Keywords: local blockage, blocked channel, nearby channel, cross flow, coolant temperature, cladding temperature, fuel temperature, nuclear power, nuclear energy, lead-cooled reactors, nuclear reactors, nuclear accidents, core design optimisation
Local blockage analysis of lead-cooled next-generation nuclear power reactors
Local blockage is one of the possible severe accidents which may occur in lead (Pb)- or lead-bismuth (Pb-Bi)-cooled fast reactors due to their high melting temperature. In a local blockage accident, the local temperature increase cannot significantly reduce power level inherently. Therefore, this type of accident should be considered in the core design optimisation of those reactors. In the present study, a local blockage accident analysis of Pb-Bi-cooled long-life fast reactors has been performed. In this analysis, we consider two models. In the first model, we consider that the cross-flow can only come from two nearby channels. In the second model, we consider that the cross-flow can come from four nearby channels. The results show that for the present long-life Pb-Bi reactors which are of medium power density level, local blockage will increase the coolant, cladding and pellet temperatures but still within their limits.