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Meeting 2023 TMS Annual Meeting & Exhibition
Symposium Ceramic Materials for Nuclear Energy Research and Applications
Presentation Title Atomistic-scale simulations used to simulate creep in oxide fuel
Author(s) Conor Galvin, Aritra Chakraborty, Laurent Capolungo, David Andersson, Michael Cooper
On-Site Speaker (Planned) Conor Galvin
Abstract Scope Doped UO2, which has been doped to produce a larger grain size than conventional UO2, has been proposed as an advanced fuel candidate. The larger grain sizes provide improved operational fuel behavior for fission gas retention and pellet-cladding interactions due to improved mechanical properties. One such property is creep. Using molecular dynamics, we predict information at the atomistic scale that can be used to develop a mechanistic UO2 creep model for use in longer time/length-scale codes. The ultimate objective of the model is to better describe the grain size dependence and impact of doping. Previously, we found that Nabarro-Herring creep was too low to capture the experimentally observed creep rates. Therefore, other mechanisms which have distinct grain size dependences have been explored. In this work, we investigate the concentration, segregation and diffusivity of various defects at grain boundaries in UO2 to explore their impact on Coble creep.
Proceedings Inclusion? Planned:
Keywords Nuclear Materials,

OTHER PAPERS PLANNED FOR THIS SYMPOSIUM

Advanced Characterization and Modeling of Nanoprecipitates in Spent Nuclear Fuel
Atomic scale simulation of amorphous intergranular films in nuclear fuel materials
Atomistic-scale simulations used to simulate creep in oxide fuel
Atomistic and mesoscale modeling of fission gas and fission products diffusivity in TRISO fuel kernels
Atomistic investigation of radiation-induced defects in ThO2
Cluster dynamics modeling of defects and fission gas in Gd doped UO2 under irradiation
Comparison of the electronic transport of UN and ThN versus ThC
Comprehensive characterization of damage in ion irradiated ceramics for validation of atomistic models
Defect chemistry and radiation stability of (Gd & Zr) co-doped UO2
Development of UC/UO2 Composite Fuels for Light Water Reactors
Diffusion properties in uranium-plutonium mixed oxides: atomic scale investigation of the effect of composition and chemical disorder
Emulation of Microstructures and Tritium Behavior in Lithium Aluminate by Ion Irradiation
Hidden defect evolution mechanism in ThO2 revealed by atomistic modeling
High-Entropy Carbide Ceramics: New Materials for Extreme Environments in Nuclear Energy Applications
Impact of Resonance Scattering on the Thermal Conductivity of ThO2
Improving uranium mononitride behaviour using uranium diboride addition
Irradiation- and dopant-induced structural changes in ceramic nuclear fuels probed via elastic and optical properties
Low-temperature fabrication of ceramic tritium breeder materials, for enhanced control of microstructure and phase formation
Microstructural characterization of neutron irradiated concrete minerals
Microstructural, Mechanical and Thermal Characterization of High Entropy Carbide Ceramics
Modelling the melting temperature of CrUO4 to assess its behaviour during the sintering of Cr-doped UO2
Multiphysics Modeling of High Burnup UO2 at Mesoscale
Multiscale Modeling for High-burnup Structure Formation in UO2
Oxidation Behavior and Mechanisms of the SiC Coating in TRISO Fuel Particles
Phase equilibria and thermodynamics of tri-carbide fuels for nuclear thermal propulsion
Predicting mechanical behavior of Uranium oxide fuel pellets using full-field defect diffusion modeling in a crystal plasticity framework
Quantifying irradiation-induced defects in SiC and WC through stored energy measurements of radiation damage
Quantifying the impact of fast interface diffusion and free surface evolution on fission gas release in UO2 using a phase-field model
Radiation Damage in Lithium Oxide, a Surrogate for Beryllium Carbide
Radiation damage of ion-irradiated high entropy ceramics
Radiation shielding ceramics with enhanced performance and scalability
Radiation studies on the TiZrNbHfTa high entropy alloy and its hydrides.
Relating Microstructural Evolution and Stoichiometry to Tritium Release from Ternary Lithium Ceramics
Revealing The Microstructure and Irradiation Effects on UO2 Fracture Via Coupled Phase-Field and MD Simulations Approach
Scanning Transmission Electron Microscopy of Nanoprecipitates in Spent UO2 Nuclear Fuel
Silica Formation on SiC Following Steam Attack
Simulation of irradiation-induced bubble over-pressurization and application in fuel performance
Soft X-ray Synchrotron Radiation Spectromicroscopy of Spent Nuclear Fuel Focused Ion Beam Sections
Surface Modification Strategies for Hydrogen Retention in Hydride Moderators
Susceptibility of Nuclear Fuel Ceramics to Oxidation and Hydridization during Off Normal Events
Thermomechanical characterization of advanced reactor alloys and composites exposed to high-temperature gas environments
Uranium Silicide Processing for Advanced Reactors
Zirconia-coated uranic fuel particles processing and in situ sintering characterisation

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