원자력발전소(원전) 내부에 설치되어 있는 주요 기기는 원전의 안정적인 운영을 돕는 주요 2차 구조 물이다. 경주 지진, 포항 지진과 같은 강한 지진이 발생하였을 때, 원전 주요 기기의 손상은 원전의 안정한 정지에 문제를 초래할 수 있다. 따라서, 원전 주요 기기의 지진응답을 저감시키기 위한 연구가 필수적으로 요구된다. 이러한 배경 아래, 본 연구에서는 원전 주요 기기의 내진성능 향상을 위하여 동 흡진장치(Dynamic Absorber)를 활용하였다. 연구에서 사용된 동흡진장치는 스프링, 댐퍼, 및 질량체로 구성된다. 이러한 동흡진장치를 설계하기 위하여 기존에 제안된 방법론들을 활용하였으며, 각 방법론 들을 기반으로 설계된 동흡진장치의 지진응답 저감효과를 비교 및 분석하였다. 구체적으로, 진동대 시 험 결과를 바탕으로 유한요소 모델을 검증하였다. 또한, 이를 기반으로 기존 동흡진장치의 설계방법론 에 따른 원전 주요 기기의 지진응답 저감 효과를 비교 및 분석하였다. 결과적으로 각 방법론들은 원전 주요기기의 가속도, 변위, 응력 응답을 평균적으로 약 30% 정도 감소시키는 효과를 보였다.

As climate change and population growth raise the likelihood of natural disasters, it becomes crucial to comprehend and mitigate these risks in vital infrastructure systems, especially nuclear power plants (NPPs). This research addresses the necessity for evaluating multiple hazards by concentrating on slope failures triggered by earthquakes near NPPs over a timeframe extending up to a return period of 100,000 years. Utilizing a Geographical Information System (GIS) and Monte Carlo Simulation (MCS), the research conducts a comprehensive fragility assessment to predict failure probability under varying ground-shaking conditions. According to the Newmark displacement method, factors such as Peak Ground Acceleration (PGA), slope angle, soil properties, and saturation ratio play significant roles in determining slope safety outcomes. The investigation aims to enhance understanding seismic event repercussions on NPP-adjacent landscapes, providing insights into long-term dynamics and associated risks. Results indicate an increase in slope vulnerability with longer return periods, with distinct instances of slope failures at specific return periods. This analysis not only highlights immediate seismic impacts but also underscores the escalating risk of slope displacement across the extended return period scales, crucial for evaluating long-term stability and associated hazards near nuclear infrastructure.

Safe radiation management is essential not only for operational nuclear power plants but also for nuclear plants to be decommissioned. When spent nuclear fuel is present on-site, meticulous radiation emergency plans are necessary to ensure safety. In Korea, numerous radiation emergency plans have been established for operational nuclear reactors. These plans delineate distinct response mitigation measures for white, blue, and red emergencies. However, clear regulations are yet to be devised for radiation emergency plans for reactors to be decommission. Therefore, this study investigated the decommissioning plan and status of Kori unit 1 to comprehensively analyze the current status of decommissioning safety in Korea. In this study, radiation emergency plans of decommissioning nuclear power plants abroad were reviewed to confirm radiation emergency action levels. Furthermore, radioactive waste treatment facilities, to be used for decommissioning reactors in Korea were evaluated. Moreover, the study assessed emergency plans (especially, emergency initiating conditions) for operational nuclear power plants in Korea for potential use in the decommissioning phase. This study proposed an emergency initiating condition that can be used for decommissioning reactors in Korea. Considering the anticipated introduction of plasma torch melting facility in Korea, this study examined the conditions of radiation emergency plans can be altered. This study identified effective measures and guidelines for managing radiological emergency initiating conditions, and effective decommissioning of nuclear power plants in Korea.

Currently, off-site dose calculations for nuclear power plants are conducted using a computer program (K-DOSE 60). The program is developed based on the regulatory guidelines of the Korea Institute of Nuclear Safety (KINS), which is a domestic nuclear regulatory agency. In this study, a domestic application of the International Atomic Energy Agency (IAEA) TRS (Technical Reports Series)-472 methodology for 3H and 14C in liquid effluents was studied. The dose-evaluation methods adopted and the program configuration for dose evaluation are described based on 3H and 14C in the liquid-effluent-evaluation module of the computer program. The accuracy of the program is verified by comparing the program-calculated results with hand calculation values. Furthermore, a comparative evaluation with LADTAP II, which is a liquid-effluent-evaluation methodology developed by the U.S. NRC (Nuclear Regulatory Commission), is performed. The result confirms that the program-calculated results for the IAEA TRS-472 methodology are consistent with the hand calculation values. Meanwhile, the result of comparative evaluation with LADTAP II indicates different results depending on the methodology used.

