Research and development with regard to severe accidents in pressurised water reactors

218 Pages·2007·3.58 MB·English

Research and development
with regard to severe
accidents in
pressurised water reactors:
Summary and outlook

Rapport IRSN-2007/83 Rapport CEA-2007/351
FOREWORD

This report was produced conjointly by the French Institute for Radiological Protection and Nuclear
Safety (IRSN) and the French Atomic Energy Commission (CEA). The French electric utility EDF also
contributed to Section 8.2.
The following individuals participated in writing or editing all or part of the report:
For the IRSN:
Ahmed Bentaïb, Hervé Bonneville, Cataldo Caroli, Bernard Chaumont, Bernard Clément, Michel Cranga,
Florian Fichot, Joëlle Fleurot, Richard Gonzalez, Vincent Koundy, Bruno Laurent, Jean-Claude Micaelli,
Renaud Meignen, Frédérique Pichereau, Delphine Plassart, Emmanuel Raimond, Anne Tenaud and Jean-
Pierre Van-Dorsselaere.
For the CEA:
Gérard Ducros, Michel Durin, Christophe Journeau, Daniel Magallon, Jean-marie Seiler, Etienne Studer
and Bruno Tourniaire.
For EDF:
William Ranval and EDF's Severe Accident Teams.
CONTENTS
CHAPTER 1 : INTRODUCTION....................................................................6
1.1 SEVERE ACCIDENTS — DEFINITION AND RESEARCH OBJECTIVES...................................................... 6
1.2 RESEACH FRAMEWORK................................................................................................ 7
1.3 REPORT OBJECTIVES.................................................................................................. 8
CHAPTER 2 : SCENARIOS.......................................................................10
2.1 ACCIDENT TYPES POTENTIALLY LEADING TO CORE MELTDOWN......................................................10
2.1.1 LEVEL 1 PSA ACCIDENT CLASSES...................................................................................10
2.1.2 MELTDOWN FREQUENCY BY CLASS IN EDF’S 900-MWE LEVEL 1 PSA ..............................................16
2.1.3 ACCIDENT PROGRESSION............................................................................................17
2.2 GENERAL PROGRESSION OF SEVERE ACCIDENTS .....................................................................19
2.2.1 PHYSICAL PHENOMENA POSTULATED FOR PWRS....................................................................19
2.2.2 DIFFERENCES BETWEEN 900-MWE AND 1300-MWE PWRS WITH REGARD TO SEVERE ACCIDENTS ..................21
2.2.3 EPR SPECIFICITIES.................................................................................................21
2.2.4 SEVERE ACCIDENT MANAGEMENT....................................................................................22
CHAPTER 3 : IN-VESSEL ACCIDENT PROGRESSION...............................................26
3.1 IN-VESSEL CORE DEGRADATION......................................................................................26
3.1.1 DEFINITION, OVERALL PHENOMENOLOGY ...........................................................................26
3.1.2 PHYSICAL PHENOMENA.............................................................................................26
3.1.3 EXPERIMENTAL PROGRAMMES, MODELLING AND COMPUTER CODES..................................................29
3.1.4 SUMMARY AND OUTLOOK...........................................................................................32
3.2 CORIUM BEHAVIOUR IN THE LOWER HEAD...........................................................................33
3.2.1 DEFINITION, OVERALL PHENOMENOLOGY ...........................................................................33
3.2.2 PHYSICAL PHENOMENA.............................................................................................33
3.2.3 EXPERIMENTAL PROGRAMMES, MODELLING AND COMPUTER CODES..................................................37
3.2.4 SUMMARY AND OUTLOOK...........................................................................................40
3.3 VESSEL RUPTURE....................................................................................................41
3.3.1 DEFINITION, OVERALL PHENOMENOLOGY ...........................................................................41
3.3.2 PHYSICAL PHENOMENA.............................................................................................41
3.3.3 EXPERIMENTAL PROGRAMMES, MODELLING AND COMPUTER CODES..................................................42
3.3.4 SUMMARY AND OUTLOOK...........................................................................................46
3.4 HIGH-PRESSURE CORE MELTDOWN...................................................................................47
3.4.1 DEFINITION, OVERALL PHENOMENOLOGY ...........................................................................47
3.4.2 PHYSICAL PHENOMENA.............................................................................................48
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3.4.3 EXPERIMENTAL PROGRAMMES, MODELLING AND COMPUTER CODES..................................................48
