Determination of neutron flux redistribution factors for a typical pressurized water reactor ex-core measurements using Monte Carlo technique - Goričanec, Tanja; Kos, Bor; Ambrožič, Klemen; Trkov, Andrej; Snoj, Luka; Kromar, Marjan (Frontiers in Energy Research, 2023)
In a typical pressurized water reactor, neutron detectors located outside the reactor core monitor reactor power. In addition, they are also used to measure the reactivity of the control rods. A novel approach to calculate the ex-core neutron detector response in a typical pressurized water reactor using the Monte Carlo technique is presented. A detailed ex-core model of the Krško nuclear power plant was developed using the Monte Carlo neutron transport code MCNP. Due to the location of the ex-core neutron detectors, the hybrid code ADVANTG is used to generate variance reduction parameters to accelerte the convergence of the results outside the reactor core. To use ADVANTG, the fixed neutron source had to be reconstructed from the criticality core calculation. This paper presents the sensitivity analysis of the response of the ex-core detectors to the neutron data libraries used, the description of the fixed neutron source and the ADVANTG parameters. It was found that a pin-wise description of the neutron source for at least two rows of fuel assemblies at the core periphery is necessary for accurate results. Our results show the importance of a correct description of the prompt neutron spectra in the high energy region and the impact this has on the response of the ex-core detectors. The method in which the prompt neutron fission spectra for important fission nuclides are weighted by the calculated reaction rates has been shown to be the best approximation, with deviations from the reference calculation within statistical uncertainty. The effect of nuclear data libraries on the response of the ex-core detector was investigated, and the difference between the ENDF/B-VII.0 and the ENDF/B-VIII.0 nuclear data libraries was ∼11%. When the deficient evaluation of the 56Fe isotope included in the ENDF/B-VIII.0 nuclear data library was replaced by the improved evaluation from the IAEA INDEN project, the differences decreased to ∼3.7%. In addition, neutron flux redistributions due to control rod movement were investigated and flux redistribution factors were updated using Monte Carlo particle transport methods. The reaction rate redistribution factors obtained with methods presented in this paper are within 1% agreement with the currently used factors.