12 Workers' safety from radiological exposure in a 1 ton/day capacity 13 spent resin treatment facility was evaluated according to the operating 14 times and outflow rate due to process related leakages. The conservative 15 annual dose based on the operating times of the workers exceeded the 16 dose limit by at least 7.38E+01 mSv for close work. The realistic dose 17 range was derived as 1.62E+01 mSv-6.60E+01 mSv. The conservative 18 and realistic annual doses for remote workers were 1.33E+01

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... v and 19 3.00E+00 mSv respectively, which were less than the dose limit. The 20 MWR was identified as the major contributor to worker exposure within 21 the 1 h period required for removal of radioactive materials. The dose 22 considering both internal and external exposures without APF was 23 derived to be 1.92E+01 mSv for conservative evaluation and 4.00E+00 24 mSv for realistic evaluation. Furthermore, the dose with APF was derived 25 as 7.27E-01 mSv for conservative evaluation and 1.51E-01 mSv for 26 realistic evaluation. Considering the APF for leakage from all parts, the 27 J o u r n a l P r e -p r o o f dose range was derived as 1.25E+00 mSv-2.03E+00 mSv for 28 conservative evaluation and 2.61E-01 mSv-4.23E-01 mSv for realistic 29 evaluation. Hence, it was confirmed that radiological safety was secured 30 in the event of a leakage accident. 31 32 34 35 J o u r n a l P r e -p r o o f 42 17 O(n, ) 14 C in heavy water in moderator systems. Carbon-14 only emits 43 beta rays, and easily diffuses, migrates, and penetrates the environment 44 compared to other nuclides. As it hardly penetrates the surface of human 45 skin, it does not cause a significant effect from the viewpoint of external 46 exposure. However, it is considered important from the viewpoint of 47 internal exposure. It is easily absorbed into the human body by the 48 inhalation or ingestion of 14 CO 2 gas or vapor, which leads to internal 49 exposure. The radioactivity concentration of Carbon-14 in spent resin 50 produced by heavy water reactors is comparable to intermediate-level 51 radioactive waste [2,3]. 52 During the operation of heavy water reactors, ion-exchange resins trap 53 radionuclides in water [4-6]. The spent resin from a nuclear power plant 54 cannot be reused because it can generate liquid radioactive waste, which 55 requires further treatment [7-9]. Therefore, the activated carbon, zeolite, 56 and spent resin produced in heavy water reactors are discharged into the 57 same storage tank in power plants, as shown in Fig. 1, and stored for 58 extended periods. The spent resin contains radionuclides such as Tritium, 59 Cobalt-60, Caesium-137, and Carbon-14. In particular, the radioactivity 60 concentration of Carbon-14 (8.06E+06 Bq/g) exceeds the disposal criteria 61 J o u r n a l P r e -p r o o f 106 to a capacity of 125 kg/h for 8 h of work per day. The maximum remaining 107 capacity of the treatment facility is 600 kg. However, for a more 108 conservative evaluation, it was assumed that 1,000 kg of spent resin was 109 inside the treatment facility and treated at 125 kg/h and that the treated 110 spent resin mixture was removed from the facility. Based on these 111 assumptions and the source term, the dose evaluation of workers was 112 performed on a 1 h basis during the treatment process. In practice, as 125 113 kg of spent resin mixture flows steadily in the facility for 8 h, the actual 114 J o u r n a l P r e -p r o o f 183 184 3.3 Internal dose evaluation method for workers 185 J o u r n a l P r e -p r o o f J o u r n a l P r e -p r o o f