Comprehensive Nuclear Materials
- 1st Edition - May 12, 2011
- Imprint: Elsevier Science
- Editors: Todd R Allen, Rudy Konings, Roger E Stoller, Shinsuke Yamanaka
- Language: English
- eBook ISBN:9 7 8 - 0 - 0 8 - 0 5 6 0 3 3 - 5
Comprehensive Nuclear Materials, Five Volume Set discusses the major classes of materials suitable for usage in nuclear fission, fusion reactors and high power accelerat… Read more
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Request a sales quoteComprehensive Nuclear Materials, Five Volume Set discusses the major classes of materials suitable for usage in nuclear fission, fusion reactors and high power accelerators, and for diverse functions in fuels, cladding, moderator and control materials, structural, functional, and waste materials. The work addresses the full panorama of contemporary international research in nuclear materials, from Actinides to Zirconium alloys, from the worlds' leading scientists and engineers.
- Critically reviews the major classes and functions of materials, supporting the selection, assessment, validation and engineering of materials in extreme nuclear environment
- Fully integrated with F-elements.net, a proprietary database containing useful cross-referenced property data on the lanthanides and actinides
- Details contemporary developments in numerical simulation, modelling, experimentation, and computational analysis, for effective implementation in labs and plants
The work will be suitable for graduate students and above studying any materials aspect of nuclear science within academia, and engineering, as well as professional nuclear engineers and research scientists
Fundamental Properties of Defects in Metals
Fundamental Point Defect Properties in Ceramics
Radiation-Induced Effects on Microstructure
Radiation-Induced Effects on Material Properties of Metals (Mechanical and Dimensional)
Radiation-Induced Effects on Material Properties of Ceramics (Mechanical and Dimensional)
The Effects of Helium in Irradiated Structural Alloys
Radiation Damage Using Ion Beams
Ab Initio Electronic Structure Calculations for Nuclear Materials
Molecular Dynamics
Interatomic Potential Development
Primary Radiation Damage Formation
Atomic-Level Level Dislocation Dynamics in Irradiated Metals
Mean Field Reaction Rate Theory
Kinetic Monte Carlo Simulations of Irradiation Effects
Phase Field Methods
Dislocation Dynamics
Computational Thermodynamics: Application to Nuclear Materials
Radiation-Induced Segregation
The Actinides Elements: Properties and Characteristics
Thermodynamic and Thermophysical Properties of the Actinide Oxides
Thermodynamic and Thermophysical Properties of the Actinide Nitrides
Thermodynamic and Thermophysical Properties of the Actinide Carbides
Phase Diagrams of Actinide Alloys
The U-F System
Zirconium Alloys: Properties and Characteristics
Nickel Alloys: Properties and Characteristics
Properties of Austenitic Steels for Nuclear Reactor Applications
Graphite: Properties and Characteristics
Neutron Reflector Materials (Be, Hydrides)
Proerties and Characteristics of SiC and SiC/SiC Composites
Proerties and Characteristics of ZrC
Properties of Liquid Metal Coolants
Uranium Oxide and MOX Production
Burnable Poison-Doped Fuel
Thermal Properties of Irradiated UO₂ and MOX
Radiation Effects in UO2
Fuel Performance of Light Water Reactors (Uranium Oxide and MOX)
Fission Product Chemistry in Oxide Fuels
Fuel Performance of Fast Spectrum Oxide Fuel
Transient Response of LWR Fuels (RIA)
Behaviour of LWR Fuel During Loss-of_Coolant Accidents
Behaviour of Fast Reactor Fuels During Transient and Accident Conditions
Core Concrete Interaction
Metal Fuel
Nitride Fuel
Carbide Fuel
Thorium Oxide Fuel
Actinide Bearing Fuels and Transmutation Targets
TRISO Fuel Production
TRISO-Coated Particle Fuel Performance
Advanced Concepts in TRISO Fuel
Inert Matrix Fuel
