Various types of tanks are used in nuclear power plants, and sludge composed of various organic substances and inorganic oxides contaminated with radioactive materials may be present at the bottom of a tank of a radioactive waste treatment device. In addition, glassy and fixative oxide contamination layers are accumulated on the inner wall of the tank depending on the tank material, usage and degree of oxidation. Such contaminated sludge is the main cause of radiation exposure to workers when dismantling nuclear power plant tanks. In addition, the waste filters generated by filtration of contaminated sludge is treated as secondary radioactive waste, and this radioactive waste not only occupies a lot of disposal space, but also the disposal cost is continuously increasing. Therefore, it is necessary to develop a technology that does not generate waste filters as much as possible. To solve this problem, NILEPLANT Co., Ltd. registered a patent named “Filtering apparatus” based on previous research and manufactured a rotary filtration membrane device through detailed design. The rotary filtration membrane device is composed of three or more multiple rotary filtration membranes, and can remove fine particles in wastewater as well as sludge accumulated inside a radioactive contamination tank. In addition, considering the site characteristics of special conditions such as nuclear power plants, it was designed to show excellent performance in removing fine particles while minimizing the area where the device is installed. The rotary filtration membrane device is designed and manufactured as a double cylinder structure that combines a hydro cyclone filter type body and an inner partition wall, and is equipped with a filter cloth-based rotary cylinder filter to process sludge through the filter cloth in addition to inertial. In addition, the patented principle enables self-backwashing without stopping the filtration process, extending the life of the filter and minimizing waste filters. The filtration performance, self-backwashing function, and sludge behavior of the rotary filtration membrane device manufactured based on the detailed design were evaluated through experiments, and improvements to obtain more effective filtration performance were derived. Accordingly, it is expected that the more improved rotary filtration membrane device can be effectively used to remove sludge generated during the dismantling of nuclear power plants in the future.

In case of Korea, unlike overseas nuclear power plants, adjacent units are located in permanently stopped nuclear power plants. Radioactive substances from airborne and liquid effluents are released into the environment from the NPP, and the radioactivity of the released substances must be reported to the regulatory authorities. Radioactive effluents are released into the environment not only in operation but also after permanent shutdown. Due to domestic conditions in which multiple units exist on the same site, it is necessary to consider radioactive effluents generated after permanent shutdown of NPPs. In particular, liquid effluent may have an increased tritium concentration due to draining the spent fuel pool. This paper summarizes the annual liquid emissions of PWR power plants that have been permanently shut down. The data was obtained from the Nuclear Regulatory Commission’s (NRC) annual radioactive effluent release report, which provides information on the annual emissions power plants into the environment. The liquid emissions of each plant were organized into an annual table, providing an overview of the amount of liquid released by each plant. This study aims to raise awareness about the potential environmental impact of permanently shut down nuclear power plants and the need for proper management of their liquid emissions. The findings of this study can used by operator, policymakers, and other stakeholders to make informed decisions regarding the decommissioning and management of nuclear power plants.

The concrete structure of a nuclear power plant is a major safety structure that performs shielding functions to block radioactive materials and radiation, heat removal, and isolation functions. Therefore, concrete structures of nuclear power plants must prove structural safety from immediately after construction to dismantling, and a representative method for this is to investigate compressive strength. The compressive strength and specimen standards of concrete structures are specified in ASTM C 42/C 42M, and samples must be obtained through core drilling in order to collect samples according to this standard. However, commercial equipment requires anchor installation work causes radiation dust generation. Even commercial products have developed equipment that does not require anchor installation work, but it can only be applied to flat walls and cannot be applied to curved walls such as bioshields. To solve this problem, a method of fixing to the scaffolding pipe was designed. The equipment developed based on this method fundamentally blocks the generation of radioactive dust. The vertical position can be adjusted using guide shafts and jack screws, and the horizontal position can be adjusted using scaffolding clamps. In addition, the distance between the installation location and the wall can be adjusted by adjusting the scaffolding clamp location of the device. Lastly, it can be rotated to the left and right, so that even on a curved wall, the sampling position can be performed perpendicular to the wall. Core drills that take specimens for measuring compressive strength use the wet type. Core drilling by wet type in radioactively contaminated concrete leads to the disposal of sludge as radioactive waste. Water supplied during core drilling is scattered in all directions by the rotation of the core drill bit, which causes radiation exposure to workers, so measures must be taken to ensure that the water does not splash and gather in one place. Nileplant Co., Ltd. has developed a sludge collection device that can be used with a core drilling device. It can be inserted into a 4-inch core drill bit to meet the specimen regulations of ASTM C 42/C 42M, and nylon resin was used as a material to withstand friction with water, and the wall of the drainage part was thickened to increase durability. Based on these results, it is expected to be able to work more quickly and safely when collecting core drilling samples of radioactively contaminated concrete or radiation and concrete.

