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Sodium-cooled fast reactors /

Sodium-cooled fast reactors is the third volume in the JSME series in thermal and nuclear power generation, which presents a comprehensive view of the latest research and activities from around the globe. Volume editors [ ...], along with their team of expert contributors, combine their knowledge an...

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Bibliographic Details
Corporate Author: ScienceDirect (Online service)
Other Authors: Morishita, Masaki (Editor), Ohshima, Hiroyuki, 1944- (Editor)
Format: eBook
Language:English
Published: London, United Kingdom ; San Diego, CA : Academic Press, [2022]
Series:JSME series in thermal and nuclear power generation ; v. 3.
Subjects:
Fast reactors.
Sodium cooled reactors.
Réacteurs rapides.
Réacteurs refroidis au sodium.
Fast reactors
Sodium cooled reactors
Electronic books.
Online Access:Connect to the full text of this electronic book
Table of Contents:
  • 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.
  • 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.