Modeling and Analysis of VENUS-2 MOX Core Experiment Based on OpenMC Monte Carlo Code

Abstract In this paper, the VENUS-2 MOX core experiment benchmark is investigated by the OpenMC Monte Carlo code. Results of OpenMC are compared with other code values as well as the experimental values. The core effective multiplication factor, the axial fission rate distributions of the fuel pins are basically consistent with experiment values. Furthermore, the VENUS-2 MOX core dosimetry experiment based on the response of different target nuclei detectors is also computed and analyzed. Different reactor dosimetry cross sections libraries are used including the IRDFF-II, IRDF-2002, ENDF/B-VI.8, IRDF-90V2 etc. The computed values of the equivalent fission fluxes are in agreement with the experimental values at most of the measurement positions, and some local results are even better than other procedures such as Al27(n, a) detector. Therefore, the OpenMC Monte Carlo code can provide good prediction accuracy for the measured values of the VENUS-2 MOX core experiment.