Modeling Tritium Retention in Graphite for Fluoride-Salt-Cooled High-Temperature Reactors. (3rd October 2021)

• Record Type: Journal Article
• Authors: Dolan, Kieran; Huang, Steven; Hackett, Micah; Hu, Lin-Wen
• Is Part Of: Nuclear technology. Volume 207:Number 10(2021)
• Page Start: 1578
• Page End: 1598
• Publication Date: 2021-10-03
• Subjects: FHR -- tritium -- molten salt -- nuclear graphite -- thermal desorption; Nuclear engineering -- Periodicals; Nuclear engineering; Nuclear Physics; Periodicals; Periodicals; 621.48
• DOI: 10.1080/00295450.2020.1829428
• Languages: English
• ISSNs: 1943-7471

Abstract

Abstract: Mitigating the release of tritium produced from neutron irradiation of molten salts containing lithium or beryllium is a technical challenge for several advanced reactor designs. In a pebble bed Fluoride-Salt-Cooled High-Temperature Reactor (FHR), tritium generated in the Li2 BeF4 (Flibe) coolant is expected to interact with the large inventory of graphite in the core. The degree to which tritium is retained in the FHR core graphite is important to understand in order to predict the tritium distribution in the reactor, operational dose rates in the plant, tritium source term, and optimal strategies to mitigate environmental release. Tritium retention in graphite is simulated in this work based on a model that considers tritium diffusion from Flibe into graphite pores as well as diffusion and trapping in graphite grains. The retention model was implemented into the TRIDENT model framework to study tritium transport at the FHR system level. Tritium permeation through the FHR primary heat exchanger was the largest source of release from the primary system, followed by tritium retention and recirculation of graphite fuel pebbles.