The Wolsong unit 1 decommissioning project is the world’s first commercial pressurized heavy water reactor decommissioning project. Although there is a lot of accumulated experience and technology for decommissioning of pressurized water reactors around the world, it can be said that there is great difficulty as there is lack of prior experience and reference materials for pressurized heavy water reactor. On the other hand, if the world’s first pressurized heavy water reactor project is completed, it is possible to enter the overseas market for pressurized heavy water reactor decommissioning. It is also a good opportunity to do so. Accordingly, the current status of operation, plans, and construction of infrastructure related to decommissioning of pressurized heavy water reactors in Canada, which can be said to be the home country of pressurized heavy water reactor, were reviewed. So, this study seeks to present considerations for entering the pressurized heavy water reactors decommissioning market in the future.

Wolsong Unit 1 nuclear power plant, which was permanently shut down in 2019, has a 678 MWe calandria vessel of the CANDU-6 type pressurized heavy-water reactor model. The calandria inside the vault is a horizontal cylindrical vessel made of stainless steel with a length of 7.8 m and a thickness of 28.6 mm. For the entire dismantling processes of a nuclear power plant, dismantling works cannot be performed using only one cutting technology and method, and when performing dismantling of a calandria vessel, various systems and components can be used for cutting and dismantling. The calandria vessel is located in a concrete compartment called a vault, and in order to safely dismantle the calandria vessel, the spread of radioactive contaminants from inside of the vault to the outside must be prevented. We designed dismantling processes using the laser cutting method to dismantle the calandria vessel and end shields. We must minimize the risk of internal radiation exposure to workers from aerosols derived from the thermal cutting processes. Therefore, we need a way to prevent secondary contamination from spreading outside the vault and within the reactor building. The path through which radioactive contaminants move is that the flying airborne products generated during the cutting process inside the vault where the calandria is located do not stay in place but spread outward through the opening of the RM-Deck structure at the top. Therefore, facilities or devices are needed to effectively prevent the spread of radioactive contaminants by blocking the expected movement path. By using these facilities or devices, it is possible to prevent the movement of radioactive aerosol particles between the location of the worker and the location of the cutting area where the calandria is located, thereby preventing internal exposure through the worker’s breathing. In addition, by using these, the cutting area where airborne pollutants are generated can be designed as an isolated work space to prevent the spread of radioactive contaminants. In this study, we propose a method of facilities for confining radioactive aerosol particles and preventing the spread of contamination when thermal cutting of the calandria vessel within the vault.

Nuclear power plants use ion exchange resins to purify liquid radioactive waste generated while operating nuclear power plants. In the case of PHWR, ion exchange resins are used in heavy water and dehydration systems, liquid waste treatment systems, and heavy water washing systems, and the used ion exchange resins are stored in waste resin storage tanks. The C-14 radioactivity concentration in the waste resin currently stored at the Wolseong Nuclear Power Plant is 4.6×106 Bq/g, exceeding the low-level limit, and if all is disposed of, it is 1.48×1015 Bq, exceeding the total limit of 3.04×1014 Bq of C-14 in the first stage disposal facility. Therefore, disposal is not possible at domestic low/medium-level disposal facilities. In addition, since the heavy water reactor waste resin mixture is stored at a ratio of about 20% activated carbon and zeolite mixture and about 80% waste resin, mixture extraction and separation technology and C-14 desorption and adsorption technology are required. Accordingly, research and development has been conducted domestically on methods to treat heavy water waste resin, but the waste resin mixture separation method is complex and inefficient, and there are limitations in applying it to the field due to the scale of the equipment being large compared to the field work space. Therefore, we would like to introduce a resin treatment technology that complements the problems of previous research. Previously, the waste resin mixture was extracted from the upper manhole and inspection hole of the storage tank, but in order to improve limitations such as worker safety, cost, and increased work time, the SRHS, which was planned at the time of nuclear power plant design, is utilized. In addition, by capturing high-purity 14CO2 in a liquid state in a high-pressure container, it ensures safety for long-term storage and is easy to handle when necessary, maximizing management efficiency. In addition, the modularization of the waste resin separation and withdrawal process from the storage tank, C-14 desorption and monitoring process, high-concentration 14CO2 capture and storage process, and 14CO2 adsorption process enables separation of each process, making it applicable to narrow work spaces. When this technology is used to treat waste resin mixtures in PHWR, it is expected to demonstrate its value as customized, high-efficiency equipment that can secure field applicability and safety and reflect the diverse needs of consumers according to changes in the working environment.