PURPOSES : The objective of this study was to review roadway management strategies that can be utilized in the event of a radiological emergency, select feasible alternatives, and simulate a portion of the West Coast network to analyze the effectiveness of these strategies. METHODS : The methodology of the study involved reviewing the relevant literature, extracting the implications, establishing an analysis procedure, and selecting an effectiveness evaluation scale. Using a national transportation database, a network was constructed using Toba, a macroscopic model. RESULTS : A reverse-flow lane system was applied to the West Coast Expressway Glory IC–Hampyeong IC (total 25 km), and a plan to increase the number of lanes was applied to the Seokgyo Street–Shinpyeong Intersection on National Route 23 (total extension 28 km). Consequently, both road management strategies were found to be effective. CONCLUSIONS : This study examined roadway management strategies that can be utilized in the event of a radiological emergency, selected feasible alternatives, and simulated a portion of the West Coast network to verify the effectiveness of these strategies. In the case of reverse flow lanes, it is most effective when applied to expressways that can restrict the entrance and exit of vehicles. In the case of increasing the number of lanes, it is most effective when applied to general roads, and institutional arrangements should be made to enable two-way traffic to use the reserved shoulder lanes.

In this study, the seasonal distribution was surveyed using acoustic in the coastal waters around nuclear power plants. Acoustic surveys were conducted in June, September, December 2022, and March 2023 in the coastal waters of Uljin-gun. According to the results of this study, zooplankton were distributed at the depths from 0 m to 50 m in the waters around nuclear power plants. Zooplankton appeared in summer (June), autumn (September), and spring (March). In the survey area, fish were distributed at the depths from 25 m to 190 m, appearing in the summer (June), autumn (September), winter (December) and spring (March). The SV of zooplankton appearing in the survey area ranged from -98.0 dB to -78.0 dB, and it exhibited a one-class in the frequency distribution of SV. The SV of fish appearing in the survey area ranged from -36.0 dB to -35.0 dB and -98.0 dB to -53.0 dB, and it exhibited two-class in the frequency distribution of SV.

South Korea’s first commercial nuclear reactor, Kori Unit 1, was permanently shut down in 2017, and preparations are currently underway for its decommissioning. After the permanent shutdown, the spent nuclear fuel from the reactor core is removed and stored in a spent fuel storage facility. Subsequently, steps are taken for its permanent disposal, and if a permanent disposal site is not determined, it is stored in an interim storage facility (or temporary storage facility). Therefore, the activation criteria for radiation emergency plans vary depending on the movement of spent nuclear fuel and the storage location. In this study, it reviewed emergency plans in the U.S. NRC Regulatory Guide (Draft) titled ‘Emergency Planning for Decommissioning Nuclear Power Reactors’ to determine the requirements for radiation emergency plans needed for decommissioned nuclear power plants. Additionally, by examining emergency plans applied to decommissioning nuclear power plants in the United States, this study identified emergency plan requirement that could be applicable to future decommissioned nuclear power plants in South Korea. This study will contribute to the establishment of appropriate radiation emergency plans for decommissioning nuclear power plants in Korea for providing accurate information on overseas cases and relevant guidelines.

