The Analysis of Partial Flow Blockage Accidents for a Sodium Cooled Fast Reactor

Abstract

Analyses were performed for flow blockage accidents postulated in a conceptual design of a 600 MWe demonstration sodium cooled fast reactor with 3 types of core designs, i.e., Uranium, *L-TRU (TRansUrium) and **M-TRU cores, using the MATRA-LMR-FB code. The analysis was addressed for the 6 sub-channel blockage which is a design basis event (DBE). The accidents for 24 and 54 sub-channel blockages were also analyzed to estimate the extent of the blockage size which could lead to sodium boiling or fuel melting. Three radial blockage positions were also taken into account in the analysis. In result, a higher maximum coolant temperature in the subassembly was obtained as the number of blocked subchannels increased. A recirculation region was usually developed right above the blockage for large blockage cases. The analysis results showed that a favorable safety margin was assured for the design basis event, i.e., the 6 sub-channel blockage accident. For the 24 and 54 sub-channel blockage cases, the peak cladding temperature limit was breached, and there was a case in which fuel melting could be threatened. * TRU extracted from LWR fuel ** Mixture of L-TRU and TRU extracted from self-recycled SFR fuel