Enthalpy measurement on (U1−xPux)O2 (x = 0, 0.18, 0.45, and 1) and analysis of heat capacity

The heat capacity of oxide nuclear fuels in high-temperature regions, which is critical for evaluating fuel thermal performance, requires extended studies due to disagreements in the literature. Furthermore, mixed-oxide (MOX) and PuO2 data are scarce compared with UO2 data, making it challenging to understand the dependence of Pu content on the properties. Herein, we measured the UO2, MOX, and PuO2 enthalpies up to 2200 K by drop calorimetry. The heat capacity was evaluated by calculating the temperature differential of the enthalpy data. The obtained enthalpy and heat capacity increased almost linearly with the increasing temperature <2000 K and continued to increase significantly at 2000–2200 K. Additionally, heat capacity behavior at higher temperatures, having a peak known as the Bredig transition, is discussed.