Composite Fuel (CERMET, CERCER)
Sphere-Pac and VIPAC Fuel
Uranium-Zirconium Hydride Fuel
Molten Salt Reactor Fuel and Coolant
Uranium Inter-Metallic Fuels (U-Al, U-Si, U-Mo)
Metal Fuel-Cladding Interaction
Ceramic Fuel-Cladding Interaction
Thermal Spectrum Control Rod Materials
Fast Spectrum Control Rod Materials
Oxide Fuel Performance Modelling and Simulation
Modeling of Fission-Gas Induced Swelling of Nuclear Fuels
Matter Transport in Fast Reactor Fuels
Modelling of Pellet Cladding Interaction
Metal Fuel Performance Modelling and Simulation
TRISO Fuel Performance Modelling and Simulation
Modeling of Sphere-Pac Fuel
Radiation Effects in Zirconium Alloys
Radiation Damage in Austenitic Steels
Ferritic Steels and Advanced Ferritic-Martensitic Steels
Radiation Effects in Nickel-Based Alloys
Radiation Damage of Reactor Pressure Vessel Steels
Radiation Effects in Refractory Metals and Alloys
Radiation Effects in SiC and SiC-SiC
Oxide Dispersion Strengthened Steels
Welds for Nuclear Systems
Radiation Effects in Graphite
Graphite in Gas-Cooled Reactors
Vanadium for Nuclear Systems
Concrete
Fracture Toughness Master Curve of BCC Steels
Ceramic Breeder Materials
Tritium Barriers and Tritium Diffusion in Fusion Reactors
Tungsten as a Plasma-Facing Material
Carbon as a Fusion Plasma-Facing Material
Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices
Physical and Mechanical Properties of Copper and Copper Alloys
Ceramic Coating as Insulators
Radiation Effects on the Physical Properties of Dielectric Insulators for Fusion Reactors
Corrosion and Compatibility
Water Chemistry Control in LWRs
Corrosion of Zirconium Alloys
Corrosion and Stress Corrosion Cracking of Ni-Base Alloys
Corrosion and Stress Corrosion Cracking of Austenitic Stainless Steels
Corrosion and Environmentally-Assisted Cracking of Carbon and Low-Alloy Steels
Performance of Aluminium in Research Reactors
Irradiation Assisted Stress Corrosion Cracking
Material Performance in Lead-Alloys
Material Performance in Molten Salts
Material Performance in Helium-Cooled Systems
Material Performance in Supercritical Water
Material Performance in Sodium
Spent Fuel Dissolution and Reprocessing Processes
Degradation Issues in Aqueous Reprocessing Systems
Spent Fuel as Waste Material
Waste Containers
Waste Glass
Ceramic Waste Forms
Metallic Waste Forms
Graphite
Minerals and Natural Analogues
Fundamental Point Defect Properties in Ceramics
Radiation-Induced Effects on Microstructure
Radiation-Induced Effects on Material Properties of Metals (Mechanical and Dimensional)
Radiation-Induced Effects on Material Properties of Ceramics (Mechanical and Dimensional)
The Effects of Helium in Irradiated Structural Alloys
Radiation Damage Using Ion Beams
Ab Initio Electronic Structure Calculations for Nuclear Materials
Molecular Dynamics
Interatomic Potential Development
Primary Radiation Damage Formation
Atomic-Level Level Dislocation Dynamics in Irradiated Metals
Mean Field Reaction Rate Theory
Kinetic Monte Carlo Simulations of Irradiation Effects
Phase Field Methods
Dislocation Dynamics
Computational Thermodynamics: Application to Nuclear Materials
Radiation-Induced Segregation
The Actinides Elements: Properties and Characteristics
Thermodynamic and Thermophysical Properties of the Actinide Oxides
Thermodynamic and Thermophysical Properties of the Actinide Nitrides
Thermodynamic and Thermophysical Properties of the Actinide Carbides
Phase Diagrams of Actinide Alloys
The U-F System
Zirconium Alloys: Properties and Characteristics
Nickel Alloys: Properties and Characteristics
Properties of Austenitic Steels for Nuclear Reactor Applications
Graphite: Properties and Characteristics
Neutron Reflector Materials (Be, Hydrides)
Proerties and Characteristics of SiC and SiC/SiC Composites
Proerties and Characteristics of ZrC
Properties of Liquid Metal Coolants
Uranium Oxide and MOX Production
