Tritium retention in nuclear graphite, system-level transport, and management strategies for the fluoride-salt-cooled high-temperature reactor

Advanced reactor concepts which use a lithium- or beryllium-bearing primary salt coolant will require technical solutions to mitigate the environmental release of tritium. One such design is the Fluoride-Salt-Cooled High-Temperature Reactor (FHR), which combines a molten Flibe (2LiF-BeF₂) salt coolant and tri-structural isotropic coated-particle fuel to produce power or process heat. Compared to current water-cooled reactors, managing tritium release from a FHR is further complicated by the mobility of tritium at high temperatures and limited knowledge of interactions between tritium and nuclear graphite in the molten fluoride salt environment. The total activity, chemical forms, and retention mechanisms for tritium in nuclear graphite were studied through thermal desorption analysis of sample materials from three in-core Flibe salt irradiations (denoted FS-1, FS-2, and FS-3) at the MIT Reactor (MITR).

Name thesis:degree_name

Doctoral

Department dc:contributor.department

Massachusetts Institute of Technology. Department of Nuclear Science and Engineering

Grantor dc:publisher

Massachusetts Institute of Technology

Year dc:date.issued

2021

Author dc:creator

• Dolan, Kieran Patrick.

Advisor dc:contributor.advisor

• Lin-wen Hu.

• Nuclear Science and Engineering.

Statement dc:rights

• MIT theses may be protected by copyright. Please reuse MIT thesis content according to the MIT Libraries Permissions Policy, which is available through the URL provided.

Licence dc:rights.uri

• http://dspace.mit.edu/handle/1721.1/7582

Language dc:language.iso

eng

Handle dc:identifier.uri

https://hdl.handle.net/1721.1/131004

OAI identifier oai:identifier

oai:dspace.mit.edu:1721.1/131004