In-vessel core degradation in LWR severe accidents
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In-vessel core degradation in LWR severe accidents

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Nuclear energy and safety

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Language English
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ISSN 1018-5593
European Commission
nuclear science
and technology
In-vessel core degradation
in LWR severe accidents
Report
EUR 16695 EN European Commission
In-vessel core degradation
in LWR severe accidents
T.J. Haste, AEA, Technology (UK)
B. Adroguer, CEA, Cadarache (F)
U. Brockmeier, University of Bochum (D)
P. Hofmann, FZK, Karlsruhe (D)
K. Müller, IKE, Stuttgart (D)
M. Pezzilli, ENEA, Rome (I)
Contract No FI3S-CT92-0001
FINAL REPORT
Research work performed in the framework of the specific programme
'Nuclear Fission Safety 1990-1994' of the European Atomic Energy Community
Reinforced concerted action on 'Reactor Safety' - Project 1 : Core degradation
Directorate-General
Science, Research and Development
1996 EUR 16695 EN Published by the
EUROPEAN COMMISSION
Directorate-General XIII
Telecommunications, Information Market and Exploitation of Research
L-2920 LUXEMBOURG
LEGAL NOTICE
Neither the European Commission nor any person acting on behalf of the Commission
is responsible for the use which might be made of the following information
Cataloguing data can be found at the end of this publication
Luxembourg: Office for Official Publications of the European Communities, 1996
ISBN 92-827-5641-6
© ECSC-EC-EAEC, Brussels · Luxembourg, 1996
Printed in Belgium IN-VESSEL CORE DEGRADATION IN LWR SEVERE ACCIDENTS:
A STATE OF THE ART REPORT
Update January 1991 - June 1995
Τ J Haste (AEA Technology. Winfrith). Β Adroguer (CEA. Cadarache).
U Brockmeier (University of Bochum), Ρ Hofmann (FZK, Karlsruhe),
Κ Müller (IKE. Stuttgart) and M Pezzilli (ENEA, Rome)
SUMMARY
In 1991 the CSNI issued a State-of-the-Art Report on In-Vessel Core Degradation, which reviewed experimental
and analytical studies for light water reactors (LWRs) of Western design. It summarised information on
experiments, computer modelling codes and their assessment, drew conclusions and recommended further work.
Fission product release and fuel-coolant interactions were specifically excluded, while areas of a mainly thermal
hydraulic nature, such as in-vessel natural circulation and debris bed dryout and rewet, were omitted or treated
only briefly. Main conclusions were that the phenomena of early phase core degradation were reasonably well
understood, but this understanding was not fully reflected in the computer codes at the time. The late phase
suffered from a lack of data and reliable models for the phenomena occurring, many of which were imperfectly
understood. Quench behaviour, with its accident management implications, was also poorly modelled.
The Commission of the European Communities (CEC) sponsors multi-partner research programmes in the field
of thermal reactor safety, in which relevant research in member countries is linked with the provision of
additional CEC funding in the framework of Re-inforced Concerted Actions (RCAs). The Core Degradation
RCA (December 1992 to June 1995) aimed to improve understanding of LWR in-vessel degraded core
phenomena; the technical programme included experiments and computer code assessment, development and
validation. To help prioritise these activities, the CSNI SOAR was initially updated to take into account
developments since that report was completed, then finally updated to consider developments during ±e period
of the RCA. Significant items highlighted initially included completion of the remaining tests in the CORA early
phase melt progression series and of the analysis of the PHEBUS SED tests, launch of the new international
PHEBUS FP programme which includes late phase phenomena, and of a new separate-effects programme on
oxidised Zircaloy quenching (single rods), independent peer reviews by the USNRC of its major modelling codes
SCDAP/RELAP5 and MELCOR. completion of the first two International Standard Problems (ISP-28 and ISP-
31) in the melt progression area, and an increased pace generally in the areas of code development, assessment
and application to plant. The range of plant applications had widened, placing new demands on code
capabilities, even in the early phase.
