Tritium thermal desorption testing of nuclear graphites irradiated at fluoride-salt-cooled high-temperature reactor conditions

The Fluoride-Salt-Cooled High-Temperature Reactor (FHR) is a next-generation nuclear plant design that combines successfully demonstrated technologies from other advanced reactor concepts such as tristructural isotropic (TRISO) coated particle fuel, molten flibe salt (LiF-BeF2) coolant, and an Air-Brayton power cycle. A prominent technical challenge for the FHR is maintaining the release of tritium generated from neutron irradiation of flibe below acceptable levels. One proposed method for partitioning tritium from the salt is through adsorption onto graphite. Demonstrating the viability of this type of tritium control system requires further experimental investigation since few studies have examined the combined effect of molten flibe, tritium, and graphite at relevant FHR temperatures. Studying tritium transport experimentally has been recently enabled by three in-core fluoride salt irradiations completed at the Massachusetts Institute of Technology Nuclear Reactor Laboratory.

Name thesis:degree_name

Master

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

2018

Author dc:creator

• Dolan, Kieran Patrick.

Advisor dc:contributor.advisor

• Lin-wen Hu and David M. Carpenter.

• Nuclear Science and Engineering.

Statement dc:rights

• MIT theses are protected by copyright. They may be viewed, downloaded, or printed from this source but further reproduction or distribution in any format is prohibited without written permission.

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/1721.1/122905

OAI identifier oai:identifier

oai:dspace.mit.edu:1721.1/122905