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  • Safety Verification Technology Development for Engineered Barrier under Deep Geological Disposal System (2021-2029)

    Safety verification technology should be developed for at least thousands of years to dispose of high-level radioactive waste, including spent nuclear fuel. Significantly, the engineered barrier's safety verification, including canister, buffer, backfill, and closure, has to be developed and estimated to dispose of high-level radioactive waste safely. In addition, the development of efficient simulation model tools is carried out for long-term assessment. Even though foreign countries make various model tools, new models considering domestic situations such as nuclear environment, geological environment, etc. should be developed. Furthermore, in this project, abnormal scenarios such as weather, seismic activity, glacial epoch, etc. are considered and assessed. As a result, the main objective of this project is to develop safety verification technology for domestic conditions.
  • Testing of solubility, sorption, and diffusion, of non-radioactive isotopes under various disposal environments (2021-2029)

    The deep disposal repository for spent nuclear fuel (SNF) consists of an artificial barrier (buffer materials) and a natural barrier (rock) to permanently isolate high-level radioactive waste from the human living environments. To verify the safety and stability of the deep disposal repository, it is necessary to secure geochemical characteristics to understand the behaviors of nuclides such as solubility, sorption, and diffusion in a multi-barrier system. In this study, studies on the solubility, sorption, and diffusion of nuclides in various disposal environments (oxidation, reduction, high temperature/brine/strong base) using non-radioactive isotopes of representative nuclides are conducted.
  • Development of high performance buffer material and long-term safety assessment(2021~2025)

    The bentonite buffer, one of the engineering barriers in the high-level waste repository, minimizes the inflow of groundwater to the repository, prevents the leakage of radionuclides to the environment, and protects the container from external shock. In general, bentonite show good removal efficiencies for cation radionuclides, however not for anion radionuclides. The development of high-performance bentonite is the key technology to enhance the safety and efficiency of the repository by increasing the removal efficiencies of radionuclides and improving thermal conductivity. This study aims to improve the cation/anion sorption properties of bentonite by various surface modification methods and decrease the diffusion of radionuclides in the repository.
  • Development of technology for C-14 treatment from spent resin and their stabilization (2019~2022)

    Radioactive spent resins can be directly immobilized to cement waste form or intermediate products resulted from decontamination or oxidation processes can also be solidified in solid matrix indirectly. In this study decomposition product of spent resin 14CO2 will be captured in saturated Ba(OH)2 solution to form Ba14CO3 which can be incorporated into modified cement matrix. Summary of proposed study is to treat the spent ion exchange resin by isolation of hazardous 14C radionuclide from resin matrix by decontamination/acid stripping, its volume reduction by advanced oxidation processes and stabilization/solidification of isolated 14C species as long-term safe and immobile waste form. Carbonate salts (BaCO3 or CaCO3) and residual spent resin from Task 2 will be encapsulated by sintering with phosphate glass frit and polyceramics waste forms, respectively.
  • Development of Nuclide Migration and Prediction Model of Nuclide Adsorption Characteristics (2019-2020)

    In order to fulfill the review requirements related to the construction and operation permit of the 2nd Phase of LILW Disposal Facility Construction, the behavioral characteristics (dissolution, adsorption, diffusion) of major nuclides in the waste for the 2nd phase site and the artificial barriers are analyzed experimentally using actual water and medium samples from the site. In addition, to reduce the long-term uncertainty of the safety of the disposal site, a model is proposed to predict the change of adsorption characteristics of the major nuclides in the groundwater conditions.
  • Development of waste package, transportation and disposal containers for decommissioining waste of nuclear power plant (2018-2023)

    Kori Unit 1 which started commercial operation for the first time in Korea, was permanently shut down last year. Considering the completion of take out of spent fuel, a total of 14,500 drums are expected to be generated at the completion of the decommissioning. Decommissioing radioactive wastes occur in large quantities in a short period of time, and they occur in various forms and levels. Therefore, it is necessary to develop containers different from operational waste container. In addition, the development of new containers and systems can reduce the administrative processing, manpower, and time required for handling and measurement in transportation containers. In this study, it is aimed to develop packaging, transportation and disposal container of decommissioning radioactive waste and to secure disposal suitability evaluation technology based on the radioactive waste transfer process.
  • Treatment of Radionuclide from CRUD using Underwater Microwave Plasma (2018-2021)

