| About this Abstract |
| Meeting |
2025 TMS Annual Meeting & Exhibition
|
| Symposium
|
Microstructural, Mechanical, and Chemical Behavior of Solid Nuclear Fuel and Fuel-Cladding Interface II
|
| Presentation Title |
High Temperature Creep Behavior of Uranium Oxide: Crystal Plasticity Modeling and Experiments |
| Author(s) |
Veerappan Prithivirajan, Deepali Patil, Dewen Yushu, Cameron Howard, Sudipta Biswas, Lingfeng He |
| On-Site Speaker (Planned) |
Veerappan Prithivirajan |
| Abstract Scope |
Uranium dioxide (UO₂), used as fuel in light water reactors, is subjected to high temperature, irradiation, and mechanical loading during service. Understanding the high-temperature mechanical behavior of UO₂ is essential for ensuring the structural integrity and safety of the reactors. This study aims to elucidate the high-temperature creep behavior of UO₂ through a comprehensive modeling and experimental approach.
We developed a dislocation density-based crystal plasticity (CP) model that is both temperature- and strain-rate-dependent, within the MOOSE framework. Micropillar compression tests were conducted on single crystal pillars across a range of temperatures to calibrate the single crystal constants and the CP model parameters. Additionally, small-scale creep testing was performed on polycrystalline samples to validate the CP model predictions. This model provides a pathway to understanding the complex material response of UO₂ under varied loading conditions, as well as phenomena such as the formation of high burnup structures of UO₂. |
| Proceedings Inclusion? |
Planned: |
| Keywords |
Computational Materials Science & Engineering, Modeling and Simulation, Other |