About this Abstract |
| Meeting |
2026 TMS Annual Meeting & Exhibition
|
| Symposium
|
Ceramics and Ceramic-Based Composites for Nuclear Applications III
|
| Presentation Title |
Preliminary Results from Post Irradiation Examination of AGR-5/6/7 TRISO Fuel |
| Author(s) |
William Cureton, Tyler Gerczak, Grant Helmreich |
| On-Site Speaker (Planned) |
William Cureton |
| Abstract Scope |
The US Advanced Gas Reactor Fuel Development and Qualification (AGR) Program has undertaken an effort to support TRISO fuel qualification through a series of fuel development and irradiation activities. The AGR Program team at Idaho National Laboratory and Oak Ridge National Laboratory (ORNL) are engaged in post-irradiation examination (PIE) and safety testing of compacts from the final planned irradiation experiment (AGR-5/6/7). Preliminary results on high temperature safety testing at temperatures at or above 1600C and destructive PIE performed at ORNL on several as-irradiated cylindrical fuel compacts are presented here. Emphasis is placed on fission product release behavior on the compact and particle scale as part of a multiscale approach. Results being obtained from AGR-5/6/7 test fuel through ongoing efforts further our understanding of how TRISO fuel responds under higher and lower irradiation temperature compared with previous AGR testing both under normal operating conditions and loss of pressurized coolant accident scenarios. |
| Proceedings Inclusion? |
Planned: |
| Keywords |
Nuclear Materials, |