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Meeting Materials Science & Technology 2019
Symposium Materials for Nuclear Applications
Organizer(s) Philip Edmondson, Oak Ridge National Laboratory
Yutai Katoh, Oak Ridge National Laboratory
Jake W. Amoroso, Savannah River National Laboratory
Levi Gardner, University of Utah
Amy Gandy, University of Sheffield
Karl Whittle, University of Liverpool
Monica Ferraris, Politecnico di Torino
Scope Rising energy demand has stimulated the design and global expansion of nuclear technologies. This expansion requires the development and understanding of novel materials that are capable of withstanding the extreme conditions associated with nuclear power generation. Knowledge of a material’s performance in environments subject to elevated temperature and pressure, high mechanical stress, extensive radiation dose, and highly corrosive environments is paramount to designing safe and efficient next generation nuclear technologies. This symposium will address a wide range of topics from fundamental and applied science aspects of nuclear materials for fission and fusion reactors to radiation detector performance in extreme conditions. Abstracts related to both advanced nuclear reactors and the lifetime extension of existing reactors will also be considered and are encouraged.
This symposium is sponsored by the Nuclear and Environmental Technology, and Engineering Ceramics Divisions of the American Ceramic Society.
Abstracts Due 04/05/2019
Proceedings Plan Definite: At-meeting proceedings

A Metallic Multilayer Composite for Use in Fluoride Molten Salt Reactors
Aging Behavior and Microstructure Evolution in Ni-Cr-Mo-W (Haynes 244) Alloy After Surface Treatment by Laser Shock Peening (LSP)
Anisotropic thermal transport in Uranium dioxide induced by dislocation
Anisotropy in thermal creep and creep life prediction of Zr-2.5%Nb pressure tube alloy
Characterization and Oxidation of Graphite and Silicon Carbide in TRISO Nuclear Fuel
Characterization of the impact of fission product inclusion on phase development in U3Si2 fuel
Combined Use of In-situ and Ex-situ TEM to Characterize Irradiation Induced Dislocation Loops in F/M Steels for Nuclear Applications
Comparison of in-situ micro- and ex-situ meso-scale tensile testing for the evaluation of mechanical properties of stainless steels
Compatibility of U3Si2 fuel with FeCrAl and SiC/SiC Based Cladding
Computational Studies of Environmental Degradation of Silicon Carbide
Corrosion Behavior of Metallic Alloys in a Molten Chloride Eutectic Salt for Nuclear Applications
Corrosion Behavior of Nanostructured Stainless Steels and High Entropy Alloys
Design of Alloy Chemistry to Mitigate Fuel-Cladding Chemical Interactions in Uranium-based Metallic Fuels
Development and Performance of High Temperature Irradiation Resistant Thermocouples
Effect of Ultrasonic Nanocrystalline Surface Modification (UNSM) on the Oxidation Behavior of Alloy 800HT in a Supercritical Carbon Dioxide (SCO2) Environment
Effects of Deposition Conditions on the Production of ZrO2 Coatings Produced by PE-CVD as Environmental Barrier Coatings for the Molten Salt Reactor
Effects of Ti and Al Additions on Irradiation Behavior of FeMnNiCr Based High-Entropy Alloys
Enhancing the Properties of a Cast FeNiMnCr10 Co-free High-entropy Alloy Through Hot Rolling
Evolution of grain boundaries in irradiated ceramics
Evolution of Microstructure, Deformation Mechanisms, and Internal Damage During Creep-Fatigue Testing of Alloy 709 (Fe-20Cr-25Ni)
Exploring TRISO layer properties and performance for multiple reactor concepts
Fundamental Studies of Tritium Diffusivity in Irradiation Defective Lithium Zirconate Pellets: A Frist Principles Density Function Theory Study
High-Entropy Carbide Ceramics for Extreme Environments
In-Situ Characterization of Zirconium Alloy Degradation to Support Nuclear Sensing Applications
In-situ Ion Irradiation Study of Silicon Carbide-Carbon Coated Nanostructured Ferritic Alloy
Investigation of Recrystallization and Grain Growth in Uranium 10 wt% Molybdenum
Materials for Capture of Uranium for Nuclear Fuel from Fertilizer
Multi-dimensional, In-Situ Mechanical Testing to Evaluate Damage and Fracture of Chromium-Coated Zirconium-based Fuel Claddings
Oxidation of TRISO particles and Matrix Graphite in Mixed Gas Atmospheres
Predicting Concrete’s Response to Irradiation
PVD Coating of Surrogate Fuels for Deep Space Nuclear Thermal Propulsion
Rapid Multiscale Simulation of Cladding Performance: Application to HT-9
Spectral thermal conductivity predictions in UO2 with Xe inclusions
Steam Oxidation and Microstructural Characterization of U3Si2 alloyed with Al, Cr, Nb, Y, and Zr
Techniques for In-Situ Monitoring of Materials Degradation in VTR Cartridge Loop Environments
Temperature Impacts on Damage Response in Mixed Carbides
The Effects of Ultrasonic Nanocrystal Surface Modification at Room Temperature and Elevated Temperatures on Residual Stress, Microstructure and Mechanical Properties of Nuclear Alloys IN600 and IN690. Harsha Venkat Sai Naralasetty1, Auezhan Amanov2, Young Sik Pyoun2, Jie Song1, Nicholas Mohr3, Seetha R. Mannava1 and Vijay K. Vasudevan1 1)Department of Mechanical and Materials Engineering, University of Cincinnati, Cincinnati, OH, USA. 2)Department of Mechanical Engineering, Sun Moon University, Asan, South Korea. 3)Electric Power Research Institute, Charlotte, NC, USA. Email address:
Thermophysical Properties of Binary Cl & F Compositions for Next Generation Molten Salt Reactors
Transmutation-induced precipitation in tungsten irradiated with a mixed energy neutron spectrum
Unveiling SiC/SiC CMC Cladding Failure Mechanisms and Hermetic Performance with In-situ 3D-Digital Image Correlation
Uranium Nitride and High Temperature Irradiation Resistant Thermocouples towards Accident Tolerant Nuclear Fuel
Using ACRT with MVB Furnace to Achieve Low-cost CZT

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