fission-product transport and release mechanisms in LWR’s under severe accident conditions. This research effort has included a very large and extensive staff and contractor effort, involving a number of national laboratories as well as nuclear industry groups, and has resulted in the development and application of several new computer codes to examine core-melt phenomena and associated source-terms involved in severe accident sequences. Work by the NRC staff has also included significant review efforts by peer reviewers, foreign partners in NRC research programs, industry groups, and the general public. Current risk assessment methods, including the latest research efforts on severe accident source terms, have been reflected in the issuance of NUREG-1150 (Ref. 2) which provides an assessment for five US nuclear power plants. Finally, the occurrence of the accident at Unit number 4 of the Chernobyl reactor in the Soviet Union on April 26, 1986 and the large accidental release of fission products resulting from it has provided further impetus to understand severe accident source terms as well as to prevent such occurrences. This paper discusses the major developments that have taken place in our understanding of this “source term” technology, starting with the simplified assumptions of WASH-1400 to the present use of detailed mechanistic codes such as VICTORIA and CONTAIN. The paper also summarizes some of the areas where additional research will add further to our understanding, as well as how programs such as the Phebus-FP can contribute to this understanding. 2. SEVERE ACCIDENT SOURCE TERM RESEARCH IN THE USA Scope of Research Efforts Beginning shortly after the accident at Three Mile Island, the NRC has sponsored numerous experimental and analytical research projects on fission product release and transport. Table 1 list the major NRC research projects conducted and makes reference to major publications resulting from that work. Table 1: Major severe accident fission product research projects sponsered by NRC Project Description

Laboratory

Reference

Out-of-Pile Release Measurements Post-Accident Chemistry Containment Aerosol Behavior TRAP/MELT Code Validation Aerosol Models for VICTORIA High Temperature Experiments VANESA Code Development Core-Concrete Aerosol Experiments ACRR Source Term Experiments VICTORIA Code Development FASTGRASS Code Development ACE Support PBF Fission Product Tests & Analysis PHEBUS Support

ORNL ORNL ORNL ORNL ORNL SNL SNL SNL SNL SNL ANL ANL INEL INEL

3, 4 5, 6 7–9 10–12 13 14–16 17 18 19 20 21 22, 23 24, 25 26

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THE PHEBUS FISSION PRODUCT PROJECT

Project Description

Laboratory

Reference

SPARC & ICEDF Code Development Source Term Reassessment Activity Coefficient Measurements LACE Support

PNL BCL BCL HEDL

27, 28 29–31 32 33

Early experiments and analytical work tended to focus on release from fuel material under high temperatures and severe accident environments. Later, aerosol deposition and transport modeling was done for behavior in the reactor coolant system and in the containment. Currently, fully integrated models are in the process of being completed for release, transport, condensation of vapors, agglomeration and settling of aerosols, and chemical reactions in the reactor coolant system (VICTORIA) and in the containment (CONTAIN with the TRENDS models). General Overview of the Phenomenology and State of Knowledge on Source Terms In-vessel source term: release from fuel and retention in RCS The Reactor Safety Study, WASH-1400 (Ref. 34) analyzed two specific reactors: Surry, a three-loop PWR with a large dry containment and Peach Bottom, a BWR with Mark I containment. For each, calculations were performed for a number of accident sequences and the results used to define a series of release categories. WASH-1400 assumed that most of the release of radionuclides from the reactor fuel occurred as it melted. Once radionuclides escaped the fuel, they were assumed to pass out of the reactor coolant system (RCS) with no attenuation. Neglect of radionuclide deposition or retention in the RCS was recognized to be unrealistic but was considered conservative. A review of the state-of-art for calculating fission product release and transport were the objectives of a major NRC initiative following the TMI-2 accident [NUREG-0772, NUREG-0773, BMI-2104]. The NRC’s Source Term Code Package (STCP), Fig 1, emerged as an integral tool for analysis of fission product transport in the RCS and containment. STCP models release from the fuel with CORSOR (Ref. 35) and fission product retention and transport in the RCS with TRAPMELT (Ref. 36). For the ex-vessel source term, the release from core-concrete is modeled by VANESA (Ref. 17). Depending on the containment type, NAUA, SPARC or ICEDF (Ref. 28) are used to model the transport and retention of fission product release

