Computer code analysis of irradiation performance of axially heterogeneous mixed oxide fuel elements attaining high burnup in a fast reactor

Abstract

Coupled computer code analyses of irradiation performance of axially heterogeneous mixed oxide (MOX) fuel elements attaining peak burnups of 98–124 GWd/t in a fast reactor were conducted. Postirradiation experiments (PIEs) of the fuel elements revealed local concentration of Cesium (Cs) near the interfaces between MOX fuel and blanket columns including internal blankets. The PIEs also showed an increase in the cladding diameters of the fuel elements. The analyses indicated that the local Cs concentration occurred as a result of Cs axial migration from the MOX fuels toward the blanket pellets near the interfaces. Swelling of the blanket pellets induced by the formation of low-density Cesium-Uranium oxide was not sufficient to cause pellet-to-cladding mechanical interaction (PCMI). The PCMI analyzed in the MOX fuel column regions was insignificant, and the cladding diameter increases were therefore caused mainly by void swelling in cladding and irradiation creep due to fission gas pressure.