About this Abstract |
| Meeting |
MS&T25: Materials Science & Technology
|
| Symposium
|
Metallic Nuclear Fuel Design, Fabrication and Characterization
|
| Presentation Title |
Thermal transport of uranium nitride (UN) after irradiation |
| Author(s) |
Zilong Hua, Emma Kindall, Md Minaruzzaman, Anshul Kamboj, Daniel Murray, Kaustubh Bawane, Amey Khanolkar, Ella Pek, 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 in pristine and irradiated states at Idaho National Laboratory. The low-dose post-irradiation samples show a rare thermal conductivity recovery. In addition, significantly-larger-than-normal dislocation loops were observed in the sample 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. |