Validation of neutron physics Monte Carlo model of Pakistan research reactor-1

Abstract

Monte Carlo model of Pakistan Research Reactor- 1 (PARR-1) is developed, using open source Monte Carlo code OpenMC, to analyze the neutron physics properties of core loading-91 and validate against experimental data. To simulate fuel assemblies at various stages of their life, burnup vector of fuel assemblies is generated using WIMS code. Properties like combined control rod worth, excess reactivity, shutdown margin, worth of thermal column, power production in individual fuel assemblies etc. are calculated and compared against experimental data. Excellent agreement is obtained for integral parameters between calculated and experimental values. This study marks the first step towards a high fidelity coupled neutron physics/thermal hydraulics system development for safety studies of PARR-1.