| 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 |
In-situ Irradiation of Uranium Carbide and Zirconium Carbide |
| Author(s) |
Rashed Almasri, Lingfeng He, Jian Gan , Adrian Wagner, Laura Hawkins, Wei-Ying Chen, Yuhan Li |
| On-Site Speaker (Planned) |
Rashed Almasri |
| Abstract Scope |
Uranium Carbide (UC) is a potential fuel with several advantages over oxide fuels. Also, Zirconium Carbide (ZrC) has a similar crystal structure and can be used as a surrogate. UC and ZrC were irradiated using the Intermediate Voltage Electron Microscope (IVEM) with Kr and Xe ions at temperatures up to 900 oC. In-situ TEM was used to show the dislocation loop nucleation and growth in the Kr-irradiated samples and the effect of irradiation temperature and dose on them. Additionally, the size and density of the Xe gas bubbles as functions of temperature and dose were visualized. The fuel swelling was calculated based on bubble size and density. The in-situ microstructures of these two carbides were compared to ion-irradiated oxide fuels. |
| Proceedings Inclusion? |
Planned: |
| Keywords |
Nuclear Materials, Ceramics, Characterization |