When the RPV head is closed, core melt accident phenomena are very similar to the sequences going on in full power mode. Therefore, the large body of guidance which is available for full power mode is basically applicable to shutdown mode with RPV closed as well. When the RPV is open, some of the L2 PSA issues become irrelevant compared to full power mode, while others come into existence. The situation is different for aspects which do not exist or which are less pronounced in sequences with RPV closed.
1.INTRODUCTION 21
2.COMPLEMENT OF EXISTING GUIDANCE FOR SHUTDOWN STATES 23
2.1DEFINITION 24
2.2INTERFACE BETWEEN L1 AND L2 PSA 29
2.3SHUTDOWN STATES L2 PSA 31
2.4ACCIDENT SEQUENCES WITH RPV CLOSED 33
2.5ACCIDENT SEQUENCES WITH RPV OPEN 35
2.5.1FISSION PRODUCT RELEASE FROM CORE MELT IN OPEN REACTOR 39
2.5.2HEAT LOAD FROM THE CORE MELT 41
2.6CONTAINMENT ISSUES FOR ACCIDENTS IN SHUTDOWN MODE 42
2.6.1MODIFIED CONTAINMENT STATUS (RPV OPEN OR CLOSED) 42
2.6.2CONTAINMENT RESPONSE ANALYSIS 45
2.6.3CONTAINMENT EVENT TREE 46
2.6.4Simultaneous accident progression in reactor and spent fuel pool 48
2.7SUMMARY FOR L2 PSA IN SHUTDOWN STATES 49
3.COMPLEMENT OF EXISTING GUIDANCE FOR SPENT FUEL DAMAGE 51
3.1INTRODUCTION 51
3.2EXISTING GUIDANCE/ METHODOLOGY FOR SPENT FUEL POOLS 55
3.3ISSUES RELATED TO SFD WITHIN EXTENDED L2 PSA 60
3.3.1IDENTIFICATION OF INITIATING EVENTS 64
3.3.2ACCIDENT SEQUENCE ANALYSIS 69
3.3.3THERMAL HYDRAULIC CALCULATIONS AND SUCCESS CRITERIA 70
3.3.4HUMAN RELIABILITY ANALYSIS 71
3.3.5FUEL DEGRADATION PROCESS IN SPENT FUEL POOLS 72
3.3.6HYDROGEN ISSUES IN SPENT FUEL POOL MELTING 74
3.3.7HEAT LOAD DUE TO SPENT FUEL POOL MELTING 76
3.3.8RELEASE PATHWAYS TO THE ENVIRONMENT IN CASE OF SPENT FUEL POOL MELTING 78
3.3.9CORRELATIONS BETWEEN ACCIDENT PROGRESSION IN SPENT FUEL POOL AND IN THE REACTOR VESSEL 81
3.3.10CORE CONCRETE INTERACTIONS FOR SPENT FUEL POOL ACCIDENTS 82
3.3.11CRITICALITY IN SPENT FUEL POOLS 82
3.3.12SAFETY ASSESSMENT OF SPENT FUEL POOL DURING DECOMMISSIONING 84
3.4SUMMARY FOR L2 PSA FOR SPENT FUEL POOLS 85
4.COMPLEMENT OF EXISTING GUIDANCE BASED ON RECENT R&D 88
4.1Recent R&D ON CORE MELT ISSUES IN GENERAL 88
4.1.1RECENT R&D ON ACCIDENTS IN REACTOR SHUTDOWN STATES 88
4.1.2ANALYSIS OF THE COMPLEXITY OF SEVERE ACCIDENT PHENOMENOLOGY BY CODE SIMULATION (ASTEC AND MELCOR) 89
4.1.3INVESTIGATION OF IN-VESSEL MELT RETENTION STRATEGY 93
4.1.4STATUS OF SOURCE TERM RESEARCH AND PERSPECTIVES FOR L2 PSA 97
4.1.5MOLTEN CORIUM CONCRETE INTERACTION (MCCI) 107
4.2RECENT R&D ON SPENT FUEL POOL ACCIDENTS 107
4.2.1CSNI STATUS REPORT ON SPENT FUEL POOL UNDER ACCIDENT CONDITIONS 108
4.2.2EXPERIMENTS WITH RELEVANCE TO SFP COOLING ACCIDENTS 108
4.2.3Detailed SIMULATION TOOLS 110
4.2.4ABILITY OF REACTOR CORE SEVERE ACCIDENT CODES TO SIMULATE SFP SEVERE ACCIDENTS 111
4.2.5ONGOING R&D ACTIVITIES 113
4.2.6ANALYSIS OF HEAVY LOAD DROPS INTO THE SFP (UJV) 117
4.3KNOWLEDGE GAPS AND FUTURE NEEDS 118
4.4SUMMARY FOR L2 PSA CONSIDERING RECENT R&D 123
