Kooperativer Bibliotheksverbund

UID:

Umfang: 1 online resource (666 pages)

ISBN: 0-12-824077-6 , 9780128240762

Serie: JSME Series in Thermal and Nuclear Power Generation ; Volume 3

Inhalt: Sodium-cooled Fast Reactors is the third volume in the JSME Series on Thermal and Nuclear Power Generation, which presents a comprehensive view of the latest research and activities from around the globe. Volume Editors Masaki Morishita and Hiroyuki Ohshima, along with their team of expert contributors, combine their knowledge and experience to provide a solid understanding of the history of SFRs and work carried out in Japan to date. This book uniquely includes case studies from these global regions to highlight SFR uses, benefits and challenges, focusing on their safety, design, operation, and maintenance. Unique to this publication, the JSME cover key technological advances which will shape power generation of the future, including developments in the use of AI for design.

Anmerkung: Intro -- Sodium-cooled Fast Reactors -- Copyright -- Contents -- Contributors -- About the authors -- Preface of JSME Series in Thermal and Nuclear Power Generation -- Preface to Volume 3: Sodium-cooled fast reactors -- Chapter 1: Introduction -- 1.1. Sodium-cooled fast reactor -- 1.2. Functions of SFR cycle and technology -- 1.3. Needs for SFRs -- 1.4. FBR development policy -- 1.5. History of FBR development in Japan -- 1.6. Overview of this volume -- References -- Chapter 2: Experimental reactor Joyo -- 2.1. Introduction -- 2.1.1. Construction and operation history of Joyo -- 2.1.2. Plant description of Joyo -- 2.2. Operation and maintenance experience -- 2.2.1. Operation and maintenance -- 2.2.1.1. Core management -- 2.2.1.2. Demonstration of Pu fuel recycle -- 2.2.1.3. Chemical analysis of sodium and cover gas -- 2.2.1.4. Reliability of sodium components -- 2.2.1.5. In-service inspection of the coolant boundary -- 2.2.2. Demonstration of SFR technologies -- 2.2.2.1. Natural circulation test -- 2.2.2.2. Fuel failure simulation tests -- 2.2.2.3. Demonstration test of self-actuated shutdown system (SASS) with a Curie point electromagnet -- 2.2.3. Replacement experience of large components in the cooling system -- 2.2.4. Development of in-vessel repair techniques in sodium fast reactor [6,7] -- 2.2.4.1. UCS replacement -- 2.2.4.2. MARICO-2 test subassembly retrieval -- 2.3. Irradiation test experience and future plan -- 2.3.1. Irradiation test technologies of Joyo [8] -- 2.3.1.1. Irradiation equipment of Joyo -- 2.3.1.2. Online irradiation equipment -- 2.3.2. Postirradiation examination (PIE) technologies -- 2.3.2.1. PIE of fuel assembly -- 2.3.2.2. PIE of fuels -- 2.3.2.3. PIE of materials -- 2.3.3. Future plan -- 2.3.3.1. Neutron spectrum tailoring -- 2.3.3.2. Lower temperature irradiation -- 2.3.3.3. High-temperature irradiation technique. , References -- Chapter 3: Prototype reactor Monju -- 3.1. Design features -- 3.1.1. Plant overview -- 3.1.2. Reactor core -- 3.1.2.1. Core design -- 3.1.2.2. Fuel subassembly design -- 3.1.3. Reactor structure -- 3.1.3.1. Reactor vessel -- 3.1.3.2. Shield plug -- 3.1.3.3. Control rod drive mechanism -- 3.1.4. Cooling systems -- 3.1.4.1. PHTS circulation pump -- 3.1.4.2. Intermediate heat exchanger -- 3.1.4.3. Steam generator -- 3.1.5. Fuel handling systems -- 3.1.6. Instrumentations -- 3.1.6.1. Neutron instrumentation -- 3.1.6.2. Failed fuel detection system -- 3.1.6.3. Sodium purity control devices -- 3.2. RandD activities for Monju