[en] Axial neutron emission rate distribution is determined on every spent fuel assembly, an axial burnup degree distribution is determined based on the value, and further, a nuclear constant for a neutron transportation calculation is determined based on the correlation between the axial burnup degree and previously corresponded nuclear constant for neutron transportation calculation. Then, an axial neutron flux distribution of a containing system during the step of containing spent fuel assemblies is measured. Neutron fluxes are calculated by neutron transportation calculation of a fixed neutron source mode by using the nuclear constant and the neutron emission rate. A portion of the nuclear constant is adjusted so that the calculated neutron fluxes are in proportion to the measured neutron fluxes, and neutron multiplication characteristics of the containing system are evaluated. A portion of the nuclear content is further adjusted axially so that the shape of the distribution as a relative change of the measured value of the axial neutron flux distribution can be realized by the calculation of the fixed neutron source mode. The subcriticality of a pool in the plant or a cask containing system can be evaluated appropriately in a case of considering credit of burnup degree. (N.H.)