About this Abstract |
| Meeting |
2027 TMS Annual Meeting & Exhibition
|
| Symposium
|
Materials and Chemistry for Molten Salt Systems
|
| Presentation Title |
Tritium Testing in Flibe to Support Kairos Power Fluoride-Salt-Cooled High-Temperature Reactor Development |
| Author(s) |
Kieran Dolan, David Carpenter, Lin-wen Hu, Masashi Shimada, Micah Hackett |
| On-Site Speaker (Planned) |
Kieran Dolan |
| Abstract Scope |
Tritium produced by neutron activation of Flibe (2LiF-BeF2) is a predominant contributor to Fluoride-Salt-Cooled High-Temperature Reactor (FHR) hypothetical accident source term and normal operation effluents. Retention and release from the reactor system can be effectively managed through engineered solutions, but proper system design requires the ability to predict the plant-level distribution of tritium. While multiple tritium transport modeling frameworks have been developed in recent years, the remaining limitation towards their utility for FHR engineering applications is fundamental transport parameter uncertainty in the Flibe environment as well as a lack of representative experimental datasets for model benchmarks. This presentation will discuss two Kairos Power collaborations which address Flibe tritium testing knowledge gaps: 1) tritium gas permeation testing at Idaho National Laboratory’s Safety and Tritium Applied Research Facility and 2) Flibe neutron beamline irradiation testing at the MIT Nuclear Reactor Laboratory supported by the Nuclear Science User Facilities (NSUF) SuperRTE program. |
| Proceedings Inclusion? |
Planned: |
| Keywords |
Nuclear Materials, High-Temperature Materials, Energy Conversion and Storage |