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Meeting 2019 TMS Annual Meeting & Exhibition
Symposium Ceramic Materials for Nuclear Energy Research and Applications
Sponsorship TMS: Nuclear Materials Committee
Organizer(s) Yongfeng Zhang, Idaho National Laboratory
Xian-ming (David) Bai, Virginia polytechnic Institute and State University
David Andersson, Los Alamos National Laboratory
Thierry Wiss, European Commission- JRC -Institute of Transuranium Elements
Scope Nuclear energy is an essential element of a clean energy strategy, avoiding greenhouse gas emissions of over two billion tons per year. Ceramic materials play a critical role in nuclear energy research and applications. Nuclear fuels, such as uranium dioxide (UO2) and mixed oxide (MOX) fuels, have been widely used in current light water reactors (LWRs) to produce about 15% of the electricity in the world. Silicon carbide (SiC) is a promising accident-tolerant cladding material and is under active research studies. Some oxide ceramics have been proposed for novel inert matrix fuels or have been extensively studied as waste forms for the immobilization of nuclear waste. Moreover, ceramics are under active studies for fusion reactor research.

This symposium focuses on experimental and computational studies of ceramics for nuclear energy research and applications. Both practical reactor materials and surrogate materials are of interest. Topics of interest include: defect production and evolution; mobility, dissolution, and precipitation of solid, volatile, and gaseous fission products; changes in various properties (e.g., thermal conductivity, volume swelling, mechanical properties) induced by microstructural evolution; and radiation-induced phase changes. Experimental studies using various advanced characterization techniques for characterizing radiation effects in ceramics are of particular interest. The irradiation techniques such as laboratory ion beam accelerators, research and test reactors, as well as commercial nuclear power reactors are all of interest. Computational studies across different scales from atomistic to the continuum are all welcome. Contributions focused on novel fuels such as doped UO2, high-density uranium fuels like uranium nitrides and silicides, and coatings for accident-tolerant fuel claddings are also encouraged. This symposium is intended to bring together national laboratory, university, and nuclear industry researchers from around the world to discuss the current understanding of the radiation response of ceramics through experiment, theory, and multi-scale modeling.
Some focused topic areas will be:
" Experimental characterization of non-irradiated and irradiated oxide ceramics
" Multi-scale modeling on microstructure evolution and physical properties in ceramics
" Thermal-mechanical properties of oxides for nuclear energy
" Non-oxide ceramics for nuclear energy
Abstracts Due 07/16/2018
Proceedings Plan Planned: Supplemental Proceedings volume
PRESENTATIONS APPROVED FOR THIS SYMPOSIUM INCLUDE

A Three-degree-of-freedom Representation of the Five-degree-of-freedom Grain Boundary Energy Space for Uranium Dioxide
A Model of Fission Gas Release and Swelling in UO2 for Engineering Fuel Analysis
An Engineering Representation of the Thermal Conductivity of a UO2 and BeO Composite Nuclear Fuel
Assessment of UO2 Based Composites Fabricated via SPS
Atomic Structure of Overstoichiometric Uranium Oxide: Insights from Molecular Dynamics Simulations with a Many Body Variable Charge Model
Characterization of Defects Structures in Fast-reactor MOX Fuels
Characterization of Intragranular Creep Deformation in Uranium Dioxide: A Multicrystal Approach
Characterization of the Hydrothermal Corrosion Behavior of SiC With and Without Corrosion Mitigation Coatings
Characterization of the Irradiation Effects in Nuclear Graphite
Characterization of U-Si Accident-tolerant Fuels Using Neutron Imaging and Diffraction
Computational Studies of Environmental Degradation of Silicon Carbide
Crystallographic and Electronic Structure in Ln-U-O Compounds
Defects and Microstructure Evolution in Oxides under Irradiation
Dynamic Structures Resulting from Ion Radiation Interactions with Porous Ceramics
Effects of Different Cation Doping on the Physical Properties of Gd2Zr2O7 Pyrochlores
Effects of Electronic Energy Loss on Irradiation Damage Production and Evolution in Ceramics
First Principles Prediction of Thermal Conductivity in Irradiated LiAlO2
Fouling Resistant, Foulant-agnostic Coatings for Nuclear Reactors and Geothermal Systems
High Density Uranium Silicide Fuels – Fabrication and Oxidation Resistance
Influence of the Miscibility Gap in the Evolution of the Microstructure in UO2-based Fuel Doped with Nd
Irradiation Effects on Nuclear Fuel
Irradiation Effects on Reactor Concrete Structures
Mechanisms for Diffusion of Uranium Interstitials in UO2
Mechanistic Mesoscale Simulation of UO2 Sintering
Mesoscale Modeling of Grain Growth in Ceramics
Microstructural and Micro-chemical Comparisons of AGR-1 and AGR-2 TRISO UCO Fuel Kernels Irradiated in the Advanced Test Reactor
Microstructural Characterization of Transmutation Nitride Fuels for Fast Reactors
Microstructural Effects on the High-temperature Oxidation Resistance of Magnetron Sputtered Cr-Al-Si-N Coatings on Zirconium Substrates
Multi-scale Modeling of Fission Gas Release in UO2 Nuclear Fuel
Nanostructured Ferritic Alloy-silicon Carbide Composites for Nuclear Applications
Neutron Irradiation Performance of Chemical Vapor Deposited SiC Fuel Systems at High Temperatures and Burnups
Phonon-based Lattice Thermal Conductivity of Uranium Dioxide
Probing the Thermodynamic and Kinetic Factors Leading to the Development of High Burnup Structure in UO2
Radiation Damage Studies in Plutonium Containing Ceramics
Radiation Effects on SiC/SiC Composites for Nuclear Energy Application
Radiation Tolerance and Helium Swelling Resistance in Amorphous SiOC
Revealing Anisotropic Swelling Trends in Irradiated Hexagonal/Trigonal Materials
Revisiting the Diffusion Mechanism of Helium in UO2: A DFT+U Study
Role of Grain Orientation and Grain Boundary Inclination during Sintering of UO2: A Phase-field Study
SiC-SiC Fiber Composites for Accident-tolerant Fuel Applications: Micromechanical Study of Radiation and Temperature Effects
Strength-suctility-irradiation Tolerance of Nanostructured Fe – Amorphous Ceramic SiOC Composites
Structural Features in Mixed Uranium Oxides with Fluorite-related Structures
Summary of In-situ Tritium Measurements from TMIST-3A
The Role of Dopant Charge State on Defect Chemistry and Grain Growth of Doped UO2
Thermochemical Investigation of (Fe,Cr,Al)3O4 Spinels
Uranium Silicide-based Nuclear Fuel Phase Relations and Computed In-reactor Thermochemical Behavior
Visualizing Stress Distribution of Irradiated and Corroded SiC Using Nano-mechanical Raman Spectroscopy
Void Dynamics in Porous Thin Films under Ion Irradiation
Water Corrosion Resistance of Modified U3Si2


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