Pressurized Heavy Water Reactors (PHWR) have stored ion exchange resins, which are used in deuteration, dehydrogenation systems, liquid waste treatment systems, and heavy water cleaning systems, in spent resin storage tanks. The C-14 radioactivity concentration of PHWR spent resin currently stored at the Wolseong Nuclear Power Plant is 4.6×10E+6 Bq/g, which exceeds the limited concentration of low-level radioactive waste. In addition, when all is disposed of, the total radioactivity of C-14, 1.48×10E+15 Bq, exceeds the disposal limit of the first-stage disposal facility, 3.04×10E+14. Therefore, it is currently impossible to dispose of them in Gyeongju intermediate- and low-level disposal facilities. As to dispose of spent resins produced in PHWR, C-14 must be removed from spent resins. This C- 14 removal technology from the spent resin can increase the utilization of Gyeongju intermediate- and low-level disposal facilities, and since C-14 separated from the spent resin can be used as an expensive resource, it is necessary to maximize its economic value by recycling it. The development of C-14 removal technology from the spent resin was carried out under the supervision of Korea Hydro & Nuclear Power in 2003, but there was a limit to the C-14 removal and adsorption technology and process. After that, Sunkwang T&S, Korea Atomic Energy Research Institute, and Ulsan Institute of Science and Technology developed spent resin treatment technology with C-14-containing heavy water for the first and second phases from 2015 to 2019 and from 2019 to the present, respectively. The first study had a limitation of a pilot device with a treatment capacity of 10L per day, and the second study was insufficient in implementing the technology to separate spent resin from the mixture, and it was difficult to install on-site due to the enlarged equipment scale. The technology to be proposed in this paper overcomes the limitations of spent resin mixture separation and equipment size, which are the disadvantages of the existing technology. In addition, since 14CO2 with high concentration is stored in liquid form in the storage tank, only the necessary amount of C-14 radioactive isotope can be extracted from the storage tank and be used in necessary industrial fields such as labeling compound production. Therefore, when the facility proposed in this paper is applied for treating mixtures in spent resin tanks of PHWR, it is expected to secure field applicability and safety, and to reflect the various needs of consumers of labeled compound operators utilizing C-14.

When a permanently-closed nuclear power plant is to be decommissioned, large structures targeted to be cut in the process include a steam generator, reactor, and reactor coolant pump (RCP). Although there are sufficient preliminary studies being done on these structures to assess the radiation exposure dose, relatively fewer studies are underway regarding pressurizers. Therefore, preliminary evaluations are required to prevent workers from being overexposed to radiation coming from a pressurizer and to avoid an unnecessary increase in the decommissioning cost. This study created a cutting scenario based on disposal drums for solid radioactive wastes. The cutting scenario was based on 200-liter and 320-liter drums for solid wastes and on the assumption that all cutting operations were done 100 centimeters away from the structure to be cut. When are cutting process of a Pressurizer is carried out per scenario, the 200-liter drum produces 272 pieces, whereas the 320-liter counterpart generates 234 pieces. Given that South Korea allocates 75,550 KRW per liter (based on 200 L) for the disposal cost, an increase in the number of drums leads to an exponential growth of the decommissioning cost, which fuels the need to establish more organized cutting strategies. Meanwhile, in terms of radiation dose, plasma, laser, and flame cutting techniques were estimated to record 0.232 mSv, 0.299 mSv, and 0.213 mSv respectively for 200 L, and 0.195 mSv, 0.251 mSv, and 0.179 mSv respectively for 320 L (based on DF-90). When compared with the annual dose limit of 100 mSv (0.0057 mSv·hr−1), the above numbers registered for both 200 L and 320 L were estimated to satisfy the dose limit, with only a negligible difference in the dose between the two capacities. The results generated from this study are expected to be utilized as a meaningful basis to identify applicable cutting techniques of a pressurizer as part of the decommissioning operation and to establish its cutting plans in compliance with ALARA.