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Criticality safety concerning the packaging of spent nuclear fuel assemblies from the transmutation of 237 Np and 241 Am in a fast reactor are investigated and compared to cases of spent fuels from commercial nuclear power stations. The investigation is based on the neutron multiplication factor calculated using the fresh and spent fuel compositions. Fuels based on a fresh composition with high amount of fissile material and low amounts of 237 Np and 241 Am resemble the fast reactor fuel, with the multiplication factor decreasing when calculated using the spent fuel compositions. However, the multiplication factor increases for targets based on a fresh composition without fissile content and high amounts of 237 Np and 241 Am, although it remains below 1 for up to 90 GWd/t burnup.

This paper deals with the calculation of induced heat generation in single layer shields due to the penetration of 6.20 MeV energy photons through graphite, concrete and aluminum as a function of the shield thickness and polar angle. The heat generation rate produced from photons emitted from different geometry radioactive sources penetrating single layer concrete of various mean free paths (mfps) is also determined. In particular the heat generation rate is calculated for 6.20 MeV photons emitted from N-16 pipe and disc geometries source and a comparison has been made in order to estimate the percentage of energy captured, and therefore saved, inside the concrete shielding layer.

In this paper the definition and calculation of angular exposure dose build up factor for combined energies disc geometry source of 1.43 and 2.75 MeV are presented. Angular contributions to the scalar build up factor for single layer shields of Al, Fe and Pb, as a function of polar angle and shield thickness (in mean free paths), are discussed.