After the decision of the Wolsong unit 1 permanent shutdown (2019), spent fuel stored in the spent fuel bay (hereafter, SFB) should be transported to a dry storage facility (MACSTOR or Canister) in order to decommission Wolsong unit 1. Accordingly, KHNP has established a shipment schedule for damaged fuel of Wolsong Unit 1 and is trying to complete the shipment according to the schedule. Wolsong is equipped with transportation casks and dry storage facilities, but baskets need to be manufactured separately. In addition, license approval is required for baskets, transport cask, and dry storage facilities for legal grounds to contain, transport, and store damaged fuels. In this paper, the initial model, upgrade model, and automation model of encapsulation equipment planned to be introduced in Canada to handle PHWR’s damaged fuel were compared, and the optimal model was selected in consideration of KHNP’s planned spent fuel shipment schedule. The PHWR’s damaged fuel encapsulation system is a system developed the PHWR’s damaged spent fuel to be handled in the same way as the existing PHWR when storing it in the dry storage facility and loading a basket for capsulation into transport cask. At the Gentilly-2 nuclear power plant in Canada, a manually operated encapsulation system was used due to the low quantity of damaged fuel, which can be encapsulated two bundles a day, and this model is an initial model. In the case of Wolsong Unit 1, it has about 300 damaged fuels, so it takes about nine months to work when using the initial model. The upgrade model developed to improve work efficiency and reliability has increased work efficiency through some automation, but it would take about eight months to process the damaged spent fuels of Wolsong Unit 1, and this model has not yet been manufactured and applied. Lastly, the automation model changed the work location outside the SFB and automated drainage/drying operations. It is easy to maintain and replace consumables because the work is carried out by lifting the damaged fuel to a shuttle outside the SFB surrounded by a shielding chimney. Considering the reduction of drainage/drying time, it is possible to save about four times as much time as the initial model. That is, if the automation model is used, it is judged that the supply of Wolsong Unit 1 can be processed in about two months. However, in terms of license, initial model and upgrade model are expected to be easier and the period is expected to be shortened. However, if licensing is carried out as soon as equipment design is completed, it is believed that the period can be shortened by parallel equipment manufacturing and licensing. It is judged that the best way to comply with the target schedule is to select an automation model with excellent work performance, develop equipment, and proceed with licensing at the same time. Accordingly, KHNP is in the process of designing equipment with the aim of using the automation model to take out damaged fuel for Wolsong Unit 1.

Heavy water primary system decontamination technology is essential to reduce worker exposure and improve safety during maintenance and decommissioning of nuclear facilities. Advanced decontamination technology development aims to secure controlled decontamination technologies that can reduce the cost of radiation exposure and dramatically reduce the amount of secondary waste generated when decontaminating large equipment and large-area facilities. We conducted a study to identify candidate corrosion inhibitors through the literature and analyze the degree of corrosion of carbon steel samples. Countries with advanced nuclear technology have developed chemical decontamination technology for the entire nuclear power generation system and applied it to the dismantling and maintenance of nuclear power plants. In the decontamination process, the corrosion oxide film must be removed. If the base metal is corroded by the decontaminant in this process, additional secondary waste is generated and treatment costs increase. Therefore, it is necessary to develop a corrosion inhibitor that inhibits the corrosion of the carbon steel base metal in the decontamination process to generate a secondary waste liquid that is favorable for waste reduction and treatment. In this presentation, a study was conducted to analyze the extent of corrosion on a carbon steel base material and identify candidate materials for corrosion inhibition testing. Samples were analyzed using optical microscopy and EPMA analysis to determine the thickness of the corroded oxide film. EPMA analysis also allowed us to map the elemental distribution of the carbon steel corrosion layer, which we plan to quantify in the future. The candidate materials for organic-based corrosion inhibitor were also selected based on their inhibition mechanism; having high electronegative elements for coordinate covalent bonding at metal surface and hydrophobic nonpolar group for preventing access of corrosive substances.The selection of candidate materials for corrosion inhibition testing was based on the mechanism of the corrosion inhibitor. Organic-based corrosion inhibitors are adsorbed by donor-acceptor interactions between metal surfaces and highly electronegative elements. Corrosion can also be inhibited by arranging hydrophobic nonpolar groups on metal surfaces in the solution direction to prevent access of corrosive substances.

It is reported that 48 pressurized heavy water reactors (PHWRs) are in operation, and 10 PHWRs including Wolsong-1 NPP have been permanently shut down in the world. In the case of PHWRs, which have been permanently ceased, they are managed through the delayed decommissioning method, but there are no cases of dismantling. Therefore, technology development is urgent for the effective decommissioning of PHWRs. Unlike PWRs, PHWRs are separated into coolant system and moderator system. Most of pipes and systems of coolant system are mainly composed of carbon steel, expect of the steam generator tubes which are composed of nickel alloy. On the other hand, the moderator system is composed of stainless steel. In the case of stainless steel, the inner layer of the oxide film is composed of chromium oxide, and the outer layer is composed of iron and nickel oxide in enriched. To remove two oxide layers, it is needs to different decontamination method, the coolant system can perform the system decontamination process through a reduction process, but in the case of the moderator system, the oxidation/reduction process is required because it has a material and oxide film similar to PWRs. In this study, this is evaluated the oxide film removal rate according to the type of stainless steel and temperature in order to remove the oxide film deposited in the moderator system. The experiments were carried out at temperatures of 60, 70, 80 and 90°C, with a concentration of 200 ppm of permanganic acid and nitric acid, and 2,000 ppm of oxalic acid, respectively. The results of the oxide film removal rate test for SUS304 showed 29% at 60°C, 38% at 70 and 80°C, and 41% at 90°C. For SUS403, the oxide film removal rate experiment results showed 62% at 60°C, 85% at 70°C, 94% at 80°C, over 99% at 90°C. The results showed that the removal efficiency of the oxide film increased as the temperature increased. Following the results of experimental, the optimum temperature of oxide removal in composed of the stainless steel material is to be 90°C for decontamination of PHWR.