To construct and operate nuclear power plants (NPPs), it is mandatory to submit a radiation environmental impact assessment report in accordance with Article 10 and Article 20 of the Nuclear Safety Act. Additionally, in compliance with Article 136 of the Enforcement Regulations of the same law, KHNP (Korea Hydro & Nuclear Power) annually assesses radiation environmental effects and publishes the results for operating NPPs. Furthermore, since the legalization of emission plans submission in 2015, KHNP has been submitting emission plans for individual NPPs, starting with the Shin-Hanul 1 and 2 units in 2018. These emission plans specify the emission quantities that meet the dose criteria specified by the Nuclear Safety and Security Commission. Before 2002, KHNP used programs developed in the United States, such as GASPAR and LADTAP, for nearby radiation environmental impact assessments. Since then, KHNP has been using K-DOSE60, developed internally. K-DOSE60 incorporates environmental transport analysis models in line with U.S. regulatory guidance Regulatory Guide 1.109 and dose assessment models reflecting ICRP-60 recommendations. K-DOSE60 is a stand-alone program installed on individual user PCs, making it difficult to manage comprehensively when program revisions are needed. Additionally, during the preparation of emission plans and the licensing phase, improvements to KDOSE60’ s dose assessment methodology were identified. Furthermore, in 2022, regulatory guidelines regarding resident dose assessments were revised, leading to additional improvement requirements. Currently, E-DOSE60, being developed by KHNP, is a network-based program allowing for integrated configuration management within the KHNP network. E-DOSE60 is expected to be developed while incorporating the identified improvements from K-DOSE60, in response to emission plan licensing and regulatory guideline revisions. Key improvements include revisions to dose assessment methodologies for H-13 and C-14 following IAEA TRS-472, expansion of dose assessment points, and changes in socio-environmental factors. Furthermore, data such as site meteorological information and releases of radioactive substances in liquid and gaseous forms can be linked through a network, reducing the potential for human errors caused by manual data entry. Ultimately, E-DOSE60 is expected to optimize resident exposure dose assessment and enhance public trust in NPP operation.

In the dismantling of nuclear power plants, various forms of radioactive gaseous waste are generated when cutting concrete and metal structures. Large amounts of radioactive dust and aerosols generated during the cutting process of each structure can cause radiation exposure to the environment around the workplace and to the radiation exposure in the body of workers. When cutting structures, water is sprayed to reduce the generation of aerosols, so early saturation of the filter is expected due to radioactive aerosols and fine particles containing a large amount of moisture. A mobile air purification device is being developed to a fast and efficient air purifier that can be used for a long time operation to protect workers from radiation exposure in high radiation areas and to minimize the amount of secondary waste generated. In this paper, the direction for a new concept of unit technology that can achieve the development purpose is described.

Thermal cutting processes that can be applied to dismantling nuclear power plants include oxygen cutting, plasma cutting, and laser cutting. According to the global trend, research projects are being carried out in various countries to upgrade laser cutting, and many studies are also being conducted in Korea with plans to apply laser cutting processes when dismantling nuclear power plants. However, with the current technology level of the laser cutting process, the maximum thickness that can be cut is limited to 250 mm. Therefore, in this study, a laser-oxygen hybrid cutting process was implemented by adding a laser heat source to the oxygen cutting process that can cut carbon steel with a thickness of 250 mm or more (RV, beam, column, beam, etc.) when dismantling the nuclear power plant. This has the advantage of improving the cutting speed and reducing the cutting width Kerf compared to conventional oxygen cutting. In this research, the laser-oxygen hybrid cutting process consisted of laser cutting to which Raycus’ 8 kW Fiber Laser power source was applied and oxygen cutting to which hydrogen was applied with Fuel Gas. The oxygen torch was placed perpendicular to the test piece, and the laser head was irradiated by tilting 35° to 70°. The effects of cutting directions on quality and performance were studied, and cutting paths were selected by comparing cutting results. Thereafter, it was confirmed that there is an optimal laser output power according to the cutting thickness by studying the effect on the cutting surface quality by changing only the laser output power under the same cutting conditions. The results of this study are expected to be helpful in the remote cutting process using laser-oxygen hybrid cutting when dismantling domestic nuclear power plants in the future.

When dismantling a power plant, a large amount of radioactive tanks are generated, and it is estimated that a significant amount of sludge will accumulate inside the tanks during long-term operation. In the process of dismantling a radioactive tanks, it is important to know the composition of the sludge because the sludge present inside must first be removed and then disposed of. In the case of certain tanks, it can be predicted that corrosion products generated due to system corrosion are the main cause of sludge formation. However, in the case of some tanks, it is not easy to predict the sludge composition because various dispersed particles in addition to corrosion products may be mixed with the wastewater. Even if it is collected and analyzed, the sludge composition can change significantly depending on the operation history, so the analysis results cannot be considered representative of the composition. In the case of LHST, surfactant components introduced during the washing and shower process, oil components and dispersed particles dissolved by the surfactant accumulate inside the tank, making sludge difficult to remove. In addition, even if it is removed by ultra-high pressure spraying, unexpected problems may occur in the subsequent treatment process due to the surfactant contained therein. Therefore, it is necessary to analyze in more detail the characteristics of sludge accumulated in LHST and prepare countermeasures. A test procedure was prepared to evaluate the characteristics of sludge accumulating in LHST. According to the test results, the long-term sludge accumulation tendency of the LHST is summarized as follows. ① Initially, the sludge settling speed increases slowly until a surface sludge layer is formed. ② After the surface sludge layer is formed, the sludge rapidly settles until the sludge layer becomes somewhat thicker. ③ When the sludge layer is formed to a certain extent, the sludge escape rate increases and the sludge accumulation rate decreases again. It is assumed that the sludge escape speed is closely related to the fluid flow speed in the relevant area. It is believed that the combined effect of these phenomena will determine the thickness of the sludge layer that will accumulate inside the tank, but it was not possible to evaluate how much the sludge layer would accumulate based on the experimental results alone. However, it can be assumed that significant sludge accumulation occurred in areas where fluid flow was minimal and sludge formation nuclei easily accumulates.

