Neutron Radiation under the Dry Storage of the Spent Nuclear Fuel
Abstract
To calculate the neutron radiation field around a ventilated container filled with spent nuclear fuel at ZNPP dry accumulated Monte Carlo simulation package used implemented in MCNP. It is shown that the main source of neutrons in the spent fuel is stored 244Cm, the contribution of (a,n) reactions is negligible. The neutron spectrum for the simulation was described by the Maxwell distribution. There is a significant difference between the neutron dose rate in the axial and radial directions due to the differences in the protective properties of the container in their respective areas. To strengthen the radiation protection from the neutron emission during storage of spent nuclear fuel with high burnup offered an additional shield, and its dimensions are optimized to ensure a significant reduction in dose. The characteristics of neutron fluxes in long-term storage of spent nuclear fuel are calculated.
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