Burnable Poison-Doped Fuel
Thermal Properties of Irradiated UO₂ and MOX
Radiation Effects in UO2
Fuel Performance of Light Water Reactors (Uranium Oxide and MOX)
Fission Product Chemistry in Oxide Fuels
Fuel Performance of Fast Spectrum Oxide Fuel
Transient Response of LWR Fuels (RIA)
Behaviour of LWR Fuel During Loss-of_Coolant Accidents
Behaviour of Fast Reactor Fuels During Transient and Accident Conditions
Core Concrete Interaction
Metal Fuel
Nitride Fuel
Carbide Fuel
Thorium Oxide Fuel
Actinide Bearing Fuels and Transmutation Targets
TRISO Fuel Production
TRISO-Coated Particle Fuel Performance
Advanced Concepts in TRISO Fuel
Inert Matrix Fuel
Composite Fuel (CERMET, CERCER)
Sphere-Pac and VIPAC Fuel
Uranium-Zirconium Hydride Fuel
Molten Salt Reactor Fuel and Coolant
Uranium Inter-Metallic Fuels (U-Al, U-Si, U-Mo)
Metal Fuel-Cladding Interaction
Ceramic Fuel-Cladding Interaction
Thermal Spectrum Control Rod Materials
Fast Spectrum Control Rod Materials
Oxide Fuel Performance Modelling and Simulation
Modeling of Fission-Gas Induced Swelling of Nuclear Fuels
Matter Transport in Fast Reactor Fuels
Modelling of Pellet Cladding Interaction
Metal Fuel Performance Modelling and Simulation
TRISO Fuel Performance Modelling and Simulation
Modeling of Sphere-Pac Fuel
Radiation Effects in Zirconium Alloys
Radiation Damage in Austenitic Steels
Ferritic Steels and Advanced Ferritic-Martensitic Steels
Radiation Effects in Nickel-Based Alloys
Radiation Damage of Reactor Pressure Vessel Steels
Radiation Effects in Refractory Metals and Alloys
Radiation Effects in SiC and SiC-SiC
Oxide Dispersion Strengthened Steels
Welds for Nuclear Systems
Radiation Effects in Graphite
Graphite in Gas-Cooled Reactors
Vanadium for Nuclear Systems
Concrete
Fracture Toughness Master Curve of BCC Steels
Ceramic Breeder Materials
Tritium Barriers and Tritium Diffusion in Fusion Reactors
Tungsten as a Plasma-Facing Material
Carbon as a Fusion Plasma-Facing Material
Beryllium as a Plasma-Facing Material for Near-Term Fusion Devices
Physical and Mechanical Properties of Copper and Copper Alloys
Ceramic Coating as Insulators
Radiation Effects on the Physical Properties of Dielectric Insulators for Fusion Reactors
Corrosion and Compatibility
Water Chemistry Control in LWRs
Corrosion of Zirconium Alloys
Corrosion and Stress Corrosion Cracking of Ni-Base Alloys
Corrosion and Stress Corrosion Cracking of Austenitic Stainless Steels
Corrosion and Environmentally-Assisted Cracking of Carbon and Low-Alloy Steels
Performance of Aluminium in Research Reactors
Irradiation Assisted Stress Corrosion Cracking
Material Performance in Lead-Alloys
Material Performance in Molten Salts
Material Performance in Helium-Cooled Systems
Material Performance in Supercritical Water
Material Performance in Sodium
Spent Fuel Dissolution and Reprocessing Processes
Degradation Issues in Aqueous Reprocessing Systems
Spent Fuel as Waste Material
Waste Containers
Waste Glass
Ceramic Waste Forms
Metallic Waste Forms
Graphite
Minerals and Natural Analogues
- Edition: 1
- Published: May 12, 2011
- Imprint: Elsevier Science
- Language: English
- eBook ISBN: 9780080560335
TA
Todd R Allen
Dr. Todd Allen is an Associate Professor in the Department of Engineering Physics at the University of Wisconsin – Madison since 2003. Professor Allen's research expertise is in the area of materials related issues in nuclear reactors, specifically radiation damage and corrosion. Dr. Allen has established the Extreme Environment Laboratory equipped with facilities for a wide array of high temperature studies as well as the ion beam laboratory for radiation damage studies. Dr. Allen is also the Scientific Director for the Advanced Test Reactor National Scientific User Facility at Idaho National Laboratory, a position that he holds in conjunction with his faculty position at the University of Wisconsin.
