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Meeting 2017 TMS Annual Meeting & Exhibition
Symposium Materials and Fuels for the Current and Advanced Nuclear Reactors VI
Sponsorship TMS Structural Materials Division
TMS: Corrosion and Environmental Effects Committee
TMS: Nuclear Materials Committee
TMS: Mechanical Behavior of Materials Committee
Organizer(s) Ramprashad Prabhakaran, Pacific Northwest National Laboratory
Dennis Keiser, Idaho National Laboratory
Raul Rebak, GE Global Research
Clarissa Yablinsky, Los Alamos National Laboratory
Scope Globally, significant efforts are ongoing to meet the growing energy demand with the increased use of nuclear energy. Extensive work is being performed to develop materials and fuels for the advanced nuclear reactors. In addition, efforts are also ongoing to extend the life of existing nuclear power plants. Scientists, engineers, and students at various national laboratories, universities, and industries are working on a number of materials challenges for the nuclear energy systems. The objective of this symposium is to provide a platform for these researchers to congregate, exhibit and discuss their current research work, in addition to sharing the challenges and solutions with the professional community and thus, shape the future of nuclear energy.

Abstracts are solicited in (but not limited to) the following topics:

• Nuclear reactor systems
• Advanced nuclear fuels - fabrication, performance, and design
• Advanced nuclear fuels - properties and modeling
• Advanced structural materials - fabrication, joining, properties, and characterization
• Lifetime extension of reactors - nuclear materials aging, degradation, and others
• Experimental, modeling, and simulation studies
• Fundamental science of radiation-material interactions
• Irradiation effects in nuclear materials
• Materials degradation issues - stress corrosion cracking, corrosion, creep, fatigue, and others
• Design of materials for extreme radiation environments
• Radiation measurement techniques and modeling studies
• Nuclear waste - disposal, transmutation, spent nuclear fuel reprocessing
Abstracts Due 07/17/2016
Proceedings Plan Planned: Supplemental Proceedings volume
PRESENTATIONS APPROVED FOR THIS SYMPOSIUM INCLUDE

