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Meeting 2014 TMS Annual Meeting & Exhibition
Symposium Materials and Fuels for the Current and Advanced Nuclear Reactors III
Sponsorship TMS Structural Materials Division
TMS/ASM: Nuclear Materials Committee
Dennis Keiser, Idaho National Laboratory
Raul Rebak, GE Global Research
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 07/15/2013
Proceedings Plan Planned: Metallurgical and Materials Transactions

A Unified Viscoplastic Constitutive Model for Creep Damage Analysis in the Welded Joints of Modified 9Cr-1Mo Steel
Ab Initio Enhanced CALPHAD Modeling of Actinide Rich Metallic Nuclear Fuels
Accident Tolerant Fuels for Light Water Reactors
Advanced Electron Microscopic Examination Aided in the Identification of Silver and Palladium in Irradiated TRISO Coated Particles
Alloy Selection for Accident Tolerant Fuel Cladding in Commercial Light Water Reactors
Aspects of Dynamic Strain Aging in HT-9 Steel
Atomistic Investigation of Ionic Conductivity in Chorimum-doped Urania Fuel
Characterization of Hot Deformation Behavior of Zr–1Nb Alloy
Characterization of Nanostructured Ferritic Alloy Atomized with Yttrium and a Controlling Oxygen Content
Characterization of U-7Mo Alloy Microstructure Irradiated to High Fission Density
Correlation of Crystallographic Texture of Zr-excel Pressure Tube Materials with Thermal Creep Behavior
Corrosion Behavior of Alumina-forming Austenitic Steels in Supercritical Carbon Dioxide
Design of an In-reactor Experiment to Measure Tritium Release Kinetics and Speciation from LiAlO2-based Breeding Materials
Development and Testing Advanced Ferritic Steels for Fast Reactor Applications
Development of Fabrication Process Methodologies for Ceramic Fuel Pellets
Development of Nanostructured Ferritic Alloys Containing Lanthana-based Nanoparticles via Spark Plasma Sintering
Development of Phase Constituents and Microstructure in Monolithic U-Mo Fuel Plate Assembly during Hot Isostatic Pressing
Effect of Ball Milling Temperature on the Ultra Fine Grained Microstructure of Oxide Dispersion Strengthened Steel
Effect of Creep Deformation on Surface Degradation of Alloy 617 at 800C in Impure He Environments.
Effect of Laser Shock Peening on SCC Behavior of Alloy 600 in Sulfur Bearing Solutions
Effect of Milling and Precipitation Reinforcement on Grain Growth in Oxide-dispersion Strengthened Steels
Effect of Mo Content, Fabrication Method, Heating Process, and Fission Density on Recrystallization Kinetics and Performance of U-Mo Alloy Fuel Dispersion in Al Matrix during Irradiation
Effect of Nanocrystalline Grain Size on Mechanical Property Variation during Irradiation of Electrodeposited Nickel Coatings
Effect of Ni on Formation of Intermetallic Phases in Highly Irradiated Reactor Pressure Vessel Steels
Effects from Cr Concentration on Stability against Inter-diffusion between Lanthanides and Fe-Cr Alloys
Electronic Structure Calculations of Structure and Chemistry of the Y2O3/Fe Interface
Elevated Temperature Compression Testing of the U-10 wt% Mo Alloy: Impact of Homogenization Treatments
Evaluation of Vanadium Carbide for Mitigating Fuel Cladding Chemical Interaction
Fabrication and Properties of High Thermal Conductivity UO2, UO2-SiC, UO2-Diamond, and UO2-CNTComposites Using Spark Plasma Sintering
Finite Element Analysis of the Rolling of U10Mo Alloy: Parametric Study on Rolling Process Parameters
Grain Boundary Diffusion of Ag in Polycrystalline Silicon Carbide in TRISO Fuel Particles
Grain Boundary Engineering of Alloy 617 through Cold Deformation and Annealing
High-density Fuel Development for High Performance Research Reactors at TUM
High Energy X-ray Diffraction Study of Deformation Behavior of Alloy HT9
High Resolution Transmission Microscopy Characterization of an Oxide Dispersion Strengthened Steel Ball-milled Powder
High Temperature Fracture Toughness Testing for Advanced Reactor Applications
Impacts of Hydrogen in Unirradiated Zircaloy Nuclear Cladding under Dry Storage Conditions
Influence of Neutron Irradiation on the Segregation of Alloying Elements in Zirconium Alloys
Insights into Atomic Scale Microstructures of Alloys under Corrosive Environments
Integration of a Viscoplastic Self Consistent Plasticity Model with Finite Element Framework MOOSE
Interactions between Gliding Dislocations and Different Types of the Irradiation-induced Loops in α-iron: Molecular Dynamics Simulations and Dislocation Dynamics Simulations Comparison
