In general, after the decommissioning of nuclear facilities, buildings on the site can be demolished or reused. The NSSC (Nuclear Safety and Security Commission) Notice No. 2021-11 suggests that when reusing the building on the decommissioning site, a safety assessment should be performed to confirm the effect of residual radioactivity. However, in Korea, there are currently no decommissioning experiences of nuclear power plants, and the experiences of building reuse safety assessment are also insufficient. Therefore, in this study, we analyzed the foreign cases of building reuse safety assessment after decommissioning of nuclear facilities. In this study, we investigated the Yankee Rowe nuclear power plant, Rancho Seco nuclear power plant, and Hematite fuel cycle facility. For each case, the source term, exposure scenario, exposure pathway, input parameter, and building DCGLs were analyzed. In the case of source term, each facility selected 9~26 radionuclides according to the characteristics of facilities. In the case of exposure scenario, building occupancy scenario which individuals occupy in reusing buildings was selected for all cases. Additionally, Rancho Seco also selected building renovation scenario for maintenance of building. All facilities selected 5 exposure pathways, 1) external exposure directly from a source, 2) external exposure by air submersion, 3) external exposure by deposited on the floor and wall, 4) internal exposure by inhalation, and 5) internal exposure by inadvertent ingestion. For the assessment, we used RESRAD-BUILD code for deriving building DCGLs. Input parameters are classified into building parameter, receptor parameter, and source parameter. Building parameter includes compartment height and area, receptor parameter includes indoor occupancy fraction, ingestion rate, and inhalation rate, and source parameter includes source thickness and density. The input parameters were differently selected according to the characteristics of each nuclear facility. Finally, they derived building DCGLs based on the selected source term, exposure scenario, exposure pathway, and input parameters. As a result, it was found that the maximum DCGL was 1.40×108 dpm/100 cm2, 1.30×107 dpm/100 cm2, and 1.41×109 dpm/100 cm2 for Yankee Rowe nuclear power plant, Rancho Seco nuclear power plant, and Hematite fuel cycle facility, respectively. In this study, we investigated the case of building reuse safety assessment after decommissioning of the Yankee Rowe nuclear power Plant, Rancho Seco nuclear power plant, and Hematite fuel cycle facility. Source terms, exposure scenarios, exposure pathways, input parameters, and building DCGLs were analyzed, and they were found to be different depending on the characteristics of the building. This study is expected to be used in the future building reuse safety assessment after decommissioning of domestic nuclear power plants. This work was

In the field of 3H decontamination technology, the number of patent applications worldwide has been steadily increasing since 2012 after the Fukushima nuclear accident. In particular, Japan has a relatively large number of intellectual property rights in the field of 3H processing technology, and it seems to have entered a mature stage in which the growth rate of patent applications is slightly reduced. In Japan, tritium is being decontaminated through the Semi-Pilot-class complex process (ROSATOM, Russia) using vacuum distillation and hydrogen isotope exchange reaction, and the Combined Electrolysis Catalytic Exchange (CECE, Kurion, U.S.) process. However, it is not enough to handle the increasing number of HTOs every year, so the decision to release them to the sea has been made. Another commercial technology in foreign countries is the vapor phase catalyst exchange process (VPCE) in operation at the Darlington Nuclear Power Plant in Canada. This process is a case of applying tritium exchange technology using a catalyst in a high-temperature vapor state. The only commercially available tritium removal technology in Korea is the Wolseong Nuclear Power Plant’s Removal Facility (TRF). However, TRF is a process for removing HTO from D2O of pure water, so it is suitable only for heavy water with high tritium concentration, and is not suitable for seawater caused by Fukushima nuclear power plant’s serious accident, and surface water and groundwater contaminated by environmental outflow of tritium. Until now, such as low-temperature decompression distillation method, water-hydrogen isotope exchange method, gas hydrate method, acid and alkali treatment method, adsorption method using inorganic adsorbent (zeolite, activated carbon), separator method using electrolysis, ion exchange adsorption method using ion exchange resin, etc. have been studied as leading technologies for tritium decontamination. However, any single technology alone has problems such as energy efficiency and processing capacity in processing tritium, and needs to be supplemented. Therefore, in this study, four core technologies with potential for development were selected to select the elemental technology field of pilot facilities for treating tritium, and specialized research teams from four universities are conducting technology development. It was verified that, although each process has different operating conditions, tritium removal performance is up to 60% in the multi-stage zeolite membrane process, 30% in the metal oxide & electrochemical treatment process, 43% in the process using hydrophilic inorganic adsorbent, and 8% in the process using functional ion exchange resin. After that, in order to fuse with the pretreatment process technology for treating various water quality tritium contaminated water conducted in previous studies, the hybrid composite process was designed by reflecting the characteristics of each technology. The first goal is to create a Pilot hybrid tritium removal facility with 70% tritium removal efficiency and a flow rate of 10 L/hr, and eventually develop a 100 L/hr flow tritium removal system with 80% tritium removal efficiency through performance improvement and scale-up. It is also considering technology for the postprocessing process in the future.

