The thermal conductivity of U-Mo alloys for research reactor fuel was studied. Thermal conductivity is one of the most important properties concerning design, performance analysis, and safety evaluation of these alloys as research reactor fuel. Thermal conductivity data of unirradiated U-Mo alloy collected from literature were examined, analyzed, and used to develop a correlation as a function of Mo content and temperature. For irradiated U-10Mo alloy, a thermal conductivity correlation was developed as a function of fission density and temperature. This model considers the effects of porosity growth by fission gas bubbles, solid fission product accumulation, and buildup of grain boundaries due to recrystallization during irradiation.