Affiliations and expertise
University of Wisconsin-Madison, WI, USARK
Rudy Konings
Rudy Konings graduated from Utrecht University with an MSc in Earth Sciences in 1985. He then joined the Netherlands Energy Research Foundation ECN as researcher in the field of thermodynamics of nuclear materials and defended his PhD at the University of Amsterdam in 1990. He stayed at ECN and subsequently at NRG (Nuclear Research and Consultancy Group) working on nuclear fuel-related issues, his last role being head of the unit "Fuel, Actinides and Isotopes". In 1999 he joined the European Commission's Joint Research Centre in Karlsruhe (formerly the Institute for Transuranium Elements). There he currently is the head of the Nuclear Fuel Safety Unit, responsible for the nuclear fuel related research, comprising fuel synthesis, characterization, and accident behavior. In addition he is part-time professor at the Delft University of Technology, holding the chair "Chemistry of the nuclear fuel cycle". His research interests are nuclear fuels and actinide materials, with particular emphasis on high temperature chemistry and thermodynamics. He is a (co)author of more than 300 scientific publications in peer-reviewed journal and 14 book chapters, has been editor of Journal of Nuclear Materials (2009-2012), and editor-in-chief of the first edition of Comprehensive Nuclear Materials.
Affiliations and expertise
Head of the Nuclear Fuel Safety Unit, European Commission's Joint Research Centre, Karlsruhe, GermanyRS
Roger E Stoller
Dr. Stoller is currently a Distinguished Research Staff Member in the Materials Science and Technology Division of the Oak Ridge National Laboratory, and serves as the ORNL Program Manager for Fusion Reactor Materials for ORNL. He joined ORNL in 1984, and is actively involved in research on the effects of radiation on structural materials and fuels for nuclear energy systems. His primary expertise is in the area of computational modeling and simulation. He has authored or co-authored more than 100 publications and reports on the effects of radiation on materials, as well as having edited the proceedings of several international conferences.
Affiliations and expertise
Oak Ridge National Laboratory, TN, USASY
Shinsuke Yamanaka
Shinsuke Yamanaka has been a Professor in the Division of Sustainable Energy and Environmental Engineering, Graduate School of Engineering, Osaka University since 1998. Prof. Yamanaka has studied the thermophysics and thermochemistry of nuclear fuel and materials. His research for the hydrogen behavior in LWR fuel cladding is notable among his achievements and he received the Young Scientist Award (1980) and the Best Paper Award (2004) from Japan's Atomic Energy Society. Prof. Yamanaka has been the program officer of Japan's Science and Technology Agency since 2005 and the visiting professor of Fukui University since 2009. He is also the associate dean of the Graduate School of Engineering at Osaka University.
Affiliations and expertise
Osaka University, Osaka, JapanRead Comprehensive Nuclear Materials on ScienceDirect