Integrity of pressure vessel lower head in case of corium relocation: identification of loads and failure modes

Fukushima Daiichi accident emphasized that the complex phenomenology of In Vessel Corium Retention (IVCR) and Ex Vessel Corium Cooling/stabilization (EVCC) strategies has to have a highest priority in the programs of R&D. The IVCR foresees the flooding of the reactor cavity before the relocation of the corium in the pressure vessel lower head occurs. The cooling of the corium is performed by means natural circulation of the water in inside the reactor cavity in regimen of nucleate boiling. The maximum heat flux is limited by the critical heat flux which determines the passage from the nucleate to the film boiling. Another limit on the maximum heat flux is imposed by the maximum heat flux by conduction through the lower head material. This limit is defined by the minimum thickness which assures the structural integrity of the vessel. A common assumption is that the corium relocation occurs in a depressurized primary system. Nevertheless a minimum value of pressure cannot be excluded for the presence of the evaporating water or for localized steam explosions. The lower head is loaded by the hydrostatic pressure caused by the corium weight. Moreover the thermal gradient in the thickness produces bending thermal stresses. This paper considers potential ranges of loads and identifies the appropriate failure modes and the correspondent material characteristics. A numerical analysis is also performed by a FEM code considering a degraded configuration determined by the interaction with the corium.