In-Vessel Core Degradation [validation martrix] (csni-r2000-21)

351 Pages·2000·4.14 MB·English

Nuclear Safety
NEA/CSNI/R(2000)21
February 2001
I
n-Vessel Core Degradation
Code Validation Matrix
Update 1996-1999
Report by an
OECD/NEA
Group of Experts
October 2000
OECD Nuclear Energy Agency
Le Seine Saint-Germain - 12, boulevard des Îles
F-92130 Issy-les-Moulineaux, France
Tél. [phone] - Fax [phone]
Internet: http://www.nea.fr
N U C L E A R • E N E R G Y • A G E N C Y
Additional copies of this CD-ROM, and paper copies, can be obtained from:

Dr. Jacques Royen
Nuclear Safety Division
OECD Nuclear Energy Agency
Le Seine - Saint Germain
12 Boulevard des Iles
F-92130 Issy-les-Moulineaux
France

E-mail: [email protected]
IN-VESSEL CORE DEGRADATION CODE VALIDATION MATRIX
NEA/CSNI/R(2000)21
CONTENTS OF CD-ROM
Ch_0 IN-VESSEL CORE DEGRADATION
CODE VALIDATION MATRIX
Update 1996-1999
Ch_1 INTRODUCTION
Ch_2 ACCIDENT SEQUENCES FOR LWRs
Ch_3 IDENTIFICATION OF PHENOMENA AND DEGREE OF
PHYSICAL UNDERSTANDING
Ch_4 EXPERIMENTAL DATABASE
Ch_4tab EXPERIMENTAL DATABASE - TABLE
Ch_5 VALIDATION MATRIX
Ch_5tab VALIDATION MATRIX
Ch_6 CONCLUSIONS AND RECOMMENDATIONS
Ch_7 ACKNOWLEDGEMENTS
X_0 APPENDIX A: Summary sheets for experiments
X-INT INTEGRAL TEST FACILITY
X_INT SEPARATE EFFECTS TEST FACILITIES
IN-VESSEL CORE DEGRADATION
CODE VALIDATION MATRIX
Update 1996-1999
31. October 2000
Authors
K Trambauer - GRS Garching, Germany
T J Haste - EC, Joint Research Centre, Ispra, Italy
B Adroguer - IPSN, Cadarache, France
Z Hózer - AEKI, Budapest, Hungary
D Magallon - EC, Joint Research Centre, Ispra, Italy
A Zurita - EC, DG Research, Brussels, Belgium
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IN-VESSEL CORE DEGRADATION
CODE VALIDATION MATRIX
Update 1996-1999
K Trambauer (GRS, Garching), T J Haste (JRC, Ispra), B Adroguer (IPSN, Cadarache),
Z Hózer (AEKI, Budapest), D Magallon (JRC, Ispra) and A Zurita (EC, DG Research)
EXECUTIVE SUMMARY
In 1991 the Committee on the Safety of Nuclear Installations (CSNI) issued a State-of-the-Art
Report (SOAR) on In-Vessel Core Degradation in Light Water Reactor (LWR) Severe
Accidents. Based on the recommendations of this report a Validation Matrix for severe
accident modelling codes was produced. Experiments performed up to the end of 1993 were
considered for this validation matrix. To include recent experiments and to enlarge the scope,
an update was formally inaugurated in January 1999 by the Task Group on Degraded Core
Cooling, a sub-group of Principal Working Group 2 (PWG-2) on Coolant System Behaviour,
and a selection of writing group members was commissioned. The present report documents
the results of this study.
The objective of the Validation Matrix is to define a basic set of experiments, for which
comparison of the measured and calculated parameters forms a basis for establishing the
accuracy of test predictions, covering the full range of in-vessel core degradation phenomena
expected in light water reactor severe accident transients. The emphasis is on integral
experiments, where interactions amongst key phenomena as well as the phenomena themselves
are explored; however separate-effects experiments are also considered especially where
these extend the parameter ranges to cover those expected in postulated LWR severe accident
transients. As well as covering PWR and BWR designs of Western origin, the scope of the
review has been extended to Eastern European (VVER) types. Similarly, the coverage of
phenomena has been extended, starting as before from the initial heat-up but now proceeding
through the in-core stage to include introduction of melt into the lower plenum and further to
core coolability and retention to the lower plenum, with possible external cooling. Items of a
purely thermal hydraulic nature involving no core degradation are excluded, having been
covered in other validation matrix studies. Concerning fission product behaviour, the effect of
core degradation on fission product release is considered, but not its detailed mechanisms;
fission product transport is also outside the scope and has been covered in other validation
matrix studies.
