Analysis of neutron flux distribution of 160 MWth PWR NPP reactor using MCNP

Abstract

One of the important parameters for assessing the safety of nuclear reactors is the neutron flux distribution in the nuclear power plant core. The safety analysis, the analysis of component life due to the radiation damage process, and source-term and radiation safety analysis could use the reactor core’s neutron flux distribution data. This study aims to determine the neutron flux distribution of the 160 MWth PWR reactor core. The analysis of the neutron flux distribution of the reactor core was carried out using MCNP. The MCNP is one of the most reliable and widely used computer programs based on the monte-carlo method with a 3-dimensional object model. However, the output of the MCNP is in normalized value, so a multiplier is required to obtain the absolute neutron flux value. The calculation results of the maximum thermal, epithermal and fast neutron flux are 4.6E+13 n/cm2/s, 7.9E+13 n/cm2/s and 9.5E+13 n/cm2/s respectively. The core average fast neutron is 4.3E13 n/cm2/s while the NuScale document standard is 3.4E13 n/cm2/s. So the core average fast neutron flux calculation is relatively close to the standard NuScale document. The results of flux distribution analysis show that the flatness of the neutron flux in the whole core will impact the flatness of the temperature distribution and the flatness of the fuel burn-up.