Figure 14.1 shows the reactor pressure vessel (RPV) of the RBWR. The common plant specifications of the RBWR and the latest commercial BWR, the ABWR, are listed in Table 14.1. The rated thermal power, electric power, diameter of the RPV, and core pressure are identical for both reactor plants. Figure 14.2 shows a horizontal cross-sectional view of the RBWR core configuration, which is composed of 720 hexagonal fuel bundles and 223 Y-type control rods. The axial configuration uses the parfait core concept in which an internal blanket of depleted uranium oxide is placed between the upper and lower fissile zones of the TRU oxides.
Various design concepts of the RBWR core have been proposed. Recent core designs have focused on TRU management. The RBWR-AC is the break-even reactor that can burn depleted uranium by using TRUs extracted from the spent fuel bundles of LWRs without decreasing the amount of TRUs. The RBWR-TB is the TRU burner that can fission almost all the TRUs, leaving only the minimum critical mass of TRUs, by repeating their recycling and collecting. The RBWR-TB2 is a modified version of the TRU burner. The RBWR-TB2 is designed to be able to burn
Fig. 14.1 Reactor pressure vessel of the resourcerenewable boiling water reactor (RBWR) 
Table 14.1 Plant specifications 
Fig. 14.2 Horizontal cross-sectional view of the RBWR core configuration 
Fig. 14.3 Utilization concept of the RBWR-AC, -TB, and -TB2
TRUs from LWR spent fuels, whereas the RBWR-TB is designed as a burner for the TRUs from the RBWR-TB itself, assuming the RBWR-TB would be utilized when the TRU usefulness is exhausted and almost all should have been fissioned. Figure 14.3 shows the utilization concept of the RBWR-AC, -TB, and -TB2.
In core designs for the RBWR-AC, -TB, and -TB2, keeping charged TRU composition preserved at every operation cycle is mandatory. This criterion ensures the multi-recycling capability, fission, and recycling process of TRUs can be continued while maintaining the criticality and fulfilling the various operation constraints, such as sufficient reactor shutdown margin and negative void reactivity coefficient. As mentioned in the Introduction, the multi-recycling capability is achieved by hardening the neutron energy spectrum and promoting the transmutation of 238U to fissile plutonium using the hexagonal tight fuel lattice, which has a H/U less than that of the conventional BWR square fuel lattice. Figure 14.4 shows the relationship between the volume ratio of water to fuel and the breeding ratio in the RBWR-AC, -TB, -TB2, and the conventional BWR. Because the RBWR-AC and -TB need to continue operation cycles without feeding fissile materials other than those contained in the discharged fuel from themselves, the volume ratios of water to fuel are set lower than those of the RBWR-TB2 and the conventional BWR.
In the following sections, the core calculation method is described first, and then each type of RBWR is described.
Core Calculation Method
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 . In the burn-up calculation, 45 actinides from
Fig. 14.4 Relationship between water to fuel volume ratio and fissile breeding ratio
228Th to 253Es 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.