About this Abstract |
| Meeting |
2026 TMS Annual Meeting & Exhibition
|
| Symposium
|
Metallic Fuels - Design, Fabrication, and Characterization
|
| Presentation Title |
Thermal transport of uranium nitride (UN) after irradiation |
| Author(s) |
Zilong Hua, emma kindall, Narayan Poudel, Volodymyr Buturlim, Anshul Kamboj, Shuxiang Zhou, Daniel Murray, Kaustubh Bawane, Amey Khanolkar, Ella Pek, Md Minaruzzaman, Jennifer Watkins, Marat Khafizov, David Hurley |
| On-Site Speaker (Planned) |
Zilong Hua |
| Abstract Scope |
Nuclear fuel thermal conductivity is one of the most critical physical properties, which directly impacts reactor safety and efficiency. This presentation summarizes our recent investigation on thermal transport behavior of UN before and after proton irradiation. The low-dose post-irradiation samples show a rare thermal conductivity recovery. In addition, dislocation loops that are significantly larger than expectation were observed in the samples after irradiated at 800℃. Through a comprehensive analysis of phonon and electron scattering mechanisms, we investigated the defect generation and evolution within UN throughout the service life, and how they change the fuel’s thermal conductivity. |
| Proceedings Inclusion? |
Planned: |
| Keywords |
Nuclear Materials, Mechanical Properties, Characterization |