Structural Materials for Generation IV Nuclear Reactors

 
 
Woodhead Publishing
  • 1. Auflage
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  • erschienen am 27. August 2016
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  • 684 Seiten
 
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978-0-08-100912-3 (ISBN)
 

Operating at a high level of fuel efficiency, safety, proliferation-resistance, sustainability and cost, generation IV nuclear reactors promise enhanced features to an energy resource which is already seen as an outstanding source of reliable base load power. The performance and reliability of materials when subjected to the higher neutron doses and extremely corrosive higher temperature environments that will be found in generation IV nuclear reactors are essential areas of study, as key considerations for the successful development of generation IV reactors are suitable structural materials for both in-core and out-of-core applications. Structural Materials for Generation IV Nuclear Reactors explores the current state-of-the art in these areas. Part One reviews the materials, requirements and challenges in generation IV systems. Part Two presents the core materials with chapters on irradiation resistant austenitic steels, ODS/FM steels and refractory metals amongst others. Part Three looks at out-of-core materials.

Structural Materials for Generation IV Nuclear Reactors is an essential reference text for professional scientists, engineers and postgraduate researchers involved in the development of generation IV nuclear reactors.


  • Introduces the higher neutron doses and extremely corrosive higher temperature environments that will be found in generation IV nuclear reactors and implications for structural materials
  • Contains chapters on the key core and out-of-core materials, from steels to advanced micro-laminates
  • Written by an expert in that particular area
  • Englisch
  • Cambridge
Elsevier Science
  • 24,64 MB
978-0-08-100912-3 (9780081009123)
0081009127 (0081009127)
weitere Ausgaben werden ermittelt
  • Front Cover
  • Structural Materials for Generation IV Nuclear Reactors
  • Related titles
  • Structural Materials for Generation IV Nuclear Reactors
  • Copyright
  • Contents
  • List of contributors
  • Woodhead Publishing Series in Energy
  • Introduction
  • 1 - Introduction to Generation IV nuclear reactors
  • 1.1 Introduction: the need for new nuclear systems
  • 1.2 Generation IV requirements and technical challenges
  • 1.2.1 Development of sustainable nuclear energy
  • 1.2.2 Maintaining or increasing competitiveness
  • 1.2.3 Improving and enhancing safety and reliability
  • 1.2.4 Ensuring proliferation resistance and physical protection
  • 1.3 Generation IV systems fulfilling these requirements
  • 1.3.1 Gas-cooled systems
  • 1.3.2 Liquid metal-cooled systems
  • 1.3.3 Molten salt
  • 1.3.4 Water-cooled systems
  • 1.4 Conclusion
  • References
  • 2 - Corrosion phenomena induced by liquid metals in Generation IV reactors
  • 2.1 Introduction to the liquid metals selected for Generation IV reactors
  • 2.2 Thermal, physical, and chemical properties of the liquid metals
  • 2.2.1 Solubility limits in liquid metal
  • 2.2.1.1 Solubility of metallic and nonmetallic elements in liquid Na
  • 2.2.1.2 Solubility of metallic and nonmetallic elements in liquid Pb and Pb-Bi
  • 2.3 The impact of structural material corrosion on reactor operation
  • 2.4 Parameters affecting corrosion in the liquid metal and experimental procedures
  • 2.4.1 Dissolution process
  • 2.4.2 Oxidation process
  • 2.4.3 Influence of temperature
  • 2.4.4 Influence of flow velocity
  • 2.4.5 Influence of carbon
  • 2.5 Corrosion under reactor conditions: mass transfer, experimental data, and modeling
  • 2.6 Impact of corrosion on mechanical strength of the structural material
  • 2.6.1 Impact of Na on mechanical properties of reference structural materials
