Neutron flux calculations for simplified PRR-1 SATER model using OpenMC

Abstract

The Philippine Research Reactor-1 (PRR-1) Subcritical Assembly for Training, Education, and Research (SATER) was modeled using the OpenMC Monte Carlo code to analyze its behavior during fuel loading and to verify its subcriticality. OpenMC, an open-source code for neutron transport simulations, was employed due to its adaptability and accessibility. Two sets of simulations were performed: (a) calculation of neutron flux in the location of the three neutron detectors during the fuel loading, and (b) neutron flux mapping of the reactor core with a full fuel load configuration. Results demonstrate the increase in neutron flux in the core with the increasing number of fuel rods loaded. It is also observed that the relative position of the detectors from the fuel rods has a significant impact on the neutron flux in the assembly. From the fuel loading calculations, neutron multiplication factor keff was estimated to be 0.96067 ± 0.00243 using 1/M technique, confirming the subcriticality of the system. The attenuating effect of water is also evident in the full fuel-loaded configuration of the reactor with the neutron flux exceeding 8 × 10³ cm⁻²s⁻¹ at the location of the neutron source while at a distance of 20 cm from the source, the flux level is decreased by a factor of 8. This study demonstrates the viability of OpenMC as a simulation tool for evaluating the reactor physics features of the PRR-1 SATER.