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Meeting 2015 TMS Annual Meeting & Exhibition
Symposium Materials and Fuels for the Current and Advanced Nuclear Reactors IV
Sponsorship TMS Structural Materials Division
TMS: Nuclear Materials Committee
TMS: Corrosion and Environmental Effects Committee
TMS: Mechanical Behavior of Materials Committee
Dennis Keiser, Idaho National Laboratory
Raul Rebak, GE Global Research
Clarissa A. 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 and others
Abstracts Due 07/15/2014
Proceedings Plan Planned: A print-only volume

A Comparison of High-Intensity Neutron Sources for Fusion Materials Development
A Formed Can Approach to Hot Isostatically Press-Bonding Aluminum Cladding to Monolithic Uranium-10 wt. pct. Molybdenum Fuel Plates
A Unified Microstructurally-Based Physical Model of Low Flux-High Fluence Irradiation Embrittlement of Reactor Pressure Steels
Advanced Investigations on the Strengthening Mechanisms in Austenitic ODS Stainless Steels
Advanced ODS FeCrAl Alloys for Accident-Tolerant Fuel Cladding
Alpha Decay-Induced Helium and Defect Accumulation in Ceramic Nuclear Waste Forms
Alternative Approaches for the Zr Coating of Low Enrichment U-10Mo High Performance Research Reactor Fuel
Apatite-Based Ceramic Waste Forms by High Energy Ball Milling and Spark Plasma Sintering for Iodine Confinement
As Fabricated Microstructures of Diffusion Barriers on U-Mo Dispersion Particles in Al-Matrix
Assessing and Modelling the Performance of Waste Forms and Container Metals for Long-Term Disposal
Atomistic-Informed Phase Field Model for Predicting Cr Segregation to Sinks in Irradiated Fe-Cr Alloys
Characterization of a Bending Fatigue Mini-Specimen Technique (Krouse Type) of Nuclear Materials
Characterization of Delta Hydride Precipitates in Pure Zr and Zr-4
Characterization of Irradiation Effects in Nanoscale Stable Precipitation-Strengthened Steels
Computational Study of Energetics and Defect-Ordering Tendencies for Rare Earth Elements in Uranium Dioxide
Corrosion of 316L Stainless Steel in Primary Water During In-Situ Proton or Electron Irradiation
Cracking of Irradiated CF-3 Cast Austenitic Stainless Steel with 24% Ferrite
Creep Resistance and Material Degradation of a Candidate Ni-Mo-Cr Corrosion Resistant Alloy for Application in a Molten Salt Nuclear Reactor
Density Functional Theory Study on the Behavior of Vanadium Carbide as a Diffusion Barrier within the Fuel/Cladding System of a Fast Reactor
Diffusion Kinetics, Interface Compound Formation and Radiation Responses of U-Fe and U-Ni Diffusion Couples
Diffusional interactions in U-Mo vs. AA6061 Diffusion Couples Annealed at 600 and 550C
Effect of Burn-Up on the Thermal Conductivity of Uranium-Gadolinium Dioxide Up to 100 GWd/tHM
Effect of Heavy Ion Irradiation on Microstructural Evolution in CF8 Cast Austenitic Stainless Steel at 300, 350 and 400°C
Effects of Alloying and Processing Modifications on Precipitation and Strength in P92-like Alloys
Energetics Associated With the Interaction between Embrittlement Species and Grain Boundaries in Alpha-Iron
Engineering Challenges in the Down Selection of the TREAT Pre-Conceptual Low-Enriched Fuel System Concepts and Design
Enhanced Thermal Conductivity of Uranium Dioxide-Diamond Composite Fuel
Evaluation of the Biaxial Thermal Creep of hydrided Zircaloy-4 Cladding for Interim Dry Storage of Spent Nuclear Fuel
Evolution of Phase Constituents and Microstructure in Hot Isostatic Pressed Monolithic U-Mo Fuel Plates in AA6061 Cladding with Zr Diffusion Barrier
Evolution of Phase Constituents and Microstructure in the U-10wt.%Mo Alloy with Various Zr Additions after Heat Treatment at 900, 650, and 560C
Fission Product Distribution Patterns as a Comparative Characterization Tool for TRISO Fuel Performance
Fuel-Cladding Chemical Interaction Effects in U, Pu-Based Fuels and Cladding
Fully Coupled Multiphysics Simulation of Fast Reactor Mixed Oxide Fuels Performance under Extreme Conditions
Gradation of Gamma Lithium Aluminate under Simulated Storage Conditions
Graphene-UO2 Composites for Accident Tolerant Nuclear Fuel
High Energy Xe Ion Irradiation Study of U-Mo/Al Dispersion Fuel
Hydrogen Embrittlement Testing of a Zirconium Based Alloy
Hydrogen Induced Degradation Processes in ZrH2-U Fuel
In Situ Study of Defect Migration Kinetics in Nanoporous Ag with Enhanced Radiation Tolerance
Interdiffusion and Reaction between U-Mo and Zr at 650C as a Function of Time
Interdiffusion between Lanthanides and Cladding through the Vanadium Carbide Coating Obtained From Low-Temperature Chemical Vapor Deposition
Investigating the Effect of Oxide Texture on Corrosion Performance and Hydrogen Pickup in Zirconium Alloys
K10: Reactivity Suppression of Liquid Sodium by Suspended Nanoparticles
