Approaches and tools for severe accident analysis for

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INTERNATIONAL ATOMIC ENERGY AGENCY
VIENNA
ISBN 978–92–0–101608–9
ISSN 1020–6450
08-28011_P1327_covI-IV.indd 1 2008-11-04 13:54:31
APPROACHES AND TOOLS FOR
SEVERE ACCIDENT ANALYSIS
FOR NUCLEAR POWER PLANTS
The following States are Members of the International Atomic Energy Agency:
AFGHANISTAN GUATEMALA PAKISTAN
ALBANIA HAITI PALAU
ALGERIA HOLY SEE PANAMA
ANGOLA HONDURAS PARAGUAY
ARGENTINA HUNGARY PERU
ARMENIA ICELAND PHILIPPINES
AUSTRALIA INDIA POLAND
AUSTRIA INDONESIA PORTUGAL
AZERBAIJAN IRAN, ISLAMIC REPUBLIC OF
QATAR
BANGLADESH IRAQ
REPUBLIC OF MOLDOVA
BELARUS IRELAND
ROMANIA
BELGIUM ISRAEL
RUSSIAN FEDERATION
BELIZE ITALY
SAUDI ARABIA
BENIN JAMAICA
SENEGAL
BOLIVIA JAPAN
SERBIA
BOSNIA AND HERZEGOVINA JORDAN
SEYCHELLES
BOTSWANA KAZAKHSTAN
BRAZIL KENYA SIERRA LEONE
BULGARIA KOREA, REPUBLIC OF SINGAPORE
BURKINA FASO KUWAIT SLOVAKIA
CAMEROON KYRGYZSTAN SLOVENIA
CANADA LATVIA SOUTH AFRICA
CENTRAL AFRICAN LEBANON SPAIN
REPUBLIC LIBERIA SRI LANKA
CHAD LIBYAN ARAB JAMAHIRIYA SUDAN
CHILE LIECHTENSTEIN SWEDEN
CHINA LITHUANIA SWITZERLAND
COLOMBIA LUXEMBOURG SYRIAN ARAB REPUBLIC
COSTA RICA MADAGASCAR TAJIKISTAN
CÔTE D’IVOIRE MALAWI THAILAND
CROATIA MALAYSIA THE FORMER YUGOSLAV
CUBA MALI REPUBLIC OF MACEDONIA
CYPRUS MALTA TUNISIA
CZECH REPUBLIC MARSHALL ISLANDS
TURKEY
DEMOCRATIC REPUBLIC MAURITANIA
UGANDA
OF THE CONGO MAURITIUS
UKRAINE
DENMARK MEXICO
UNITED ARAB EMIRATES
DOMINICAN REPUBLIC MONACO
UNITED KINGDOM OF
ECUADOR MONGOLIA
GREAT BRITAIN AND
EGYPT MONTENEGRO
NORTHERN IRELAND
EL SALVADOR MOROCCO
UNITED REPUBLIC
ERITREA MOZAMBIQUE
ESTONIA MYANMAR OF TANZANIA
ETHIOPIA NAMIBIA UNITED STATES OF AMERICA
FINLAND NEPAL URUGUAY
FRANCE NETHERLANDS UZBEKISTAN
GABON NEW ZEALAND VENEZUELA
GEORGIA NICARAGUA VIETNAM
GERMANY NIGER YEMEN
GHANA NIGERIA ZAMBIA
GREECE NORWAY ZIMBABWE
The Agency’s Statute was approved on 23 October 1956 by the Conference on the Statute of
the IAEA held at United Nations Headquarters, New York; it entered into force on 29 July 1957.
The Headquarters of the Agency are situated in Vienna. Its principal objective is “to accelerate and
enlarge the contribution of atomic energy to peace, health and prosperity throughout the world’’.
SAFETY REPORTS SERIES No. 56
APPROACHES AND TOOLS FOR
SEVERE ACCIDENT ANALYSIS
FOR NUCLEAR POWER PLANTS
INTERNATIONAL ATOMIC ENERGY AGENCY
VIENNA, 2008
COPYRIGHT NOTICE
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of the Universal Copyright Convention as adopted in 1952 (Berne) and as
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© IAEA, 2008
Printed by the IAEA in Austria
October 2008
STI/PUB/1327
IAEA Library Cataloguing in Publication Data
Approaches and tools for severe accident analysis for nuclear power
plants. - Vienna : International Atomic Energy Agency, 2008.
p. ; 24 cm. - (Safety reports series, ISSN 1020-6450 ; no. 56)
STI/PUB/1327
ISBN 978–92–0–101608–9
Includes bibliographical references.
1. Nuclear power plants -- Safety measures. — 2. Nuclear power
plants -- Accidents. I. International Atomic Energy Agency. II. Series.
IAEAL 08–00537
FOREWORD
Severe accidents at nuclear power plants involve very complex physical
phenomena that take place sequentially during various stages of accident
progression. Computer codes are essential tools for understanding how the
reactor and its containment might respond under severe accident conditions.
