Analysis of Kobayashi benchmark with indigenous Monte Carlo neutron transport code PATMOC

The solution of the neutron transport equation is the basic input for the reactor physics design of a nuclear reactor system. Because of the complexities of geometry and cross-section data, the neutron transport equation is generally solved using numerical methods. One of the difficulties in the solution using these methods concerns the accuracy of distribution of neutron flux in the system containing void regions in a highly absorbing medium. The difficulty arises because of the issue in the continuity of flux at the void and material interface. There is a sudden and large change in the flux at this interface. Kobayashi benchmarks are widely used problems for testing the ability and accuracy of reactor physics codes to compute neutron flux distribution on such systems. An indigenous Monte Carlo neutron transport code, named PATMOC, has been designed and developed at Bhabha Atomic Research Centre (BARC) for reactor physics design applications. The Kobayashi benchmarks have been used to test the PATMOC code to verify its accuracy in flux computation for systems with voids in-between high absorbing materials. Our results show that the PATMOC estimated values of flux distribution in the system compare very well with the reference results provided with the benchmark. In this paper we present the detailed results and analyses of this benchmark with PATMOC code.