An outline of the calculation methods used for the core design is as follows. Group constants of 12 energy groups for the core neutronic calculation were evaluated for the horizontal cross section of the fuel bundle lattice by the Monte Carlo calculation code with 190 energy groups [5]. In the burn-up calculation, 45 actinides from

Volume ratio (H2O/fuel)

* Ratio of discharged amount of fissile materials to charged amount of fissile materials

Fig. 14.4 Relationship between water to fuel volume ratio and fissile breeding ratio

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Th to Es and 84 fission products (83 nuclides treated explicitly and 1 lumped fission product) were treated. In the core neutronic calculation, the 12-energy group, three-dimensional neutron flux was obtained by solving the diffusion equa­tion with 1 mesh for each fuel bundle in the horizontal direction and 34 meshes in the vertical direction.

In the thermal hydraulic calculation, the in-channel coolant flow rate, the two-phase flow pressure drop, and the axial void fraction distribution were calcu­lated based on the power distribution obtained by the core neutronic calculation, so that the pressure drops between fuel bundles were balanced. The core neutronic calculation and the thermal hydraulic calculation were iterated until the power distribution and in-channel coolant flow distribution converged.

The void reactivity coefficient was evaluated by decreasing the core coolant flow rate to 95 % of the rated flow and dividing the change of the neutron multiplication factor by the change of core averaged void fraction, from the respective values at the rated flow.