The codes are used as a tool to support engineering judgement, based on which
specific measures to mitigate the effects of severe accidents are designed. They
are also used to determine accident management strategies and for
probabilistic safety assessment. It is very important to use these sophisticated
tools in accordance with certain rules derived from knowledge accumulated
worldwide.
Severe accidents are addressed in the IAEA Safety Standards Series: in
the Safety Requirements for Safety of Nuclear Power Plants (Safety Standards
Series Nos NS-R-1 and NS-R-2) and the Safety Guide on Safety Assessment
and Verification for Nuclear Power Plants (Safety Standards Series No. NS-G-
1.2). These recommend that severe accident sequences be identified, using a
combination of engineering judgement and probabilistic methods, to determine
those sequences for which reasonably practicable preventive or mitigatory
measures should be evaluated and potentially implemented.
The IAEA Safety Report on Accident Analysis for Nuclear Power Plants
(Safety Reports Series No. 23) provides practical guidance for performing
accident analysis based on present good practices worldwide. All the steps
required to perform such an analysis are covered in the report, for example, the
selection of initiating events, acceptance criteria, computer codes and
modelling assumptions, preparation of input data and presentation of
calculation results. It covers both design basis accidents and beyond design
basis accidents, including severe accidents, however, only basic guidance is
provided for severe accident analysis. Therefore, a specific publication on
accident analysis for severe accidents is needed and is provided through this
publication.
The objective of this publication is to provide a set of suggestions, using
current good practices worldwide, on how to perform deterministic analyses of
severe accidents using the available computer codes.
This publication provides a description of factors important to severe
accident analysis, an overview of severe accident phenomena and the current
status in their modelling, categorization of available computer codes, and
differences in approach for various applications of severe accident analysis. The
report covers both the in-vessel and ex-vessel phases of severe accidents, and is
consistent with the IAEA Safety Report on Accident Analysis for Nuclear
Power Plants. It can be considered a complementary report specifically devoted
to the analysis of severe accidents.
Although the publication does not explicitly differentiate between
various reactor types, it has been written essentially on the basis of the
available knowledge and databases developed for light water reactors. Its
application is, therefore, oriented mainly towards pressurized water reactors
and boiling water reactors and, to a more limited extent, Russian WWER type
reactors and pressurized heavy water reactors.
The IAEA officer responsible for this publication was S. Lee of the
Division of Nuclear Installation Safety.
EDITORIAL NOTE
Although great care has been taken to maintain the accuracy of information
contained in this publication, neither the IAEA nor its Member States assume any
responsibility for consequences which may arise from its use.
The use of particular designations of countries or territories does not imply any
judgement by the publisher, the IAEA, as to the legal status of such countries or territories,
of their authorities and institutions or of the delimitation of their boundaries.
The mention of names of specific companies or products (whether or not indicated
as registered) does not imply any intention to infringe proprietary rights, nor should it be
construed as an endorsement or recommendation on the part of the IAEA.
CONTENTS
1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.1. Background . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
1.2. Objectives and scope of the report . . . . . . . . . . . . . . . . . . . . . . . 4
1.3. Structure of the report . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5
2. IMPORTANT IN-VESSEL PHENOMENA . . . . . . . . . . . . . . . . . . . 5
2.1. Thermohydraulics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
2.1.1. Natural circulation of steam and non-condensable
gases . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6
2.1.2. Reflooding of hot, damaged cores . . . . . . . . . . . . . . . . . 8
2.2. Oxidation of core materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10
2.2.1. Zircaloy oxidation in steam . . . . . . . . . . . . . . . . . . . . . . . 10
2.2.2. Oxidation of B C in steam . . . . . . . . . . . . . . . . . . . . . . . . 12
4
2.3. Loss of core geometry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12
2.3.1. Ballooning and rupture of the cladding . . . . . . . . . . . . . 13
2.3.2. Liquefaction and relocation of control and structural
materials . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13
2.3.3. Liquefaction and relocation of zircaloy cladding . . . . . 14
2.3.4. Liquefaction and slumping of the fuel . . . . . . . . . . . . . . 15
2.3.5. Relocation of molten pool materials into the lower
plenum . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16
2.3.6. Fragmentation of embrittled core materials . . . . . . . . . 17
2.4. Heating and failure of the lower head . . . . . . . . . . . . . . . . . . . . 18
2.5. Other factors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19
2.5.1. Impact of alternative core/vessel designs . . . . . . . . . . . . 19
2.5.2. Recriticality . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22
2.5.3. Fission product release and transport . . . . . . . . . . . . . . . 22
3. IMPORTANT EX-VESSEL PHENOMENA . . . . . . . . . . . . . . . . . . 23
3.1. Containment thermohydraulics . . . . . . . . . . . . . . . . . . . . . . . . . . 24
