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Meeting 2023 TMS Annual Meeting & Exhibition
Symposium Ceramic Materials for Nuclear Energy Research and Applications
Presentation Title Predicting Mechanical Behavior of Uranium Oxide Fuel Pellets Using Full-field Defect Diffusion Modeling in a Crystal Plasticity Framework
Author(s) Aritra Chakraborty, Conor Oscar Galvin, Michael W.D. Cooper, Laurent Capolungo
On-Site Speaker (Planned) Aritra Chakraborty
Abstract Scope Uranium-oxide (UO2) presents a great technological interest as a nuclear fuel for pressurized water reactors. Under the operating conditions of high temperatures and low-to-moderate stresses, several microstructural changes (formation of hydrogen, noble gases, etc.) can occur leading to diffusion of multiple species in the fuel pellets. This work aims to quantify the contribution from these diffusion mediated processes on the overall mechanical behavior of these pellets. As a first step, through a coupled chemo-mechanical model in a crystal plasticity framework, we predict creep response for UO2 fuel pellets considering the local defect concentration, grain size, and stoichiometry. The underlying crystal plasticity framework also accounts for the plasticity due to dislocation glide and climb— affected by the local dislocation density and defect concentration. With such full-field models local hot spots of vacancy supersaturation can be identified, acting as potential sites for damage nucleation, thus capturing failure in these systems.
Proceedings Inclusion? Planned:
Keywords Ceramics, Nuclear Materials, Modeling and Simulation

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
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
Exploring Irradiation-induced Phase Evolution in WC
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
O-1: Uranium Silicide Processing for Advanced Reactors
O-31: Radiation Damage in Lithium Oxide, a Surrogate for Beryllium Carbide
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 the Impact of Fast Interface Diffusion and Free Surface Evolution on Fission Gas Release in UO2 Using a Phase-field Model
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
Zirconia-coated Uranic Fuel Particles Processing and In Situ Sintering Characterisation

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