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About this Symposium
Meeting 2017 TMS Annual Meeting & Exhibition
Symposium Ceramic Materials for Nuclear Energy Research and Applications
Sponsorship TMS Structural Materials Division
TMS: Advanced Characterization, Testing, and Simulation Committee
TMS: Energy Committee
TMS: Nuclear Materials Committee
Organizer(s) Xian-Ming Bai, Virginia Tech
Yongfeng Zhang, Idaho National Laboratory
Maria Okuniewski, Purdue University
Donna Guillen, Idaho National Laboratory
Marat Khafizov, Ohio State University
Thierry Wiss, European Commission- JRC -Institute of Transuranium Elements – Germany
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.

Topic 1: Experimental characterization of non-irradiated and irradiated oxide ceramics
Topic 2: Multi-scale modeling on microstructure evolution and physical properties in ceramics
Topic 3: Thermal-mechanical properties of oxides for nuclear energy
Topic 4: Non-oxide ceramics for nuclear energy
Topic 5: Nanostructured ceramics for nuclear energy (joint topic with "Nanostructured materials for nuclear applications II")
Abstracts Due 07/17/2016
Proceedings Plan Planned: Supplemental Proceedings volume
PRESENTATIONS APPROVED FOR THIS SYMPOSIUM INCLUDE

A TEM Study of Microstructure of Hi-Nicalon Type S SiC Composite beyond Ultimate Shear Strength
Alpha-damage Formation in Mixed Americium-uranium Compounds
Anisotropic Thermal Conductivity and Interface Resistance in Pyrolytic Carbon Coated Zirconia Particles
Atomistic Simulation of Swift Heavy Ion Irradiation Effects in UO2 and CeO2
Ceramic Materials for Nuclear Energy Research and Applications
Correlation Between Particle Size and Grain Size Distributions in Single/Multiphase Ceramic Oxide Surrogate Materials
Effect of Burn-up on the Thermal Conductivity of Fast Reactor MOX Fuel
Evaluation of Creep Behavior of UO2 at Sub-grain Length Scales
Evolution of Irradiation Defects in Ti2AlC Ceramics During Heavy Ion Irradiation
Five-dimensional Representation of Grain Boundary Energies in UO2
High Burn-up Nuclear Fuel, Impact of Fission Gases
Highlights of Ceramic Nuclear Fuel Research within the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program
In-Situ Tritium Measurements from γ-LiAlO2 Pellets Irradiated in TMIST-3A
In Situ Synchrotron Characterization of the Field Assisted Sintering of UO2
Ionization-Induced Damage Annihilation in Silicon Carbide
Irradiation-induced Recrystallization in UO2: A Phase Field Study
Irradiation Dependent Deformation and Thermal Properties of SiC and SiO2 Measured by Using Nanomechanical Raman Spectroscopy
Irradiation Effects on Electrochemical Performance of TiO2 Anode
Micro-Mechancial Interphase Property Evaluation for SiC-SiC Composites
Microstructural Characterization of the Processes, Stability, and End-of-Range Effects in Heavily Irradiated Pyrochlores
Modeling the Effect of Percolation on Fission Gas Release in UO2 Nuclear Fuels
Molecular Dynamics Simulations of Thermal Transport in Uranium Dioxide with Intrinsic Defects and Fission Products
Multi-scale Coupled Radiation Damage and Heat Transport Modeling for Dispersed Nuclear Fuels
Multi-scale Modeling of Fracture Behavior in SiC with a Phase Field Fracture Model
Neutron Irradiated SiC Advanced Analysis to Understand Fission Product Transport: Safety Tested TRISO Coated Particles
One-Dimensional String-like Relaxation in Actinide Oxides
Phase Field Modeling of Uranium Dioxide Sintering and Densification
Probing Oxygen Defects in Ion Irradiated Actinide and Analogue Oxides Using Neutron Total Scattering
Processing Routes for Improving Purity and Theoretical Density of UN Microspheres
Progress in Development of Non-Oxide Ceramic Nuclear Fuels
Radiation-Stability of Zirconium Carbide and Nitride Ceramics for Advanced Fuel Cycles
Radiation Damage on UO2 and UN
Role of Ion Species in Radiation Effects of Lu2Ti2O7
Sensitivity Analysis and Uncertainty Quantification of the MARMOT Mesoscale Fuel Performance Code
Study of Oxide Dispersion Strengthened 316L Austenitic Steel by Mechanical Milling
Study of Point and Extended Defects in Fluorite UO2 with Variable Charges Empirical Potentials
The Roles of Surfaces, Chemical Interfaces, and Disorder on Plutonium Incorporation in Pyrochlores
Theoretical and Experimental Investigation of the Interrelationship Between Radiation Damage and Ionic Transport in Pyrochlore
Thermal-Mechanical Properties of Sintered UO2
Thermal Transport Properties of Uranium Dioxide from Atomistic Simulations
Thermoelectric Properties of Doped and Pure UO2 at High Temperatures


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