The report initially provides brief overviews of the main LWR severe accident sequences and
of the dominant phenomena involved. The experimental database is then summarised, with test
conditions, phenomena covered and parameter ranges being presented in concise standard
tabular formats to aid comparison amongst the different experimental series. These data are
then cross-referenced against a condensed set of the phenomena and test condition headings
presented earlier, judging the results against a set of selection criteria and identifying key tests
of particular value. Areas where data are still required are identified. Finally, the main
conclusions and recommendations are listed.
The main body of the report is supplemented by an appendix which summarises the
experiments listing the most important references for data and evaluation reports for each test
and/or relevant series, indicating the availability of the data for further analysis, while in
addition evaluating the strengths and weaknesses of the experiments in providing data for code
validation. Overall, the report is designed to help code developers in choosing tests to help
formulate and validate new models, by summarising the relevant data in one place in a
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convenient form, and to aid reviewers and users of codes to judge whether validation of a
particular programme is sufficient for given plant applications, by checking that the relevant
phenomena and parameter ranges have been covered.
Two categories of key tests are defined, in merit order:
• Category 1 experiments are amongst the best qualified for code validation in their field.
ISPs normally fall into this category. Data are well documented and boundary conditions
are well defined (these conditions may be relaxed if there are specific unique features).
Category 1 tests are strongly recommended for the validation of system codes (depending on
their specific objectives);
• Category 2 experiments are well qualified for code validation, and could be used to
increase the degree of confidence in a code's suitability for a given application. The
experiment may not be unique, but valuable in the sense of parameter range.
In general, no key test assignments are made for separate effects experiments, since they are
unlikely to be used directly for system code validation. Exceptions are made in the case of
bundle ballooning and fuel coolant interaction (FCI) experiments as these have some integral
characteristics.
Twelve tests of the core degradation integral experiments and one reactor accident were
selected as Category 1: CORA-13, CORA-28, CORA-33, CORA-W2, Phebus B9+,
PBF SFD-1.4, ACRR ST-1, ACRR DF-4, LOFT LP-FP-2, Phebus FPT1, ACRR MP-1&2, and
TMI-2.
Twenty-one tests of the core degradation integral experiments were selected as Category 2:
CORA-2, CORA-5, CORA-12, CORA-15, CORA-17, CORA-31, CORA-30,
Phebus SFD C3+, Phebus SFD AIC, NRU FLHT-5, ACRR-DF-2, Phebus FPT0, Phebus FPT4
(provisional), Sandia XR1-2, SCARABEE BF1, ACRR DC-1, CODEX-AIT1, CODEX-AIT2,
QUENCH-01, QUENCH-03, and QUENCH-04.
Of the bundle separate effects experiments four tests were selected: REBEKA-6 and
NRU MT-4 for category 1 and for category 2 Phebus 218 and MRBT B6.
Of the fuel-coolant interaction experiments seven tests were selected: FARO L-14,
FARO L-28, and KROTOS K-44 for category 1 and for category 2 FARO L-11, FARO L-31,
FARO L-33, and KROTOS K-58.
Concerning the late phase, the reactor situation is best covered by the following separate
effects test: BALI, RASPLAV Salt, SIMECO, BENSON Rig and the CYBL facility.
Based on the completeness of the experimental data base necessary for code validation, and
reflecting the degree of understanding of phenomena, it is considered that the following
parameters or processes are not fully covered or understood only in a limited way:
• Effects of high burn-up, MOX, quenching at high temperature, and air ingress on core
degradation;
• Transition from early to late phase, crust failure with subsequent slumping of melt into the
lower plenum and quenching;
• Thermal loads to RPV by phase separation in molten pool including metallic layers and
their consequences on vessel failure.
Continual updating of the matrix is considered worthwhile on a regular basis to monitor the
adequacy and completeness of the experimental database for the code validation. This is
particularly important regarding the late phase, which is currently less well covered than the
early phase.
