Abstract
The propagation of γ-rays emitted from a ventilated storage container with spent nuclear fuel is calculated using a model of a cylindrical surface source. The parameters of a protective structure which are adequate for safe operating conditions are determined. It is shown that because the average energy of the radiation decreases after passing through 30-cm concrete protection the dose rate drops rapidly away from the storage facility. It is found that the combined effect of the angular distribution of the radiation from the containers and their arrangement scheme makes it possible to replace massive transport gates with a sluice system. Then the dose rate near the outer gates of the sluice system is kept at the natural background level while the gate structure can be consider lightened.
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