KAERI has been developing a new decontamination process that does not contain any organic chemicals in the decontamination solution and minimizes the use of ion exchange resin in the solution as a purifying step. The process is hydrazine based reductive metal ion decontamination for decommissioning (HyBRID) and consists of N2H4, H2SO4 and Cu+ ions. The primary system of the LWR is composed of materials with high corrosion resistance, such as stainless steel and Inconel, but among the materials, the feeder and header of the primary system of the PHWR are composed of carbon steel (SA106Gr.B) with low corrosion resistance. Therefore, when decontamination of PHWRs, attention should be paid to corrosion of carbon steel. Since Fe3O4, a contaminating oxide film formed on the surface of carbon steel dissolves faster than ferrite or chromite formed on the surface of Inconel or stainless steel, the base material is exposed to the solution and is corroded during decontamination. When a large amount of iron ions is eluted into the decontamination agent due to corrosion of carbon steel, not only the soundness of the base metal is adversely affected, but also the amount of decontamination waste increases. The purpose of this study is to develop inhibitors that can minimize corrosion of carbon steel when decontamination of PHWRs using the HyBRID decontamination process. CG, 570S and PP3 were selected as corrosion inhibitors. In addition, corrosion tests of carbon steel were conducted in the HyBRID solution with corrosion inhibitors. The best corrosion inhibitors and optimal operating conditions were selected, and HyBRID decontamination agents with corrosion inhibitors were much better in corrosion resistance than existing commercial decontamination agents.

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.

In this study, an aerosol process was introduced to produce CaCO3. The possibility of producing CaCO3 by the aerosol process was evaluated. The characteristics of CaCO3 prepared by the aerosol process were also evaluated. In the CaCO3 prepared in this study, as the heat treatment proceeded, the calcite phase disappeared. The portlandite phase and the lime phase were formed by the heat treatment. Even if the CO2 component is removed from the calcite phase, there is a possibility that the converted CO2 component could be adsorbed into the Ca component to form a calcite phase again. Therefore, in order to remove the calcite phase, carbon components should be removed first. The lime phase was formed when CO2 was removed from the calcite phase, while the portlandite phase was formed by the introducing of H2O to the lime phase. Therefore, the order in which each phase formed could be in the order of calcite, lime, and portlandite. The reason for the simultaneous presence of the portlandite phase and the lime phase is that the hydroxyl group (OH−) introduced by H2O was not removed completely due to low temperature and/or insufficient heating time. When the sufficient temperature (900°C) and heating time (60 min) were applied, the hydroxyl group (OH−) was removed to transform into lime phase. Since the precursor contained the hydrogen component, it could be possible that the moisture (H2O) and/or the hydroxyl group (OH−) were introduced during the heat treatment process.

The permanent shutdown of Wolseong 1, PHWR (Pressurized Heavy Water Reactor) was decided. Accordingly, there is need for C-14 treatment technology to spent resin generated by PHWR in classified Medium Level Radioactive Waste by C-14 specific activity. However, spent resin by PHWR is mixed and stored with activated carbon and zeolite (mixture), not a single storage, and separation from the mixture must be carried out in advance for C-14 treatment in the spent resin. This study developed a C-14 treatment facility that combined with the technology of separating spent resin from spent resin mixture by PHWR NPP and the technology of C-14 treatment for disposal. The C-14 treatment facility consists of spent resin separation (Part 1) and treatment of separated spent resin. (Part 2) Part 1 is applied with a process of separating the mixed and stored spent resin from the spent resin mixture by applying a drum screen method. In the case of Part 2, spent resin treatment process for desorbing and collecting C-14 nuclides in the separated spent resin using microwave reactor was applied. Except for the adsorbent used to collect C-14 detached in the process of separating and treating spent resin, no additional material is introduced into the facility, and thus secondary waste is significantly reduced. In addition, pollution prevention banks at the bottom of the facility and a sealed automated circulation system were applied to prevent unexpected leakage and diffusion of radioactive materials and ensure stability of workers. Currently, the C-14 treatment facility has been verified for spent resin separation and spent resin treatment using simulated spent resin mixture, and the facility will be demonstrated and verified for field applicability. According to derived results, it is believed that it will be possible to apply the C-14 treatment facility when decommissioning of PHWR.