3.4.4 SUMMARY AND OUTLOOK...........................................................................................54
CHAPTER 4 : PHENOMENA LIABLE TO INDUCE EARLY CONTAINMENT FAILURE ...................63
4.1 DIRECT CONTAINMENT HEATING.....................................................................................63
4.1.1 DEFINITION, OVERALL PHENOMENOLOGY ...........................................................................63
4.1.2 PHYSICAL PHENOMENA.............................................................................................63
4.1.3 EXPERIMENTAL PROGRAMMES, MODELLING AND COMPUTER CODES..................................................65
4.1.4 SUMMARY AND OUTLOOK...........................................................................................70
4.2 HYDROGEN RISK AND MITIGATION STRATEGIES......................................................................72
4.2.1 HYDROGEN RISK — BACKGROUND...................................................................................72
4.2.2 OVERALL PHENOMENOLOGY AND PROBLEM AREAS...................................................................72
4.2.3 EXPERIMENTAL PROGRAMMES.......................................................................................74
4.2.4 MODELLING AND SIMULATION CODES................................................................................79
4.2.5 SUMMARY AND OUTLOOK...........................................................................................80
4.3 STEAM EXPLOSIONS..................................................................................................82
4.3.1 DEFINITION, OVERALL PHENOMENOLOGY ...........................................................................82
4.3.2 PHYSICAL PHENOMENA.............................................................................................83
4.3.3 EXPERIMENTAL PROGRAMMES, MODELLING AND COMPUTER CODES..................................................87
4.3.4 SUMMARY AND OUTLOOK...........................................................................................91
CHAPTER 5 : PHENOMENA THAT COULD LEAD TO A LATE CONTAINMENT FAILURE...............98
5.1 MOLTEN CORIUM-CONCRETE INTERACTION (MCCI).................................................................98
5.1.1 BACKGROUND TO MCCI AND THE OBJECTIVES OF THE R&D PROGRAMME...........................................98
5.1.2 PHENOMENOLOGY OF MCCI AND THE RESULTING PROBLEMS........................................................98
5.1.3 EXPERIMENTAL PROGRAMMES, MODELLING AND COMPUTER CODES.................................................100
5.1.4 SUMMURY AND OUTLOOK..........................................................................................106
CHAPTER 6 : IN-VESSEL AND EX-VESSEL RETENTION AND COOLING OF CORIUM................111
6.1 IN-VESSEL RETENTION.............................................................................................111
6.1.1 DEFINITION, PHENOMENOLOGY AND SAFETY CHALLENGES..........................................................111
6.1.2 IN-VESSEL RETENTION BY FLODING THE PRIMARY CIRCUIT..........................................................112
6.1.3 IN-VESSEL RETENTION BY FLOODING THE RACTOR PIT..............................................................115
6.2 COOLING OF THE CORIUM UNDER WATER DURING THE MOLTEN CORIUM-CONCRETE INTERACTION................124
6.2.1 DEFINITION AND OVERALL PHENOMENOLOGY.......................................................................124
6.2.2 EXPERIMENTAL PROGRAMMES, MODELLING AND COMPUTER CODES.................................................124
6.2.3 SUMMARY AND OUTLOOK..........................................................................................128
6.3 CORIUM SPREADING (EPR)........................................................................................129
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6.3.1 DEFINITION AND OVERALL PHENOMENOLOGY.......................................................................129
6.3.2 PHYSICAL PHENOMENA INVOLVED..................................................................................129
6.3.3 EXPERIMENTAL PROGRAMMES, MODELLING AND COMPUTER CODES.................................................130
6.3.4 SUMMARY AND OUTLOOK..........................................................................................134
6.4 EX-VESSEL CORE CATCHER........................................................................................135
6.4.1 TYPES OF EX-VESSEL CORE CATCHER ..............................................................................135
6.4.2 PHYSICAL PHENOMENA INVOLVED..................................................................................137
6.4.3 EXPERIMENTAL PROGRAMMES, MODELLING AND COMPUTER CODES.................................................139
6.4.4 SUMMARY AND OUTLOOK..........................................................................................144
CHAPTER 7 : RELEASE AND TRANSPORT OF FISSION PRODUCTS...............................152
7.1 IN-VESSEL RELEASE OF FISSION PRODUCTS........................................................................152
