Abstract
A novel nuclear fuel concept has been proposed, consisting of UO₂ particles suspended in a liquid metal alloy. To evaluate its feasibility, the compatibility of Zircaloy-4 cladding with candidate eutectic PbSn and PbBiSn alloys—as well as with pure Pb, Bi, and Sn—was investigated through 1000-hour exposures at 400 °C and 600 °C. Severe degradation of the Zircaloy-4 specimens was observed in eutectic PbSn at 600 °C, whereas oxidation and oxygen diffusion were evident after exposure at 400 °C. At 600 °C, the formation of Zr₃Fe precipitates at the oxide–alloy interface was identified in specimens exposed to Pb and PbBiSn. Thermodynamic analysis suggested that the observed intermetallic formation resulted from the destabilization of Laves phases due to oxygen ingress at elevated temperatures.
