Radiobiological Impact Evaluation Within Monte-Carlo Shielding Calculations of CANDU Spent Fuel

Abstract

The radiobiological effect on the human health of CANDU spent fuel is assessed using Monte Carlo shielding estimates. The examination of spent fuel occurs after it has been discharged from the reactor. A specific cooling interval is considered, with the radiation dose rates that characterize the used fuel being of interest. Two kinds of fuel were studied in a CANDU standard fuel bundle with 37 fuel components: natural uranium (NU) fuel and slightly enriched uranium (SEU) fuel. The fuel burnup was simulated using the ORIGEN-S algorithm, and the photon sources describing the wasted fuel were retrieved. A generic stainless steel shipping cask type B was used for spent fuel transfer, and radiation doses at the cask wall and in the air up to 8 m away from the shipping cask were computed using the Monte Carlo MORSE-SGC algorithm. To ensure nuclear safety and radiation protection, spent fuel must be maintained in temporary wet cooling storage for six months. The projected dosage rates were modest, allowing for the safe handling of the used fuel shipping cask. The corresponding dosages on human body organs for the two considered spent fuels were estimated without and with shielding. Due to the varied sensitivity and reaction of organs/tissue, the effective dosage was evaluated for the human body by applying a tissue-weighting factor; these weighting factors are not equal, and functional coefficients specified by ICRP are used. The equivalent doses calculation modeling findings for the present study underlined the complete effectiveness of the applied shielding and attained the acceptable dose level