Among the twenty six nuclear power plants in Korea, twenty four plants are currently in operation excluding the two permanently shut-down Kori #1 and Wolsung #1 plants. The decommissioning process includes many tasks such as cutting, decontamination, disposal and treatment. Among the tasks, because cutting is one of the tasks performed close to the target structure, there is a possibility for the workers to be exposed excessively to the radiation. There are representative large structures such as steam generators, nuclear reactors, reactor coolant pump, and pressurizer, made of metals, and radioactive concrete, made of concrete. Especially, compared to the trend of research to manage the radiation exposure of steam generators that are directly connected to pressurizers, the trend of research to manage the radiation exposure of pressurizers to workers is not satisfactory. Moreover, although there have been many studies on radioactive concrete, the studies to manage the radiation exposure to workers with a systematic cutting scenario are insufficient. In this study, radioactive concrete, a representative large structure made of concrete, was selected as the target for evaluation. The conditions for evaluation were cutting speed (1~10 m2/hr) and the time for cutting (permanent shutdown~30 years after the shutdown). A cutting scenario was developed by applying the situation for abrasive decontamination beforehand and Hot-to-Cold and Cold-to-Hot, and effort was made to derive a reasonable plan. The evaluation result derived were hourly radiation dose distribution of 1.19~0.103 mSv/hour and 1.29~0.0113 mSv/hour for a scenario without abrasive decontamination (in the order of Hot to Cold, Cold to Hot), and hourly radiation dose distribution of 0.547~0.0479 mSv/hour and 0.608~0.0522 mSv/hour for a scenario with abrasive decontamination. The maximum value of collective dose derived was 1.54E+04 mSv at the cutting time of permanent shutdown with cutting speed of 1 m2/hour in the Cold to Hot scenario before abrasive decontamination, and the minimum value derived was 5.15E+01 mSv at the cutting time of 30-year after the permanent shutdown with cutting speed of 10 m2/hour in the Hot to Cold scenario after abrasive decontamination.

Wolsong unit 1, the first PHWR (Pressurized Heavy Water Reactor) in Korea, was permanent shut down in 2019. In Korea, according to the Nuclear Safety Act, the FDP (Final Decommissioning Plan) must be submitted within 5 years of permanent shutdown. According to NSSC Notice, the types, volumes, and radioactivity of solid radioactive wastes should be included in FDP chapter 9, Radioactive Waste Management, Therefore, in this study, activation assessment and waste classification of the End shield, which is a major activation component, were conducted. MCNP and ORIGEN-S computer codes were used for the activation assessment of the End shield. Radioactive waste levels were classified according to the cooling period of 0 to 20 years in consideration of the actual start of decommissioning. The End shield consists of Lattice tube, Shielding ball, Sleeve insert, Calandria tube shielding sleeve, and Embedment Ring. Among the components composed for each fuel channel, the neutron flux was calculated for the components whose level was not predicted by preliminary activation assessment, by dividing them into three channel regions: central channel, inter channel, and outer channel. In the case of the shielding ball, the neutron flux was calculated in the area up to 10 cm close to the core and other parts to check the decrease in neutron flux with the distance from the core. The neutron flux calculations showed that the highest neutron flux was calculated at the Sleeve insert, the component closest to the fuel channel. It was found that the neutron flux decreased by about 1/10 to 1/20 as the distance from the core increased by 20 cm. The outer channel was found to have about 30% of the neutron flux of the center channel. It was found that no change in radioactive waste level due to decay occurred during the 0 to 20 years cooling period. In this study, activation assessment and waste classification of End Shield in Wolsong unit 1 was conducted. The results of this study can be used as a basis for the preparation of the FDP for the Wolsong unit 1.

