연구분야
HOME > 연구실소개 > 연구분야
  • Uptake mechanism for radioactive iodine species to natural organic material (NOM) (2011-2012)

    Radioactive iodines are released into environments during nuclear fuel processing, nuclear weapon tests and nuclear accidents such as Chernobyl and Fukushima. Exposure to radioactive iodine leads to increase of metabolic disorders, mental retardation, and thyroid cancer in humans. Understanding transport of radioactive iodine in the environment is important to predict the contamination extent, and evaluate the risk to human and ecological systems. The fate and transport of iodine depends on the iodine speciation, because chemical interactions of iodine with geologic medium differ depending on which iodine species are present in environments. A variety of inorganic and organic forms are found in nature including iodide (I-), iodate (IO3-), and organic iodine. Iodine species are strongly assimilated to natural organic matter (NOM) in aquifer sediments. Black carbon (BC) is one type of particulate NOM formed by incomplete combustion of fossil fuels and pyrolysis in forest fires. The BC has great sorption capacities with nonlinearity for organic contaminants because of high surface area from ubiquitous micropores. Based on these studies, we hypothesize that particulate NOM (BC) can be also effective sorbent for iodine species such as general NOM (humic materials) in aquatic and groundwater environments. The goal of this study is to determine the uptake characteristics of particulate NOM (i.e., BC) for different iodine species such as iodide and iodate. Specific objectives are to compare the uptake of iodine species to humic substance and BC as representative of general and particulate NOM, respectively, and to understand uptake mechanism of iodine species to the BC.
    • 1. Choung S, Wooyong Um, M Kim and MG Kim. 2013. "Uptake mechanism for iodine species to black carbon", Environmental Science and Technology, 47, 10349-10355.
  • Effects of Radiation and Temperature on Iodide Sorption by Surfactant-Modified Bentonite (2012-2013)

    The mobility of radioactive iodine is strongly controlled by iodine speciation under certain pH-Eh conditions. Dominant iodine species are anionic forms, such as iodide (I-), iodate (IO3-), and organo iodine in the natural environment. Iodide, which is mainly found under reducing conditions in deep radioactive waste repositories, possesses a low sorption affinity for geologic media because the surfaces of many minerals are mostly negatively charged at circumneutral pH condition. Bentonite has been regarded as a qualified backfill material for engineered barriers in radioactive waste disposal sites. However, bentonite is ineffective for sorbing anionic radionuclides such as iodine species, because it has both a permanently negatively charged surface like other aluminosilicate minerals and a pH-dependent negative surface charge caused by deprotonation of the surface hydroxyl group under high pH conditions. Therefore, surface modification of bentonite has been proposed to enhance its sorption capacity for anionic radionuclides. The main goal of this study was to address the potential use of SMB as a material for a final engineered barrier to prevent the release of radioactive iodide under similar subsurface conditions within and from underground geological repositories.
    • 1. Choung S, Minkyung Kim, Jung-Suk Yang, and Wooyong Um. 2014. “Effects of Radiation and Temperature on Iodide Sorption by Surfactant-Modified Bentonite”, Environmental Science and Technology, 48(16), 9684-9691.
  • Development of iron phosphate ceramic waste form to immobilize radioactive waste solution (2011-2013)

    To develop an iron phosphate ceramic (IPC) waste form using converter slag obtained as a by-product of the steel industry as a source of iron instead of conventional iron oxide. Both synthetic off-gas scrubber solution containing technetium-99 (or Re as a surrogate) and LiCl-KCl eutectic salt, a final waste solution from pyrochemical processing of spent nuclear fuel, were used as radioactive waste streams. The IPC waste form was characterized for compressive strength, reduction capacity, chemical durability, and contaminant leachability. Consequently, the IPC waste form developed using converter slag has the potential capability for immobilizing radioactive waste solutions. The large amount of iron, high reduction capacity, and low cost make the convert slag better and more economical as a substitute for the conventional iron oxides in the production of the IPC waste form.
    • 1. Choi, J., Um, W., & Choung, S. (2014). Development of iron phosphate ceramic waste form to immobilize radioactive waste solution. Journal of Nuclear Materials, 452(1-3), 16-23.
  • Liquid Scintillation Counting Methodology for 99Tc Analysis (2013-2014)