3.1.1. Sources and sinks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25
3.1.2. Pressurization and depressurization . . . . . . . . . . . . . . . . 26
3.1.3. Gas/water interflows between containment
compartments (transport) . . . . . . . . . . . . . . . . . . . . . . . . 27
3.1.4. Heat and mass transfer . . . . . . . . . . . . . . . . . . . . . . . . . . . 27
3.1.5. Hydrogen issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28
3.2. Aerosol and iodine behaviours . . . . . . . . . . . . . . . . . . . . . . . . . . 32
3.2.1. Aerosol behaviour. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32
3.2.2. Iodine behaviour . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 33
3.3. Melt behaviour . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35
3.3.1. Melt release . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 35
3.3.2. Direct containment heating . . . . . . . . . . . . . . . . . . . . . . . 36
3.3.3. Melt coolability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 38
3.3.4. Fuel coolant interactions . . . . . . . . . . . . . . . . . . . . . . . . . 39
3.3.5. Molten corium concrete interaction . . . . . . . . . . . . . . . . 40
3.3.6. Relocation of melt and spreading . . . . . . . . . . . . . . . . . . 42
3.3.7. Interactions with refractory and sacrificial materials . . 43
3.4. Selected engineering systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43
3.4.1. Systems impacting transport . . . . . . . . . . . . . . . . . . . . . . 43
3.4.2. Systems impacting containment leakages . . . . . . . . . . . 44
3.4.3. Safety engineering systems . . . . . . . . . . . . . . . . . . . . . . . 44
4. STATUS IN THE MODELLING OF IN-VESSEL
PHENOMENA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44
4.1. Modelling of individual phenomena . . . . . . . . . . . . . . . . . . . . . . 45
4.1.1. Thermohydraulics . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45
4.1.2. Heat transfer . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 51
4.1.3. Fission and decay heat . . . . . . . . . . . . . . . . . . . . . . . . . . . 52
4.1.4. Oxidation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52
4.1.5. Material interactions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 54
4.1.6. Material relocation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 56
4.1.7. Lower plenum behaviour . . . . . . . . . . . . . . . . . . . . . . . . . 58
4.1.8. Fission products . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 62
4.2. Implications of modelling options and categories of codes. . . . 63
4.3. Uncertainties in modelling severe accident phenomena. . . . . . 65
4.3.1. Well understood phenomena (high level of
knowledge) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 65
4.3.2. Phenomena not fully understood (medium level of
knowledge) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 65
4.3.3. Phenomena only partially understood (low level of
knowledge) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66
5. STATUS IN THE MODELLING OF EX-VESSEL
PHENOMENA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 67
5.1. Containment thermohydraulics . . . . . . . . . . . . . . . . . . . . . . . . . . 67
5.1.1. Sources and sinks . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 69
5.1.2. Pressurization and depressurization . . . . . . . . . . . . . . . . 70
5.1.3. Gas/water cell interflows . . . . . . . . . . . . . . . . . . . . . . . . . 70
5.1.4. Heat and mass transfer . . . . . . . . . . . . . . . . . . . . . . . . . . . 70
5.1.5. Hydrogen issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 71
5.2. Aerosol and iodine behaviours . . . . . . . . . . . . . . . . . . . . . . . . . . 75
5.2.1. Aerosol behaviour . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 76
5.2.2. Iodine behaviour . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 80
5.3. Melt behaviour . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 84
5.3.1. Melt release . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 84
5.3.2. Direct containment heating . . . . . . . . . . . . . . . . . . . . . . . 85
5.3.3. Coolability . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 87
5.3.4. Fuel coolant interactions . . . . . . . . . . . . . . . . . . . . . . . . . 89
5.3.5. Molten corium concrete interaction . . . . . . . . . . . . . . . . 90
5.3.6. Relocation of melt and spreading . . . . . . . . . . . . . . . . . . 92
5.3.7. Interactions with refractory and sacrificial materials . . 93
5.4. Modelling of selected engineering systems . . . . . . . . . . . . . . . . 94
5.4.1. General description . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 94
5.4.2. Systems impacting transport . . . . . . . . . . . . . . . . . . . . . . 94
5.4.3. Systems impacting containment leakages . . . . . . . . . . . 95
5.4.4. Safety engineering systems . . . . . . . . . . . . . . . . . . . . . . . 95
6. USE OF COMPUTER CODES FOR THE ANALYSIS
OF SEVERE ACCIDENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 97
6.1. Classification of severe accident computer codes . . . . . . . . . . . 97
6.1.1. Fast running integral codes . . . . . . . . . . . . . . . . . . . . . . . 98
6.1.2. Detailed codes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 99
6.1.3. Special (dedicated) codes . . . . . . . . . . . . . . . . . . . . . . . . . 99
6.1.4. Capability assessment of present generation
computer codes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 100
6.2. Requirements for modelling severe accident phenomena . . . . 101
6.2.1. General requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . 101
6.2.2. Required level of sophistication of models for
different applications . . . . . . . . . . . . . . . . . . . . . . . . . . . . 103
6.3. Verification and validation of computer codes . . . . . . . . . . . . . 107
6.4. User qualification and user effect on accident analysis . . . . . . 113