LWR SEVERE ACCIDENT SOURCE TERM RESEARCH IN THE USA

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Fig. 1: STCP

from the RCS and from core-concrete interaction into the containment, and subsequent release of fission products to the environment consistent with the state of the containment. CORSOR is a simple temperature correlation (Arrhenius form) of the results of out-of-pile release experiments. Implicit assumptions are made about the chemical form of the 18 species treated, but they are invariant and cannot be adjusted to the changing accident conditions (different ratio of steam and hydrogen environment). Barium, for example, can exist either as an oxide or as a metal in the fuel debris. At the same temperature, a choice of one vs. the other, depending upon the oxidizing state of the environment, can mean a difference in the predicted release rate using CORSOR of about a factor of 500. Furthermore, no provision is made in CORSOR for predicting releases during melting or eutectic formation, from rubble beds or molten pools, or under extremely oxidizing conditions accompanying air ingress after vessel breach. Table 2 shows some STCP results for the fractions of initial core inventory released to the reactor vessel prior to pressure vessel failure for a PWR and BWR and for both high and low pressure sequences. Deficiencies in CORSOR were identified in NUREG-0956, and improved modelling of fission product release from fuel is being developed and implemented in the VICTORIA code (Ref. 20), which is NRC’s most current and sophisticated method for in-vessel release and transport. Particular attention has been devoted in the STCP to radionuclide retention in the RCS. This is an area where significant additional model development has taken place since WASH-1400. The expertise developed for aerosol transport in fast reactor safety programs was applied to the behavior of radionuclides in light water reactor coolant systems. This was done via the TRAPMELT code, which calculates aerosol and vapor transport within the RCS. Table 2: STCP results for fraction of initial core inventory released to vessel prior to RTV failure Radionuclide Group Surry (high pressure) TMLB’ Surry (low pressure) V Peach Bottom (high pressure) Peach Bottom (low pressure) TC2 TC1 NG I Cs Te Sr Ba Ru

0.98 0.98 0.98 0.46 7×10−4 0.013 10–6

1.0 1.0 1.0 0.63 1.5×10−3 0.03 3×10−6

0.87 0.87 0.87 0.62 5×10−4 0.01 10−6

0.92 0.92 0.91 0.3 6×10−4 0.01 8×10−7

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THE PHEBUS FISSION PRODUCT PROJECT

Radionuclide Group Surry (high pressure) TMLB’ Surry (low pressure) V Peach Bottom (high pressure) Peach Bottom (low pressure) TC2 TC1 Ce La

0 10−7

0 2×10−7

0 10−7

0 3×10−8

Using TRAPMELT, it was found that there could be significant retention of released radionuclides in the RCS for many, but not all important sequences. For some accidents analyzed, TRAPMELT calculated that less than 20% of the radioactive material released from the degraded reactor core emerged from the RCS. A convenient way to describe the overall effect of retention in the RCS is to indicate the fraction of materials released from the fuel which is released from the vessel. STCP results for SURRY and Peach Bottom are shown in Table 3. A comparison of these values indicates that retention in the RCS is primarily a function of the RCS pressure. Low pressure sequences are characterized by rapid blowdown of the RCS with little gravitational settling, the dominant mechanism for aerosol deposition. On the other hand, for high pressure accident sequences, fission products released from the fuel are retained in the RCS with high efficiency (except for noble gases). While the PWR results show a fairly regular trend toward increasing release fraction with decreasing RCS pressure, trends among the BWR data are less clear. Reduction in RCS retention for volatile materials in BWR accident sequences illustrates the effect of revaporization because of fission product heating of the structures where fission products had originally deposited. Table 3: STCP results for fraction of initial core inventory released from vessel into cotainment Radionuclide Group Surry (high pressure) TMLB’ Surry (low pressure) V Peach Bottom (high pressure) TC2

