| About this Abstract |
| Meeting |
2024 TMS Annual Meeting & Exhibition
|
| Symposium
|
Ceramics and Ceramic-based Composites for Nuclear Fission Applications
|
| Presentation Title |
Aspects of Graphite Performance in Molten Salt |
| Author(s) |
Raluca Scarlat |
| On-Site Speaker (Planned) |
Raluca Scarlat |
| Abstract Scope |
Graphite and TRISO fuel graphite matrix material are employed as structural materials and moderators. These materials perform structural functions, heat transfer functions, moderator and reflector functions during reactor operation, in some cases they serve for radioisotope retention, during accident scenarios and ultimate disposal. These functions need to be maintained with irradiation and while immersed in the molten salt and the cover gas above the molten salt, at temperatures from room temperature to 700 oC and above. A large body of knowledge is already available for high temperature performance of graphite in gas-cooled reactors, and a smaller body of knowledge is available for high temperature performance of graphite in salt-cooled and liquid-fueled reactors. This talk will provide examples of graphite behavior that is specific to its operation in molten salt: tritium uptake from molten salt, wetting by molten salt, tribology and wear under molten salt. |
| Proceedings Inclusion? |
Planned: |
| Keywords |
Nuclear Materials, High-Temperature Materials, Pyrometallurgy |