Source terms and shielding studies related to the spent fuel storage of the R-A reactor

The paper presents the methods developed and calculations carried out to analyse the photon source term and equivalent dose rates of the Vinca R-A reactor spent fuel elements load inside the aluminium barrels. An analysis methodology for the radiological characteristics evolution, based on the SAS2H, MCNP-4C/ORIGEN2.1 and KENOV. a/ORIGEN2.1 utility codes is presented along with information representing the validation of the methods and geometrical models. The sequence SAS4 with the MORSE-SGC code, and the MCNP-4C, a Monte Carlo code has been used for solving the shielding problem. Gamma dose rates estimated in radial direction from the aluminium barrels with spent fuel elements are presented and discussed. (author)