Methodology for calculating dose rates from storage cask arrays using MCNP

Dry cask storage is being used at nuclear facilities to increase storage capacity for spent nuclear fuel. Its contribution to the overall site radiation dose rate must be calculated and be shown to be in accordance with 10 CFR 72.104. This specifies that the annual dose equivalent to the whole body of any real individual beyond the controlled area (off-site) must not exceed 25 mrem/yr from all plant operations. Therefore, it is essential that the dose rate from an array of storage casks be calculated accurately, including radiation interaction with surrounding casks as well as scattering by the ground and air. The three-dimensional, continuous-energy MCNP Monte Carlo code was chosen for these calculations because of its accuracy in radiation transport and its ability to correctly model the cask array geometry.