4.5Safety Research areas identified in the NEA-SAREF-Project 124
5.CONCLUSION AND RECOMMENDATIONS 127
5.1COMPLEMENTARY GUIDANCE FOR LEVEL 2 PSA FOR THE SHUTDOWN STATES OF REACTORS 127
5.2COMPLEMENT OF EXISTING GUIDANCE FOR SPENT FUEL DAMAGE 130
5.3Recent R&D in L2 PSA 133
5.4List of General Recommendations 136
6.List of References 138
7.List of Tables 146
8.List of Figures 147
9.APPENDICES 148
9.1 Practical containment analysis during shutdown states for French PWR (EDF) 148
9.1.1 IDENTIFICATION 148
9.1.2QUANTIFICATION 148
9.2 LEVEL 1 SHUTDOWN STATES PSA 149
9.3Examples of ACCIDENT PROGRESSION in shutdown state 151
9.3.1An example from Belgian PWRs 151
9.3.2An Example from German PWR’s 155
9.3.3An example from Spanish BWR (Mark-III containment) 156
9.3.4An example from Swedish NPP’s 157
9.3.5An example from Swiss NPP’s (CCA) 161
9.3.6An example from Ukrainian VVER’s (SSTC) 172
9.3.7EXAMPLE FROM BULGARIA (INRNE) 177
11.Examples of ACCIDENT PROGRESSION in spent fuel pool 189
11.1Example from GRS for Spent fuel pool accident in a PWR 189
11.2ASETC Calculations OF SFP ACCIDENTS IN L2 PSA (IRSN) 193
11.2.1CONTEXT IN FRANCE 193
11.2.2900 AND 1300 MWE PWR SFP CONFIGURATIONS STUDIED WITH ASTEC 194
11.2.3PHYSICAL PHENOMENA 196
11.2.4SUMMARY OF ASTEC RESULTS AND LIMITATIONS OF THE CODE 197
11.2.5CONCLUSIONS 198
11.3Example Spent fuel pool in Sweden (LRC) 198
11.4Example Spent fuel pool in France (EDF) 200
11.5Example Spent fuel pool IN JAPAN (jansi) 200
11.6EXAMPLE SPENT FUEL POOL IN HUNGARY (NUBIKI) 201
11.7Example Spent fuel pool in Ukraine (SSTC) 208
11.8EXAMPLE FROM BULGARIA (INRNE) 211
ACWS
|
Auxiliary Cooling Water System
|
AECL
|
Atomic Energy of Canada Limited
|
AEKI
|
Atomic Energy Research Institute (Hungary)
|
AFW
|
Auxiliary Feed Water
|
ALPS
|
Advanced Liquid Processing System
|
ASTEC
|
Accident Source Term Evaluation Code
|
BBN
|
Bayesian Belief Networks
|
BDBA
|
Beyond Design Basis Assessment
|
BEEJT
|
Benchmark Exercise on Expert Judgment Techniques
|
BRUA
|
Steam Relief Valve to Atmosphere
|
BRUK
|
Steam Dump Valve to Condenser
|
BWR
|
Boiling Water Reactor
|
CAV
|
Cavity Package
|
CCI
|
Corium Concrete Interaction
|
CDF
|
Core Damage Frequency
|
CET
|
Containment Event Tree
|
CFD
|
Computational Fluid Dynamics
|
CFF
|
Containment Failure Frequency
|
CFR
|
Code of Federal Regulations
|
CFVS
|
Containment Filtered Venting System
|
CHRS
|
Containment Heat Removal System
|
CIS
|
Containment Isolation System
|
CNFW
|
Condensate And Feedwater System
|
CO
|
Carbon Monoxide
|
COR
|
Core Behaviour Package
|
CRD
|
Control Rod Drive Pumps
|
CSNI
|
Committee on the Safety of Nuclear Installations
|
CSS
|
Containment Spray System
|
CVH
|
Control Volume Hydrodynamics
|
DBA
|
Design Basis Assessment
|
DCH
|
Direct Containment Heating
|
ECCS
|
Emergency Core Cooling System
|
EOP
|
Emergency Operating Procedures
|
EPRI
|
Electric Power Research Institute
|
ERG
|
Emergency Response Guidelines
|
EUR
|