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CONTENTS LIST
Executive Summary i
Contents List iii
List of Tables vii
List of Figures ix
Abbreviations xi
Glossary xv
1. Introduction 1
1.1 Background 1
1.2 Objectives and Scope 1
1.3 Report Structure 2
References 3
2. Accident Sequences for LWRs 5
2.1 Plant Types 5
2.2 Plant Status and Initiating Event 7
2.2.1 PWR Accident Sequences 7
2.2.2 BWR Accident Sequences 8
2.2.3 Stand-by and Shutdown Conditions 9
2.3 Degraded Core Accident Progression 9
References 14
3. Identification of Phenomena and Degree of Physical Understanding 17
3.1 Fission and Decay Heat 18
3.2 Fluid State 18
3.3 Initial Core Damage 19
3.4 Oxidation and Hydrogen Generation 21
3.5 Fission Product Release 22
3.6 Core Degradation and Melt Progression 22
3.7 Core Debris in The Lower Plenum 25
References 28
4. Experimental Database 29
For Each Facility, Four Sub-Sections:
4.x.y.1 Objectives
4.x.y.2 Facility Description
4.x.y.3 Test Description
4.x.y.4 Processes Quantified
4.1 Integral Facilities 30
4.1.1 NIELS 30
4.1.2 CORA 30
4.1.3 PHEBUS-SFD 31
4.1.4 PBF-SFD 33
4.1.5 NRU-FLHT 35
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4.1.6 ACRR-ST 36
4.1.7 ACRR-DF 37
4.1.8 LOFT-LP-FP 38
4.1.9 PHEBUS-FP 40
4.1.10 ACRR-MP 43
4.1.11 SANDIA-XR 44
4.1.12 TMI-2 46
4.1.13 SCARABEE 47
4.1.14 ACRR-DC 48
4.1.15 CODEX 49
4.1.16 QUENCH 50
4.1.17 FARO 51
4.1.18 KROTOS 53
4.2 Separate Effects Facilities 54
4.2.1 Clad Ballooning 54
4.2.2 Materials Interactions 55
4.2.2.1 Material Oxidation 56
4.2.2.2 Structural Material Interactions 57
4.2.2.3 Metal/Ceramic Interactions 58
4.2.3 Reflood 58
4.2.3.1 JAERI 58
4.2.3.2 Fz Karlsruhe 59
4.2.4 Melt Pool Thermal Hydraulics 61
4.2.4.1 Technische Universität Hannover 61
4.2.4.2 Ohio State University (1) 61
4.2.4.3 Ohio State University (2) 62
4.2.4.4 AEA Technology (AEAT) 63
4.2.4.5 COPO I 64
4.2.4.6 COPO II 64
4.2.4.7 ACOPO 65
4.2.4.8 University of California, Los Angeles (UCLA) 66
4.2.4.9 BALI 66
4.2.4.10 RASPLAV AW200 67
4.2.4.11 RASPLAV Salt 68
4.2.4.12 SIMECO 68
4.2.5 Gap Thermal Hydraulics 69
4.2.5.1 CHFG Experiments 69
4.2.5.2 BENSON Test Rig 70
4.2.5.3 CTF 71
4.2.5.4 CORCOM 71
4.2.6 Ex-vessel Thermal Hydraulics 72
4.2.6.1 SULTAN 72
4.2.6.2 SBLB Facility 73
4.2.6.3 CYBL 73
4.2.6.4 ULPU 74
4.2.7 Gap Formation 75
4.2.7.1 FOREVER 75
4.2.7.2 LAVA 76
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4.2.8 Fuel Coolant Interaction 77
4.2.8.1 WFCI 77
4.2.8.2 MAGICO-2000 78
4.2.8.3 SIGMA-2000 79
References 79
Tables 4.1.1 to 4.2.8 81
Figures 4.1.1 to 4.1.18 179
5. Validation Matrix 205
5.1 Matrix Organisation 205
5.2 Selection Criteria 205
5.2.1 Data and Documentation 205
5.2.2 Boundary Conditions 206
5.2.3 Dominant Characteristics 206
5.2.4 Key Test 206
5.3 Cross-Reference Matrix 207
5.4 Justification of Key Tests 208
5.4.1 Category 1 Core Degradatation Integral Experiments 208
5.4.2 Category 2 Core Degradatation Integral Experiments 209
5.4.3 Bundle Separate Effects experiments 213
5.4.4 Fuel-Coolant Interaction Multi Effects Experiments 213
5.5 Late Phase Separate Effects Experiments 214
5.5.1 Melt Pool Thermal-hydraulics 214
5.5.2 Gap Thermal-hydraulics 215
5.5.3 Ex-vessel Thermal-hydraulics 215
5.5.4 Gap Formation 215
5.5.5 Fuel-Coolant Interaction Separate Effects Experiments 215
5.6 Identification of Remaining Experimental Needs 216
References 217
Legend to Tables 5.1 to 5.4 218
Tables 5.1 to 5.5 219
6. Conclusions and Recommendations 225
7. Acknowledgements 227
Appendix A : Summary Sheets for Experiments A - 1
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