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Meeting 2024 TMS Annual Meeting & Exhibition
Symposium Ceramics and Ceramic-based Composites for Nuclear Fission Applications
Presentation Title Elucidating the Effect of Radiation-induced Defect Accumulation on Swelling in UN Using In-situ TEM Irradiation
Author(s) Maria Kosmidou, Adrien J. E. Terricabras, Caitlin A. Kohnert, Joshua T. White, Erofili Kardoulaki
On-Site Speaker (Planned) Maria Kosmidou
Abstract Scope Uranium mononitride (UN) is under consideration as an advanced nuclear fuel alternative to UO2. During irradiation, the fuel swells due to the formation and growth of fission gas bubbles. A key limiting factor for its in-reactor application is runaway swelling; a phenomenon where at high temperatures (and burnups) changes to the mobility of defects combined with microstructural changes cause the swelling of the fuel to rapidly accelerate. It is thought that the accumulation of irradiation-induced defects and fission gas play a key role. Irradiations in test reactors are time intensive and costly due to handling hazardous irradiated nuclear material. In-situ TEM irradiations afford accelerated irradiation and damage evolution in nuclear materials to expedite and safely study irradiation damage in novel systems. In this work, we aim to in-situ study the nucleation and evolution of defects (e.g. dislocation loops, cavities) in UN by Kr ion irradiations at reactor relevant temperatures.
Proceedings Inclusion? Planned:

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

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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