For decontamination and quantification of trace amount of tritium in water, an efficient separation technology capable of enriching tritium in water is required. Electrolysis is a key technology for tritium enichment as it has a high H/T and D/T separation factors. To separate tritium, it is important to develop a proton exchange membrane (PEM) electrolyzer having high hydrogen isotope separation factor as well as high electrolyzer cell efficiency. However, there has not been sufficient research on the separation factor and cell efficiency according to the composition and manufacturing method of the membrane electrode assembly (MEA) Therefore, it is necessary to study the optimal composition and manufacturing method of the MEA in PEM electrolyzer. In this study, the H/D separation factor and water electrolysis cell efficiency of PEM electrolyzer were analyzed by changing the anode and cathode materials and electrode deposition method of the MEA. After the water electrolysis experiment using deionized water, the D/H ratio in water and hydrogen gas was measured using a cavity ring down spectrometer and a mass spectrometer, respectively, and the separation factor was calculated. To calculate the cell efficiency of water electrolysis, a polarization curves were obtained by measuring the voltage changes while increasing the current density. As a result of the study, the water electrolyzer cell efficiency of the MEA fabricated with different anode/cathode configurations and electrode formation methods was higher than that of commercial MEA. On the other hand, the difference in H/D separation factor was not significant depending on the MEA fabrication methods. Therefore, using a cell with high cell efficiency when the separation factor is the same will help construct a more efficient water electrolysis system by lowering the voltage required for water electrolysis.

The decommissioning of Korea’s nuclear power facilities is expected to take place starting with the Kori Unit 1 followed by the Wolsong Unit 1. In Korea, since there is no experience of decommissioning, considerations of site selection for the waste treatment facilities and reasonable selection methods will be needed. Only when factors to be considered for construction are properly selected and their effects are properly analyzed, it will be possible to operate a treatment facility suitable for future decommissioning projects. Therefore, this study aims to derive factors to be considered for the site selection of treatment facilities and present a reasonable selection methodology through evaluation of these factors. In order to select a site for waste treatment facilities, three virtual locations were applied in this study: warehouse 1 to warehouse 3. Such a virtual warehouse could be regarded as a site for construction warehouses, material warehouses, annexed building sites, and parking lots in nuclear facilities. If the selection of preliminary sites was made in the draft, then it is necessary to select the influencing factors for these sites. The site of the treatment facility shall be suitable for the transfer of the waste from the place where the dismantling waste is generated to the treatment facility. In addition, in order for construction to take place, interference with existing facilities and safety should not be affected, and it should not be complicated or narrow during construction. Considering the foundation and accessibility, the construction of the facility should be economical, and the final dismantling of the facility should also be easy. In order to determine one final preferred plan with three hypothetical locations and five influencing factors, there will be complex aspects and it will be difficult to maintain consistency as the evaluation between each factor progresses. Therefore, we introduce the Analytic Hierarchical Process (AHP) methodology to perform pairwise comparison between factors to derive an optimal plan. One optimal plan was selected by evaluating the three virtual places and five factors of consideration presented in this study. Given the complexity and consistency of multiple influencing factors present and prioritizing them, AHP tools help users make decisions easier by providing simple and useful features. Above all, it will be most important to secure sufficient grounds for pairwise comparison between influencing factors and conduct an evaluation based on this.

