Thermal conductivity of (U,Pu,Am)O solid solutions

Morimoto, Kyoichi ; Kato, Masato   ; Kashimura, Motoaki; Abe, Tomoyuki

Plutonium and uranium mixed oxide (MOX) fuel with high Pu content have been developed as a fuel of fast reactor (FR). In a current research, there is few research of the thermal conductivity evaluation of the MOX fuel with over 20% Pu content, and there is no research of thermal conductivity of the MOX fuel containing Am. In this work, to examine the influences of density, O/M ratio and Am content on the thermal conductivity of MOX fuel, the thermal diffusivities of the MOX fuel with 30% Pu content were measured, and the thermal conductivities of these MOX samples were evaluated.