[en] The refuelling analysis for the Pu-fuelled heavy water reactor ''Fugen'' has been performed with a coarse mesh, three-dimensional diffusion code, POLESTAR-3D. The program reported previously was based on a modified one-group model, but its two-group version has been added recently, and both programs were used for this study. For the direct verification of this code, the calculated thermal neutron flux distribution was compared with the values measured with a scanning in-core monitor during the power operation using the initial core and twice refuelled cores. First, the two-group, coarse mesh, three-dimensional calculation was compared with the measurement, and it was found that the calculation was sufficiently accurate. Next, the modified one-group calculation was compared with the measurement. It was found that the accuracy was quite reasonable with only small difference from the two-group case. The modified one-group model takes in consideration the reflector transient for the correction of neutron flux distribution. The core management and the monitor layout, the analytical model and the comparison between the calculated and measured flux distributions are reported. The computing time is also discussed. (Kako, I.)