This study presents a rapid and sequential radiochemical separation method for Pu and Am isotopes in radioactive waste samples from the nuclear power plant with anion exchange resin and TRU resin. After radionuclides were leached from the radioactive waste samples with concentrated HCl and HNO3, the sample was allowed to evaporate to dryness after filtering the leaching solution with 0.45 micron filter. The Pu isotopes were separated in HNO3 medium with anion exchange resin. For leaching solution passed through anion exchange column, the Am isotopes were separated with TRU resin. The purified Pu and Am isotopes were measured by alpha spectrometer, respectively, after micro-precipitation of neodymium. The sequential radiochemical separation of Pu and Am isotopes in radioactive waste samples using anion exchange resin and TRU resin was validated with ICP-MS system.

Chelating agents, such as EDTA, NTA, and citric acid, possess the capacity to establish complexes with radionuclides, potentially enhancing the migration of such radionuclides from the disposal sites. Hence, quantification of these chelating agents in radioactive wastes is required to ensure secure disposal protocols. The determination of chelating agents in radioactive wastes is mainly composed of two steps, e.g. extraction and detection. However, there are little information on the extraction of the chelators in various radioactive wastes. We endeavored to optimize the extraction conditions for citric acid (CA) found within concrete, a prevalent component in the context of dismantled waste materials. Given the inherent high solubility of CA in water, we applied an aliquot of deionized water to the concrete and conducted a one-hour ultrasonic leaching procedure to facilitate chelate extraction. Subsequently, following the preparation of the concrete leachate via vacuum filtration and centrifugation to yield a clarified solution, we quantified the concentration of CA within the solution using Ion Chromatography (IC). To enhance leaching efficiency, we examined the % recovery variation with respect to the pH of the leaching solution. The optimized extraction method will be applied to diverse chelating agents and radioactive waste categories.

For the disposition of radioactive wastes generated from nuclear power plant, radioisotope inventory must be analyzed to determine an activity concentration of radionuclides. Radionuclides in low- and intermediate-low-level of radioactive wastes, however, can be easily classified to easyto- measure (ETM) and difficult-to-measure (DTM) nuclides. ETM nuclides are gamma emitting nuclides that is relatively easy to measure because they do not need to be destroyed for the preprocessing. On the other hands, DTM nuclides are alpha and beta emitting nuclides that need to be destroyed for the preprocessing and also need chemical separation. Currently, measurement methods for DTM nuclides are developed and in this paper measurement methods of Fe-55, Ni-59, Ni-63, Sr-90 and Tc-99 will be introduced.

A new annual dose evaluation system called E-DOSE has been developed. The system is based on the methodology of the previous version, K-DOSE60, which uses the dose evaluation methods of the International Commission on Radiological Protection (ICRP-60). However, E-DOSE is coded in ABAP to be compatible with the KHNP’s enterprise resource planning (ERP) system, SAP. This allows E-DOSE to use the real-time data from SAP, which minimizes the need for user intervention. The socio-environmental data, which was previously managed by the staff of each plant sites, can now managed in the system in a centralized manner. This is a significant improvement over the previous system, as it reduces the risk of errors and makes it easier to track and manage data. The system also automatically generates the reports required by regulations. EDOSE is expected to minimize the occurrence of human errors in preparing and managing the input data. This is because the system uses the data from SAP, which is less prone to errors than manually entered data. Additionally, the automatic generation of reports reduces the risk of errors in report preparation. E-DOSE is also expected to improve work efficiency. This is because the system automates many of the tasks involved in annual dose evaluation, such as data entry, calculation, and report generation. Overall, E-DOSE is a significant improvement over the previous annual dose evaluation system. It is more efficient, accurate, and user-friendly.

When occurring at a nuclear power plant (NPP) by accidents, accurate prediction and identification of the process of radioactive material dispersing into atmosphere is important to protect public and environment. Atmosphere dispersion of radioactive materials is significantly influenced by wind direction and wind speed. The government and nuclear operator continuously monitor wind data at nuclear sites through meteorological tower to prepare for such accidents involving the release of radioactive materials. The purpose of this study is to construct wind rose diagrams at 5 NPP sites (Kori, Saewool, Wolsong, Hanbit, Hanul). Wind roses serve as invaluable tool for identifying wind patterns in each region and visualizing wind directions. This can be utilized to predict the dispersion pathway and extent range of radioactive materials carried by the wind. This program will take on the role of establishing appropriate evacuation routes or shelter locations for residents when reliable wind data is not immediately available during an NPP accident. The wind data used in the study was collected from a meteorological tower located at the NPP site, and measurements were taken at 1-hour intervals for each operation over a period of ten years. The collected data underwent preprocessing, followed by the development of Python code to render the wind rose diagrams in an interpretable format. The future direction of this study will be focused on enhancing this program by integrating geographical mapping capabilities. With these advancements, it will become feasible to superimpose shelter positions on a map in accordance with prevailing wind directions. These improvements will contribute to the development of additional protective measures for residents and the proposal of alternative shelter options in response to potential radioactive material releases.