European Utility Requirements
|
FASTNET
|
Fast Nuclear Emergency Tools
|
FCVS
|
Filtered Containment Venting System
|
FHB
|
Fuel Handling Building
|
FL
|
Flow Paths
|
FMEA
|
Failure Mode Effect Analysis
|
FP
|
Fission Product
|
FPCS
|
Fuel Pool Cooling System
|
FWS
|
Fire Water System
|
GDC
|
General Design Criteria
|
HRA
|
Human Reliability Analysis
|
IAEA
|
International Atomic Energy Agency
|
IE
|
Initiating Event
|
ISTP
|
International Source Term Program
|
IVMR
|
In-Vessel Melt Retention
|
KNPP
|
Kozloduy Nuclear Power Plant
|
LERF
|
Large Early Release Frequency
|
LLOCA
|
Large-Break Loss Of Coolant Accident
|
LOCA
|
Loss Of Coolant Accidents
|
LOOP
|
Loss Of Offsite Power
|
LPP
|
Low Pressure Pump
|
LPSD
|
Low Power And Shutdown
|
LPSIS
|
Low-Pressure Safety Injection System
|
LTO
|
Long Term Operation
|
LWR
|
Light Water Reactors
|
MAAP
|
Modular Accident Analysis Program
|
MCCI
|
Molten Core Concrete Interaction
|
MCCI
|
Molten Corium Concrete Interaction
|
MCP
|
Main Coolant Pump
|
MCR
|
Main Control Room
|
MELCOR
|
Methods For Estimation Of Leakages And Consequences Of Releases
|
NCO
|
Niigataken-Chuetsu-Oki
|
NPP
|
Nuclear Power Plant
|
NRA
|
Nuclear Regulation Authority (Japan)
|
NRC
|
Nuclear Regulatory Commission
|
NUREG
|
Nuclear Regulatory Commission Regulation (USA)
|
OECD
|
Organisation For Economic Co-Operation And Development
|
OPEX
|
Operating Experience
|
PAR
|
Passive Autocatalytic Recombiner
|
PDS
|
Plant Damage State
|
PHWR
|
Pressurizer Heavy Water Reactors
|
PIRT
|
Phenomena Identification And Ranking Table
|
PORV
|
Power Operated Relief Valve
|
POS
|
Plant Operating State
|
PRV
|
Pressure Relief Valves
|
PSA
|
Probabilistic Safety Assessment
|
PWR
|
Pressurised Water Reactor
|
R&D
|
Research And Development
|
RASTEP
|
Rapid Source Term Prediction
|
RC
|
Release Categories
|
RCPB
|
Reactor Coolant Pressure Boundary
|
RCS
|
Reactor Coolant System
|
RHR
|
Residual Heat Removal
|
RHRS
|
Residual Heat Removal System
|
RN
|
Radionuclide Behaviour Package
|
RPV
|
Reactor Pressure Vessel
|
RUSET
|
Ruthenium Separate Effect Tests
|
RV
|
Reactor Vessel
|
RWST
|
Refuelling Water Storage Tank
|
SAFEST
|
Severe Accident Facilities For European Safety Targets
|
SAMG
|
Severe Accident Management Guideline
|
SAREF
|
Safety Research Opportunities Post-Fukushima
|
SARNET
|
Severe Accident Research Network
|
SASA
|
Severe Accident Sequence Analysis
|
SBO
|
Station Black Out
|
SC
|
Shutdown Cooling
|
SFD
|
Spent Fuel Damage
|
SFP
|
Spent Fuel Pool
|
SG
|
Steam Generators
|
SGTR
|
Steam Generator Tube Rupture
|
SNAP
|
Symbolic Nuclear Analysis Package
|
SPSA
|
Shutdown PSA
|
SRV
|
Safety Relief Valves
|
SSC
|
System Structure And Components
|
SSM
|
Swedish Radiation Safety Authority
|
ST
|
Source Term
|
STCS
|
Shutdown And Torus Cooling System
|
TBICWS
|
Turbine Building Intermediate Cooling Water
|
TCS
|
Torus Cooling System
|
TS
|
Technical Specification
|
VVER
|
Water Water Energetic Reactor (Russian Design)
|