design -- 3.2.1. Design history -- 3.2.2. Sodium components -- 3.2.2.1. Reactor vessel -- 3.2.2.2. Shield plug -- 3.2.2.3. Control rod drive mechanism -- 3.2.2.4. Circulation pump -- 3.2.2.5. Intermediate heat exchanger -- 3.2.2.6. Steam generator -- 3.2.2.7. Sodium handling technologies -- 3.2.3. Development of elevated temperature structural design guide -- 3.2.3.1. Structural tests -- 3.2.3.2. Elevated temperature structural design guide -- 3.2.4. Reactor core and fuel -- 3.2.4.1. Research on MOX fuel subassembly -- 3.2.4.2. Development of PNC316 for fuel cladding material -- 3.2.4.3. Reactor physics -- 3.2.5. Safety -- 3.2.5.1. Sodium leak and fire -- 3.2.5.2. Sodium-water reaction -- 3.2.5.3. Fuel failure criterion -- 3.2.5.4. Core disruptive accident -- 3.3. Operation and maintenance -- 3.3.1. Operation -- 3.3.2. Maintenance -- 3.3.2.1. Maintenance program -- 3.3.2.2. Maintenance and repair experiences -- 3.3.3. Accidents and failures -- 3.3.3.1. Thermal displacement of SHTS piping -- 3.3.3.2. Pressure drop of water-steam system flash tank -- 3.3.3.3. Secondary sodium leak accident -- 3.3.3.4. In-vessel transfer machine dropping -- 3.4. RandD achievements in Monju -- 3.4.1. Achievements in commissioning. , 3.4.1.1. Major commissioning steps -- 3.4.1.2. Comprehensive system function tests -- 3.4.1.3. System startup test -- 3.4.1.4. Power operation -- 3.4.1.5. Resumed SST -- 3.4.2. Design validation through commissioning -- 3.4.2.1. Neutronic design validation -- 3.4.2.2. Thermal hydraulic design validation -- 3.4.2.3. Component design validation -- 3.4.2.4. Development of ISI technology -- 3.4.3. Safety evaluation -- 3.4.3.1. Safety margins -- 3.4.3.2. Probabilistic risk assessment -- 3.4.3.3. Seismic back-check -- 3.4.3.4. Safety improvement following the 1F Accident -- Reference -- Chapter 4: Demonstration and commercial plant design study -- 4.1. DFBR: Demonstration reactor project lead by utilities -- 4.1.1. Background of DFBR design study -- 4.1.2. Outline of DFBR design -- 4.1.2.1. Study to improve economic efficiency -- 4.1.2.2. Study to improve safety [4,5] -- 4.1.2.3. Influence of the Monju accident and further rationalization of the design -- 4.1.3. Feasibility study on practical application strategies -- 4.2. JSFR developed in the FaCT project -- 4.2.1. Design requirement on commercial concepts -- 4.2.1.1. Introduction -- 4.2.1.2. Development targets and design requirements -- 4.2.1.3. Design approach to meet design targets and requirements -- 4.2.2. Core -- 4.2.2.1. Basic concept -- 4.2.2.2. Reference core specification -- 4.2.2.3. Detailed core design highlights -- 4.2.2.4. Metal fuel core alternative -- 4.2.3. Safety design -- 4.2.3.1. Development goals -- 4.2.3.2. Safety design concept -- 4.2.3.3. Safety evaluations -- 4.2.4. Reactor cooling system -- 4.2.4.1. Overview -- 4.2.4.2. Reactor structure -- 4.2.4.3. Two-loop primary cooling system -- 4.2.4.4. Integrated IHX-pump -- 4.2.4.5. Secondary system -- 4.2.4.6. SG -- 4.2.4.7. Decay heat removal system -- 4.2.5. Balance of plant and reactor building -- 4.2.5.1. Fuel handling system. , 4.2.5.2. Reactor building layout -- 4.2.5.3. Steel plate-reinforced concrete structure -- 4.2.5.4. Seismic isolation design -- 4.2.6. In-service inspection and repair -- 4.2.6.1. ISI program -- 4.2.6.2. Repair program -- 4.2.6.3. Design accommodation -- 4.2.7. Selection of demonstration reactor specifications -- 4.2.7.1. Requirements on demonstration reactor -- 4.2.7.2. Reactor structure -- 4.2.7.3. Cooling