Mechanical behaviour of macro-dispersed inert matrix fuels

Abstract Macro-dispersed inert matrix fuels were irradiated in the high flux reactor in Petten. These fuels consisted of UO 2 inclusions embedded in the inert matrices MgO, MgAl 2 O 4 , Y 3 Al 5 O 12 , CeO 2- x and Y 2 O 3 . The uranium burn-up reached 17.1–19.8% FIMA after an irradiation period of 198.9 days. The sample temperature was about 700–1000 K. Room temperature indentation measurements were performed in the inert matrices before and after irradiation to determine the Vickers hardness and the fracture toughness. The volume swelling of the UO 2 inclusions has been determined. Pellets of UO 2 inclusions embedded in MgO, MgAl 2 O 4 and Y 3 Al 5 O 12 show cracks in the matrix between these inclusions after neutron irradiation. A model is used to describe the fracture behaviour of these inert matrix fuels.