7.1.1 DEFINITION AND OVERALL PHENOMENOLOGY.......................................................................152
7.1.2 PHYSICAL PHENOMENA INVOLVED..................................................................................154
7.1.3 EXPERIMENTAL PROGRAMES, MODELLING AND COMPUTER CODES...................................................157
7.1.4 SUMMARY AND OUTLOOK..........................................................................................161
7.2 TRANSPORT OF FISSION PRODUCTS IN THE PRIMARY AND SECONDARY CIRCUITS...................................163
7.2.1 DEFINITION AND OVERALL PHENOMENOLOGY.......................................................................163
7.2.2 PHYSICAL PHENOMENA INVOLVED..................................................................................163
7.2.3 EXPERIMENTAL PROGRAMMES, MODELLING AND COMPUTER CODES.................................................166
7.2.4 SUMMARY AND OUTLOOK..........................................................................................167
7.3 EX-VESSEL RELEASE OF FISSION PRODUCTS........................................................................169
7.3.1 DEFINITION AND OVERALL PHENOMENOLOGY.............................................................169
7.3.2 PHYSICAL PHENOMENA INVOLVED..........................................................................169
7.3.3 EXPERIMENTAL PROGRAMMES, MODELLING AND COMPUTER CODES..................................................170
7.3.4 SUMMARY AND OUTLOOK..........................................................................................170
7.4 BEHAVIOUR OF THE AEROSOLS WITHIN THE CONTAINMENT BUILDING.............................................171
7.4.1 DEFINITION AND OVERALL PHENOMENOLOGY.......................................................................171
7.4.2 PHYSICAL PHENOMENA INVOLVED..................................................................................171
7.4.3 EXPERIMENTAL PROGRAMMES, MODELLING AND COMPUTER CODES.................................................172
7.4.4 SUMMARY AND OUTLOOK..........................................................................................173
7.5 THE CHEMISTRY OF FISSION PRODUCTS............................................................................174
7.5.1 DEFINITION AND OVERALL PHENOMENOLOGY.......................................................................174
7.5.2 PHYSICAL PHENOMENA INVOLVED..................................................................................174
7.5.3 EXPERIMENTAL PROGRAMMES, MODELLING AND COMPUTER CODES.................................................181
7.5.4 SUMMARY AND OUTLOOK..........................................................................................183
CHAPTER 8 : SEVERE ACCIDENT CODES – CURRENT DEVELOPMENT AND VALIDATION STATUS ..190
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8.1 ASTEC............................................................................................................190
8.1.1 CURRENT CAPABILITIES OF THE CODE..............................................................................191
8.1.2 VALIDATION STATUS...............................................................................................193
8.1.3 CODE UPGRADE PROSPECTS .......................................................................................196
8.2 MAAP.............................................................................................................197
8.2.1 CURRENT CAPABILITIES OF THE CODE..............................................................................197
8.2.2 VALIDATION STATUS...............................................................................................199
8.2.3 CODE UPGRADE PROSPECTS .......................................................................................200
8.3 MELCOR..........................................................................................................201
8.3.1 CURRENT CAPABILITIES OF THE CODE..............................................................................201
8.3.2 VALIDATION STATUS...............................................................................................202
8.3.3 CODE UPGRADE PROSPECTS .......................................................................................204
CHAPTER 9 : CONCLUSION...................................................................208

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CHAPTER 1 : INTRODUCTION
1.1 SEVERE ACCIDENTS — DEFINITION AND RESEARCH OBJECTIVES
In this report, the term “severe accident” refers to an event causing significant damage to reactor fuel
and resulting from more or less complete core meltdown. Such accidents are highly unlikely in light of
the preventive measures implemented by operators. However, they are the focus of considerable
research because the release of radioactive products into the environment would have serious
consequences. This research also reflects a commitment to the defence-in-depth approach.
