Advanced reactor applications that use a molten fluoride salt coolant and graphite moderator are under consideration as next-generation energy technologies. For molten salts with lithium or beryllium, such as flibe (2LiF-BeF2), the production of tritium from neutron irradiation is a significant technical challenge. Understanding the expected quantities and mechanisms for tritium retention in graphite is important for designing tritium management strategies in these advanced reactors. In this work, the tritium content of IG-110U graphite from a 2013 in-core flibe irradiation experiment was measured by leaching in water and thermal desorption. Five total samples were tested, with an average measured tritium content per salt-contacting surface area of 3.83 ± 0.25 Ci/m2. The tritium measured from the thermal desorption experiments was primarily in a water-insoluble form. Compared to the overall tritium generation during the irradiation, the total amount of retention in graphite predicted by the desorption measurements is significant.