Steady State Calculations of the PWR MOX/UO2 Core with the Monte Carlo Code MCNP6

This paper presents the steady-state analysis results of the OECD/NEA and U.S. NRC PWR MOX/UO2 (MOX: Mixed Oxide) Core Transient Benchmark with the modern MCNP6 Monte Carlo code based on the ENDF/B-VII.1 evaluated nuclear data library. The purpose was to verify an MCNP6 model proposed for calculations of a heterogeneous MOX/UO2 fuelled PWR core, which has different neutronic characteristics from the popular homogeneous ones loaded with the UO2 fuel due to its partial loading of the MOX fuel. The effective neutron multiplication factors, assembly power distributions, and control rod worths calculated using MCNP6 showed a reasonable agreement within 390 pcm, 6%, and 175 pcm, respectively, with the available benchmark data. The discrepancies between the MCNP6 results and the benchmark data were also discussed. Consequently, these results obtained with MCNP6 and ENDF/B-VII.1 can be considered as a new full-core heterogeneous transport solution to supplement for the available benchmark solutions at the steady-state conditions.