system -- 4.2.7.4. Reactor building -- 4.2.7.5. Comparison of demonstration reactor output -- 4.3. Design improvement of JSFR -- 4.3.1. Update on requirements and conditions -- 4.3.1.1. Safety requirements -- 4.3.1.2. Maintenance requirements -- 4.3.1.3. Updates on design conditions -- 4.3.2. Safety improvement -- 4.3.2.1. DHRS -- 4.3.2.2. Reactor building -- 4.3.3. Reactor cooling system update -- 4.3.3.1. Reactor structure -- 4.3.3.2. Primary main piping -- 4.3.3.3. Integrated pump-IHX -- 4.3.3.4. Steam generator -- 4.3.3.5. Alternative design reducing RandD loads -- 4.4. Pool-type SFR -- 4.4.1. Design concept -- 4.4.2. Reactor structure under severe seismic conditions -- 4.4.2.1. Seismic loading -- 4.4.2.2. Structural characteristics of RVs -- 4.4.2.3. Three-dimensional seismic isolation system -- 4.4.2.4. Structural intactness based on thermal hydraulic analysis -- 4.4.3. Safety design -- 4.4.3.1. Measures for the safety design concept -- 4.4.3.2. Applicability evaluation of the SASS -- 4.4.3.3. Decay heat removal system -- References -- Chapter 5: Key technologies for future sodium-cooled fast reactors -- Nomenclature -- 5.1. Safety -- 5.1.1. Introduction -- 5.1.2. Development of SDC/SDG -- 5.1.2.1. Introduction -- 5.1.2.2. SDC -- 5.1.2.3. SDG on safety approach -- 5.1.2.4. SDG on SSCs -- 5.1.2.5. SFR safety design concepts in a manner consistent with SDC/SDGs -- 5.1.3. Self-actuated shutdown system (SASS) -- 5.1.3.1. Introduction. , 5.1.3.2. Development program -- 5.1.3.3. Development goals -- 5.1.3.4. Test results -- 5.1.4. Severe accident -- 5.1.4.1. Background and objective -- 5.1.4.2. The initiating phase -- 5.1.4.3. The transition phase -- 5.1.4.4. Material relocation phase and heat-removal phase -- 5.1.4.5. Concluding remarks -- 5.1.5. Sodium combustion -- 5.1.5.1. Chemical reaction of sodium combustion -- 5.1.5.2. Types of sodium combustion -- 5.1.5.3. Experimental and numerical research -- 5.1.6. Sodium-water reaction -- 5.1.6.1. Overview -- 5.1.6.2. Research and development in recent years -- 5.1.6.3. Summary -- 5.1.7. Source term -- 5.1.7.1. Overview -- 5.1.7.2. Experimental studies at JAEA -- 5.1.7.3. Analytical study at JAEA -- 5.1.7.4. Summary -- 5.2. Sodium component development -- 5.2.1. Introduction -- 5.2.2. Component development plan for JSFR -- 5.2.3. AtheNa facility specifications -- 5.2.4. GIF collaboration -- 5.2.5. Current status -- 5.3. Reactor core physics -- 5.3.1. Calculation codes and methods -- 5.3.1.1. Features of reactor core physics calculation for fast reactors -- 5.3.1.2. Modules and methods -- 5.3.1.3. Cross-section sensitivity analysis -- 5.3.1.4. Integrated code system -- 5.3.2. Experimental validation database -- 5.3.3. Data assimilation method -- 5.3.4. Future developments -- 5.4. Fuel and materials -- 5.4.1. Introduction -- 5.4.2. MOX fuel development -- 5.4.3. MOX fuel performance -- 5.4.3.1. Irradiation behavior -- 5.4.3.2. Fuel performance code -- 5.4.4. Core material development -- 5.4.4.1. Modified-type 316 stainless steel -- 5.4.4.2. Ferritic steel -- 5.4.4.3. Oxide dispersion-strengthened steel -- 5.5. Thermal-hydraulics -- 5.5.1. Plant dynamics thermal-hydraulics -- 5.5.1.1. Plant dynamics analysis -- 5.5.1.2. Multilevel simulation by 1D3D coupling analysis -- 5.5.1.3. Neutronics-related issues. , 5.5.2. Fuel subassembly thermal-hydraulics (FS-TH).

Weitere Ausg.: Print version: Morishita, Masaki Sodium-Cooled Fast Reactors San Diego : Elsevier Science & Technology,c2022

Sprache: Englisch