    As the operating period of nuclear power plants increases, radioactive contaminated metal waste is generated from the facility. In the process of decommissioining and decontamination of nuclear power plants, radionuclides such as cobalt (60Co) are deposited on metal surfaces such as CRUD in a primary system. Until now, Chemical decontamination using low concentration organic acids is known to be effective for contaminated metal wastes. However, chemical decontamination method produces waste streams such as secondary decontamination effluents containing non-degradable organics and non-reusable resins. It is necessary to reduce the amount of secondary wastes and safely preserve the environment by eliminating these non- degradable organics through optimum decomposition reaction. In this study, it is aimed to develop a cobalt (60Co) removal system that exists in the CRUD and decontamination effluent more effectively than the conventional method using the underwater micro-plasma technology.
  • Development of contaminated soil decontamination technology for dismantled NPP’s using selective nuclide adsorption technique and microbubble vortex breakdown based on high pressure cleaning (2018-2021)

    Radionuclides released to the environment for a variety of reasons are slowly contaminated by rainwater or snow through the surface of the earth. According to a few reports, it contaminated up to 1 m depending on the surrounding environment and soil. Several months after the Fukushima nuclear power plant accident, radioactivity was measured in contaminated soil samples. Most radioactivity was contained in fine soil below 75um. In the case of decommissioning the actual nuclear concrete building starting with Kori Unit 1, a considerable amount of radioactive materials is attached to the concrete surface composed of sand or gravel. Furthermore, in order to efficiently remove contaminated soil, it is necessary to evaluate the characteristics and grain size of the soil around the target plant, and then to develop a decontamination apparatus and process by making a simulated soil containing a standard radioactive material.
  • Solidification of Wastes from Decommissioning Processes (2018 - 2019)

    During the chemical decontamination for primary reactor of nuclear power plants (NPPs), ion-exchange resin wastes can be considerably released. In addition, there are some problems to dispose these organic wastes originated from ion-exchange resin. To solve these problems, the inorganic material, as the adsorbent, can become promising way to remove the radioactive isotope of waste solution; then, it can be chemically solidified to be stable before putting into intermediate and low level radioactive waste disposal site, and this process can increase the receptivity efficiency. In this project, the main objectives are (1) the development of novel inorganic adsorbent, (2) the volume reduction of waste forms, and (3) the disposal suitability increase in waste forms. As a result, our laboratory is conducting the research and experiments as follows. In case of the (1), chalcogels and cancrinite are used as the adsorbent to remove the radioactive isotope (Co, Fe, Ni, Cr, Mn). In case of the (2), vitrification and ceramicrete processes are used for the solidification.
  • Advanced Nuclear Environment Research Center (ANERC) (2017-2022)

    The objective of this research is the development of decontamination method for reducing the volume of contaminated concrete and soil waste after nuclear power plant (NPP) decommissioning, the characteristic of ceramicrete waste forms for the long-term in-situ sustainability of waste forms, the evaluation of disposal safety and suitability of decontamination waste decommissioning, and the development of soil remediation and environment assessment method. In addition, the expert training and education of graduate students will be accomplished for NPP decommissioning and decontamination field through the research in the Advanced Nuclear Environmental Research Center.Theseresearch results are planned to provide the new decontamination method for waste(concrete and soil) volume reduction applied to the NPP decommissioning market growing in the world to increase of economic profit and the novel research case that bioremediation of radioactive contaminated soil and environment assessment method for site restoration of NPP.
  • Tritium separation using quantum sieve method (2017 - 2020)

    This study investigates the tritium separation mechanism through quantum sieve method. Based on the quantum sieve concept, different isotopes that have different masses can be separated by different diffusivities passing through the selected pores. The current research work will develop a modified quantum sieve system combined with electrolytic enrichment using novel catalysts, distillation, or isotopic exchange method to increase the separation factor higher than the values before. The unseparated vapors will be coupled with composite process that contains distillation method, improved electrolysis, gas phase quantum sieving and isotopic exchange. Using this coupled cyclic process tritium separation efficiency will be improved and separation cost and operation energy will be reduced.
  • Safety Case / Risk Management for Decommissioning and Decontamination including Spent Fuel Management (2017-2020)

