Bare (U, Th)O2 fuel pellets were irradiated in a graphite structure to evaluate the potential of this type fuel for high-temperature gas-cooled reactors. The maximum fuel temperature was 1650°C at fuel pellet centers and 1370°C at fuel pellet-tographite interfaces. The experiment was terminated when fission-gas release rates increased by an order of magnitude and the radial temperature gradient from the fuel pellet centers to outer edges increased from 335 to 390°C. Postirradiation evaluation showed no evidence of chemical reaction or incompatibility between the fuel and the surrounding graphite. The graphite underwent no significant changes, but most of the fuel pellets were severely fractured. Burnup (2.4% heavy metal) was below that where the fuel swelling would be expected and optical measurements of two intact pellets showed no dimensional changes.