Determination of gamma ray and neutron dose rate on the outer surface of existing containers with irradiated fuel in the RA reactor spent fuel storage (Revision 1, June 2007)

This report presents the methods developed to analyse: the fuel burnup of the RA reactor spent fuel elements with high (80 % 235U) enriched UO2; photon and neutron source terms; and equivalent dose rates on the outer surfaces of existing containers with irradiated fuel elements in the spent fuel storage of the RA reactor. The fuel burnup analysis was performed with combined one-dimensional (1D) and three-dimensional (3D) full core fuel burnup procedure SAS2H/KENO-V.a, based on the KENO-V.a and SAS2H codes from the SCALE-4.4a code system. The photon, neutron and beta particle source terms analysis was founded: on the application of design-oriented SAS2H sequence (from the SCALE-4.4a code system) for 1D geometry models; and reference methodologies MOCUP (MCNP-4C/ORIGEN2.1) and KWO2 (KENO-V.a/ORIGEN2.1) in 3D geometry models of the RA reactor unit cells. The sequence SAS4 with the MORSE-SGC code, and the MCNP-4C code were used for solving the shielding problem. Results representing the verification and experimental validation of the methods and geometrical models are included. Finally, the basic data about: fuel burnup of the RA rector spent fuel elements; heavy nuclides inventory; activity of most important fission products; photon, neutron and beta (electron) source terms for these fuel elements; and photon and neutron equivalent dose rates for water (or dry air) reflected existing containers with the RA reactor spent fuel elements, obtained with presented methodologies are summarized. (author)