  • 2.6.1.1 Effect of carbon in Na
  • 2.6.1.2 Effect of oxygen in Na
  • 2.6.1.3 Na and neutron irradiation synergetic effects
  • 2.6.1.4 Impact on component design
  • 2.6.2 Impact of Pb and Pb-Bi on mechanical properties of reference structural materials
  • 2.6.2.1 Tensile properties in Pb and Pb-Bi
  • 2.6.2.2 Creep and creep-to-rupture properties
  • 2.6.2.3 Low cycle fatigue properties
  • 2.6.2.4 Pb/Pb-Bi and irradiation fields
  • 2.6.2.5 Impact on component design
  • 2.7 Corrosion mitigation
  • 2.7.1 Corrosion mitigation for SFR
  • 2.7.2 Corrosion mitigation for LFR
  • 2.8 Conclusions
  • References
  • 3 - Corrosion phenomena induced by gases in Generation IV nuclear reactors
  • 3.1 Corrosion of IHX alloys in impure helium of a VHTR system
  • 3.1.1 VHTR atmosphere
  • 3.1.2 Chemical reactivity between metallic surfaces and VHTR helium at high temperature
  • 3.1.3 Rapid carburization/decarburization of alloys in improper helium
  • 3.1.4 Long-term evolution of alloys in chemically controlled oxidizing helium
  • 3.2 Corrosion phenomena in supercritical CO2
  • 3.2.1 Mild steels
  • 3.2.2 9-12Cr ferritic-martensitic steels
  • 3.2.3 Austenitic steels and nickel-base alloys
  • 3.3 Concluding remarks
  • References
  • 4 - Corrosion phenomena induced by supercritical water in Generation IV nuclear reactors
  • 4.1 Introduction
  • 4.1.1 Historical perspective
  • 4.1.2 SCWR materials requirements
  • 4.1.3 Corrosion allowance
  • 4.1.4 Environmentally assisted cracking
  • 4.2 What is supercritical water?
  • 4.2.1 Corrosion product and impurity transport
  • 4.2.2 Water radiolysis
  • 4.2.3 SCW density
  • 4.3 Test methodologies
  • 4.4 General corrosion in SCW
  • 4.4.1 Effects of key parameters
  • 4.4.2 Reproducibility
  • 4.4.3 Mechanisms and modeling
  • 4.4.3.1 Empirical and phenomenological models
  • 4.4.3.2 Deterministic models
  • 4.5 Environmentally assisted cracking
  • 4.5.1 Effects of key variables
  • 4.5.1.1 Environmental factors
  • 4.5.1.2 Material factors
  • 4.5.1.3 Mechanical factors
  • 4.5.1.4 Irradiation factors
  • 4.5.2 Mechanisms and modeling
  • 4.6 Summary
  • References
  • 5 - Corrosion phenomena induced by molten salts in Generation IV nuclear reactors
  • 5.1 Introduction: molten salts in Generation IV nuclear reactors
  • 5.2 Requirements and molten salt mixtures available
  • 5.3 Corrosion processes in molten salts
  • 5.4 Salt chemistry control
  • 5.4.1 Salt purification
  • 5.4.2 Determination of redox potentials in corrosion studies
  • 5.4.2.1 Dynamic beryllium reference electrode
  • 5.4.2.2 Electroreduction of U(IV) to U(III)
  • 5.5 Materials and corrosion data for different reactor systems and components
  • 5.5.1 Metallic materials in the fuel salt
  • 5.5.2 Metallic materials in coolant salts
  • 5.6 Conclusion
  • References
  • 6 - Mechanical behavior of structural materials for Generation IV reactors
  • 6.1 Introduction
  • 6.2 Mechanical properties of F-M steels
  • 6.2.1 Mechanical strength at high amplitudes
  • 6.2.2 Microstructure of tempered martensitic steels
  • 6.2.3 Microstructural changes during strain at high temperature
  • 6.3 Analysis of the macroscopic behavior of martensitic steels for low loads
  • 6.3.1 Cyclic strain
  • 6.3.2 Fatigue-relaxation and fatigue-creep
  • 6.3.3 Creep
  • 6.4 Microstructural changes during the strain of martensitic steels at low loads
  • 6.4.1 Cyclic strain
  • 6.4.2 Creep
  • 6.5 Elements of a martensitic steel softening model
  • 6.5.1 Analytic simulation of cyclic softening
  • 6.5.2 Creep simulations
  • 6.5.3 Polycrystalline modeling
  • 6.6 Damage and fracture in fatigue and creep
  • 6.6.1 Pure fatigue
  • 6.6.2 Fatigue-relaxation and fatigue-creep
  • 6.6.3 Damage and fracture in creep
  • 6.6.3.1 Experimental data
  • 6.6.3.2 Necking: observations and predictions
  • 6.6.3.3 Long-term creep and intergranular damage
  • 6.6.4 Further recommended work in tempered martensite-ferritic steels
