| About this Abstract |
| Meeting |
2022 TMS Annual Meeting & Exhibition
|
| Symposium
|
Advanced Characterization and Modeling of Nuclear Fuels: Microstructure, Thermo-physical Properties
|
| Presentation Title |
Thermal Transport Behavior of Pristine and Zirconium-doped Alpha-Uranium |
| Author(s) |
Zilong Hua, David Hurley |
| On-Site Speaker (Planned) |
Zilong Hua |
| Abstract Scope |
The U-Zr based metallic fuel is a promising candidate for the next generation of fast reactors. The structure-induced thermal anisotropy of alpha-U has been reported, but accurate knowledge and its dependence on temperature is still lacking. In this talk, recent research results of thermal conductivity in pristine alpha-U are reported. Experimental data in a wide temperature range are compared with DFT and MD simulation to investigate the transport mechanisms. In addition, the influence from different mass percentages of Zr-doping is studied to validate the modeling work that can be used to predict the in-reactor performances with irradiation-induced point defects. These new insights are expected to aid in fuel performance modeling and fuel design. |
| Proceedings Inclusion? |
Planned: |
| Keywords |
Mechanical Properties, Nuclear Materials, |