Severe accidents are generally caused by a cooling failure within the core, which prevents proper
evacuation of residual power. Within a matter of hours, fuel element structure is damaged as the
result of multiple dysfunctions, arising from equipment and/or human error, including the failure of
safety procedures. A series of complex phenomena then occur, according to various scenarios and
depending on the initial conditions of the accident and operator actions. These scenarios might
ultimately lead to a loss of containment integrity and to the risk of large radioactive releases outside
the containment building.
For the purposes of this document, “early releases” are those liable to occur before all the measures
aimed at protecting the general public can be implemented.
The physical phenomena involved in severe accidents are extremely complex; they generally demand
the development of specific research.
The research in this area thus aims to further understanding of these physical phenomena and reduce
the uncertainties surrounding their quantification, with the ultimate goal of developing models that
can be applied to reactors. These models, grouped together in computer codes, should allow predicting
severe accident progression.
In this field, there is no way to conduct experiments on a real-world scale and reproduce all the
possible situations. That is why elementary tests must be used, allowing each physical phenomenon to
be studied separately, followed by global tests confirming the interactions between phenomena. All
testing must be done on a scale within the technical and financial capacities of the facilities used,
while remaining sufficiently representative to allow extrapolation to the reactor scale.
This research addresses reactors currently in operation as well as future reactors. The underlying
phenomena are the same for existing pressurised-water reactors and those in the planning phases.
However, for existing plants, severe accidents were not a design consideration. Consequently,
modifications to these facilities are limited and the research in this area is primarily aimed at limiting
the potential impact of severe accidents. Specifically, there are two complementary research
orientations: 1) characterising releases and studying modes of containment failure, and 2) developing
methods to deal with the consequences.
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For the future EPR (European Pressurised-water Reactor), the French Nuclear Safety Authority has set
the objective of significantly reducing the risk of radioactive releases for all conceivable accident
situations, including core meltdown accidents. Specific design measures must be implemented in order
to practically eliminate accidents with the potential for large early releases, and to reduce the impact
of accidents involving low-pressure core meltdown. Research conducted within this framework should
allow these objectives to be reached.
1.2 RESEACH FRAMEWORK
Most countries with well-developed nuclear power sectors, such as the United States, Japan, Germany,
Belgium, Canada, South Korea, Switzerland, Sweden and Russia, also have severe accident research
programmes. In general, each country has focused on one or more particular aspects of the issue; the
field is too vast to allow investigation of all phenomena by any one national programme.
In France, the first major research programmes addressing severe accidents were initiated in the early
1980s, following the accident at the Three Mile Island Unit 2 reactor in 1979. Due to the size of its
nuclear fleet, France undertook to develop its own programmes in nearly all areas related to severe
accidents. Safety research in France is primarily conducted by the IRSN, the CEA, EDF and AREVA. All
these entities develop or help develop simulation codes, and the IRSN, the CEA and AREVA (in
Germany) all have testing facilities. Certain actions are also outsourced to research centres or
component manufacturers.
Research in the area of severe accidents involves very substantial human and financial resources as
well as collaboration between nuclear stakeholders, industry groups, research centres and safety
authorities, at both the national and international levels. In France, the IRSN, the CEA, EDF and AREVA
have initiated joint programmes on numerous subjects (involving bipartite, tripartite or even
quadripartite agreements) and participate in international programmes, especially those initiatives
supported by the European Commission through its Framework Programmes for Research and
Development (FPRD), or those conducted under the auspices of the OECD (Organisation for Economic
Co-operation and Development).