The domestic Nuclear Power Plant (NPP) decommissioning project is expected to be carried out sequentially, starting with Kori Unit 1. As a license holder, in order to smoothly operate a new decommissioning project, a process in terms of project management must be well established. Therefore, this study will discuss what factors should be considered in establishing the process of decommissioning NPPs. Various standards have been proposed as project management tools on how to express the business process in writing and in what aspects to describe it. Representatively, PMBOK, ISO 21500, and PRICE 2 may be considered. It will be necessary to consider IAEA safety standards in the nuclear decommissioning project. GSR part 6 and part 2 can be considered as two major requirements. GSR part 6 presents a total of 15 requirements, including decommissioning plans, general safety requirements until execution and termination. GSR part 2 presents basic principles for securing the safety of nuclear facilities, and there are a total of 14 requirements. Domestic regulatory guidelines should be considered, and there will be largely laws and regulations related to the decommissioning of nuclear facilities, guidelines for regulatory agencies, and guidelines and regulations related to HSE. The Nuclear Safety Act, Enforcement Decree, Enforcement Rules, and NSSC should be considered in the applicable law for nuclear facilities. Since the construction and operation process has been established for domestic decommissioning project, there will be parts where existing procedures must be applied in terms of life cycle management of facilities and the same performance entity. As a management areas classification in the construction and operation stage, it seems that a classification similar to Level 1 and Level 2 should be applied to the decommissioning project. This study analyzed the factors to be considered in the management system in preparing for the first decommissioning project in Korea. Since it is project management, it is necessary to establish a system by referring to international standards, and it is suggested that domestic regulatory reflection, existing business procedures, and domestic business conditions should be considered.

In the case of decommissioning nuclear facilities in Korea, the dismantling activities will be initiated after obtaining approval from the regulatory agency for the Final Decommissioning Plan (FDP). The contents to be described in the FDP are presented in the notice of the Nuclear Safety and Security Commission, and among them, it is suggested to estimate and provide the basis for the decommissioning cost. The Work Breakdown Structure (WBS) is used for schedule management for the project, and the performance activities can be used as a cost management structure as well as schedule management. In order to easily manage the process and cost, the WBS structure can be normally used, and at this time, there might be a connection with ISDC if necessary. Therefore, this study aims to examine the link between activities from the WBS structure to ISDC in the decommissioning project. In general, the activities assumed as a WBS structure in this study in carrying out the decommissioning project were derived at the Level 1 and Level 2. Activities at Level 1 can be classified into project management costs, controlled area dismantling, conventional area dismantling, site remediation, waste treatment facility, construction/service, R&D, waste treatment and disposal, and characterization. For Level 2 activities, a cost activities embodied in Level 1 was derived. ISDC was developed by the OECD Nuclear Energy Agency (NEA) decommissioning cost estimation group, which improves ambiguous cost systems and presents common cost items for direct comparison between international decommissioning projects. The ISDC consists of Level 1, Level 2, and Level 3, where Level 1 represents the principal activity, Level 2 represents the activity group, and Level 3 represents the typical activity. The cost categories for typical activity at Level 3 consist of labour, investment, expenses, and contingency. In this study, the connection between WBS and ISDC was shown, and a comparison was made at Level 2. Directly, one-on-one matches have difficulties, and as much as possible, they were organized into similar items. We arranged the Level 2 linked to ISDC based on WBS. If there is a difficulty in one-on-one matches, it seems that the accurate cost calculation of ISDC items should consider the impact of additional cost distribution. Therefore, in order to calculate ISDC costs, it seems necessary to organize cost items of WBS in consideration of the ISDC.

The nuclear power plant decommissioning project inevitably considers time, cost, safety, document, etc. as major management areas according to the PMBOK technique. Among them, document management, like all projects, will be an area that must be systematically managed for the purpose of information delivery and record maintenance. In Korea, where there is no experience in the decommissioning project yet, data management is systematically managed and maintained during construction and operation. However, if the decommissioning project is to be launched soon, it is necessary to prepare in consideration of the system in operation, what difference will occur from it in terms of data management, and how it should be managed. As a document that can occur in the decommissioning project, this study was considered from the perspective of the licensee. Therefore, the types of documents that can be considered at Level 1 can be divided into (1) corresponding documents, (2) project documents, (3) internal documents, and (4) reference materials. Four document types are recommended based on Level 1 for the classification of documents to be managed in the decommissioning of nuclear facilities. In this study, documents to be managed in the decommissioning project of nuclear facilities were reviewed and the type was to be derived. Although it was preliminary, it was largely classified into major categories 1, middle categories 2, and 3 levels, and documents that could occur in each field were proposed. As a result, it could be largely classified into corresponding documents, project documents, internal documents, and reference materials, and subsequent classifications could be derived. Documents that may occur in the decommissioning project must be managed by distinguishing between types to reduce the time for duplication or search, and the capacity of the storage can be efficiently managed. Therefore, it is hoped that the document types considered in this study will be used as reference materials for the decommissioning project and develop into a more systematic structure.