In the case of decommissioning of a nuclear power plant, it is expected that a significant amount of VLLW and LLW that need to be disposed of are also expected. Conventional reduction technology is a method of extracting or removing radionuclides from waste, but this project is being carried out for the purpose of obtaining a reduction effect through the development of a material that treats another radioactive waste using radioactive waste. In this paper, the technology of impregnating LiOH capable of adsorbing radiocarbon to the gas filter material manufactured from concrete and soil waste as raw materials and the radiocarbon removal performance were reviewed. In this study, a raw material of ceramic filter was prepared by mixing concrete and soil waste with a powder of 40 m or less, and after sintering at 1,250°C, 5wt% to 40wt% of LiOH is impregnated with a filter capable of adsorbing carbon dioxide. was prepared. The prepared filter used ICP-OES and XRD to confirm the LiOH deposition result, and the concentration of carbon dioxide discharged through the carbon dioxide adsorption device was confirmed. It was possible to obtain the result that the amount of adsorption was changed depending on the flow rate of carbon dioxide supplied and the amount of material. Through this, it was possible to confirm the possibility of power generation in the adsorption performance of gas. In this study, after crushing waste concrete and waste soil, powders of 40 m or less were mixed with other additives to prepare raw materials for ceramic filters, and sintered at 1,250°C to manufacture filters. 5wt% to 40wt% of LiOH was impregnated on the prepared filter to give functionality to enable carbon dioxide adsorption. The results of LiOH deposition were confirmed using ICP-OES and XRD, and the change in the concentration of carbon dioxide emitted through a separately prepared adsorption device was confirmed. It was possible to obtain the result that the amount of adsorption was changed according to the flow rate of carbon dioxide supplied and the amount of material, and the possibility of developing a material for radioactive waste treatment using radioactive waste was confirmed when the porosity and specific surface area of the filter material were increased.

Regulations on the concentration of boron discharged from industrial facilities, including nuclear power plants, are increasingly being strengthened worldwide. Since boron exists as boric acid at pH 7 or lower, it is very difficult to remove it in the existing LRS (Liquid Radwaste System) using RO and ion exchange resin. As an alternative technology for removing boron emitted from nuclear power plants, the electrochemical boron removal technology, which has been experimentally applied at the Ringhal Power Plant in Sweden, was introduced in the last presentation. In this study, the internal structure of the electrochemical module was improved to reduce the boron concentration to 5 mg/L or less in the 50 mg/L level of boron-containing waste liquid. In addition, the applicability of the electrochemical boron removal technology was evaluated by increasing the capacity of the unit module to 1 m3/hr in consideration of the actual capacity of the monitor tank of the nuclear power plant. By applying various experimental conditions such as flow rate and pressure, the optimum boron removal conditions using electrochemical technology were confirmed, and various operating conditions necessary for actual operation were established by configuring a concentrated water recirculation system to minimize secondary waste generation. The optimal arrangement method of the 1 m3/hr unit module developed in this study was reviewed by performing mathematical modeling based on the actual capacity of monitor tank and discharge characteristics of nuclear power plant.