The radioactive cesium, released from the normal operation or the accidental operation of nuclear facilities, should be regularly monitored for environmental regulatory compliance. The 135Cs/137Cs isotopic ratios, potentially useful for long-term tracking Cs transport in seawater, can be used as a tool of understanding how radionuclides are transported from different nuclear production source terms and distributed in the ocean. The ultra-high sensitive mass spectrometers (TIMS, SF-ICP-MS and TQ-ICP-MS) have been used to measure the 135Cs/137Cs isotopic ratios. However, the radiochemical separation of Cs from the seawater matrix is essential for the analysis of Cs using the mass spectrometers. An automated radiochemical procedure for the separation of Cs in seawater was developed for the analysis of 135Cs/137Cs isotopic ratios using a sequential column chromatography with AMPPAN and AG50Wx8 cation exchange resins. National Instrument’s LabVIEW is a graphical programming language and a powerful tool for the instrument control. A virtual instrument system for the automated separation of cesium isotopes was developed by the state machine of the fundamental design patterns in LabVIEW. In this study, the conceptual designs of an automated separation system of cesium isotopes, its virtual instrument system based on the LabVIEW state machine architectures and an automated radiochemical procedure were described for the purification of cesium isotopes at trace levels found in seawater discharged from the various nuclear facilities.

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.

Plasma melting technology uses electrical arc phenomena such as lightning to create hightemperature sparks of about 1,600 degrees or more to meet waste disposal requirements through treatment and reduction without distinguishing radioactive waste generated during nuclear power plant operation and dismantling according to physical characteristics. Decommissioning radioactive waste scabbed concrete occurs when polishing and cutting the contaminated structure surface to a certain depth. In this study, Scabbed concrete containing paint was confirmed for volume reduction and disposal safety using plasma treatment technology, and it is planned to be verified through continuous empirical tests.

Recently, BNS (Best System) developed a system for evaluation and classification of soil and concrete wastes generated from nuclear power plant decommissioning. It is composed of various modules for container loading, weight measurement, contamination evaluation, waste classification, stacking, storage and control. The contamination evaluation module of the system has two sub modules. One is for quick measurement with NaI (Tl) detector and the other is for accurate measurement with HPGe detector. The container used at the system for wastes handling has capacity of 100 kg and made of stainless steel. According to the measurement result of Co-60 and Cs-137, the waste is classified as waste for disposal or waste for clearance. Performance of the system was demonstrated using RM (Reference Material) radiation source. This year, necessity of system improvement was suggested due to revised operation requirements. So, the system should show throughput of more than 1 ton/hr and Minimum Detectable Activity (MDA) of less than 0.01 Bq/g (1/10 of criteria for regulatory clearance) for Co-60 and Cs-137. And soil waste become main target of the system. For this, the container used for soil waste handling should have capacity of 200 kg. As a result, material for the container need to be changed from stainless steel to plastic or FRP (Fiber Reinforced Plastics). And large area detector should be introduced to the system to enhance processing speed of the system. Additionally, container storage rack and conveyor system should be modified to handle 200 kg capacity container. Finally, moving path of the container will be redesigned for enhanced throughput of the system. In this paper, concept development of the system was suggested and based on that, system development will be followed.

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.

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.

Wet solid wastes including spent ion exchange resins, evaporator concentrates and sludges require solidification to transform wastes into an acceptable solid, monolithic form for final disposal. The development of the process control program for the solidification of radioactive sludges generated at nuclear power plants has been in progress to provide reasonable assurance that the solidified product will meet the established waste acceptance criteria for solidified waste. A mobile solidification system to produce the solidified waste in the size of a 200 L drum was used, which adopts the in-line mixing method where the waste and binder are mixed and then transferred to the disposable container. To simulate radioactive sludges, non-radioactive sludges are synthesized and the specimens are prepared by using them. The qualification tests on the prepared specimens including the compressive strength test, the thermal cycling test, the irradiation test, the leach test, the immersion test, etc. have been performed to qualify recipes for a range of waste compositions. The results of the tests will be analyzed and discussed.

The radioactive waste generated within radiation-controlled areas is classified and processed according to relevant laws and regulations based on contamination levels. In cases where such radioactive waste complies with the legally defined clearance concentration or dose criteria, it is disposed of as non-radioactive waste by means of incineration, reclamation, recycling, etc. Within radiation controlled areas, various consumables are periodically replaced to ensure the proper operation of the area. It is necessary to have appropriate disposal methods for these consumables. In particular, waste items such as fire extinguishers, fluorescent lamps, batteries, and pressure vessels (hereinafter referred to as “Special Waste Type”), which may contain hazardous substances within their internal components and contents, should be considered for appropriate disposal methods that comply with nuclear safety and environmental laws. In the present case, the specified special waste type do not come into direct contact with radiation sources, and they have impermeable surfaces, which significantly reduces the risk of external contamination infiltrating the interior. However, the current method of clearance is not suitable for these items (Typically, nuclear energy-related business operators are required to classify clearance target waste based on internal and external components and demonstrate compliance with the criteria. Nevertheless, for special waste type, it is difficult to separate and measure internal and external components within the radiation-controlled area). In this case, the Clearance Procedure for special waste type applied to Korea Atomic Energy Research Institute was introduced. Additionally, we have extracted considerations for future domestic clearance of the type.