As part of the Sixth Framework Programme, a Network of Excellence called SARNET (Severe Accident
Research Network of Excellence) and coordinated by the IRSN, was set up to optimise the use of
resources in the area of severe accident research. SARNET brings together 49 participants from 19
different countries, either members of the European Union or new candidates for membership, as well
as Switzerland. It aims to improve scientific knowledge, to define and oversee research programmes,
to ensure the sustainability of outcomes and to disseminate information. Two of its “integration”
activities are the ASTEC code (see Chapter 8) and methodologies for Level 2 Probabilistic Safety
Assessments (PSA-2).
Many international collaborations have also been set up in the framework of the OECD. Through the
Committee on the Safety of Nuclear Installations (CSNI), the OECD coordinates the launch and oversight
of research programmes with the goal of building consensus on scientific and technical questions of
general interest. Subjects are most often selected based on the general recommendations of expert
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reports on issues which have yet to be fully resolved, and on programmes or facilities which will
involve international collaboration. In particular, the report [1.2_1] can be cited in reference; it is
currently being updated. Since the OECD lacks its own budget for this type of action, it relies on
contributions from participants. In the area of simulation tools, the CSNI facilitates the creation of
expert workgroups to establish validation matrices. It also oversees ISP (International Standard
Problems [1.2_2]) which compare experimental results for a given problem with the outcomes of
various teams using various computer codes. Finally, State-of-the-Art Reports are produced on subjects
of general interest (e.g. hydrogen distribution, hydrogen combustion, aerosol behaviour). These SOARs
provide a complete vision of a given problem by reviewing current knowledge and the remaining
uncertainties, and by presenting recommended orientations.
1.3 REPORT OBJECTIVES
This document reviews the current state of research on severe accidents in France and other countries.
It aims to provide an objective vision, and one that's as exhaustive as possible, for this innovative field
of research. It will help in identifying R&D requirements and categorising them hierarchically.
Obviously, the resulting prioritisation must be completed by a rigorous examination of needs in terms
of safety analyses for various risks and physical phenomena, especially in relation to Level 2
Probabilistic Safety Assessments. PSA-2 should be sufficiently advanced so as not to obscure physical
phenomena that, if not properly understood, might result in substantial uncertainty. It should be noted
that neither the safety analyses nor PSA-2 are presented in this document.
This report describes the physical phenomena liable to occur during a severe accident, in the reactor
vessel and the containment. It presents accident sequences and methods for limiting impact. The
corresponding scenarios are detailed in Chapter 2. Chapter 3 deals with in-vessel accident progression,
examining core degradation (3.1), corium behaviour in the lower head (3.2), vessel rupture (3.3) and
high-pressure core meltdown (3.4). Chapter 4 focuses on phenomena liable to induce early
containment failure, namely direct containment heating (4.1), hydrogen risk (4.2) and steam
explosions (4.3). The phenomenon that could lead to a late containment failure, namely molten core-
concrete interaction, is discussed in Chapter 5. Chapter 6 focuses on problems related to in-vessel and
ex-vessel corium retention and cooling, namely in-vessel retention by flooding the primary circuit or
the reactor pit (6.1), cooling of the corium under water during the corium-concrete interaction (6.2),
corium spreading (6.3) and ex-vessel core catchers (6.4). Chapter 7 relates to the release and transport
of fission products (FP), addressing the themes of in-vessel FP release (7.1) and ex-vessel FP release
(7.3), FP transport in the primary and the secondary circuit (7.2), aerosol behaviour in the containment
(7.4) and FP chemistry (7.5). Finally, Chapter 8 presents a review of development and validation
efforts for the main severe accident codes: ASTEC, MAAP and MELCOR.
In Chapters 3—7, for each of the theme areas, the phenomena involved are reviewed. The major
relevant experiments are then briefly described, including recent, ongoing and future projects. The
key models and specific codes (except for integral codes) used to simulate the phenomena in question
are also discussed. Finally, the state of current knowledge is reviewed and an outlook for the future is
presented, especially regarding experimental programmes and the development of modelling tools.
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CHAPTER 1 REFERENCES

[1.2_1] Nuclear Safety Research in OECD Countries, Major Facilities and Programmes at Risk
NEA/2001
[1.2_2] CSNI International Standard Problems, Brief description (1975-1999) NEA/CSNI/R(2000)5

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