PRELIMINARY RESULTS FOR OPR1000/APR1400 WHOLE-CORE ANALYSIS WITH NEUTRON TRANSPORT CODE STREAM

Abstract

Preliminary results of Korean OPR1000 and APR1400 reactor problems have been generated by using whole core neutron transport analysis code STREAM. STREAM has been developed to analyze Korean type commercial reactors as well as a benchmark published Westinghouse type reactor with the direct neutron transport calculation and multi-physics coupling. The OPR1000 and APR1400 reactor cores have been selected to validate STREAM. Various reactor design parameter such as critical boron concentration (CBC), peaking factors, and assembly power distribution have been compared to measured data and that of design code. STREAM shows ~20 ppm error in CBC calculation. It is validated that the peaking factors and fuel assembly power distribution from STREAM agree with measurement data within error criteria and design margin.