Kori Unit 1 nuclear power plant is a pressurized water reactor type with an output of 587 Mwe, which was permanently shut down on June 18, 2017. Currently, the final decommissioning plan (FDP) has been submitted and review is in progress. Once the FDP is approved, it is expected that dismantling will begin with the secondary system, and dismantling work on the primary system of Kori Unit 1 will begin after the spent nuclear fuel is taken out. It is expected that the space where the secondary system has been dismantled can be used as a temporary storage place, and the entire dismantling schedule is expected to proceed without delay. The main equipment of the secondary system is large and heavy. The rotating parts is connected to a single axis with a length of about 40 meters, and is complexly installed over three floors, making accessibility very difficult. A large pipe several kilometers long that supplies various fluids to the secondary system is installed hanging from the ceiling using a hanger between the main devices, and the outer diameter of the pipe is wrapped with insulation material to keep warm. In nuclear secondary system decommissioning, it is very important to check for radiation contamination, establish and implement countermeasures, and predict and manage safety and environmental risks that may occur when cutting and dismantling large heavy objects. So we plan to evaluate the radiation contamination characteristics of the secondary system using ISOCS (In- Situ Object Counting System) to check for possible radioactive contamination. According to the characteristics results, decommissioning plans and methods for safe dismantling by workers were studied. In addition, we conducted research on how to safely dismantle the secondary system in terms of industrial safety, such as asbestos, cutting and handling of heavy materials and so on. This study proposes a safe decommissioning method for various risks that may occur when dismantling the secondary system of Kori Unit 1 nuclear power plant.

After the major radioactivation structures (RPV, Core, SG, etc.) due to neutron irradiation from the nuclear fuel in the reactor are permanently shut down, numerous nuclides that emit alpha-rays, beta-rays, gamma-rays, etc. exist within the radioactive structures. In this study, nuclides were selected to evaluate the source term for worker exposure management (external exposure) at the time of decommissioning. The selection of nuclides was derived by sequentially considering the four steps. In the first stage, the classification of isotopes of major nuclides generated from the radiation of fission products, neutron-radiated products, coolant-induced corrosion products, and other impurities was considered as a step to select evaluation nuclides in major primary system structures. As a second step, in order to select the major radionuclides to be considered at the time of decommissioning, it is necessary to select the nuclides considering their half-life. Considering this, nuclides that were less than 5 years after permanent suspension were excluded. As a third step, since the purpose of reducing worker exposure during decommissioning is significant, nuclides that emit gamma rays when decaying were selected. As a final step, it is a material made by radiation from the fuel rod of the reactor and is often a fission product found in the event of a Severe accident at a nuclear power plant, and is excluded from the nuclide for evaluation at the time of decommissioning is excluded. The final selected Co-60 is a nuclide that emits high-energy gamma rays and was classified as a major nuclide that affects the reduction of radiation exposure to decommissioning workers. In the future, based on the nuclide selection results derived from this study, we plan to study the evaluation of worker radiation exposure from crud to decommissioning workers by deriving evaluation results of crud and radioactive source terms within the reactor core.