    Technetium-99 (99Tc) is a redox sensitive element and exists in many oxidation states in the environment. The most common found and environmentally stable Tc specie is TcO4-. An analogues specie of Tc is Rhenium (Re) which is ReO4- specie similar to TcO4-. Our current research activities are focused to discover alternate Re species that are produced from ReO4- reduction and are environmentally stable. During our research many species were discovered that are stable. We discovered a cyclic bi-metallic Re-Sn oxide that exists in solid form, and is hard to dissolve. We also discovered ReO as mobile aqueous stable specie. Apart from these oxide species, we discovered a number of Re-EDTA complexes that are formed during ReO4- reduction in the presence of EDTA. These complexes can combine together to form multiple molecular weight complexes. Similar research are ongoing to discover more species that are possible candidates for Re/Tc transport and environmental persistence. Accurate analysis of 99Tc in environmental sample depends upon analysis highly depends upon many factors. Our research found a LSC methods for 99Tc analysis that has significant savings in analysis cost and time. Method efficiency was measured in the presence of 1.9 to 11 900 ppm total dissolved solids. The resultant quench curve proved to be effective for quantifying spiked 99Tc activity in deionized water, tap water, and groundwater, seawater, and urine samples. Counting efficiency was found to be 91.66% for Ultima Gold LLT, and Ultima Gold. Relative error in spiked 99Tc samples was ±3.98% in Ultima Gold and Ultima Gold LLT. Minimum detectable activity was determined to be 25.3 and 22.7 mBq for Ultima Gold LLT and Ultima Gold respectively. A pre-concentration factor of 1000 was achieved at 100 °C for 100% chemical recovery.
    • 1. Khan, M.; Um, W., "Liquid Scintillation Counting Methodology for (99)Tc Analysis: A Remedy for Radiopharmaceutical Waste". Anal Chem 2015, 87, (17), 9054-60.
  • Effect of seawater intrusion on radionuclide transport under subsurface environment (2013-2016)

    Most of the nuclear power plants (NPPs) and their yard storage facilities are located on or close to the shoreline because the NPPs need for cooling water. In a case of a severe accident, the radionuclides may leave the storage sites and transport in the subsurface environment in which nearby seawater can introduce through shattered fractures and cause the salinization of primary groundwater aquifers. Changes in porewater salinity may trigger ion exchange and dissolution-precipitation reactions that will affect the solubility and mobility of radionuclides associated with subsurface sediments.
    Radioactive strontium (90Sr, t1/2=28.1 years), as one of the significant components of radioactive waste, is electrostatically bound to the matrix surfaces as outer-sphere complexes with ion exchange as the dominant sorption mechanism.
    90Sr has sensitive sorption behavior so that the released 90Sr into the groundwater due to the wreck of storage tanks can be highly mobile, and transport behavior will easily change depending on the salinity degree. In our laboratory, the exchangeable/competitive 90Sr sorption reaction and transport behavior have been evaluated using the change convention models, e.g., Vanselow, Gaines-Thomas, and Cernik at a different degree of ionic strength assuming seawater intrusion in groundwater. We are also studying to develop site-specific ion exchange model to quantify and predict the mobility of 90Sr under subsurface environment.
  • Microbial communities in underground disposal facility (2013-2017)

    The microbial communities in underground disposal facility for the low- and intermediate-level waste (LILW) are important in long-term safety in terms of gas generation and radionuclide transport. In fact, radionuclides could be released to the nearby environments due to geochemical reaction and microbial activity during the life-span of the disposal facility. Moreover, as LILW consist of organic waste, the gas generation inside disposal facility which results from microbial degradation of LILW becomes great concern. However, the bacteria in underground disposal facility remain unexplored and previous studies have been only focused on the culture-dependent method which can provide limited information.
    • 1. Choung S, Wooyong Um, Seho Choi, AJ. Francis, Sungpyo Kim, Jin beak Park, and Suk-Hoon Kim. 2014. “Biogeochemical Changes at Early Stage After the Closure of Radioactive Waste Geological Repository in South Korea”, Annals of Nuclear Energy, 71, 6-10.
  • CRUD chemical decontamination (2015-2016)