A Grand-Potential Phase Field Model for Bubble Formation and Growth in U-Si Fuel
A Modified Embedded-Atom Method Interatomic Potential for U-Si
An Integrated Simulation for Deformation and Irradiation-Induced Grain Growth in, U-10 wt%Mo
Atomic Scale Behavior of Beryllium in Zirconium
Characterization and Simulation of Wear-tested Zirconium Alloy Surfaces
Characterization of Nuclear Fuels by Neutron Diffraction and Energy-resolved Neutron Imaging
Charged Particle Irradiation Studies of High Dose Precipitation in Reactor Pressure Vessel Steels
Cluster Dynamics Modeling of Cu Precipitation Hardening in Reactor Pressure Vessel Steels
Complex SiC-SiC Composite Structures for Nuclear Applications
Continuum-level Modeling of Irradiation Damage Cascades with Explicit Microstructure Representation
Creep-Fatigue Deformation of 9Cr-1MoV Steel and Weldments
Creep Fatigue Crack Growth of T91: Test Design and Data Analysis
Damage Rate Dependence of Oxide Evolution on Zircaloy-4 under Simultaneous Irradiation-corrosion Experiment
Density Functional Theory Investigation of Defect and Fission Gas Diffusion in U3Si2
Determination of Material Properties of Ion-irradiated and Corroded Zircaloy-4 by Using Nanomechanical Raman Spectroscopy
Effect of Different Processing Routes on the Microstructure and Texture of 14YWT Alloy
Effect of Grain Morphology on Gas Bubble Swelling in UMo Fuels – A 3D Microstructure Dependent Booth Model
Effects of Ion-irradiation Damage on Mechanical Behavior in Silicon Carbide
Effects of Thermal Aging and Neutron Irradiation on Cast Austenitic Stainless Steels
Electron Backscatter Diffraction Analysis of Irradiated U-Mo Plate Fuel for the US High Performance Research Reactor Development Program
Eutectoid Transformation Kinetics of As-Cast U - 8 wt% Mo Established by In Situ Neutron Diffraction
Evolution of Stress and Fracture During Oxidation of Zirconium Alloys
Exposing the Mechanisms of Pellet-Cladding Interaction Using Atomistic Simulation
Fabrication and Characterization of TRISO Particles Using 800Ám Uranium Nitride and Surrogate ZrO2 Kernels
Fission Product Electron Microscopy Analysis of Post Irradiated TRISO-coated Particles from the Second Advanced Gas Reactor Experiment
G-11: A Composite Waste Form for Electrochemical Processing Wastes
G-14: Fission Gas Release and Swelling Model of Uranium Nitride Based on the Rate Ttheory
G-16: Simulation of Constituent Redistribution and Fuel Restructuring in MOX Fuel
G-17: Thermodynamic Properties of Strontium-Bismuth Alloys for Electrochemical Separation of Strontium
G-1: Effects of Added Molybdenum on Corrosion of 3l6L Stainless Steel
G-2: Fabrication and Microstructures of Burnable Absorber-cored Oxide Pellets for Advanced Nuclear Fuel
G-3: Diffusion Studies in the Development of an FCCI Barrier for High-Burnup Metallic Nuclear Fuel
G-4: Irradiated Materials Characterization Laboratory at Idaho National Laboratory
G-5: Quantification of the Stress-Stabilization of Tetragonal ZrO2
G-6: Steam Oxidation Resistance of Silicide and Aluminide-coated Refractory Metals
G-7: Advanced Electron Microscopy of Fission Products in Irradiated TRISO Fuel
G-8: Phase Field Modeling of Fission Gas Behavior in Metallic Nuclear Fuel
G-9: Asymptotic Expansion Homogenization of Thermal Conductivity and Elasticity of Irradiated Hafnium-Aluminum Composite Performed on Reconstructed and Synthetic Microstructures
Grain Boundary Complexions in SiC and Their Relevance in Silver Diffusion in TRISO Particles
Grain Boundary Damage Precursors Leading to Intergranular SCC Initiation of Cold-Worked Alloy 600 and Alloy 690 in PWR Primary Water
High Temperature Fuel Cladding Chemical Interactions between Unirradiated TRIGA Fuels and 304 Stainless Steel
Impact of the Neutron Irradiation on the Structure and Properties of the 6061 Al Alloy Produced by Ultrasonic Additive Manufacturing
In-situ High Energy X-ray Characterization of Neutron Irradiated HT-UPS Stainless Steel under Tensile Deformation
Interdiffusion and Reaction between U and Zr
Investigation of Property-Property Correlations for Irradiated Steels
Irradiation-Induced Microstructure of Proton Irradiated Commercial Austenitic Alloys
Mechanical Property Measurements of a New Metal Matrix Material for Nuclear Reactor Applications
Microstructural Analysis of Electrochemically Formed Zirconium Coatings for Uranium-Molybdenum Nuclear Fuels
Microstructural Characterization and Thermal Properties of Metallic Pu-Zr Systems
Microstructural Characterization of AA6061-AA6061 HIP Bonded Cladding Interface
Microstructural Development of UMo-Al Dispersion Fuels after Thermal Annealing
Microstructural Evolution of Thermal Recovery in Ti3AlC2-Ti5Al2C3 and Ti3SiC2
Microstructural Heterogeneity of Deformed and Annealed FeCrAl Alloys with Nb Addition
Microstructure Evolution during Spark Plasma Sintering of Nuclear Fuel Pellets and Their Large-scale Manufacturability
Mitigation of IASCC Susceptibility in a BWR-irradiated 304L Stainless Steel Utilizing Post-irradiation Annealing
Model of Thermal Conductivity Reduction Due to Point Defect Accumulation in Ion Irradiated UO2
Modeling Activation and Radionuclide Decay in Proton Irradiated Zirconium Alloys
Monte Carlo Modeling of Recrystallization Processes in α-Uranium
Nanoscale Structural and Compositional Analysis of U-10Mo Fuels
Neutron Irradiation-induced Creep of IG-110 Nuclear Graphite
On Silver Transport in 3C-SiC
Peuget: How Ion Beam Irradiations Simulate the Radiation Aging of Nuclear Glass
Phase Field Modeling of PWR Cladding Corrosion with the HOGNOSE Code
Post Irradiation Electron Microscopy Examination of UCO Fuel Kernels from TRISO Coated Particles
Preliminary Post Irradiation Examination SEM Analysis of AGR 2 UO2 and UCO TRISO Fuel Particles
Production of Fully Ceramic Microencapsulated Fuel for Test Reactor Irradiation
Property Evolution Due to Thermal Aging of Cast Duplex Stainless Steels As Measured by Multi-Scale Mechanical Methods
Recrystallization Texture in U10Mo Alloy
Reduced Modulus and Hardness of Uranium-molybdenum Solid Solution as a Function of Mo Composition and Related Phase Transformations
Results of Microstructural Characterization Focused on the U-10Mo/Zr Diffusion Barrier Interface in Irradiated Monolithic Fuel Plates
Role of Localized Deformation and Grain Boundary Plane Orientation on Crack Initiation in Irradiated Stainless Steels
Seamless Thin-wall Tube Production of ATF Wrought FeCrAl Alloys
Sensitivity Analysis on the Temperature of U–Mo/Al Plate-type Dispersion Fuel
Small Scale Mechanical Testing of UO2 at Elevated Temperatures
Study on Texture Evolution of As-hydrided Zircaloy-4 Cladding under Low Temperature Biaxial Creep Test
Study on the Microstructure and Mechanical Behavior of the New Type SA508-IV Reactor Pressure Vessel (RPV) Steel by Different Methods
The Effect of Low-fluence Neutron Irradiation on Cast Austenitic Stainless Steels
The Increase in Fatigue Crack Growth Rates Observed for Zircaloy-4 in a PWR Environment
The Recovery of Irradiation Damage for Zircaloy-2 and Zircaloy-4 Following Irradiation at Higher Temperatures of 377-410C
The Role of Stoichiometry on Ordering Phase Transformations in Ni-Cr Alloys for Nuclear Applications
Thermodynamic Modeling and Continuum Scale Fuel Performance Simulations
Utilizing In-situ Microtensile Testing to Evaluate Mechanical Property Changes Due to Ion-beam Irradiation
Wear Results for Zirconium Alloys and Their Oxides
Wear Study Comparison of Accident Tolerant FeCrAl Cladding, Zircaloy-2 and SS304 against X750


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