Calculation of the neutronic behavior of minor actinides burning in a thermal research reactor using the MCNPX 2.6 code

Abstract

Abstract Due to the reduction of accessible uranium resources as well as waste proliferation issues, researchers are looking for more suitable approaches, such as replacement of uranium as breeding fuels. Among the practical fuel matrixes, the thorium fuel matrix is favored for its naturally abundant and minor actinide proliferation resistance. Monte Carlo computational methods are widely used to successfully simulate neutronic behavior of nuclear reactors. Calculation of some neutronic and dynamic parameters of a 37-assembly simulated research reactor consisting of thorium oxide fuel and 1 minor actinide pin have been carried out in the present work using the MCNPX 2.6 code.