Investigation of U-Zr-O and Fe-Zr-O systems by a laser heating approach

International audience; During a severe accident in a nuclear reactor, the nuclear fuel (UO$_2$ or MOx) may interact with the Zircaloy cladding and with the metallic structures inside the vessel (spacer grids, neutronic absorbers cladding) forming a solid/liquid mixture called corium. At this stage of the accident, the temperature inside the damaged reactor can exceed 2700K. In this framework, the knowledge of the high temperature properties of corium are paramount for the comprehension of how possible accidental scenarios proceed. In particular, the U-(Pu)-Zr-Fe-O system has been considered in the present study, as a simplified version of the real corium. The aim of this study has been the acquisition of high temperature data on some key sub-systems of corium. This paper is focused on the results obtained on the U-Zr-O and Fe-Zr-O ternary systems. Liquidus and solidus temperatures have been obtained by using the laser heating setup at JRC-ITU.