Various radionuclides are released and contaminate soils by the nuclear accidents, nuclear tests and disposal of radioactive waste. Among radionuclides, 137Cs is a harmful radioactive element that emits high-energy β particles and γ rays with a half-life of 30.2 years. 137Cs is difficult to extract because it is fixed to soil particles. For the volume reduction technology development of contaminated soil, this study tried to evaluate the irreversible Cs adsorption capacity of granite-originated soil. The soil sample used in the study was collected from C horizon of the soil developed in Mesozoic mica granite. The soil texture, mineralogy, organic content, pH, EC, cation exchange capacity (CEC), water-soluble cation and anion content of the soil samples were determined. A kinetic adsorption experiment and an isotherm adsorption experiment were performed to understand the overall Cs adsorption characteristics using 133Cs. The desorption of Cs by 0.1 mM KCl was also tested for the sample spiked with 133Cs and 137Cs. The soil sample showed a pH of 6.73, EC of 24.50 μS cm-1, and CEC of 1.34 cmolc kg-1, organic matter of 0.53% and sandy loam in texture. Quartz, feldspar and mica were identified as the major mineral components of bulk sample. The clay fraction consists of mica, hydroxyl-interlayer vermiculite (HIV), vermiculite and kaolinite. In the kinetic adsorption experiment, the Cs adsorption showed fast adsorption rates at the initial stage (6 hours) regardless of the 133Cs concentration, and the adsorption equilibrium state was reached after 48 hours. It was the most suitable for the pseudo second-order model. The 133Cs adsorption increased nonlinearly from low to high concentration, which was well match with the dual site Langmuir model. As a result of the desorption experiment, desorption was not performed up to 1.1 mg kg-1 in the presence of competitive ions K+, which is about 0.035% of CEC calculated by the isotherm model. The adsorption of Cs was controlled by frayed edge sites (FES) at a low concentrations and by basal sites or interlayer sites at a high concentration. Irreversible Cs fixation of by FES may be contributed by mainly weathered mica, and when these minerals are separated from the granite origin soil, the possibility of reducing the contamination concentration and volume of radioactive soil waste can be expected.

With the increase of temporarily-stored radioactive waste in Korea, the disposal of radioactive waste in a deep geological repository, which is located in crystalline rock at a depth of hundreds of meters below the ground level, has received great attention nowadays. To ensure the permanent isolation of radionuclides from the human and surrounding ecosystems, the safety assessment for the high-level radioactive waste disposal facilities is essential. For the reliable safety assessment of fractured rock, it is especially important to input proper hydraulic properties of fractures such as aperture and hydraulic conductivity, which can directly affect the fluid flow and radionuclide transport. Meanwhile, it has become important to consider sudden fault behavior caused by an earthquake with the recent occurrence of high-intensity earthquakes in the Korean Peninsula. The sudden fault behavior can induce the changes of the hydraulic properties of fractures. Since the changes of the hydraulic properties directly affects to the radionuclide transport in the fractured rock, it is important to estimate the effect of earthquake-induced stress change on hydraulic properties of fractures in the perspective of long-term safety assessment. In this study, the effect of an earthquake on the hydraulic properties of fractures was explored by a numerical approach. The static Coulomb stress change after the earthquake was calculated using software ‘Coulomb 3’ developed by United States Geological Survey (USGS) with the assumption for several mechanical properties such as Young’s modulus, Poisson’s ratio and effective coefficient of friction. The final stress after earthquake occurrence was calculated as the sum of the initial stress and the stress change. Thereafter, the normalized transmissivity of fracture after the earthquake was calculated using the final stress from the stress-transmissivity relationship. Using the methodology for calculating fracture transmissivity change induced by the earthquake developed in this study, the effect of several factors, such as the earthquake magnitude and the distance between fracture and epicenter, was additionally explored. The newly developed methodology will be applied to the processbased total system performance assessment framework (APro) being developed by KAERI, and this study is expected to be helpful for the safety assessment considering long-term evolution phenomena including earthquakes.

