Determining the optimum thickness of a hot cell for safe processing of spent fuel plate and cobalt source

To design a hot cell, it is essential to consider all safety requirements and radiation protection acceptance criteria. In this research, the design of a hot cell with specific geometric dimensions and materials was simulated using MCNP code. Then, the gamma dose rate was calculated for a ۶۰Co source with ۱.۸×۱۰۱۳ Bq activity and a spent fuel plate with ۹۰% burnup and a cooling time of ۳۰ days to determine the appropriate shielding thickness. In these calculations, the source intensity and the gamma spectrum of the spent fuel plate were obtained using ORIGEN code. According to the references, the gamma dose rate criterion of ۱۰ μSv.h-۱ was used to determine the thickness of the hot cell wall, which is made of barite concrete with ۳.۳۵ g.cm-۳ density and a combination of concrete and paraffin, in different orientations. The results indicate that the necessary optimal thicknesses of shielding for different locations are ۸۰, ۶۵ and ۷۵ cm respectively regarding the irradiation safety criteria.

کلیدواژه ها:

Hot cell ، Shielding calculations ، Co-۶۰ source ، Fuel burnup ، Optimal thickness ، MCNP and ORIGEN codes