Calculation of Unmitigated Small Break Loss of Coolant Accident for the IRIS Reactor

Abstract

Preliminary probabilistic risk assessment (PRA) analyses have shown that the IRIS (International Reactor Innovative and Secure) reactor has very low core damage frequency, but in the frame of evaluating accident sequences in IRIS relevant for revising the need for relocation and evacuation measures some severe accident sequences should be defined. Systematic approach based on PRA results was historically used for identification of the sequences and, subsequently, explicit deterministic calculation of a representative sequence was then performed. Calculation methodology is based on using the coupled RELAP5-GOTHIC code to provide the boundary conditions for the severe accident calculation by means of the ASYST code.

The limiting severe accident scenario analyzed in the paper was hypothetical reactor pressure vessel break at the active core bottom elevation with passive safety systems available. Preliminary studies demonstrated that the accident sequence is highly dependent on the break position along the reactor pressure vessel outside surface and by moving the break downwards the core loses more water and, hence, its temperature rises faster. The break size is 4-inch in diameter which corresponds to the size of the piping in the chemical volume and control system. The reactor core heat-up, cladding oxidation, core degradation and core melt progression processes are similar to those obtained in the analyses severe accident progression in light water reactors.