  • 6.7 Recent progresses concerning long-term creep and fatigue behavior of austenitic stainless steels
  • 6.7.1 Microstructure
  • 6.7.2 Short-term and long-term creep lifetime predictions
  • 6.7.2.1 State of art
  • 6.7.2.2 Necking simulation
  • 6.7.2.3. Intergranular damage prediction
  • 6.7.2.4 Very long-term creep failure of other austenitic stainless steels
  • 6.7.3 Low-stress regime of creep strain rate
  • 6.7.3.1 Change of slope in the Norton diagram
  • 6.7.3.2 Larson-Miller approach
  • 6.7.4 Advanced austenitic stainless steels
  • 6.7.5 Comparison with tempered martensite-ferritic steels
  • 6.7.6 Pure fatigue and fatigue-relaxation properties
  • 6.8 Conclusions and recommended further work
  • Acknowledgments
  • References
  • 7 - Irradiation effects in Generation IV nuclear reactor materials
  • 7.1 Introduction
  • 7.2 Radiation damage process
  • 7.2.1 Irradiation-induced point and line defects in steels
  • 7.2.2 Volume defects in steels
  • 7.2.3 Radiation-induced segregation in steels
  • 7.2.4 Irradiation-induced precipitation
  • 7.2.5 Radiation-induced amorphization
  • 7.3 Advances in characterization of defects in irradiated materials
  • 7.4 Mesoscale modeling of radiation damage
  • 7.4.1 Cluster dynamics modeling of void and dislocation loop growth
  • 7.4.2 Kinetic Monte Carlo modeling of phase precipitation in alloys
  • 7.4.3 Phase field modeling of patterned structure formation under irradiation
  • 7.4.4 Rate theory modeling of irradiation-induced segregation in alloys
  • 7.5 Summary
  • Acknowledgments
  • References
  • 8 - Irradiation-resistant austenitic steels as core materials for Generation IV nuclear reactors
  • 8.1 Introduction
  • 8.2 Austenitic steels and Generation IV systems
  • 8.2.1 Overview of the austenitic core structures of interest in Generation IV systems
  • 8.2.2 Requirements, design rules, and durability challenges for the associated austenitic materials: the example of SFR core stru ...
  • 8.2.3 The different irradiation-resistant austenitic grades studied in the different national programs of materials development f ...
  • 8.3 Out-of-pile characteristics of reference austenitic steels
  • 8.3.1 Physical and mechanical properties
  • 8.3.2 Aging and microstructural studies
  • 8.3.3 Corrosion properties and resistance to the process environment
  • 8.3.3.1 MOX-clad chemical interaction
  • 8.3.3.2 Reprocessing capabilities
  • 8.4 In-pile and postirradiation mechanical properties of reference austenitic steels
  • 8.4.1 Survey of the bibliography on the steels belonging to the 300-series
  • 8.4.2 Data on 15/15Ti and derivatives
  • 8.5 Swelling and irradiation creep properties of reference austenitic steels
  • 8.5.1 General insight about swelling, irradiation creep phenomena
  • 8.5.2 Influence of the irradiation parameters
  • 8.5.2.1 Basic parameters: temperature, dose, defect production rate
  • 8.5.2.2 Other irradiation parameters: temperature and swelling gradients, their consequence on the differences of behavior between ...
  • 8.5.3 Influence of the metallurgical parameters
  • 8.5.3.1 The nature of the basic matrix and the role of the major elements
  • 8.5.3.2 Role of main additive elements known to increase the swelling resistance
  • Effects of main swelling inhibitors at low temperature
  • Generalization of the effect of main additive elements to the entire irradiation temperature range
  • Special role of stabilizing elements
  • Advantage of a multistabilization in the presence of phosphorus
  • 8.5.3.3 Influence of the other specification elements
  • Elements V, Co, Sn, Sb, Ge
  • Mo and Mn
  • N and B
  • 8.5.3.4 Influence of the final metallurgical state
  • The importance of the quality of last annealing treatment
  • The importance of the value of final cold working
  • 8.6 Development of advanced austenitic materials designed to increase the in-pile duration of core structures of Generation IV ...