In nuclear power plant (NPP) decommissioning, ventilation and purification of the building atmosphere are important to create a working environment, ensure worker safety, and prevent the release of gaseous radioactive materials into the environment. The heating, ventilation, and air conditioning (HVAC) system of each building is maintained, modified, or newly installed. In this study, based on APR1400, operation strategies were presented in case of ventilation abnormalities in the reactor containment building (RCB), where highly radioactive particles and high dust are most frequently generated during NPP decommissioning. For research, it was assumed that the entire RCB atmospheric ventilation during decommissioning would use the RCB purge system of the existing NPP and perform continuous ventilation. Additionally, it is assumed that areas where high radiation particles and high dust occur locally, such as reactor containers or internal segments, are sealed with tents and purified using a HEFA filter of a temporary portable HVAC, and a exhaust flow path is connected to the discharge duct of the existing RCB purge system. The possibility of abnormal occurrence was largely divided into two cases. First, when large amounts of uncontrolled pollutants are released into the atmosphere inside the RCB, discharge to the environment is stopped manually or automatically by a modified engineered safety function activation signal (ESFAS). Afterwards, the RCB purge system should be operated in recirculation mode to sufficiently purify the RCB atmosphere with a HEPA filter. Second, when the first train of the low volume purge system is not running due to a failure, standby train should be operated. If both low volume purge trains fail, a high volume purge system is used. Intermittent purge operation is preferred due to large capacity during high volume purge operation. In cases where it is not possible to operate all purge systems due to common issues such as power supply, atmospheric sampling is performed to determine whether to proceed with the work inside RCB.

The purpose of this report is to provide a summary of the Phase 1 Final Status Survey (FSS) Final Report results and overall conclusions which conduct that the Zion Nuclear Power Station (ZNPS) facility and site meets the 25 mrem(0.25 mSv)per year release criterion as established in Nuclear Regulatory Commission Regulation (NRC) 10 CFR 20.1402 “Radiological Criteria for Unrestricted Use”. The FSS results provided assessment and summarize that any residual radioactivity results in a Total Effective Dose Equivalent (TEDE) to an Average Member of the Critical Group (AMCG) that does not exceed 25 mrem per year, and the residual radioactivity has been reduced to levels that are as low as reasonably achievable (ALARA). The release criterion is translated into site-specific Derived Concentration Guideline Levels (DCGLs) for assessment and summary. ZionSolutions, a decommissioning service provider, estimates that a total of four (4) FSS Final Reports be generated and submitted to the NRC during the decommissioning project. ZionSolutions established the Characterization/License Termination (C/LT) Group, within the Radiation Protection division, with sufficient management and technical resources to fulfill project objectives. The C/LT Group is responsible for the safe completion of all surveys related to characterization and final site closure. Approved site procedures and detailed Technical Support Documents (TSD) direct the FSS process to ensure consistent implementation and adherence to applicable requirements. The development and planning phase was initiated in 1999 by the “Zion Station Historical Site Assessment” (HSA) and the initiation of the characterization process for FSS. Develop the information necessary to support FSS design, including the development of Data Quality Objectives (DQOs) and survey instrument performance standards. DQOs are qualitative and quantitative statements derived from the DQOs process that clarify technical and quality objectives. The next step, FSS design utilizes the combination of traditional scanning surveys, systematic sampling protocols and investigative/judgmental methodologies to evaluate survey units relative to the applicable release criteria for open land sample plans. To aid in the development of an initial suite of potential radionuclides of concern for the decommissioning of ZNPS, the analytical results of representative characterization samples collected at the site were reviewed. At this FSS design step, the Radionuclides of Concern (ROC) are determined. As Co-60 and Cs-137 account for 99.5% of the analysis results of concrete core sampling data form ZNPS’s Containment Building and Auxiliary Building, they are determined and used as the basic ROC in the survey design. Additionally, site information is described and Historical Site Assessment (HSA) is performed. Data collected for the initial HSA will be used to establish the initial regional survey unit and corresponding MARSSIM classification. Next, an assessment of the collected data is performed using the DQO process, and a survey methodology is established by selecting a sampling method and measuring instrumentation. These result judgments provide guidance for C/LT Engineer to interpret findings using the Data Quality Assessment (DQA) process, which analysis Recorded data, Missing values, Deviation from established procedure, and Analysis flags. In conclusion, FSS is the process used to demonstrate that the ZNPS facility and site comply the radiological criteria for unrestricted use specified in 10 CFR.20. The purpose of FSS Sample Plan is to describe the methods to be used in planning, designing, conducting, and evaluating the FSS.

