[en] A neutron flux distribution in the fuel bundle is calculated by a multi-bundle calculation which simulates a circumstance in the periphery of fuel bundles in a case where they are loaded actually in a reactor core, while taking the adjacent bundles into consideration. Local power distribution of the fuel bundles is calculated by the neutron flux distribution. A concentration of fuel materials and a distribution of burnable poison concentration satisfying the limiting values relative to the maximum local power can be determined by the local power distribution. (N.H.)