    The purpose of this study is to develop a harmonized approach to safety assessment and to define the elements of safety assessment for decommissioning including the application of a graded approach. In addition, this study is to investigate the practical applicability of the methodology and performance of safety assessments for the decommissioning of various types of phase/facility. This study focus on developing methodology for safety assessments and risk management for decommissioning activities and the development of a regulatory approach for reviewing safety assessments for decommissioning activities including spent fuel management and as a basis for regulatory decision making.
  • Radiological Characterization / Radionuclide transport analysis for Decommissioning and Decontamination (2017-2020)

    Radionuclides of interest resulting from neutron activation, fission and the presence of actinides from fuel are essential information for Decommissioning and Decontamination. The research and development in this field should be performed through simulations using computer codes, and typical methods of sampling and measurement, with references to specialized literature. This study focus on characterization and radionuclide transport mechanism for Decommissioning and decontamination. Another important aspect of this study is to derive the influence factor of radioactive inventory on decommissioning planning and strategy.
  • Development of continuous treatment system for decommissioining radiowaste based on mechanochemistry process (2017-2020)

    Following the decommissioning phase, various radioactive wastes are generated in large quantities through decontamination that remove contaminated radioactivity from nuclear facilities and buildings. In this study, the disposal cost increases proportionally as the quantity of wastes to be disposed increases. For this reason, there is a need for technology to compress the volume of decommissioining waste using mechanochemical energy. It also develops disposal technologies that can easily dispose of wastes by stabilizing liquid and non-degradable wastes so as to replace the liquid waste and non-degradable wastes with disposal restrictions in HIC(High Integrity Container).
  • Solidification of radioactive waste after decontamination (2017-2019)

    Development and characterization of solidified waste forms (cement and geopolymer) compatible with the disposal environment for the secondary radioactive waste sludges containing radioactive Co and Cs after chemical decontamination. The solidification method using cement and geopolymer waste form will be developed and used for radioactive wastes generated from decontamination process of nuclear facilities and NPP in the world. In addition, the results of this study can be used for safe storage of radioactive wastes in radioactive waste repository. The results of this study will be used to solidify radionuclides in secondary waste generated after chemical decontamination process in nuclear power plants. The successful results of the proposed waste form development are expected to be applied for one of the decommissioning technologies nuclear power plants including Kori-1 plant.
  • 3H removal from groundwater (2017-2019)

    Tritium (3H) is a radioactive isotope of hydrogen. Naturally occurring tritium is extremely rare on earth but it can be produced by irradiating lithium metal or lithium bearing ceramic pebbles in a nuclear reactor. The tritium which from nuclear power plants is harmful to human health when inhaled, ingested via food or water, or absorbed through the skin, because tritium is a low energy beta emitter.
    The objective of our project is the evaluation and development of the tritium removal method using protonic manganese oxide spinel adsorbent (PMOS). The tritium can be removed by PMOS, and PMOS which adsorb the tritium need to be separated. In this research, tritium removal technology is developing by using hybrid method with PMOS and mesoporous polymer which synthesized silica-cation surfactant. In this research, tritium removal capacity will be evaluated using adsorption experiment and adsorption mechanism will be investigated. Tritium removal experiment will be done using ion exchange membrane in laboratory-scale. Pilot-scale reactor for tritium removal experiment will be done.
  • Transport simulation using GoldSim (2017-2019)

    GoldSim is the Monte Carlo simulation software solution for dynamically modeling complex systems in engineering, science and business and it is the premier tool in the world for carrying out probabilistic performance assessments of proposed and existing radioactive waste management sites. These performance assessments all utilize the GoldSim Radionuclide Transport (RT) Module, which includes specialized and powerful features to facilitate simulation of radionuclide transport within a range of environmental media. The RT Module can accurately and efficiently model complex processes such as decay and ingrowth of reaction/decay products, solubility constraints, sorption onto porous media, release from engineered barriers, diffusive transport, and transport of contaminants on particulates. The main purpose of this study is to investigate transport simulation of radioactive iodine under Wolsung repository natural condition after disposal using GoldSim Radionuclide Transport Module, which helps to understand the role of selected parameters in the near-field region of the final repository and to prepare an own complex model of the repository behavior considering the advection, diffusion, and retardation.
  • Solidification of radioactive waste after decontamination of primary system (2017 - 2019)

