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Severe accidents in sodium fast reactors : safety study approach, prevention and mitigation by design /

Severe accidents in Sodium Fast Reactors: Safety Study Approach, Prevention and Mitigation by Design is a unique presentation of research work from the Sodium Fast Reactor Severe Accident team at CEA. Focusing on in-core and out-core severe accident phenomena, this book analyzes severe accident scen...

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Gardado en:
Detalles Bibliográficos
Outros autores: Bertrand, Frédéric (Editor), Bachrata, Andrea (Editor), Seiler, Nathalie (Editor), Journeau, Christophe (Editor)
Formato: Electrónico eBook
Idioma:inglés
Publicado: London : Academic Press, [2025]
Subjects:
Sodium cooled reactors > Accidents.
Nuclear reactor accidents.
Réacteurs refroidis au sodium > Accidents.
Réacteurs nucléaires > Accidents.
Nuclear reactor accidents
Sodium cooled reactors > Accidents
Acceso en liña:ScienceDirect
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Table of Contents:
  • Front Cover
  • Severe Accidents in Sodium Fast Reactors
  • Front matter
  • About the cover page
  • Severe Accidents in Sodium Fast Reactors
  • Copyright
  • Tribute
  • Contents
  • Contributors
  • About the book and chapter editors
  • About chapter editors
  • Acknowledgments
  • 1
  • Introduction
  • 1.1 Objectives of the book
  • 1.2 Mitigation of severe accidents, a key issue for SFR
  • 1.3 Presentation of book content
  • 1.4 Scope limits
  • 1.5 Presentation of editors and authors
  • Acronyms
  • References
  • 2
  • Dynamic and historical vision of severe accident approach in SFR former projects
  • 2.1 Introduction
  • 2.2 SFR features: interests and importance of severe accident issues
  • 2.2.1 Past and current rationales for development of fast neutron reactors
  • 2.2.2 Advantages and drawbacks of liquid sodium as reactor coolant
  • 2.2.3 Overview of the main design characteristics of sodium-cooled fast reactors
  • 2.2.3.1 Main design characteristics
  • 2.2.3.2 Main issues of severe accidents in sodium-cooled fast reactors
  • 2.3 General concepts on SFR safety approach
  • 2.3.1 Consideration of severe accident in the French projects of sodium-cooled fast reactors
  • 2.3.1.1 Evolution of the consideration of severe accidents
  • 2.3.1.2 Consideration of severe accident in SuperphEnix
  • 2.3.1.3 Consideration of severe accident for SPX2 and EFR projects
  • 2.3.1.4 Consideration of severe accidents for ASTRID project
  • 2.3.2 Consideration of severe accident in Japanese SFR projects
  • 2.4 Main physical features of an SFR related to severe accident
  • 2.4.1 Core behavior and neutron physics features
  • 2.4.1.1 Core reactivity stability around the nominal operating point
  • 2.4.1.2 Core reactivity effect occurring during a severe accident
  • 2.4.2 Core design features for core melting prevention
  • 2.4.3 SFR severe accident sequence general description.
  • 2.4.4 Issues related to sodium specificities
  • 2.4.4.1 Cooling capability (conduction, natural convection, etc.)