  • 8.6.1 Overview of the programs developed in national laboratories
  • 8.6.1.1 British research and development
  • 8.6.1.2 American research and development
  • 8.6.1.3 Japanese research and development
  • 8.6.1.4 The initial European cooperation programs
  • 8.6.2 The present CEA program of development of an advanced austenitic steel
  • 8.6.2.1 Behavior of samples irradiated in capsule
  • 8.6.2.2 Behavior of fuel pins
  • Global behavior
  • 8.6.3 Main intermediate conclusions
  • 8.7 Conclusion
  • Glossary
  • References
  • 9 - Irradiation-resistant ferritic and martensitic steels as core materials for Generation IV nuclear reactors
  • 9.1 Introduction
  • 9.2 Use of ferritic-martensitic steels in fast reactors and future Generation IV reactors
  • 9.3 Irradiation effects in ferritic-martensitic steels
  • 9.3.1 Microstructural evolution
  • 9.3.1.1 Irradiation-induced dislocation microstructures
  • 9.3.1.2 Swelling behavior up to high doses
  • 9.3.1.3 Radiation-induced segregation, precipitation under irradiation
  • Radiation-induced segregation
  • Precipitation under irradiation
  • 9.3.2 Evolution of mechanical properties
  • 9.3.2.1 Irradiation creep
  • 9.3.2.2 Irradiation-induced hardening, effect on tensile behavior
  • 9.3.2.3 Irradiation-induced embrittlement, modifications of impact, and fracture toughness properties
  • 9.4 Advanced ferritic-martensitic steels with improved thermal creep resistance
  • 9.5 Summary
  • Abbreviations
  • References
  • 10 - Oxide dispersion-strengthened/ferrite-martensite steels as core materials for Generation IV nuclear reactors
  • 10.1 Introduction
  • 10.2 Nanosized oxide particle control
  • 10.2.1 Dissociation and precipitation
  • 10.2.2 Precipitation coherency
  • 10.3 Development of oxide dispersion-strengthened steels in Japan
  • 10.3.1 Martensitic (9, 11)Cr oxide dispersion-strengthened steels
  • 10.3.2 Ferritic (12-15) Cr oxide dispersion-strengthened steels
  • 10.3.3 Manufacturing
  • 10.3.4 Tensile, creep strength, and Charpy properties
  • 10.4 Development of oxide dispersion-strengthened steels in France
  • 10.4.1 Chemical composition and microstructure
  • 10.4.2 Mechanical properties of ferritic oxide dispersion-strengthened alloys
  • 10.4.3 Fabrication route of oxide dispersion-strengthened tubes
  • 10.4.3.1 Martensitic oxide dispersion-strengthened steels
  • 10.4.3.2 Ferritic oxide dispersion-strengthened steels
  • 10.5 Development of other oxide dispersion-strengthened steels
  • 10.6 Joining
  • 10.6.1 Joining in Japan [69,70]
  • 10.6.2 Joining in France
  • 10.7 Environmental compatibility
  • 10.7.1 Corrosion in a sodium environment [72,73].