The type of radioactive waste that may occur in the process of NPP dismantling can be classified into solid, liquid, gas, and mixed waste. Most of the radioactive waste generated during the dismantling of a NPP is metal solid waste, but liquid radioactive waste is also a very important factor in terms of radiation environmental impact assessment. In the case of liquid radioactive waste, it is necessary to calculate the generation amount in order to design liquid radioactive waste processing system of Radioactive Waste Treatment Facility (RWTF). Depending on the amount of liquid radioactive waste generated, the type of liquid radioactive waste processing system included in the RWTF is different. In addition, in order to apply to the domestic RWTF, it is important to secure the site area occupied by the each system, the liquid radioactive waste treatment capacity of the system, and how to secure circulating water used for dilution and discharge of liquid radioactive waste. Therefore, this review aims to suggest an optimal method for the treatment system for liquid radioactive waste included in RWTF of Wolseong.

The type of radioactive waste that may occur in the process of nuclear power plant dismantling can be classified into solid, liquid, gas, and mixed waste. The amount of these wastes must be defined in the Final Decommissioning Plan for approval of the licensing. Also, in the case of liquid radioactive waste, it is necessary to calculate the generation amount in order to treat radioactive waste at a Radioactive Waste Treatment Facility (RWTF) or on-site. In this regard, there is no Code and Standard for the amount of liquid radioactive waste generated during NPP are dismantled, but ANSI/NS-55.6 describes the amount of liquid radioactive waste generated from a light water reactor type NPP. This code is applied to nuclear power-related facilities such as domestic NPP and radioactive waste disposal facility. Therefore, this review intends to suggest an application plan for domestic NPP decommissioning through codes for liquid radioactive waste expected to generate during nuclear power plant decommissioning.

Radioactive carbon dioxide (14CO2) capture using innovative materials is desirable due to associated radiological hazards, and growing climate change. Mineral carbonation technology (MCT) is amenable to irreversibly capture CO2. Typically, MCT is attractive because capturing carbon through the chemical reaction between alkaline earth metal ions and CO2 forms insoluble and significantly stable carbonates. However, most applications of MCT have an intrinsic restriction regarding their operational conditions since no forward reaction occurs within realistic time scales. Thereby, the CO2 capture performance, such as CO2 capacity and carbonation reaction rate, of MCTs and their applications are severely restricted by the difficulty of operations under mild conditions. For example, natural minerals require aggressive carbonation reaction conditions e.g. high pressure (≥ 20 bar), high temperature (> 373 K), and pH-adjusted carrier solutions. To overcome such obstacles, the fabrication of alkaline earth oxides impregnated into an amorphous glass structure have been recently developed. They show enhanced rates of dissolution of alkaline earth metal ions and carbonation reaction due to the loosely packed glass structure and the generation of a surface coating silica gel, consequently facilitating CO2 capture under mild conditions. In this presentation, we report the synthesis and application of a crystallized glass tailored by controlled heat treatment for CO2 capture under mild conditions. The controlled heat treatment of an alkaline earth oxide-containing glass gives rise to a structural transformation from amorphous to crystalline. The structural characterizations and CO2 capture performance, including CO2 capacity, carbonation reaction rate, and the dissolution rate of alkaline earth metal ion, were analyzed to reveal the impact of controlled heat treatment and phase transformation.

