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The enthalpy increments of ( U 1 - y , Am y ) O 2 - x mixed oxides with y = 0.0877 and 0.1895 and x = 0.01–0.03 were measured using drop calorimetry in the temperature range 425–1790 K and the heat capacity was obtained as differential of the obtained enthalpy increments with respect to temperature. The thermal diffusivity was measured using the laser flash technique from 500 to 1550 K. The thermal conductivity was calculated from the measured thermal diffusivity, density and heat capacity. Measured enthalpy increments of the ( U 1 - y , Am y ) O 2 - x solid solutions are very close to the end members, indicating no excess contribution. The derived heat capacities for the two intermediate compositions are slightly higher than that of UO2 and in a good agreement with literature data of AmO2 up to 1100 K. For the thermal conductivity of (U, Am)O2−x mixed oxides a correlation using the classical phonon transport model in crystal structures is proposed.

Nitride compounds are currently considered as possible fuels for advanced nuclear reactors. Among the relevant properties to be investigated, thermophysical quantities such as heat capacity and thermal diffusivity/conductivity are key to ensure safe operation of the fuel in the reactor. Both physical and chemical phenomenological experimental activities are included in this effort. The experimental procedures were optimized to limit oxidation of materials in presence of even small quantities of oxygen and moisture, especially during the measurements at high temperature. Heat capacity and thermal diffusivity/conductivity were measured for ZrN and (Zr,Pu)N compounds. A consistent set of results was obtained from low- and high-temperature Cp measurements for both materials. The excess Cp of solution was estimated. Increasing trends with temperature were observed for the thermal diffusivity and conductivity of both materials in the temperature range considered, confirming the metallic behavior of this class of materials. The presence of Pu causes an almost constant decrease of the thermal conductivity compared to the pure ZrN matrix.

In this work the three binary phase diagrams LiF–ZrF 4 , BeF 2 –ZrF 4 and UF 4 –ZrF 4 were thermodynamically assessed. The three ternary phase diagrams LiF–BeF 2 –ZrF 4 , BeF 2 –ZrF 4 –UF 4 and LiF–ZrF 4 –UF 4 were approximated on the basis of binary data. A pseudo-ternary LiF–BeF 2 –ZrF 4 system with constant amount of UF 4 set to 0.83 mol% was calculated as well, typical for molten salt reactor fuel. Based on this diagram a suggestion for the composition of the fuel in this system was made.