Peach Bottom (low pressure) TC1

NG I Cs Te Sr Ba Ru Ce La

0.98 0.81 0.81 0.13 0.8 0.8 0.8 0 0.76

1 0.22 0.21 0.62 0.26 0.26 0.26 0 0.3

N/A N/A N/A N/A N/A N/A N/A N/A N/A

1 0.9 0.8 0.15 0.62 0.6 0.6 0 0.78

Since the fractional releases tabulated in Tables 2 and 3 are correlated in a phenomenological sense, it is more reasonable to present the results in terms of the fraction of initial core inventory released from the vessel into the containment at, or before, vessel failure. This is shown in Table 4. (Individual values in Table 4 may not precisely equal the product of Tables 2 and 3, since these represent mean values of distributions.) The estimated fractional releases depend strongly on the volatility of the fission products, as might be expected. Volatile fission products Iodine and Cesium have similar releases. The difference between semivolatile fission products Sr and Ba are not great. Low volatile fission products Ce and La show similar behavior. For bypass sequences due to the multiple steam generator tube ruptures, STCP predicts very little retention in the RCS or in the secondary side of the steam generator system. Because these predicted releases are high, bypass sequences at Surry and Sequoyah dominated the risk. The radionuclide release for the Surry bypass sequence is shown in Fig. 2. The figure shows that the uncertainty ranges for the source terms for these sequences are large. Table 4: STCP results for fraction of initial core inventory released from the vessel into containment (puff release) Radionuclide Group Surry (high pressure) TMLB’ Surry (low pressure) V Peach Bottom (high pressure) Peach Bottom (low pressure) TC2 TC1 NG I Cs Te Sr Ba Ru

0.98 0.22 0.21 0.28 2×10−4 3×10–3 3×10–7

N/A N/A N/A N/A N/A N/A N/A

0.96 0.92 0.79 0.2 4×10−4 7×10−3 7×10−7

0.9 0.75 0.74 0.04 5×10−4 9×10−3 7×10−7

LWR SEVERE ACCIDENT SOURCE TERM RESEARCH IN THE USA

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Radionuclide Group Surry (high pressure) TMLB’ Surry (low pressure) V Peach Bottom (high pressure) Peach Bottom (low pressure) TC2 TC1 Ce La

0 44×10−8

N/A N/A

0 6×10−8

0 7×10–8

In the STCP analyses, CORSOR assumes iodine and cesium to be in the form of CsI and CsOH, and Te to be in elemental form. These species are transported from the core as vapor. In the RCS, fission products may condense as aerosols or react with surfaces. However, TRAPMELT does not account for chemical reactions of CsI. Several processes are known to alter the chemical form of iodine (e.g. reactions of CsI with borates, metal surfaces). TRAPMELT is basically an aerosol code that treats inert particles. In a rudimentary way, however, TRAPMELT treats some chemical reactions using empirical deposition velocities, but deposition is irreversible. True chemical reactions are not modeled, and revaporization cannot occur. Chemical reactions between aerosols and vapors are not modeled either in TRAPMELT, nor is reentrainment calculated. These processes though are being modeled in VICTORIA. In-vessel source term: revaporization from reactor coolant system Deposition of radioactive material in the RCS has focused on the stability of the deposited radioactive material, particularly whether continued decay heating induces substantial revaporization. As long as the reactor core has not penetrated the reactor vessel, TRAPMELT considers revaporization of the condensed radionuclides, and assumes iodine and cesium to be in especially volatile forms. It probably over-predicts revaporization of deposited materials because the code has no ability to predict evolution of deposited materials to less volatile chemical forms (i.e. does not take into account the change in the chemical form of deposited CsOH). It underpredicts the revaporization of Ba, Ru and Te simply because TRAPMELT contains no chemistry to allow these species to become vapors. The issue has become of particular concern as models of core degradation have evolved to predict that, for some sequences, natural circulation of gases through the reactor core may also heat structures in the RCS to substantially higher temperatures than had been previously predicted. Radioactive material deposited on surfaces within the PWR RCS and BWR reactor vessel can also be reevolved after vessel failure because of self-heating. In NUREG-1150, two cases were considered for the PWRs: one vs. two holes in the RCS (i.e., opening in the vessel due to melt penetration of vessel bottom and/or failure of RCS piping at certain location due to high temperature natural circulation). The latter case offers the opportunity of a “chimney effect” and a greatly different environment. The “chimney effect” provides a natural circulation of air through the RCS following failure of the RPV. The circulation is driven because gases in the RCS are heated by deposited radionuclides and retained fuel. The effect is of some significance since current analyses of core degradation indicate substantial fractions of the core could be retained within the RPV after vessel failure. Evidence from TMI-2 suggests that as much as half of the core material may have stayed within the original confines after the rest of the core had melted and drained into the lower plenum. This remaining fuel in the core region could be exposed to air once the plenum has been breached. The analysis considered the balance of buoyancy forces and pressure drop to determine heat transfer from radionuclides and core debris to the gas, and the chemical thermodynamics of revaporization. The results are most strongly affected by the thermo-chemistry of the deposited radionuclides and the geometry of breaches in the RCS. Releases for three elemental groups: iodine, cesium, and tellurium were considered. The results show that the fractional release is greatest for iodine and least for tellurium. Fig. 3 illustrates the distribution for the release of iodine for the PWR case with two holes in the RCS. The range for this case, which produces the greatest release, is from 0 to 70 percent, with a median release of 20 percent. Cesium release fractions were comparable to the iodine values, but slightly less. The median release of tellurium was 0 percent for all cases, but the upper bound varied from 20 to 60 percent. This skewed distribution is indicative of a general belief that there will be little or no revaporization of tellurium, but it recognizes that substantial revaporization cannot be ruled out. Recognizing that TRAPMELT is inadequate to address revaporization from the RCS, the VICTORIA code is incorporating such a model. Ex-vessel source term In low pressure accident scenarios where the reactor vessel fails, high-temperature core debris may fall into the reactor cavity where it interacts with structural concrete. At high temperatures (approximately 1,300–1,500°C), concrete decomposes, and the ablation products commonly include water vapor and carbon dioxide as well as the refractory oxides CaO and SiO2. The liquefied oxidic components of the concrete mix with the uranium oxide fuel and metallic oxides of the debris. Typically, the core debris is initially all or partially molten; gases released at the debris-concrete interface bubble through the debris pool reducing some low-vapor-pressure oxides like La2O3 to high-vapor-pressure forms like LaO. These more volatile forms then vaporize into the bubble volume thus releasing fission product species that were not released in the vessel. Aerosols are

