Abstract: For the heat pipe cooled space reactor, region-dependent homogenized cross sections in the predefined 26 group structure were generated with the OpenMC code based on the R-Z geometric model of the reactor core. The neutron transport calculation was performed with SARAX, which was a deterministic neutronic analysis code for fast spectrum reactors. The calculation results were compared with those obtained with MVP. The generation procedure of the homogenized cross sections was verified and the capability of SARAX for the neutronic analysis of heat pipe reactors was demonstrated.