| About this Abstract |
| Meeting |
2024 TMS Annual Meeting & Exhibition
|
| Symposium
|
Ceramics and Ceramic-based Composites for Nuclear Fission Applications
|
| Presentation Title |
Beryllium Carbide Tolerance to Radiation Damage for Advance Reactor Moderators |
| Author(s) |
Diego Alejandro Muzquiz, Stephen Raiman |
| On-Site Speaker (Planned) |
Diego Alejandro Muzquiz |
| Abstract Scope |
Graphite is a commonly proposed moderator for advanced reactors, including molten salt reactors, despite its low moderating cross section and dimensional instability under irradiation. Beryllium carbide (Be2C) is an attractive alternative to graphite moderators because of its high melting point, moderating efficiency, and environmental compatibility. However, its behavior under neutron irradiation is not yet known.
For this work, a new experiment was designed to safely irradiate beryllium-containing samples at The Michigan Ion Beam Laboratory. Using this new capability, Be2C samples were irradiated with 9 MeV carbon ions at varying doses from 5 dpa to 15 dpa, at temperatures up to 700℃. Samples were characterized with TEM to measure effects of irradiation on sample microstructures. This talk will present the results gathered from all experiments along with images from SEM and TEM to better understand Be2C response to radiation damage. Future irradiation experiments and material compatibility will also be discussed. |
| Proceedings Inclusion? |
Planned: |
| Keywords |
Ceramics, Nuclear Materials, High-Temperature Materials |