The soils contaminated with radionuclides such as Cs-137 and Sr-90 should be solidified using a binder matrix, because radioactively contaminated soils pose environmental concerns and human health problems. Ordinary Portland cement has been widely used to solidify various radioactive wastes due to its low cost and simple process. In this study, simulant soil waste was solidified using cement waste form. The soils were collected around ‘Kori Nuclear Power Plant Unit 1’ and they were contaminated with the prepared simulant liquid waste containing Fe, Cr, Cs, Ni, Co, and Mn. The water-to-dry ingredients (W/D) ratio of cement waste form was 0.40. The cement paste was poured into a cubic mold (5×5×5 cm) and then cured for 28 days at room temperature. The 28-day compressive strength, water immersion, and EPA1311-toxicity characteristic leaching procedure (TCLP) tests were performed to evaluate the structural stability of cement waste form. The compressive strength was not proportional to soil waste loading, and the lowest compressive strength (4±0.1 MPa) was achieved in cement waste form containing 50wt% soil waste. After the water immersion test for 90 days, the compressive strength of cement waste form with 50wt% soil waste increased to 7.5±0.6 MPa, meeting the waste form acceptance criteria in the repository. It is believed that long-term water immersion test contributed to the additional curing and hydration reaction, resulting in the enhanced compressive strength. As a result of the TCLP test, the released amount of As, Ba, Cd, Cr, Pb, Se, Co, Cs, and Sr was less than the domestic and international standards. These results imply that cement waste form can be a promising candidate for the solidification of radioactive soil wastes.

Kori-1 and Wolseong-1 nuclear power plants were permanently shut down in June 2017 and December 2019, and are currently in the preparation stage for decommissioning. In this regard, it is necessary to secure nuclear power plant decommissioning capacity in preparation for the domestic decommissioning marketplace. To address this, the Korea Research Institute of Decommissioning (KRID) was established to build a framework for the development of integrated nuclear decommissioning technology to support the nuclear decommissioning industry. The institute is currently under construction in the Busan-Ulsan border area, and a branch is planned to be established in the Gyeongju area. Recently, R&D projects have been launched to develop equipment for the demonstration and support verification of decommissioning technology. As part of the R&D project titled “Development and demonstration of the system for radioactivity measurement at the decommissioning site of a nuclear power plant”, we introduce the plan to develop a radioactivity measurement system at the decommissioning site and establish a demonstration system. The tasks include (1) measurement of soil radioactive contamination and classification system, (2) visualization system for massive dismantling of nuclear facilities, (3) automatic remote measurement equipment for surface contamination, and (4) bulk clearance verification equipment. The final goal is to develop a real-time measurement and classification system for contaminated soil at the decommissioning site, and to establish a demonstration system for nuclear power plant decommissioning. The KRID aims to contribute and support the technological independence and commercialization for domestic decommissioning sites remediation of nuclear power plant decommissioning site by establishing a field applicability evaluation system for the environmental remediation technology and equipment demonstration.

Canada’s Pickering Unit 3 was performed a three-stage decontamination from June to August 1989 in preparation for pressure tube replacement. The first step was a reducing CAN-DECON treatment to dissolve the magnetic film inside the reactor, which was applied following partial defueling of the reactor core. The second step was an oxidative dilute alkaline permanganate treatment to remove the chromium-rich oxides of the stainless steel parts. And the final CAN-DECON step was applied continuously after completely removing fuel from the reactor core. In situ pipe gamma-ray spectroscopy techniques were applied to measure radioactivity within feeder piping during various stages of Pickering Unit 3 decontamination. Measurements were performed at a maximum dose rate of 5 mSv/h, and both the detector and the scanned feeder pipe were properly shielded from other neighboring pipes. 60Co was the dominant radionuclide in feeder piping prior to decontamination. And radionuclides 103Ru, 95Zr, 95Nb, 59Fe, 140La and 124Sb were detected. The Co-60 radioactivity was 2.09×105 Bq/cm2 before decontamination and 3.11×103 Bq/cm2 after decontamination in the inlet feeder pipe T18. And in the outlet feeder pipe P21, it is 2.56×104 Bq/cm2 before decontamination and 2.04×103 Bq/cm2 after decontamination.

