| About this Abstract |
| Meeting |
2023 TMS Annual Meeting & Exhibition
|
| Symposium
|
Microstructural, Mechanical and Chemical Behavior of Solid Nuclear Fuel and Fuel-cladding Interface
|
| Presentation Title |
Uncertainty Quantification of Thermal Performance of UO2 Fuel Pellets |
| Author(s) |
Robert Annewandter |
| On-Site Speaker (Planned) |
Robert Annewandter |
| Abstract Scope |
Predicting thermal performance of a UO2-Zr system at high discharge burn up poses a challenge for nuclear fuel performance codes if they rely on empirical models. For a well established nuclear fuel performance code (TRANSURANUS) a statistical approach is adopted to quantify uncertainties in thermal transfer from pellet to cladding arising at high discharge burn-up. The adopted uncertainty quantification is supported by MD simulations to understand the thermal performance of High Burnup Structures in the rim region, and to quantify the effect of stoichiometry on thermal conductivity. Uncertainty modelling will be used to target the most rapid reduction of uncertainty of fuel behaviour possible. |
| Proceedings Inclusion? |
Planned: |
| Keywords |
Nuclear Materials, Computational Materials Science & Engineering, Modeling and Simulation |