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Meeting 2024 TMS Annual Meeting & Exhibition
Symposium Ceramics and Ceramic-based Composites for Nuclear Fission Applications
Sponsorship TMS Structural Materials Division
TMS: Materials Characterization Committee
TMS: Mechanical Behavior of Materials Committee
TMS: Nuclear Materials Committee
Organizer(s) Dong (Lilly) Liu, University of Oxford
Assel Aitkaliyeva, University of Florida
Anne A. Campbell, Oak Ridge National Laboratory
Konstantina Lambrinou, University of Huddersfield
Cynthia Adkins, Idaho National Laboratory
Scarlett Widgeon Paisner, Los Alamos National Laboratory
Scope Ceramics and ceramic-based composites play an important role in nuclear industry as they can be used to generate nuclear power and dispose of radioactive nuclear waste. For instance, nuclear graphite has been used widely in gas-cooled reactors, either in prismatic designs or pebble-bed configuration, as a fast-neutron moderator as well as structural components. Graphite composites and SiC-based ceramics are also used as matrix for TRISO fuels (pellets or pebbles). SiC ceramic-matrix composites in tubular shape, on the other hand, have been investigated as an alternative to conventional Zircaloy as accident tolerant fuels. Lastly, ceramics, such as borosilicate glass, are also adopted in the immobilization/storage of nuclear waste. The nano-/micro-structure and the thermal/mechanical properties of these materials evolve with irradiation in service, and it is critical to understand the underlying mechanisms via experimental and modelling methods. It is therefore essential that a symposium brings together experts/scientists across the world to share knowledge and experience on these materials to inspire novel and transformative ideas. The primary topics of interest to this symposium are:

• Fuels: UO2, UCO, MOX, and TRISO (stand-alone particles or embedded in graphite or SiC matrix)
• Nuclear graphite: reactor core components or as matrix material for TRISO
• Waste management: borosilicate glasses and other relevant materials
• Experimental characterization: microstructural evolution, degradation behaviours
• Properties: thermal and mechanical properties
• Modelling of ceramic degradation mechanisms and properties

Abstracts Due 07/15/2023
Proceedings Plan Planned:

Advanced Fuel for Integrating Nuclear SMRs with Renewables
Analyzing ZrN and LiF-doped ZrN as a Shield Material for Fusion and Space Reactors
Aspects of Graphite Performance in Molten Salt
Beryllium Carbide Tolerance to Radiation Damage for Advance Reactor Moderators
Correlating Atomic Structure with Elastic Properties in Non-textured Pyrocarbon
Elucidating the Effect of Radiation-induced Defect Accumulation on Swelling in UN Using In-situ TEM Irradiation
Evaluating the Strength of TRISO Pebbles via Drop Tests and Nondestructive Techniques
Exploring Hydrogen Absorption with High-density Fuels
From First-principles, Modeling the Effect of Point-defect Phonon Scattering on the Thermal Conductivity of Oxide Fuels
Impact of Grain Boundary and Surface Diffusion on Predicted Fission Gas Bubble Microstructural Evolution Behavior and Release in UO2 Nuclear Fuel
In Situ High-temperature 3D Imaging of the Damage Evolution in a SiC Nuclear Fuel Cladding Material
Mechanical and Thermophysical Properties of ZrC, NbC, and TaC Binary Carbide Surrogate Fuels for Nuclear Thermal Propulsion Systems
Microstructure, Mechanical Properties, and Residual Stresses of Cr-Coated SiC Fuel Cladding for Light Water Reactors
Multiphysics and Multiscale Modeling of the Mechanical Properties of the Porous Pyrocarbon Buffer Layer in TRISO Particle Fuel
Multiphysics and Multiscale Modeling of Micro- and Macro- cracking in High Burnup UO2 Fuels
Next Steps in TRISO Fuel Technology Development
Radially Resolved Thermo-physical Modelling in High Burnup Oxide Nuclear Fuel
Radiation Effects and Corrosion of Silicon Carbide in Nuclear Reactor Environments
Radiation Effects in Ceramics for Immobilization of Actinide-containing Nuclear Waste
Reduced Processing Temperature of Advanced Ceramic Composites
Response of the U3Si2 + 50 wt%UB2 Composite Alloyed with Al, Al2O3, Y and Y2O3 in High-Temperature Oxidizing Atmospheres
Steam Corrosion of Cr- and Zr-containing Uranium Nitride Fuels: Mechanistic Insights from In Situ Neutron Diffraction
Thermal Conductivity Suppression in Uranium-doped Thorium Dioxide Due to Phonon Resonant Scattering
Thermo-mechanical Characterization of a Novel Alumina-Yttria Ceramic Coating for Lead Fast Reactors
Thermodynamics of Complex Carbides for Nuclear Applications in Extreme Environments
Thermophysical Properties of Solid Solution Carbide Fuels for Nuclear Thermal Propulsion
Thermophysical Properties of Uranium Nitride-metal Composite Fuels
Unraveling the Influence of Charge Effect on Defect Recombination in ThO2
Uranium Enrichment Homogeneity Study on HALEU/LEU UO2 Fuel Pellets
Using In Situ Neutron Powder Diffraction to Study the Thermal Expansion of Fission Product Doped UN

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