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...
Shranjeno v:
| Drugi avtorji: | , , , |
|---|---|
| Format: | Elektronski eKnjiga |
| Jezik: | angleščina |
| Izdano: |
London :
Academic Press,
[2025]
|
| Teme: | |
| Online dostop: | ScienceDirect |
| Oznake: |
Označite
Brez oznak, prvi označite!
|
MARC
| LEADER | 00000cam a2200000 i 4500 | ||
|---|---|---|---|
| 001 | on1477810438 | ||
| 003 | OCoLC | ||
| 005 | 20260915095222.0 | ||
| 006 | m o d | ||
| 007 | cr cnu---unuuu | ||
| 008 | 241209s2025 enk o 000 0 eng d | ||
| 040 | |a OPELS |b eng |e rda |e pn |c OPELS |d OCLCO |d OCLCL |d SFB |d OCLCF |d OCLCO | ||
| 020 | |z 9780323954358 | ||
| 035 | |a (OCoLC)1477810438 | ||
| 050 | 4 | |a TK9203.L5 | |
| 082 | 0 | 4 | |a 621.48/35 |2 23/eng/20241209 |1 https://id.oclc.org/worldcat/ddc/E4bQRtd8BdjdTKD99dMmxh4Jgp |
| 245 | 0 | 0 | |a Severe accidents in sodium fast reactors : |b safety study approach, prevention and mitigation by design / |c edited by Frédéric Bertrand [and 3 more]. |
| 264 | 1 | |a London : |b Academic Press, |c [2025] | |
| 300 | |a 1 online resource. | ||
| 336 | |a text |b txt |2 rdacontent | ||
| 337 | |a computer |b c |2 rdamedia | ||
| 338 | |a online resource |b cr |2 rdacarrier | ||
| 520 | |a 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 scenarios and mitigation, related calculations and tools, and key outcomes from important experimental programs. The book describes the state-of-the-art on severe accident research which will help to inform and direct further pre-conceptual Generation IV SFRs. Lessons learned from the pre-conceptual phase of the French ASTRID project are highlighted alongside different approaches to demonstrate robustness. To support the demonstration of concepts explored, a special section dedicated to calculation tools and their validation offers readers a unique set of tools to guide and inform their own research work. Readers will gain a deep understanding of methodologies to illustrate approaches as well as an overview of safety approaches from key countries based on the editor's rich international collaboration. | ||
| 588 | 0 | |a Online resource; title from PDF title page (ScienceDirect, viewed December 9, 2024). | |
| 505 | 0 | |a 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. | |
| 505 | 8 | |a 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. | |
| 505 | 8 | |a 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. | |
| 505 | 8 | |a 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. | |
| 505 | 8 | |a 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& -- D prospects for future tool platform -- Acronyms -- Appendix -- A5.1 R& -- 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. | |
| 650 | 0 | |a Sodium cooled reactors |x Accidents. | |
| 650 | 0 | |a Nuclear reactor accidents. | |
| 650 | 6 | |a Réacteurs refroidis au sodium |0 (CaQQLa)201-0233321 |x Accidents. |0 (CaQQLa)201-0380249 | |
| 650 | 6 | |a Réacteurs nucléaires |x Accidents. |0 (CaQQLa)201-0315983 | |
| 650 | 7 | |a Nuclear reactor accidents |2 fast |0 (OCoLC)fst01040715 | |
| 650 | 7 | |a Sodium cooled reactors |x Accidents |2 fast |0 (OCoLC)fst01124057 | |
| 700 | 1 | |a Bertrand, Frédéric, |e editor. |1 https://id.oclc.org/worldcat/entity/E39PCjxjdXQ6QK9gd88C4M6vgX | |
| 700 | 1 | |a Bachrata, Andrea, |e editor. | |
| 700 | 1 | |a Seiler, Nathalie, |e editor. | |
| 700 | 1 | |a Journeau, Christophe, |e editor. | |
| 856 | 4 | 0 | |3 ScienceDirect |u https://www.sciencedirect.com/science/book/9780323954358 |