K12: The Effects of Irradiation on China RPV Steel Cleavage Fracture Behavior
K13: Irradiation Damage in Ultra-Fine and Large Grain Zirconium Materials
K14: Enhanced Irradiation Tolerance of Ultrafine Grained T91 Steel Processed by Equal Channel Angular Extrusion
K15: Serrated Flow in 9–11Cr Ferritic/Martensitic Steels
K17: Development of Nuclear Quality Components Using Metal Additive Manufacturing
K18: Effect of Strain and Degree of Sensitization in TGSCC Susceptibility of Stainless Steel in High Temperature
K19: Interaction of Selected MAX Phases with Pure Sodium
K20: Interaction of Selected MAX Phases with Pyrolytic Carbon and Silicon Carbide
K21: Physical Properties and Corrosion Studies of Titanium Aluminum Carbide Coatings
K22: Ductile-Phase-Toughened Tungsten Laminates for Plasma-Facing Materials
K24: Early Stage Corrosion Study of Zircaloy-4 Inside a Transmission Electron Microscope
K25: Damage Evolution in Irradiated SiC: Modeling and Experimental Study
K26: Characterization of a Bending Fatigue Mini-Specimen Technique (Krouse Type) of Nuclear Materials
K27: Corrosion Studies on U-Mo Fuel for Research Reactor Applications
K28: Investigation of Tungsten-Yttrium Based Structural Materials for Nuclear Reactor Applications
K9: Validation and Numerical Simulation for Shrinkage Porosity of an X12 Steel Ingot
Magnetic Cr-Doped Fe-Fe Oxide Core-Shell Nanoparticles for Used Nuclear Fuel Separation
Magnetic Nanosorbents for Recycling Spent Nuclear Fuel
Mechanical Properties and Microstructural Stability of Oxide Dispersion Strengthened Alloy 617
Microstructure Characterization of 12Cr ODS Steel after Creep Rupture Test at 700oC
Microstructure Evolution and Microstructure-Strength Correlation Predictions in Nuclear Materials Based on Grain Boundary Structure-Fractal Dimension Correlations
Microstructures Observed in U-Mo Dispersion Fuel with Magnesium Matrix
Mitigation of Oxidation of LWR Zircaloy Cladding in High Temperature Steam via FeCrAl Coatings and Chromium Oxide Buffer Layers
Modeling the Homogenization Process for As-Cast U-10Mo
Modelling Silicide Fuel for Improved Accident Tolerance in Current and Next Generation Light Water Reactors
Molecular Dynamics Simulations of Zirconium/Zirconium-Hydride Interface with the Charge Optimized Many Body (COMB) Potential
Neutron Irradiation Studies on Friction Stir Processed ODS Alloys
Ni2Cr-Type Long-Range Ordering in IN690 and Ni-Cr Binary Alloys
Out-Of-Pile Test of the Effectiveness of Chemical Immobilization of Lanthanide in U-Zr Alloy Fuel
Oxidation of Zircaloy-4 in Simulated PWR Environments during In-Situ Proton Corrosion-Irradiation
Performance of Ultrafine-Grained Tungsten under ELMs-Like Transient Heat Loads of ITER
Phase-field Modelling of Gas Bubble Swelling and Its Impact on Thermo-mechanic Properties in UMo Metal Fuels
Post-irradiation Annealing of a BWR-Irradiated 304L Stainless Steel and Resultant Mitigation of IASCC Susceptibility
Pre-conceptual Development and Characterization of an Extruded Graphite Composite Fuel for the TREAT Reactor
Quantification of the Variability in Physical and Mechanical Properties of Nuclear-Grade Graphites
Radiation-Induced Microstructural Effects in Nickel-Chromium Binary Alloys
Radiation-Tolerant Nanoceramic Coatings for Next Generation Nuclear Systems
Radiation Effects on the Thermophysical Properties of a New Neutron Absorbing Material
Sample Environment for In Situ Corrosion Studies of Zirconium and Advanced Steel Cladding Alloys in Extreme Environments
Simulating Changes in Raman Spectra of Point Defects in UO2 from Lattice Dynamics
Solute Distributions in Oxide and Sub-Oxide Layers during Corrosion of Zirconium Alloys
Supercritical Carbon-Dioxide System for Materials Corrosion Testing
Synchrotron X-Ray Diffraction Characterization to Elucidate Oxidation of Advanced Steel Cladding Alloys; APMT and Alloy-33
Tensile and Fracture Toughness Properties of 14Cr-3W-0.3Ti-0.2Y (FCRD NFA-1)
The Deformation Behaviours of Long-Term Thermal Aged Duplex Stainless Steels Studied By In-Situ High-Energy X-ray Diffraction and In-Situ Scanning Electron Microscope
The Effect of Applied Stress on C-Component Dislocation Loops in Zr-Based Alloys
The Lift Out Mechanical Testing of Highly Irradiated Structural Materials
The Role of Localized Deformation in IASCC Initiation of Neutron Irradiated Austenitic Stainless Steel
The Role of Stress-State on the Deformation and Fracture Mechanism of Hydride and Non-Hydrided Zircaloy-4
The Role of the Interactions of Dislocation Channels with Grain Boundaries in the Irradiation Assisted Stress Corrosion Cracking Mechanism
Thermal Ageing Experiments of Ferritic-ODS Alloys
Thermal Conductivity of U-Mo Fuel as a Function of Burnup
Thermal Degradation of Cast Duplex Stainless Steels

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