  • 10.7.2 Compatibility with Pb and Pb-Bi
  • 10.7.3 Cladding internal corrosion
  • 10.7.4 Cladding behavior during fuel processing-recycling
  • 10.8 Irradiation
  • 10.8.1 Joyo and BOR-60 irradiation [82,83,90-94]
  • 10.8.2 Phenix irradiation
  • 10.8.3 Simulated irradiation by charged particles
  • 10.9 Conclusion
  • References
  • 11 - Refractory metals as core materials for Generation IV nuclear reactors
  • 11.1 Refractory metals for nuclear application
  • 11.2 V and its alloys
  • 11.2.1 Introduction
  • 11.2.2 Fabrication, joining technology, and fundamental properties
  • 11.2.3 Corrosion and compatibility of V-alloys in various coolants
  • 11.2.4 Irradiation effects
  • 11.2.5 Potential improvements of the properties
  • 11.2.6 Summary of V-alloys
  • 11.3 Nb, Ta, Mo, W, and their alloys
  • 11.3.1 Introduction
  • 11.3.2 Alloy production, fabrication, and welding
  • 11.3.3 High-temperature mechanical properties
  • 11.3.4 Compatibility issues
  • 11.3.5 Radiation effects
  • 11.3.6 Advanced alloys
  • 11.4 Summary
  • References
  • 12 - SiCf/SiC composites as core materials for Generation IV nuclear reactors
  • 12.1 Introduction
  • 12.2 Potential use in Generation IV systems
  • 12.2.1 Very-high-temperature gas-cooled reactors
  • 12.2.2 Gas-cooled fast reactors
  • 12.2.3 Molten salt-cooled reactors
  • 12.2.4 Sodium-cooled fast reactors
  • 12.2.5 Lead-cooled fast reactors
  • 12.3 Fabrication and role of each constituent of the SiCf/SiC composite and matrix filling technologies
  • 12.3.1 Fiber
  • 12.3.2 Interphase
  • 12.3.3 Matrix and matrix-filling technologies
  • 12.3.3.1 Chemical vapor infiltration
  • 12.3.3.2 Liquid silicon infiltration
  • 12.3.3.3 Polymer impregnation/pyrolysis
  • 12.3.3.4 Slurry Impregnation and Hot Press
  • 12.3.4 Joining
  • 12.4 Behavior of the SiCf/SiC composite in operating conditions
  • 12.5 Codes and standards
  • 12.6 Summary
  • Acknowledgments
  • References
  • 13 - Carbon/carbon materials for Generation IV nuclear reactors
  • 13.1 Introduction
  • 13.2 Potential use in Generation IV systems
  • 13.3 Fabrication and role of each constituent of C/C composites and matrix filling technologies
  • 13.3.1 Classification of C/C composites
  • 13.3.2 Carbon fibers
  • 13.3.3 Reinforcement structures
  • 13.3.4 Matrix densification and posttreatment
  • 13.4 Behavior of C/C in operating conditions
  • 13.4.1 Behavior of graphite materials
  • 13.4.2 Behavior of C/C
  • 13.4.2.1 Behavior of the fibers and architectures
  • 13.4.2.2 Behavior of the matrices
  • 13.4.2.3 Behavior of the composites
  • 13.5 Standards and codes
  • 13.5.1 Standards
  • 13.5.2 Codes
  • 13.6 Conclusions
  • References
  • 14 - Graphite as a core material for Generation IV nuclear reactors
  • 14.1 Introduction
  • 14.2 Nuclear graphite grades, their manufacture, microstructure, and properties
  • 14.3 Nuclear graphite irradiation-induced dimensional and property changes
  • 14.3.1 Irradiation fluence units
  • 14.3.2 Irradiated material property data
  • 14.3.3 Dimensional change
  • 14.3.4 Coefficient of thermal expansion
  • 14.3.5 Thermal conductivity
  • 14.3.6 Young's modulus
  • 14.3.7 Strength
  • 14.3.8 Irradiation creep
  • 14.3.9 Influence of strain on the coefficient of thermal expansion
  • 14.4 Component structural integrity
  • 14.5 Thermal oxidation in fault conditions
  • 14.6 Dealing with irradiated graphite waste
  • 14.7 Advances in the treatment of graphite and carbowastes
  • 14.8 Molten salt reactors-graphite
  • 14.9 Discussion and conclusions
  • References
  • 15 - Absorber materials for Generation IV reactors
  • 15.1 Introduction: neutron absorbers for Generation IV reactors
  • 15.1.1 The Generation IV project: a short presentation
  • 15.1.2 Neutron absorbers in generation II-III and prototypic reactors
