Computed dose map of a fuel material handling facility using radiation transport

Fuel cycle facilities handle a large amount of fissile and radioactive substances leading to external and internal radiation exposure. It is necessary to minimize the dose received without affecting the productivity of the facility. The estimation of radiation dose in areas with limited or full occupancy is an important requirement during the design and planning stage of such a facility. Computer programs provide a convenient tool to estimate the radiation dose. Three main categories of computer codes are used for dose estimation namely Point Kernel method based codes, Monte Carlo radiation transport based codes and deterministic radiation transport codes. In the present study we are using a Monte Carlo radiation transport code to estimate neutron and gamma dose inside a fuel material handling facility