| About this Abstract |
| 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 |