Application of OpenMC code to VVER-1000 MOX Сore computational benchmark

This paper presents the results of numerical modeling of the VVER-1000 benchmark with a mixed loading of uranium and MOX fuel, developed by the Nuclear Energy Agency of the Organisation for Economic Co-operation and Development (NEA OECD), using the OpenMC code. OpenMC is a Monte Carlo-based neutron transport simulation code that employs Python as a user interface. The neutron cross sections for various nuclides required for the calculations are obtained from the ENDF/B-VII.1 and ENDF/B-VIII.0 evaluated nuclear data libraries. The use of two different nuclear data versions made it possible to assess the sensitivity of key neutronic parameters of the VVER-1000 reactor to the choice of cross-section libraries. Calculated values of the effective multiplication factor and fission reaction rates in the fuel assemblies for several stationary reactor states are compared with benchmark results obtained using MCU, RADAR, and MCNP codes (published by the NEA OECD and, in particular, by researchers from the Turkish Atomic Energy Authority). Additionally, a cross-verification of the results is carried out using an independent model of the VVER-1000 benchmark developed in the Serpent code. The results demonstrate good agreement with other high-precision codes and confirm the correctness of the OpenMC and Serpent models. The presented methodology is applicable for further modeling and analysis of new core configurations in various reactor systems