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Meeting Materials Science & Technology 2017
Symposium Materials for Nuclear Energy Applications
Jake Amoroso, Savannah River National Laboratory
Aladar Csontos, Nuclear Regulatory Commission
Kevin Fox, Savannah River National Laboratory
Yutai Katoh, Oak Ridge National Laboratory
Bill Lee, Imperial College of London
Josef Matyas, Pacific Northwest National Laboratory
Raul B Rebak, GE Global Research
Cory Trivelpiece, Savannah River National Laboratory
Scope Global expansion of nuclear power calls for the accelerated development of structural materials and fuels for current and advanced reactor concepts. Materials used in the nuclear reactors must perform reliably for extended service times under extreme conditions of temperature, mechanical stresses, high radiation fields, and corrosive environments. This symposium will address topics ranging from fundamental and applied science and engineering aspects of materials development for fission and fusion reactors to radiation detectors. Additionally, the focus will be on lifetime extension of existing nuclear reactors and meeting the challenges of future high temperature reactor concepts. The radiation response of a wide range of materials including metallic alloys, ceramics, and composite materials and corrosion mechanisms in reactor-relevant coolants will be of interest. The mechanical behavior of reactor structural materials and recent advances in testing and characterization of these materials will be of interest. The symposium will also provide a platform for research and development activities at the interfaces of materials and manufacturing for reactor applications, including thermomechanical processing, additive manufacturing, and joining. Finally, advances in computational approaches at the atomistic and meso-scales are enabling simulations of radiation damage and other degradation mechanisms across multiple length and time scales are especially encouraged.
Abstracts are solicited in (but not limited to) the following topics:
• Corrosion and flow-assisted corrosion
• Fundamental science of radiation damage and defect processes
• Irradiation effects in materials
• In situ examination of irradiation damage
• Materials for extreme environments
• Accelerated development and deployment of materials
• Advanced characterization of reactor materials
• Atomistic scale modeling of radiation induced defects and structure evolution
• Small scale mechanical and thermal characterization of irradiated materials
• Advanced nuclear fuels
• Accident tolerant cladding and fuels materials
• Advanced structural materials
• Advances in manufacturing of reactor components
• Nanomaterials & Nanotechnology in nuclear engineering
• Aging and degradation studies
• Non-destructive evaluation
• Materials for radiation detection
• Theory, modeling and simulation of radiation effects
• Hybrid Energy: Nuclear and other energy sources
This Symposium is proposed by NETD-ACerS
This Symposium is co-sponsored by NACE International
Abstracts Due 03/31/2017
Proceedings Plan Undecided

Advanced Characterization of Highly Irradiated Nuclear Fuel from a Commercial LWR
Advanced ODS FeCrAlYZr Alloys for Accident-tolerant Fuel Cladding
Assessing the Irradiation-damage Tolerance of 316L after High-Strain-Rate Severe Plastic Deformation
Assessment of Aging Degradation Mechanisms of Alloy 709 for Sodium Fast Reactors
Assessment of Creep-fatigue Behavior and Damage Mechanisms of Alloy 709 Under Accelerated Conditions
Atom Probe Tomography and Transmission Electron Microscopy Investigations of Nano-precipitate Nucleation in ODS FeCrAl Alloys
Axial Temperature Uniformity and Diametrical Dilatometry Testing in the Gleeble 3500 to Study Phase Transformations in Zr-2.5Nb
Chemically-biased Defect Diffusion in Concentrated Solid-solution Alloys
Computational Studies of UO2, UN and Zr Materials for Pellet-cladding Interactions
Corrosion Mechanism of Incoloy 800H Alloy in High Temperature Thermal Energy Storage MgCl2-KCl Salts
