The Pebble-Bed Fluoride-Salt-Cooled High Temperature Reactor (PB-FHR) is an advanced nuclear reactor concept that combines high-temperature and low-pressure fluoride salt coolants with particle-encapsulated nuclear fuel. FHRs deliver heat in the 600 - 700°C range, and this temperature range makes it possible to couple nuclear heat to commercially available gas turbines for open air power conversion cycles, which enables combined-cycle efficiencies of 65% and provides the capability of natural gas co-firing for power peaking. One of the technical challenges that is most important in the commercialization if FHRs is the management of tritium. Under neutron-irradiation, the metallic constituents of the salt generate tritium, for which the salt has very low solubility. At 600oC and above, tritium readily diffuses through metals, and the salt-to-air metal heat exchanger is a pathway for the release of tritium to the atmosphere. In order to manage tritium release, two complementary approaches are being studied: tritium permeation barriers on metallic tubes, and methods for the extraction of tritium from the salt coolant. The fuel elements of FHR are made of a graphitic material in which the encapsulated fuel particles are embedded. It would be advantageous if the graphitic material of the FHR fuel elements could serve as an effective in-situ tritium sink. The Scarlat group studies the ability of this graphitic material to absorb hydrogen isotopes from the liquid fluoride salts. This presentation will provide an overview of tritium management in FHR, and will present the ongoing research at University of Wisconsin Madison on understanding the transport mechanisms of hydrogen isotopes from liquid fluoride salts into graphitic materials.