Reactor behavior comparisons for two liquid metal-cooled fast reactors during an event of loss of coolant

Abstract

The aim of this paper is to present a fast-numerical tool based on the average channel approach to analyze the response of fast reactors under unprotected loss of flow events. The comparison of the transient behavior of a Lead-cooled Fast Reactor and a Sodium-cooled Fast Reactor during a loss of flow event is presented. The coolant flow to the core inlet was reduced to 90%, 70% and 50% of the nominal value. The parameters compared were power, fuel, cladding, and coolant temperatures, as well as heat removal. In order to compare the results between both reactors, the values of neutronic density and removed heat obtained during the transient were normalized with respect to the values in the steady state, as a result, the percentages of increase or decrease of the parameters selected for the model were analysed. In the case of analysis of fuel, gap, clad and coolant, the increments of temperature are presented. With the obtained results, the capacity of the coolants for the removal of the heat generated during the transient can be identified. Therefore, lead has a better capacity to remove heat in fast nuclear reactors.