The core characteristics are both neutronic and thermal. Neutronic characteristics are obtained by the core nuclear design. In the nuclear design, the core configura­tion, the refueling plan and the plutonium enrichment are determined so that the core safely generates the designed thermal power, based on the plant and fuel basic specifications, throughout the plant life. Also, the core reactivity, breeding perfor­mance, neutron flux distribution, burnup characteristics, control rod worth, and reactivity coefficients are evaluated. The design point of those parameters are determined and coordinated with other design points as described in Fig. 4.4.

The nuclear design calculation methods appear earlier in Chap. 2.1. In the basic calculations of the nuclear design of fast reactors, multi-dimensional, multi-group diffusion calculation codes are mainly utilized.

[1] Multi-group reactor constants

In the nuclear design calculations of fast reactors, a wide range of neutron energies from thermal neutrons (0.025 eV) to fast neutrons (up to about 10 MeV) needs to be treated. The wide energy range is divided into many groups. The set of multi-group reactor constants, which are the averaged cross section of each nuclide within each energy group, are utilized for each group. Based on those sets of multi-group reactor constants, prepared by processing evaluated nuclear data files such as JENDL, the multi-group reactor constants for the design are made using the core configuration, the material composition and the temperature data of the actual core as input.

[2] Nuclear design calculations

The neutronic characteristics, such as the core reactivity, power distribution, burnup characteristics, and control rod worth, are evaluated using the multi­group reactor constants, the core geometry, etc. The reactivity coefficients are calculated by multi-dimensional diffusion and perturbation codes.

Owing to the recent progress in computer performance, the following improvements have been made.

• The number of energy groups is increased for better calculation accuracy. Three-dimensional calculation codes with detailed modeling of core geom­etry are mainly used.

• In cases for which the neutronic characteristics must be accurately calcu­lated in geometries with strong heterogeneity or sharp spatial change in the neutron flux, multi-dimensional transport calculation codes are utilized and then the result of the diffusion calculation is corrected if necessary.

• In cases for which the neutronic characteristics must be accurately calcu­lated in complex geometries, Monte-Carlo codes are utilized.

[3] Validation by calculating mock-up critical experiments

Mock-up critical experiments for simulating fast reactor cores have been conducted and their measurement data have been utilized for evaluating the accuracy of core nuclear design calculation methods. Based on those evalua­tions, the design calculation results can be corrected and the design specifica­tions are then determined with accompanying prediction errors. In Japan, the major mock-up critical experiments done so far are those by the Fast Critical Assembly (FCA) for Joyo, the MOZART experiments [12] for Monju, and the JUPITER experiments [13] for a demonstration FBR plant.