  • 15.1.3 Present reflections and developments status
  • 15.1.4 Materials resources and needs
  • 15.2 Scaling the neutron absorbers
  • 15.2.1 Nuclear properties
  • 15.2.1.1 Microscopic absorption cross-sections of the elements
  • 15.2.1.2 Macroscopic cross-sections of materials
  • 15.2.1.3 Neutron absorption products
  • 15.2.2 Materials properties
  • 15.2.2.1 Boron carbide
  • 15.2.2.2 Ag-In-Cd
  • 15.2.2.3 Hafnium
  • 15.3 Behavior under irradiation of neutron absorber materials
  • 15.3.1 Boron carbide
  • 15.3.1.1 Thermal water reactors
  • 15.3.1.2 Fast neutron reactors
  • 15.3.2 Ag-In-Cd alloy
  • 15.3.3 Hafnium
  • 15.3.4 Other materials
  • 15.3.4.1 Dysprosium titanate
  • 15.3.4.2 Hafnium compounds
  • Hafnium hydride
  • Hafnium dioxide (hafnia)
  • 15.3.4.3 Transition metal diborides
  • 15.3.4.4 Composites materials
  • 15.4 Conclusion: for a better definition of the needs
  • Abbreviations
  • References
  • 16 - Advanced irradiation-resistant materials for Generation IV nuclear reactors
  • 16.1 Introduction
  • 16.2 Identification of potential advanced irradiation-resistant materials
  • 16.2.1 Scientific bases for irradiation resistance
  • 16.2.1.1 High point defect sink strength
  • 16.2.1.2 Low vacancy mobility
  • 16.2.1.3 Radiation-resistant matrix phase
  • 16.2.2 Candidate advanced irradiation-resistant materials
  • 16.3 Basic properties
  • 16.3.1 Next-generation steels
  • 16.3.2 High-entropy alloys
  • 16.3.3 Multilayer metallic nanocomposites
  • 16.3.4 Ceramic composites
  • 16.3.5 MAX phase ceramics
  • 16.3.6 Bulk metallic glasses
  • 16.4 Fabrication and joining
  • 16.5 Experimental feedback and possible applications
  • 16.6 Future trends and conclusions
  • Acknowledgment
  • References
  • 17 - Conventional austenitic steels as out-of-core materials for Generation IV nuclear reactors
  • 17.1 Introduction
  • 17.2 General overview of austenitic steels in Generation IV frame
  • 17.2.1 AISI 300 series: type 304 and 316 grades
  • 17.2.2 Alloy 800 series
  • 17.3 Choice of austenitic steel grades for future French SFR out-of-core components
  • 17.4 Basic physical, thermal, and mechanical properties
  • 17.5 Fabrication and joining
  • 17.5.1 Product forms
  • 17.5.2 Processing and thermal-mechanical treatments
  • 17.5.3 Joining techniques
  • 17.6 Long-term mechanical behavior in operating conditions
  • 17.6.1 Creep resistance
  • 17.6.2 Creep-fatigue resistance
  • 17.6.3 Thermal aging
  • 17.6.3.1 Microstructural evolution
  • 17.6.4 Mechanical properties degradation
  • 17.7 Corrosion and oxidation behavior
  • 17.7.1 Sodium compatibility, wastage
  • 17.7.2 Lead and lead-bismuth eutectic compatibility
  • 17.7.3 Other systems
  • 17.8 Low-dose irradiation
  • 17.9 Codes and standards
  • 17.10 New alloy development
  • 17.11 Conclusions
  • Glossary
  • References
  • 18 - Conventional ferritic and martensitic steels as out-of-core materials for Generation IV nuclear reactors
  • 18.1 Introduction-attractive characteristics for Generation IV nuclear plants
  • 18.2 Pedigree of materials
  • 18.3 Application and challenges
  • 18.4 Evaluation technologies
  • 18.4.1 Longer service life
  • 18.4.1.1 Creep data acquisition and extrapolation
  • 18.4.1.2 Stability of microstructure
  • 18.4.1.3 Creep-fatigue
  • 18.4.1.4 Aging effects
  • 18.4.2 Welded joints
  • 18.4.3 Environmental effects
  • 18.5 Fabrication technologies
  • 18.6 Code qualification
  • 18.6.1 ASME Boiler and Pressure Vessel Code
  • 18.6.2 RCC-MRx code
  • 18.6.3 JSME Fast Reactor Code
  • 18.7 Summary
  • References
  • Index
  • A
  • C
  • D
  • E
  • F
  • G
  • H
  • I
  • J
  • K
  • L
  • M
  • N
  • O
  • P
  • R
  • S
  • T
  • U
  • V
  • W
  • Back Cover

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