For safe management of spent nuclear fuels, they should be delivered to repository or waste disposal site. As the amount of spent nuclear fuel transportation is expected to increase in the future due to the provision of an intermediate storage facility, the necessity to secure transportation cask is emerging. In order to secure the spent nuclear fuel transportation cask, it is necessary to analyze the regulatory processes for domestic and foreign spent nuclear fuel transportation cask. In this study, the regulatory processes for domestic and foreign spent nuclear fuel transportation cask was analyzed. In this study, the IAEA, US, and Korea spent nuclear fuel transportation cask regulatory processes were analyzed. The domestic and foreign spent nuclear fuel transportation cask regulatory processes consist of design phase, manufacturing phase, and operation phase. In the design stage, the transport requirements are designed in accordance with the safety requirements of international organizations and countries. The application to be submitted when applying for approval should include a safety analysis report, evidence proving compliance with safety requirements et al. In the manufacturing stage, it is a stage to check whether the safety requirements are satisfied before the first use after manufacturing the transportation cask. Inspections include welding inspection, leakage inspection, shielding inspection, and thermal inspection. In the operation stage, it is a stage of periodically performing inspections for continuous maintenance of the package when the transportation cask is used. The inspection items to be performed are similar to the manufacturing stage and typically include performance inspection of components and leakage inspection. In this study, domestic and foreign spent nuclear fuel transportation cask regulatory processes were analyzed. It was found that the domestic and foreign spent nuclear fuel transportation cask regulatory processes consist of the design phase, the manufacturing phase, and the operation phase. The results of this study can be used as basic data for policy decision-making for the spent nuclear fuel cask.

Monitoring a state that intentionally hides its nuclear activity via open-source information is akin to looking through a black box. Direct information on the state’s nuclear activity remains in the dark, leaving scholars to speculate how much nuclear material or warheads are being produced. Nevertheless, a state’s nuclear program consists of a complex network that ranges from producing weapon-grade nuclear materials by operating its nuclear facilities to securing resources to fund these activities. These indirect activities allow a narrow window of opportunity for researchers to map a state’s activity that sometimes may not be directly linked to nuclear activity per se but is significant to maintaining and operating its nuclear program. These may include malicious cyberattacks to steal or launder cryptocurrency and facilitating cooperation with fellow rogue states that do not comply with the NPT and nuclear nonproliferation regime. The problem lies in how researchers can map this network. Much of the literature that uses text analysis uses data from either (1) formal statement, reports, and documents or (2) journal articles to extract relations between topics that is otherwise difficult to surmise. This study, however, analyzes news articles containing keywords related to a states’ nuclear activity such as international sanctions, trade activities, other states’ policy etc. While news articles fail to live up to the academic rigor of journal articles and unlike formal documents may sometime contain misinformation or incorrect facts, they are a valuable medium to show the day-to-day activity of a state. Although bias may exist as particular news articles may or may not be chosen for text analysis, by using articles collected from 2021 to 2022, this study argues it is enough data to show a short-term trend in nuclear activity.

For countering nuclear proliferation, satellite imagery is being used to monitor suspicious nuclear activities in inaccessible countries or regions. Monitoring such activities involves detecting changes over time in nuclear facilities and their surroundings, and interpreting them based on prior knowledge in terms of nuclear proliferation or weaponisation. Therefore, analysts need to acquire and analyze satellite images periodically and have an understanding of nuclear fuel cycle as well as expertise in remote sensing. Meanwhile, as accessibility of satellite information has been increasing and accordingly a large amount of high-resolution satellite images is available, a lack of experts with expertise in both fields to perform satellite imagery analysis is being concerned. In this regard, the Institute of Korea Nonproliferation and Control (KINAC) has developed a prototype of semi-automatic satellite imagery analysis system that can support monitoring of potential nuclear activities to overcome the limitations of professionals and increase analysis efficiency. The system provides a satellite imagery database that can manage acquired images, and the users can load images from the database and analyze them in stages. The system includes a preprocessing module capable of resizing, correcting and matching images, a change detection module equipped with a pixel-object-based change detection algorithm for multi-temporal images, and a module that automatically generates reports with relevant information. In particular, this system continuously updates open-source information database related to potential nuclear activities and provides users with an integrated analytics platform that can support their interpretation by linking related images and textual information together. As such, the system could save time and cost in processing and interpreting satellite images by providing semi-automated analytic workflows for monitoring potential nuclear activities.