Steady State Calculations of the PWR MOX/UO2 Core with the Monte Carlo Code MCNP6 - Dung, Nguyen Thi; Phu, Tran Viet; Hartanto, Donny; Lan, Luu Thi; Thuan, Mai Huu; Viet Ha, Pham Nhu (VNU Journal of Science: Mathematics - Physics; Vol 37 No 3DO - 10.25073/2588-1124/vnumap.4621, 2021)
This paper presents the steady-state analysis results of the OECD/NEA and U.S. NRC PWR MOX/UO 2 (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/UO 2 fuelled PWR core, which has different neutronic characteristics from the popular homogeneous ones loaded with the UO 2 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.