With the rapid growth of nuclear power in China, a large number of dry wastes, which mainly include the high efficiency particulate air filters (glass fiber), cotton, polyethylene, and absorbent paper with low-level radioactivity and high volume, will be produced during the operation and maintenance of the nuclear power plants. Thermal plasma treatment is a world acceptable technology to incinerate and immobilize radioactive wastes, owing to the high volume reduction factor and the excellent chemical durability of the vitrified waste form. China has developed thermal plasma technology for the treatment of dry wastes from nuclear power plants for more than 15 years and the pilot plant has been constructed. This work will concentrate on the formulation of waste glass fiber to adapt to the vitrification process. A three-component (glass fiber-CaO-Na2O) constrained-region mixture experiment was designed and their viscosity data was mainly studied. The quadratic Scheffé model was used to plot the component effect on melting temperature. The retentions of simulated nuclides, such as Co, Sr, and Cs in the glasses were analyzed. In addition, the glass fiber as a glass matrix to immobilize residual ashes from the thermal plasma gasification of cotton, polyethylene, and absorbent paper was investigated as well.

When decommissioning a nuclear power plant, a large amount of radioactive waste is generated simultaneously. Therefore, efficient treatment of radioactive waste is crucial to the success of the decommissioning process. An utility or decommissioning contractor of NPP often build separate radioactive waste treatment facilities (RWTF) to handle this waste. In Korea, RWTFs are planned to be built for the decommissioning of the Kori Unit 1 and Wolsong Unit 1. In this study, we introduce an application case of using process simulation to derive the optimal layout design and investment plan for a radioactive waste treatment facility. In particular, the steam generator is the largest and most complex device processed in RWTF. Therefore, it is necessary to reflect the large equipment processing area that can treat steam generators in the design of RWTF. In this study, Siemens’ Plant Simulation® was used to derive an optimization plan for the dismantling area of large equipment in RWTF. First, a virtual facility was built by modeling based on the steam generator dismantling process and facilities developed by Doosan Enerbility. This was used to pre-validate the facility investment plan, discover wasteful factors in the logistics waste streams, and evaluate alternatives to derive, validate, and apply appropriate improvement alternatives. Through this, we designed a layout based on the optimal logistics waste streams, appropriate workstations, and the number of buffer places. In addition, we propose various optimization measures such as investment optimization based on optimal operation of facility resources such as facilities and manpower, and establishment of work standards.

Various cutting technologies such as thermal and mechanical are being researched and developed to dismantle shutdown nuclear power plants. Each technology has the following advantages and disadvantages. The thermal cutting method has low reaction force and fast cutting speed, but secondary waste such as fume, dross, and fine dust is generated. The mechanical cutting method has the advantage of low generation of secondary waste such as fume, dross, and fine dust, but has the disadvantage of increasing the size of the device due to its large reaction force. In this study, the performance of plasma milling robot cutting technology for nuclear power plant materials was evaluated. First, before applying plasma auxiliary milling to the robot, tests were conducted on SUS 316 L and Alloy 600 to secure processing conditions such as plasma torch output and transfer speed. The test have shown that the mechanical strength was decreased of each material at the output power of the plasma torch of 4.4 and 8.4 kW, the transfer speed of 200 and 100 mm/min. Based on the test results, a plasma milling was attached to the robot and tested, and it was confirmed that even a small robot with a load of 140 kg can cut without any major problems.

Prevention of radiation hazards to workers and the environment in the event of decommissioning nuclear power plants is a top priority. To this end, it is essential to continuously perform radiation characterization before and during decommissioning. In operating nuclear power plants, various detectors are used depending on the purpose of measurement. Portable detectors used in power plants have excellent portability, but there is a limit to the use of a single measuring device alone to quantify radioactive contamination, nuclide analysis, and ensure representation of measurement results. In foreign countries, gamma-ray visualization detectors are being actively used for operating and decommissioning nuclear power plants. KHNP is also conducting research on the development of gamma-ray visualization detectors for multipurpose field measurement at decommissioning nuclear power plants. It aims to develop detectors capable of visualizing radioactive contamination, analyzing nuclides, estimating radioactivity, and estimating dose rates. To this end, we are developing related software according to the development process by purchasing sensors from H3D, which account for more than 75% of the US gamma-ray visualization detector market. In addition, field tests are planned in the order of Wolsong Unit 1 and Kori Unit 1 with Research reactor in Gongneung-dong in accordance with the progress of development. The detector will be optimized by analyzing the test results according to various gamma radiation field environments. The development detector will be used for various measurement purposes for Kori unit 1 and Wolsong