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Meeting 2021 TMS Annual Meeting & Exhibition
Symposium Ceramic Materials for Nuclear Energy Research and Applications
Sponsorship TMS Extraction and Processing Division
TMS Structural Materials Division
TMS Light Metals Division
TMS: Advanced Characterization, Testing, and Simulation Committee
TMS: Energy Committee
TMS: Nuclear Materials Committee
Organizer(s) Xian-Ming Bai, Virginia Polytechnic Institute and State University
Yongfeng Zhang, University of Wisconsin
Larry Aagesen, Idaho National Laboratory
Vincenzo V. Rondinella, Jrc-Ec
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. The topics of interest include but are not limited to: 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.
Abstracts Due 07/20/2020
Proceedings Plan Planned:

A Model of Grain Boundary Energy Anisotropy in Uranium Dioxide Nuclear Fuel
A Thermo-mechanical Coupled Phase Field Dynamic Fracture Model and Its Application in UO2
Comprehensive Treatment of Thermal Transport Under Irradiation in ThO2
Development of Hydrothermal Corrosion Barrier Coatings for High-density Nuclear Fuels
Development of Yttrium Hydride for High Temperature Moderator Application
Effect of UV and Gamma Irradiation on the Hydrothermal Corrosion of Ion-irradiated SiC
Electron Microscopy Characterization of the Fuel-cladding Interaction in Annular Fast Reactor MOX
Evolution of Ion Irradiated Nitride Ceramics Properties for Coated Particle Fuel Systems
Exotic Magneto-elastic Properties in Uranium Dioxide
Hydrothermal Corrosion of Silicon Carbide
Hydrothermal Corrosion Study of Additive Manufactured SiC Fibers
Impact of Dislocation Loops on Thermal Conductivity of CeO2
In-situ Measurement of Tritium Release from Lithium Aluminate Under Neutron Irradiation
Influence of Dose Rate and Temperature on Mass Transport in Hematite
Ionization Effects on Damage Accumulation Behavior in SiC
Irradiation Damage in High-entropy Carbide Ceramics
Irradiation Effects on Zirconium Alloy Oxides and Their Impacts on In-reactor Corrosion Rates
Microstructural Analysis and Micro-mechanical Testing on Xenon-irradiated Uranium Dioxide
Microstructural and Fission Products Analysis from Irradiated UO2 Fuel Using Atom Probe Tomography
Microstructural Characterization of Radiation Effects in 3D printed SiC
Microstructure and Chemical States of Fission Products in Irradiated AGR-1 and AGR-2 TRISO Particle UCO Fuel Kernels
Modeling of Pressure-driven Inter-granular Fracture in High Burnup Structure UO2 during LOCA Using A Phase-field Approach
Molecular Dynamics Investigations of AlN-based Piezoelectric Ceramics under Irradiation
Multiscale Modeling of High Burn-up Structure (HBS) Formation and Evolution in UO2
On the Role of Neutron Irradiation Damages on Fission Products Transport in the SiC Layer of TRISO Fuel Particles
Oxidation Behavior of TRISO Fuel Materials
Phase-field Modeling of Bubble Growth During High Burn-up Structure Formation in UO2
Radiation Tolerance of Nanoporous Gadolinium Titanate
Radiolytic Damage and Hydrogen Generation at Carbide – Water Interfaces
TEM Characterization of Dislocation Loops in Ion-irradiated Single Crystal ThO2
TMIST-3A Post-irradiation Examination
Towards a Model of Coupled Irradiation and Corrosion

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