External radiation of a container used for dry storage of spent VVER-1000 nuclear fuel from the Zaporozhie nuclear power plant

The program MCNP (Monte Carlo Modeling of radiation transfer) is used to calculate the characteristics of external neutron and γ radiation from a ventilated dry-storage container for spent nuclear fuel. Data are obtained on the spatial, energy, and angular distribution of the neutron and γ-ray flux outside the container and the dependence of the dose rate on the storage time of the spent fuel is determined. It is shown that γ-rays make the main contribution to the dose rate on the side surface of the container and neutrons do on the cover. The computed dose rate is 1.4 times higher than the measured value on an individual loaded container at the Zaporozhie nuclear power plant.