Interatomic Potentials Accuracy: How Do They Bridge the Scales? U-Mo Fuel Case
Ion Irradiation Enhanced Interdiffusion in Uranium-iron System
Irradiation Effects on Fission Product Behavior in PyC and SiC
L38: A Theoretical Model of Corrosion Rate Distribution in Liquid LBE Flow Loop at Higher Temperature Ranges
L39: Comparison of EAM and MEAM Interatomic Potentials for Metallic Uranium
L40: Dynamics of Deformation Localization and Dislocation Channeling in Irradiated Austenitic Stainless Steels
L41: Hardness Recovery under Isochronal Annealing of Highly Irradiated RPV Steels
L42: Interdiffusion between of Mg and AA6061 Aluminum Alloy
L43: Irradiation Effect of P92 Steel during Ions Irradiations at Elevated Temperature
L44: Remote Exterior Condition Monitoring System for Spent Nuclear Fuel Dry Storage Containers
Lab-scale Methods to Enable the Selection of Nuclear Fuel Concepts for Development
Line Dislocation Dynamics Simulation of Fundamental Dislocation Properties in Zirconium
Material Characterization of Zr Nuclear Fuel Clad Tubes via Imperfection Modeling
Material Selection for Accident Tolerant Fuel Cladding
Materials Challenges in Next Generation Nuclear Reactors
Mechanical Properties and Microstructure of Ultrafine-grained Zircaloy-4 processed through Multiaxial Forging
Mechanical Properties of Irradiated T91 Alloy from the MEGAPIE Experiment
Micro-texture Development in Relation to Total Circumferential Elongation: Burst Test Performance in Zircaloy-4 Clads
Microstructural Characteristics of As-fabricated Monolithic U-Mo Nuclear Fuels
Microstructure Evolution in Advanced Ferritic-martensitic Steels Following Friction Stir Welding
Microstructure of Aluminum Matrix in Composite Absorber Block Material
Modifying Ceramic Fuel Pellets to Improve UO2 Properties
Molten Salts and Nuclear Energy
Nano-particles for Spent Nuclear Fuel Separation
On the Intermetallic Phases Formed between U, Pu-based Fuels and Fe-based Alloys
Oxidation of Alloy 617 in Controlled Impure Helium Environments at High Temperatures
PWSCC of Alloy 600 with Water Environment
Radiation Induced Hardening in Iron and Ferritic Alloys: An Atomic-scale View
Similar and Dissimilar Friction Stir Welding of ODS and RAFM Steels
Sink Strengths of Grain Boundaries in Irradiated Nanocrystalline Materials
Spectroscopic Real Time Monitoring of Molten Salts in Nuclear Electrorefiner Systems
Stable Storage of He in Nanometer-scale Interfacial Platelets
Steel Corrosion Tests in Flowing Lead-bismuth Eutectic in LANL DELTA Loop
Stress Corrosion Crack Initiation of Alloy 690 in Subcritical and Supercritical Water
Structure and Properties of Modified High Nitrogen Austenitic Stainless Steels
Study of Ordering Transformation in Ni-based Superalloy 690
Synchrotron Study on Loading Partitioning with Phase Development in an Austenitic 304 ODS
Synthesis of U3Si2 by High-energy Ball Milling
The Effect of Time, Temperature and Processing on the Microstructure Development in U-10 wt% Mo
The Effects of Processing on Precipitate Distribution and Mechanical Properties of a Nanostructured Ferritic Alloy (NFA)
The Fuel Fabrication Capability and Uranium-molybdenum Alloy: An Overview
The Nanoparticle-matrix Orientation Relationship and the Strengthening Mechanism in Austenitic ODS Stainless Steels
The Recovery of Irradiation Damage for Zircaloy-2 and Zircaloy-4 Following Low Dose Neutron Irradiation at Nominally 358°C
Thermal Aging Effect on Fracture Toughness of Modified 9Cr-1Mo Steel for Advanced Reactor Applications
Thermal Stability of Ultrafine Grained Austenitic ODS Steel
Thermal Stability of Uranium-rich U-Mo Alloys for Advanced Nuclear Fuels
Thermal Transport in Uranium Dioxide from First Principles
Thermo-mechanical and Microstructural Characterization of Molybdenum-alloy/Zirconium Alloys/FeCrAlY Composite Tubing for Fuel Cladding of Light Water Reactors
Thermo-mechanical Processed Two-dimensional Linear Plane-strain Machining of 316L Austenitic Stainless Steel for Improved SCC Resistance
Thermodynamic and Kinetic Modeling of Oxide Precipitation in Nanostructured Ferritic Alloys
Thermodynamic Modeling of Precipitate Phases in Austenitic Steels
Thermodynamics of U-Mo-Zr Alloys: Application to RERTR Nuclear Fuels
Transition in Creep Mechanisms in HANA-4 Zirconium Alloy
Understanding the Mechanisms for Amorphization Resistance in ZrC

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