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THE PHEBUS FISSION PRODUCT PROJECT

Fig. 2: Release fraction for containment bypass at Surry

Fig. 3: Revaporization release fraction for iodine, PWR case with two openings.

formed when the bubbles exit the upper surface and fragment. If an overlying water pool exists, a considerable amount of the aerosols may be scrubbed and kept out of the containment atmosphere. STCP uses CORCON-MOD2 (Ref. 37) for modelling core-concrete interactions and VANESA for radionuclide releases driven by bubbling of reaction gases into the melt. VANESA calculates the releases by vaporization of fission products and other melt constituents from the melt into the gas bubbles. Among the factors that influence the magnitude of the ex-vessel releases are the composition and temperatures of the core debris. Concrete composition also has a major impact on the amount of aerosols entrained into the containment atmosphere. Limestone concrete produces larger gas flows and is more oxidizing compared to basaltic. Among the five plants analyzed in NUREG-1150, only Surry has basaltic concrete. CORCON-MOD3 is the latest computer code for predicting core-concrete interactions. It combines CORCON-MOD2 and VANESA together into a single code. For core-concrete interactions, the code predicts heat transfer to the containment, noncondensible and combustible gas generation, and radionuclide release and aerosol generation. CORCON-MOD3 is near completion and it is expected to be adequate for analysis of ex-vessel source terms.