Corrosion Resistance of Pure SiC and SiC-NFA Composite under High Temperature Water Vapor Conditions
Development and Qualification of a New Plate-type Low-enriched Uranium-molybdenum Fuel for High Power Research Reactors
Development of High Dose Radiation Tolerant Materials for Nuclear Applications
Development of the Advanced ODS 14YWT Ferritic Alloy for Radiation Tolerance
Dislocation Loop Dynamics in Irradiated FeCrAl Alloys
E-10: Effect of Irradiation Temperature on Microstructural Changes in Self-ion Irradiated Austenitic Stainless Steel
E-11: Experimental Characterization of the U3Si5-USi2 Phase Region
E-12: Neutron Irradiation Effect on 0.4t-CT Specimen of Alloy 690 Tested at Elevated Temperature
E-13: Recrystallization Behaviour of Nickel-based Alloys for Molten Salt Reactors
E-9: Effect of D+ Ion Irradiation on Ta under Fusion-relevant Conditions
Effect of Cascade Mixing on α' Precipitation in Irradiated Fe-Cr Alloys
Effects of Increasing Neutron Dose on Stability of MAX Phase Ti3AlC2/Ti3SiC2 Ceramics and MX/MA Impurity Phases
Evolution of Irradiation-induced Strain and Damage Recovery in Equiatomic NiFe Alloy
Evolution of Small Defect Clusters in Ion-irradiated 3C-SiC: Combined Cluster Dynamics Modeling and Experimental Study
Fretting Wear Behaviors of Surface-Modified Zr Cladding Supported by Pre-oxidized Spacer Grid
High Temperature Creep Behavior of Alloy 709
High Temperature Oxidation of SPS Sintered Cr3C2-coated SiC-NFA Composites in Water Vapor Containing Environment
In–situ High-Energy X-ray Characterization of Nuclear Reactor Materials
Increasing Nuclear Power Plant Safety with FeCrAl Cladding in Advanced Technology Fuel
Investigating Formation and Re-orientation of Multi-Phase Hydrides in Zirconium Nuclear Fuel Cladding by Phase Field Modeling
Investigations of Capacitor Discharge Welding for Attaching End Caps onto Molybdenum Fuel Rod Tubes
Irradiation-induced Damage Evolution in Concentrated Solid-solution Alloys
Microstructure Studies of U3Si2 Fuel Pellets Sintered in Argon vs. Vacuum Environment
Multiscale Irradiation Effects of Tungsten Based Materials as Plasma Facing Components
Nanodispersed Cu-Nb for Enhanced Radiation Tolerance
On Creep Deformation Behavior of Nb-containing Fe-Ni-Cr Austenitic Stainless Steel
On the 650℃ Thermostability of 9-12Cr Heat Resistant Steels Containing Different Precipitates
Oxidation and Corrosion Resistance of Zrn+1AlCn MAX Phases for Future Fission Environments
Powder Ageing and Sintering Of High Uranium Density Nuclear Fuels for Light Water Reactor Applications
Promise and Limitations of Ion Irradiations for Understanding High Dose Radiation Effects in Materials
Properties of Small Diameter SiC/SiC Composite Tubes for ATF Cladding Modeling and Design
Proton Irradiation of Pure Nickel
Pulse Electric Current Joining of Oxide-Dispersion-Strengthened Steels
Radiation Effects on SiC/SiC Composites for Advanced Accident Tolerant Fuel Cladding Tubes
Review of Technologies for Ocean Mining of Uranium
Small Scale Mechanical Testing on Materials for Nuclear Applications
Structural Characterization of Nanolayered Response in Select MAX Phase Ceramics to High Fluence Self-ion Irradiation
Structural, Chemical and Thermal Property Changes of Zirconium Diboride under Ion Beam Irradiation
Study of SPS Sintered NFA and NFA-SiC Cladding Materials under High Dose Self-ion Irradiation
Studying Synergistic Radiation Effects in TPBAR Materials with In-situ Triple Beam Irradiation TEM
Thermochemical Modeling of Candidate Accident Tolerant Fuel Systems
Thermodynamic Modeling and Compatibility of UN Fuel-FeCrAlY Cladding Materials
Tuning MoO3 Nanostructures Using Low Energy High Flux He+ Ion Irradiation
U3Si2 Instability in Both Oxidizing and Reducing Atmospheres
Understanding Chloride-induced Stress Corrosion Cracking Behavior of SS304 for Dry Storage Canisters for Spent Nuclear Fuels Storage
Westinghouse Accident Tolerant Fuel Materials

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