Advanced Monte Carlo analysis of ex-vessel neutron dosimetry for cycle 25 of Krško nuclear power plant

This paper presents the results of Monte Carlo neutron transport calculations performed for the containment building of the Krško Nuclear Power Plant. The calculations were performed using the Monte Carlo N-Particle Transport Code (MCNP) in combination with the ADVANTG hybrid code to speed up the convergence of the results outside the reactor core. The paper focuses on the analysis of the Ex-Vessel Neutron Dosimetry (EVND) program conducted during the 25^th fuel cycle at the Krško nuclear power plant with the irradiation of gradient chains and multiple foil sensor sets. The calculated reaction rates at the dosimeter locations are in good agreement with the measured values. However, deviations were observed in the thermal part of the neutron energy spectrum which cannot be neglected, and investigated were some possible reasons for the deviations: nuclear data, concrete material composition, concrete geometry, sensor position and impurities in sample.