Radiological safety evaluation of a recycling facility for dismantled concrete waste

When decommissioning a nuclear power plant, a large amount of radioactive waste will be generated. It must be disposed of in a safe and efficient way. Approximately 50% of radioactive waste is concrete waste. The bio-shield is expected to be highly activated compared to concrete structures because of its continuous radiation by neutron flux. Therefore, the spatial dose distribution of the recycling facility and the acquired doses of workers must be determined when recycling or disposing of bio-shield concrete. This study constitutes a preliminary evaluation of radiation safety for a recycling facility for dismantled concrete structures. The radiation dose was simulated using VISIPLAN code to simulate the spatial dose distribution and the external dose of radiation workers. The internal dose of radiation workers was also calculated. The most significant factor in the dose assessment of concrete waste recycling is the radioactivity value of the waste. Thus, this radioactivity value used by cutting the bio-shield is important. It is classified into three cases as follows: the section of the bio-shield that was exposed to the highest radiation levels (case 1), a vertical cut through the section closest to the nuclear reactor (case 2), and where the concrete waste exceeded the allowable concentration for the regulatory clearance (case 3). Case 3 includes parts of cases 1 and 2, as wastes with concentrations higher than the allowable limit for regulatory clearance are classified regardless of the cutting method. The annual effective doses of a worker were 4.40E+01 mSv, 9.07E+00 mSv, and 5.19E+00 mSv for cases 1, 2, and 3, respectively. We determined that the stipulated safety thresholds of 100 mSv over 5 years and no more than 50 mSv in any single year were only exceeded (specifically, the 5-year limit) for case 1, under the assumptions made in our study. We demonstrated that lead shielding is highly effective in reducing doses. Shielding corresponding to thickness of 10 mm and 20 mm, respectively, produced reductions in dose rates of 43.5% and 65.8%, respectively. The workers' annual doses in all cases except case 1 with lead 10 mm when adding lead shielding did not exceed the stipulated annual dose limit of 20 mSv. This dose evaluation of the work processes of a concrete recycling facility is expected to be applied to the actual management of workers' radiation safety.