    As nuclear power plants are getting older, gradual accumulation of CRUD in nuclear power plant is in progress, so decontamination process has been concerned. In case of Republic of Korea, the lifespan of kori-1 is expired in 2017, South Korea is preparing to enter a new era of nuclear decommissioning. Chalk River Unidentified Deposit (CRUD) is a technical term in nuclear engineering which is an accumulated material on external fuel rod cladding surfaces in nuclear power plants. It is a corrosion product which is composed of either dissolved ions or solid particles such as Ni, Fe and Co. Corrosion product can be activated through nuclear reactions to form radio-nuclides such as Cobalt-60 (60Co). It can affect to human health and environment, so decontamination process is essential for reducing occupational exposures, limiting potential releases and uptakes of radioactive materials
    In our study, we synthesize simulated CRUD, nickel ferrite which is main phase composition of outer layer of CRUD and cobalt ferrite was composed as powder foam, and corrosion product of CRUD was generated on surface of metallic coupon which is compose of SUS 304 (Ni: 8~11%, Cr: 18~20%). After simulated CRUD was synthesized, dissolution test using the simulated CRUD with different chemical reagents (Oxalic acid, Malonic acid, Citric acid, and Nitric acid) to find the good condition (pH and Temperature) for effective cobalt released.
    • W. Kim, S. Nam, S. Chang, H. Kim, and W. Um, “Removal of Chalk River unidentified deposit (CRUD) radioactive waste by enhanced electrokinetic process”, Journal of Industrial and Engineering Chemistry, 2017, xx(xx), pp xxxx-xxxx
  • Studtite formation and dissolution (2015-2016)

    Deep geological disposal is a promising method to dispose of spent nuclear fuel (SNF). Because fission products and other radionuclides may leach from the SNF, the radionuclide transport should be correctly predicted and prevented to ensure safe operation and long-term performance assessment of repositories. Studtite (UO4 4H2O) is formed by reaction between uranyl ion (UO22+) that may leak from SNF and H2O2. Therefore, the dissolution kinetics of studtite in natural environmental conditions is important factor for the performance and risk assessment of repositories. The objectives of this study are to investigate the dissolution kinetics and mechanisms of uranium release from studtite in different solution conditions. The results will be useful to assess the comprehensive transport of uranium from both nuclear waste and SNF stored in deep geological repositories.
  • Uranium removal using TBP-coated hydroxyapatite (2016-2018)

    Uranium (U) is highly toxic and radioactive element and can affect the harmful risk to human health and environment. Efficient and rapid removal of radioactive contaminants is crucial when they are released to the environment through severe nuclear accidents. The objective of this study is to develop a new adsorbent in order to get a higher removal efficiency of U from various waste streams as well as for recovery of U from seawater. Here, the combination of hydroxyapatite and Tributyl Phosphate (TBP) can be expected to produce adsorbents with enhanced adsorption properties. This results will be applicable for decontamination of radioactive contaminant and recovery of U from seawater for contributing to stable supply of raw U material for nuclear fuel.
    • 1. H. Kim, W. Um, W-S. Kim, and S. Chang, “Synthesis of Tributyl Phosphate-Coated Hydroxyapatite for Selective Uranium Removal”, Industrial & Engineering Chemistry Research, 2017, 56(12), pp 3399-3406
  • Synthesis of graphene oxide based material for removing radioactive iodine (2016-2019)

    Radioactive iodines (125/129/131I) are released into environments during nuclear fuel processing, nuclear power plant operation and severe accidents such as Chernobyl and Fukushima. Environmental concern is mostly for 129I due to its high toxicity and long half-life, t1/2 = 1.6 x 107 years. Usually, to remove the radioactive iodine, activated carbon based organic materials and silver exchanged zeolite were used as sorbents in aqueous phase. However, these materials have the limits in removal efficiency, cost, and high selectivity. Therefore, this study investigated the bismuth functionalized graphene oxide (Bi-GO) to test high removal efficiency and selectivity of iodide and iodate species under the various Cl- background solution conditions. In Summary, we demonstrate bismuth-functionalized graphene oxide synthesis as novel absorbent for iodine removal and compare the results with the commercial Ag-zeolite under different ionic strength conditions. The bismuth based materials show higher iodine removal efficiencies (≥95%) than Ag-zeolite for iodide and iodate. In contrast, the Ag-zeolite has a limitation for removal capacity for iodate.
  • 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.
  • 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.
  • 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. Bonding structure of radionuclide in activated and contaminated concrete will be investigated in this study. It can be applied to development of decontamination method and reduce the volume of concrete waste. Because Kori-1 plant is considered to start the decommissioning next year, the result of this study can be applied to the decommissioning of Kori-1 plant and decontamination of contaminated concrete.
  • 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.
  • 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.
  • 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).
  • 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.
  • 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.
  • 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.
  • 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.
  • 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.
  • 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).
  • 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.
  • 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 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.