LWR SEVERE ACCIDENT SOURCE TERM RESEARCH IN THE USA

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Release from containment Ultimately the amount of fission products released to the environment depends on the containment’s ability to withstand the various challenges which result from the evolution of a severe accident and the accompanying thermal and mechanical loads. The ability then of the containment to maintain its integrity is determined mainly by two factors: (i) the magnitude of these thermal and mechanical loads, and (ii) the response of the containment structure to those loads. From a risk perspective, containment is considered to have failed to perform its function when the leak rate of radionuclides to the environment becomes substantial. Failure can occur as the result of a structural failure of the containment, tearing of the containment liner, or a high rate of leakage through a penetration. In some accidents, loss of the containment function is independent of these loads. For example, in interfacing system LOCAs, the containment is effectively bypassed. In these sequences, check valves isolating low-pressure piping fail and the piping connected to the RCS fails outside containment. Radionuclides can escape to secondary buildings through the RCS piping without passing through the containment. For most severe accidents, if the containment function is maintained, the radiological consequences will be small. If the containment does fail, the timing of failure can be very important. The longer the containment remains intact relative to the time of core melting and fission product release from the RCS, the more time is available for radioactive material to be removed from the containment atmosphere by engineered safety features and natural deposition processes. A delay in containment failure also provides time for protective actions to be taken. Thus, in evaluating containment performance, it is convenient to designate no failure, late failure, bypass, and early failure of containment as separate categories characterizing different degrees of severity. Plants which have the option to vent the containment are represented by a separate category. In NUREG-1150, containment performance was analyzed with respect to the timing of containment failure and the magnitude of leakage to the environment. However, radionuclide release to the environment is also affected by the performance of engineered safety features. Engineered safety features typically employed in PWRs are sprays, fan coolers, and ice condensers; and in BWR’s, filters and suppression pools. Flooding of reactor cavities or pedestals may also be employed. Suppression pools are effective in the removal of radionuclides in the form of aerosols or soluble vapors. Some of the most important radionuclides, such as iodine, cesium, and perhaps tellurium, are largely released during the in-vessel release period, and directed to the suppression pool where they are subjected to scrubbing, even if containment failure has already occurred. For the Peach Bottom plant, decontamination factors ranging from 1.2 to 4,000 with a median value of 80 were calculated in the study. Depending on the timing and location of containment failure, the suppression pool may also be effective in scrubbing core-concrete releases after vessel failure. Although decontamination factors for the suppression pool are large, iodine captured in the pool will not necessarily remain there. The re-evolution of iodine was important in accident scenarios in which the containment has failed and the suppression pool is boiling. In a containment with an ice condenser, borated ice beds remove fission products from the air by processes similar to the BWR pressure suppression pools. The decontamination factor is very sensitive to the volume fraction of steam in the flowing gas, which in turn depends on whether the air-return fans are operational. With the air-return fans on, decontamination factors range from 1.2 to 20, with a median value of 3. Containment sprays are also effective in reducing airborne concentrations of fission product aerosols and vapors. In the Surry (sub-atmosphere) and Zion (large dry) designs, approximately 20 percent of core meltdown sequences were predicted to eventually result in delayed containment failure or basemat meltthrough. The effect of sprays, in those scenarios in which they are operational for an extended time, is to reduce the concentration of particulate radionuclides airborne in the containment to negligible levels in comparison to the noble gases. For shorter periods of operation sprays still have a substantial mitigative effect on the releases. The likelihood and amount of water accumulation below the reactor vessel is determined by the configuration of the reactor cavity or pedestal regions. For the Surry plant, if borated spray is not operating, the cavity will be dry at vessel failure. For Peach Bottom, there is a maximum of approximately 2 feet of water available on the pedestal and drywell floor because of the configuration of the downcomer. If a coolable debris bed is formed in the cavity or pedestal and makeup water is continuously supplied, core-concrete release fission products would be avoided. Even if molten core-concrete interactions occur, an overlying pool of water can reduce the release of radioactive material to the containment by scrubbing. Other more dynamic processes such as steam explosions, which under some circumstances can take place when molten core debris comes into contact with water will cause debris fragmentation which results in additional aerosol formation. Depending on the aerosol size, it may or may not transport far from the source of generation. An example of source terms (fractions of the core inventory of groups of radionuclides released to the environment) from NUREG-1150 for the Surry plant is shown in Table 5. Groups of release fractions are shown vs. the mean exceedance frequency. The results in this table may be contrasted with those of WASH-1400, which indicated that 70 percent of the core inventory of iodine and cesium were predicted to be released with a probability of about 10−5 per reactor-year (for release

18

THE PHEBUS FISSION PRODUCT PROJECT

category PWR-2). NUREG-1150, in contrast, indicates that releases of this magnitude would have a probability of occurrence almost two orders of magnitude lower. Some of this difference is attributed to improved estimates of containment performance since WASH-1400. Table 5: Exceedance frequencies for release fractions for Surry: all internal initators (mean values) Exceedance Freq. (per reactor year)

Release Fractions

NG

I

Cs

Te

Sr

Ba

Ru

La

Ce

10−5

7.5×10−3

3×10−4

10−8