The present report updates to June 1995 the original SOAR in areas relevant to the CEC RCA on core
degradation. It considers only the codes widely used for melt progression analysis in the CEC, keeping broadly
within the remit of the original SOAR (noting an independent review of ICARE2). but in addition the effect of
core degradation on fission product release is briefly treated. It concludes that the heat-up phase of severe
accidents is sufficiently well understood and modelled. Significant deficiencies in the modelling following the
onset of core degradation have been addressed in the CEC programme, with improvements in the capabilities
of European codes such as ICARE2 and KESS; further work is desirable in some areas involving material
interactions, for example PWR control rod modelling. Additional valuable data have become available from the
CORA and separate-effects materials interactions programmes, especially for WER-specific materials. New
data from the oxidised Zircaloy quench tests have improved understanding of the mechanisms of degraded core
quench, important for accident management. However, fully satisfactory models are not yet available, and
bundle quench experiments are now required to provide necessary additional data. While the modelling of
fission product release from degraded fuel is still poorly treated, the PHEBUS FP series is starting to produce
new data as a basis for new model development. Overall, good progress has been made in the in-vessel core
degradation area, providing a sound platform for future developments in the 4th Framework Programme.
This report was produced for the Cornmission of the European Communities under the Re-inforced Concerted
Action "Core Degradation", in part-fulfilment of Contract Number FI3S-CT92-0001 under the 3rd Framework
Programme.
(Ill) CONTENTS
INTRODUCTION 1
References 2
2. SUMMARY DESCRIPTION OF SEVERE ACCIDENT PHENOMENA 3
2.1 Heat-up Phase including Severe Accident Thermal Hydraulics
2.2 Oxidation/Hydrogen Generation 3
2.3 Chemical Interactions amongst Core Materials 4
2.4 Cladding Failure 5
2.5 Relocation and Blockage Formation 6
2.6 Late Phase · 6
2.7 Refill/Quench 7
2.8 Fission Product Release from Degraded Fuel
References 8
3. EXPERIMENTAL PROGRAMMES 13
3.1 CORA
3.2 PHEBUS SFD 29
3.3 PHEBUS FP 38
3.4 SCARABEE 42
3.5 KfK Single Rod Quench Tests4
3.6 Sandia MP Experiments5
3.7 Sandia Ex-Reactor Experiments7
3.8 TMI-2 Investigation Programmes 4
3.9 RASPLAV9
3.10 Conclusions 50
References2
(V) 4. MATERIAL DATA 65
4.1 Materials relevant to Western PWR and BWR designs 6
4.2 Materials relevant to WER designs6
4.3 Conclusions 71
References2
5. REVIEW OF MAIN CODES7
5.1 ATHLET
5.2 ICARE2/CATHARE 80
5.3 KESS5
5.4 SCDAP/RELAP56
5.5 ESTER9
5.6 MELCOR 91
5.7 Conclusions4
References5
6. CODE ASSESSMENTS 103
6.1 Comparisons with Experiments
6.1.1 International Standard Problems
6.1.1.1 ISP-28
6.1.1.2 ISP-318
6.1.1.3 ISP-36 111
6.1.2 Individual Codes4
6.1.2.1 ATHLET
6.1.2.2 ICARE2/CATHARE 123
6.1.2.3 KESS 13
6.1.2.4 SCDAP/RELAP58
(VI) 6.1.2.5 ESTER 139
6.1.2.6 MELCOR
6.2 Assessments based on Reactor Calculations 140
6.3 Code Reviews 14
6.3.1 USNRC Peer Review - MELCOR1
6.3.2 USNRC Peer Review - SCDAP/RELAP5 142
6.3.3 Independent Review - ICARE24
6.4 MELCOR Co-operative Assessment Programme
6.5 CSNI In-Vessel Core Degradation Validation Matrix 145
6.6 Conclusions ' 147
References 149
7. IDENTIFICATION OF MODELLING NEEDS 163
7.1 Heat-up Phase including Severe Accident Thermal Hydraulics 16
7.2 Oxidation/Hydrogen Generation
7.3 Chemical Interactions amongst Core Materials4
7.4 Cladding Failure 165
7.5 Relocation and Blockage Formation
7.6 Refill/Quench6
7.7 Fission Product Release from Degraded Fuel 16
7.8 Material Data 16
7.9 Late Phase7
References8
8. CONCLUSIONS AND RECOMMENDATIONS 171
8.1 Experimental Programmes and Material Dababase
8.2 Code Assessment 172
(VII) 8.3 Code Development Status and Improvement Needs 173
9. ACKNOWLEDGEMENTS 175
10. APPENDK A : DETAILED DESCRIPTIONS OF CODES7
A.l ATHLET9
A.2 ICARE2/CATHARE 18
A.3 KESS 19
A.4 SCDAP/RELAP5 207
A.5 ESTER 215
A. 6 MELCOR
(VIII)