| About this Abstract |
| Meeting |
2025 TMS Annual Meeting & Exhibition
|
| Symposium
|
Composite Materials for Nuclear Applications III
|
| Presentation Title |
Irradiation tolerance and molten salt compatibility of beryllium carbide – a candidate high temperature moderator material. |
| 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 (MSRs), 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, information of its behavior under MSR conditions is limited. For this work, novel methods were developed to safely irradiate beryllium containing samples at the Michigan Ion Beam Laboratory. Using this new capability, Be2C samples were irradiated at varying doses from 2 dpa to 30 dpa, at temperatures up to 500˚C. Along with ion beam irradiations, Be2C samples were exposed to FLiBe salt alongside graphite to assess their salt compatibility. This talk will present the results gathered from all experiments along with images from SEM and TEM to better understand Be2C performance as a moderator in MSR conditions. |
| Proceedings Inclusion? |
Planned: |
| Keywords |
Ceramics, Characterization, Nuclear Materials |