  • 2.4.4.2 Chemical reactivity
  • 2.4.4.3 Thermal loadings of structures due to sodium at high temperature
  • 2.4.4.4 Sodium two-phase behavior
  • 2.4.4.5 Molten fuel-coolant interaction
  • 2.4.4.6 Sodium impact on fission products release
  • 2.4.5 Postaccident cooling
  • 2.4.6 Main difference with LWR severe accident features
  • 2.4.6.1 Overview of LWR and SFR concepts
  • 2.4.6.2 Severe accident sequences
  • 2.4.6.3 In-vessel severe accident phenomena
  • 2.4.6.4 Severe accident management strategies
  • 2.5 Historical overview of SFR design approach regarding severe accidents
  • 2.5.1 Historical overview
  • 2.5.1.1 In France
  • 2.5.1.2 In Japan
  • 2.5.2 Core melting prevention
  • 2.5.2.1 French approach for severe accident prevention
  • 2.5.2.1.1 SuperphEnix
  • 2.5.2.1.2 SPX2 and EFR projects
  • 2.5.2.1.3 ASTRID project
  • 2.5.2.2 Core melting prevention in Japan
  • 2.5.3 Design against core melting and mechanical loadings
  • 2.5.3.1 Historical approach in France
  • 2.5.3.1.1 SuperphEnix design
  • 2.5.3.1.2 EFR design
  • 2.5.3.1.3 ASTRID design
  • 2.5.3.2 Historical approaches in Japan
  • 2.5.3.2.1 Joyo
  • 2.5.3.2.2 Monju
  • 2.5.3.2.3 DFBR
  • 2.5.3.2.4 JSFR
  • 2.5.4 Design for long-term cooling
  • 2.5.4.1 European projects
  • 2.5.4.1.1 In-vessel retention concept
  • 2.5.4.1.2 Management of the molten material relocation
  • 2.5.4.1.3 Molten materials and sodium interaction
  • 2.5.4.1.4 Impact of molten materials on the core catcher structure
  • 2.5.4.1.5 Core debris subcriticality
  • 2.5.4.1.6 Freezing of the molten materials on the core catcher
  • 2.5.4.1.7 Heat transfer from the core catcher to sodium
  • 2.5.4.1.8 Heat transfer toward the heat removing system
  • 2.5.4.2 Historical approach in Japan.
  • 2.5.5 Design mitigating fission product release
  • 2.5.5.1 Containment system in the French reactor
  • 2.5.5.1.1 SuperphEnix design
  • 2.5.5.1.2 French and European project after SuperphEnix
  • 2.5.5.2 Containment design: Historical approach in Japan
  • Glossary
  • Acronyms
  • Appendix
  • Appendix 2.3.1: Line-of-defense method
  • Appendix 2.3.2: Top-down approach for mitigation of severe accident
  • Appendix 2.3.3: Practical-elimination concept
  • Appendix 2.3.4: Lessons learned from the Fukushima-Daiichi accident for the ASTRID design
  • References
  • 3
  • Overview of ASTRID and JSFR design
  • 3.1 Overview of ASTRID design
  • 3.1.1 General objectives and features of ASTRID, innovative design options
  • 3.1.1.1 ASTRID objectives
  • 3.1.1.2 ASTRID main features
  • 3.1.1.3 Overview of innovative options
  • 3.1.2 Core design (including DCS-P) and DCS-M-TT design
  • 3.1.2.1 Dedicated SA designs regarding severe accident prevention and mitigation
  • 3.1.2.1.1 Complementary safety devices for severe accident prevention (DCS-P)
  • 3.1.2.1.1.1 Hydraulically triggered complementary safety device
  • 3.1.2.1.1.2 Thermally triggered complementary safety device
  • 3.1.2.1.2 Complementary safety devices for severe accident mitigation (DCS-M-TT)
  • 3.1.3 Core catcher design
  • 3.1.4 Reactor design
  • 3.1.4.1 Primary system
  • 3.1.4.2 Intermediate loops and sodium gas heat exchangers
  • 3.1.4.3 Power conversion system design
  • 3.1.4.4 Decay heat removal systems
  • 3.1.4.5 Reactor pit and above roof area
  • 3.1.4.6 Cover gas system and intervessel gas system
  • 3.1.4.7 Containment and reactor building
  • 3.2 Overview of JSFR design
  • 3.2.1 General objectives and features of JSFR
  • 3.2.2 Safety design
  • 3.2.3 Core design (including FAIDUS and CRGT)
  • 3.2.4 Reactor design, including double-wall tube steam generators
  • 3.2.4.1 Reactor structure.
  • 3.2.4.2 Reactor coolant system
  • 3.2.4.3 Pump-integrated intermediate heat eXchanger
  • 3.2.4.4 Steam generator (SG) (Kanda et al., 2016
  • Futagami et al., 2009)
  • 3.2.5 Containment and reactor building
  • 3.2.5.1 Containment structure
  • 3.2.5.2 Seismic isolation system
  • Acronyms
  • References
  • 4
  • Severe accident scenario consideration and study methodology for mitigation implementation by design proposed b ...