    This study shows Development and characterization of solidified waste forms (cement and geopolymer) compatible with the disposal environment for the secondary radioactive waste sludges containing radioactive Co and Cs after chemical decontamination.The solidification method using cement and geopolymer waste form will be developed and used for radioactive wastes generated from decontamination process of nuclear facilities and NPP in the world. In addition, the results of this study can be used for safe storage of radioactive wastes in radioactive waste repository and expected to be applied for one of the decommissioning technologies nuclear power plants including Kori-1 plant.
  • Study on the fate of radionuclide in radioactive waste by XAS analysis (2017 - 2018)

    Recently, decontamination and decommission of nuclear power plants became a rising issue followed by the decommissioning of Kori-1 plant. Concrete materials in nuclear facilities may become contaminated or activated by various radionuclides through different mechanism. The objective of this study is to investigate the molecular scale characterization of radionuclides interaction and bonding structure.This study will give the information of bonding structure of radionuclide in concrete waste after decommissioning of nuclear facilities. Appropriate decontamination method can be developed using the result of this study. It can reduce the volume of considerable quantities of concrete waste produced from nuclear facilities.
  • Study of contaminant removal by Fe oxide transformation process (2017 - 2018)

    The overall objectives of this proposal are to understand of 99Tc removal mechanism by co-precipitation method during iron oxide transformation process, develop the optimized technology for 99Tc removal, determine the changes of efficiency and kinetics in 99Tc removal during iron transformation with and without the presence of competing contaminant Cr(VI), and investigate a scientific mechanism for 99Tc binding mechanism and molecular-level understanding using molecular dynamic modeling. The results through this research will be also used to prepare another international project, and the students and researchers will continuously work in the field of contaminant remediation, which can apply for the fusion study between earth science and environmental and nuclear engineering subjects.
  • Iodine study in various environments (2016-2024)

    Radioactive iodine from nuclear power plants (NPPs) operation, NPPs decommissioning, and severe accident is quite dangerous element as adversely affecting human body. In addition, radioactive iodine species such as I-131, I-135, I-125, and I-129 show high mobility, toxicity, and radioactive energy. For this reason, the iodine removal has been studied by making and using inexpensive and high efficiency adsorbents. Furthermore, NPPs are usually located in waterfront to secure the mass cooling water, and radioactive waste disposal site in Korea is also located in waterfront. Because of the geological characteristic, the ground and sea water can be penetrated or flowed in the underground of NPP facilities, and it makes the unsaturation condition of the underground saturated. In case of the deterioration of drainpipe or unexpected severe accident, the radioactive nuclides like iodine indicating high mobility can be leaked due to saturation condition above. To prepare emergencies like the situation above, the radioactive iodine transport model development are required, and then, Kd value (distribution coefficient) can be obtained through this research. Also, iodine shows high volatilization depending on geological condition (pH, Eh, temperature, etc.). So far, few researches have been proceeded to set up and control iodine volatilization mechanism. For the safe management of volatile iodine after NPPs decommissioning, iodine volatilization mechanism should be investigated. As a result, our laboratory is carrying out three types of iodine studies which are (1) iodine adsorbent development, (2) iodine transport model development, and (3) iodine volatilization mechanism development.
  • Laboratory batch sorption and column tests for radionuclides at NPP sites (2012, 2013, 2014, and 2017-2018)

    The liquid radioactive wastes are stored temporarily in storage tanks at nuclear power plants (NPPs) or nuclear facilities before treatment. However, the radionuclides such as 3H, 137Cs, 90Sr and 60Co can be released to the subsurface environment due to deteriorated tanks or unexpected severe accidents during the storage and NPP operation period, respectively. In the case of accidental leaks, these radionuclides would be a potential risk and pollution source for the surrounding environments. Therefore, we studied on sorption and transport behavior of 90Sr, 137Cs and 60Co using core rock samples (surface soil, fractured rock, and bedrock) collected at various NPPs located in near the East Sea, South Korea. The respective sorption distribution coefficient (Kd) of 90Sr, 137Cs, and 60Co were evaluated to compare as a function of the solid type or varying total ionic strengths from groundwater to seawater through batch experiments. Also, saturated column experiments were used to investigate their mobility using groundwater, seawater, and mixed groundwater-seawater solutions to improve our understanding of radionuclides transport behavior beneath NPPs located nearby the ocean. The data from both experiments represent preliminary data used to develop site-specific models for prediction of 90Sr mobility under NPP sites (Shin-Kori, Kori, Wolseong, and Shin Hanwool).