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Assessment of the Photon and Neutron Source Term for the NPP Krško Spent Fuel

Accurate knowledge of the fuel nuclide inventory is important after reactor shut down, during the fuel storage and subsequent reprocessing or disposal to provide adequate shielding from the photon and neutron radiation. In this paper possibility to calculate the NPP Krško photon and neutron source term with the Serpent code has been analysed. Some deficiencies in the supplied ENDF/B-VII.0 decay library have been observed. In addition, Serpent reports only spontaneous fission rates without (α, n) and (β, n) contributions. To get neutron emission, spontaneous fission rates had to be multiplied with the average number of neutrons born for each particular nuclide manually. Comparison with the Origen code has shown acceptable agreement of the ENDF/B-VII.1 results. Influence of several factors such as fuel burnup, enrichment, temperature, moderator temperature (density), soluble boron concentration, average power, and burnable absorbers has been analysed. In addition, it was demonstrated that, except for the burnup and enrichment, averaging of all other parameters is acceptable approach. IFBA fuel should be accounted for explicitly due to relative high impact on the photon and neutron emissions.