  • 4.1 Approach for mitigation of severe accidents for ASTRID
  • 4.1.1 Elements on general safety approach
  • 4.1.2 Consideration on severe accidents at the preconceptual design stage
  • 4.2 Study methodology supporting the prevention and mitigation approach
  • 4.2.1 Introduction
  • 4.2.2 Core melting scenarios
  • 4.2.2.1 Core melting initiating sequences
  • 4.2.2.2 General description of ULOF sequences
  • 4.2.2.3 General description of UTOP sequences
  • 4.2.2.4 USAF
  • 4.2.3 Objectives and methodology of the studies
  • 4.2.3.1 Study objectives
  • 4.2.3.2 Natural behavior of the core
  • 4.2.3.3 Study of degraded state of the reactor
  • 4.2.3.4 Scenario studies from the initiating event
  • 4.2.3.5 Feedback on design
  • Glossary
  • Acronyms
  • References
  • 5
  • Severe accident calculation tools used for severe accident and mitigation studies
  • 5.1 Introduction
  • 5.2 Steady-state mechanistic tools
  • 5.2.1 Neutron physics simulation tools
  • 5.2.1.1 ERANOS tool
  • 5.2.1.1.1 Modeling of the core in the nominal state
  • 5.2.1.1.2 Validation of ERANOS
  • 5.2.1.2 TRIPOLI-4 tool
  • 5.2.1.2.1 Description of TRIPOLI-4
  • 5.2.1.2.2 Validation of TRIPOLI-4
  • 5.2.2 Thermodynamic equilibrium calculation tools
  • 5.3 Mechanistic tools for severe accident transient simulation
  • 5.3.1 Presentation of the historical computation scheme for calculating a severe accident transient in a SFR
  • 5.3.2 SAS4A
  • 5.3.3 SIMMER.
  • 5.3.3.1 Presentation of the code
  • 5.3.3.2 Description of main features of SIMMER
  • 5.3.3.2.1 Reactor representation
  • 5.3.3.2.2 Coupling between physical processes
  • 5.3.3.2.3 Fluid dynamics model
  • 5.3.3.2.4 Structure model
  • 5.3.3.2.5 Neutronics model
  • 5.3.3.2.6 Other models
  • 5.3.3.3 Validation of SIMMER
  • 5.3.3.4 Continuing validation of SIMMER-III and future validation of SIMMER V
  • 5.3.4 EUROPLEXUS
  • 5.3.4.1 Introduction
  • 5.3.4.2 The ADCR model
  • 5.3.4.3 Verification and validation test cases
  • 5.3.5 CONTAIN-LMR
  • 5.3.5.1 Validation studies
  • 5.3.6 Lessons learned from simulation needs for ASTRID studies
  • 5.4 Parametrable tools for severe accident simulations
  • 5.4.1 General overview and objectives
  • 5.4.2 Tools for the primary phase (BETINa, MACARENa, SUREX, and OCARINa)
  • 5.4.2.1 BETINa: The total instantaneous blockage accidental sequence
  • 5.4.2.2 MACARENa: Unprotected loss of flow sequence
  • 5.4.2.3 OCARINa: Unprotected reactivity insertion sequence
  • 5.4.2.4 SUREX: reactivity evolution and fuel dispersion
  • 5.4.3 Tools devoted to transition phase: The MARINa tool
  • 5.4.4 Tools devoted to expansion phase: The DETONa tool
  • 5.4.5 Tools devoted to postaccident cooling phase (LIDENa, Super-COPD)
  • 5.4.6 Chaining of parametric tools and with statistical methods
  • 5.4.6.1 Chaining of parametric tools
  • 5.4.6.2 Coupling with advanced statistical methods
  • 5.5 R&amp
  • D prospects for future tool platform
  • Acronyms
  • Appendix
  • A5.1 R&amp
  • D prospects for future tool platform
  • A5.1.1 PROCOR_Na
  • A5.1.2 SEASON
  • A.5.2 Validation table matrices of SIMMER
  • References
  • 6
  • Mitigation studies carried out for the French SFR demonstrator
  • 6.1 Introduction
  • 6.2 Range of accident consequences, natural behavior of the core
  • 6.2.1 Introduction
  • 6.2.2 Core safety design criteria.

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