The Korea Research Institute of Standards and Science has developed certified reference materials (concrete, soil, and metal radioactive liquid) for measuring gamma-emitting radionuclides to improve and maintain the quality assurance and quality control of the radioactivity measurement in decommissioning nuclear power plants. The raw materials that make up each CRM were mixed in an appropriate ratio with radionuclides. For certification and homogeneity assessment, 10 bottles were randomly selected, two sub-samples were collected from each bottle, and radionuclides were measured via HPGe gamma spectrometry. The results of the homogeneity tests using a one-way analysis of variance on the radionuclides in the CRMs fulfilled the requirements of ISO Guide 35. Coincidence summing and self-absorption correction were performed on measurement results by introducing the Monte Carlo efficiency transfer code and Monte Carlo N-Particle transport code. In concrete analysis, the reference values for five radionuclides (60Co, 241Am, 134Cs, and 137Cs) in the CRM were in the range of 15-40 Bq/kg, and the expanded uncertainty was within 10% (k = 2). In soil analysis, the reference values for the 137Cs and 60Co were 118.7 and 124.4 Bq/kg, and the expanded uncertainty was within 10% (k = 2). In metal radioactive liquid analysis, the reference values for 134Cs, 137Cs and 60Co in the CRM were in the range of 200-270 Bq/kg, and the expanded uncertainty was within 7% (k = 2).

The decommissioning of domestic Nuclear Power Plants (NPPs) in Korea is expected to begin with the Kori-1, which was permanently shutdown in 2017. In addition, Wolsong-1 has been also permanently shutdown, and another type will be the decommissioning project following Kori-1. KHNP is promoting operation and decommissioning projects as the owner of NPPs, and the Central Research Institute (CRI) has been developing a Final Decommissioning Plan (FDP) for the decommissioning license document. The FDP consists of 11 major chapters in the order of overview of the project, characteristic evaluation, safety assessment, radiation protection, decontamination & dismantlement activities, waste management, etc. The contents described in each chapter are individual chapters, but there are also parts that consider the connection with other chapters. The CRI, which develops the FDP for the first decommissioning project in Korea, has spent a lot of time and effort considering this and has been proceeding through trial and error until the present stage. Therefore, this study aims to explain the current status of FDP, a license document for domestic decommissioning projects, and the link between major input data in major chapters. It can be said that System, Structure, and Components (SSCs) subject to dismantling are considered as the scope of FDP. Chapters that perform estimations on these dismantling targets may include safety assessments, exposure dose assessments for workers and residents, and waste inventory assessments. Therefore, an important part of performing the estimation works is to consider the entire scope of decommissioning activities, and as a way, it can start from data based on the inventory data. After generating the inventory data, the waste treatment classification for the inventory is designated by reflecting the results of the characterization. In addition, for cost estimation, the cost of decommissioning project is predicted by inputting some data (i.e., UCF) such as work process, number of workers, and time required for each item with data reflected in quantity and characterization. After that, based on these inventory, characterization, and UCF data, accident scenarios and industrial safety evaluation are performed for the safety assessment. The worker exposure dose is estimated by considering the dose rate of the workspace with these data. In the case of the amount of waste, the final amount of waste is estimated by considering the factors of reduction and decontamination. In summary, the main estimation contents of FDP are evaluated by adding elements required for the purpose of each chapter from data combined with inventory, characterization, and UCF, so the contents of these chapters are based on the logic of considering the entire scope of decommissioning in common.

To effectively assess the inventory of radionuclides generated from nuclear power plants using a consistent evaluation method across diverse groups, it is imperative to analyze the similarity in radioactive distribution between these groups. Various methodologies exist for evaluating this similarity, and the application of statistical approaches allows us to establish similarity at a specific confidence level while accounting for the dataset size (degrees of freedom). Initially, if the variance characteristics of the two groups are similar, a t-test for equal variances can be employed. However, if the variance characteristics differ, methods for unequal variances should be applied. This study delineates the approach for assessing the similarity in radioactive distribution based on the analytical characteristics of the two groups. Furthermore, it delves into the results obtained through two case studies to offer insights into the assessment process.

Domestic nuclear power plants can affect the environment if multiple devices are operated on one site and even a trace amount of pollutants that may affect the environment after power generation are simultaneously discharged. Therefore, not only radioactive substances but also ionic substances such as boron should be discharged as minimally as possible. We adopted pilot CDI and SD-ELIX sytem to separating and concenrating of boron containing nulcear power plant discharge water. The boron concentration of the initial inflow water tended to decrease over time. The water quality of concentrated water also reached its peak until the initial 60 minutes, but tended to decrease in line with the decrease in the inflow water concentration. The boron removal rate was in the range of 85 to 99% with respect to the initial boron concentration of 15 to 25 mg/L. On the other hand, performance degradation due to the use of electrochemical modules is also observed, and regeneration through low ion-containing water cleaning effective. We shortened processing time by considering the optimal flow rate conditions and conductivity conditions and converting electrochemical modules into series or parallel.