Minimizing of radiation exposure for the operating and decommissioning personnel is a key indicator for safe operation of the NPP. This is reflected in the application of the ALARA (As Low As Reasonable Achievable) principle. The main objectives of radiation management during full system decontamination for NPP decommissioning are to reduce the exposure dose, prevent contamination of the body and reduce solid radioactive waste. In order to reduce exposure of workers, the dose rate should be reduced by installing a temporary shield after evaluating the dose rate for the piping, component and decontamination equipment of the decontamination path before full system decontamination. Furthermore, unnecessary exposure to radiation should be reduced by thoroughly entering and exciting the radiation area and limiting the access to the high-radiation area except for workers or persons concerned. A telemetric dosimetry system should be as installed to remotely monitor radiation levels at different locations within the decontamination flow path. Remote monitoring of radiation fields using teledosimetry worked well in assessing process effectiveness and is highly recommended. However, care must be taken to place the detectors in appropriate locations. For the prevent of body contamination, it is necessary to install a fence using a heat-resistant waterproof sheet to prevent leakage of highly radioactive contamination water. When replacing high-dose filters and ion exchange resins, it is necessary to remotely monitor to reduce the exposure dose of workers.

The Korea government decided to shut down Kori-1 and Wolsung-1 nuclear power plants (NPPs) in 2017 and 2019, respectively, and their decommissioning plans are underway. Decommissioning of a NPP generates various types of radioactive wastes such as concrete, metal, liquid, plastic, paper, and clothe. Among the various radioactive wastes, we focused on radioactive-combustible waste due to its large amount (10,000–40,000 drums/NPP) and environmental issues. Incineration has been the traditional way to minimize volume of combustible waste, however, it is no longer available for this amount of waste. Accordingly, an alternative technique is required which can accomplish both high volume reduction and low emission of carbon dioxide. Recently, KAERI proposed a new decontamination process for volume reduction of radioactivecombustible waste generated during operation and decommissioning of NPPs. This thermochemical process operates via serial steps of carbonization-chlorination-solidification. The key function of the thermochemical decontamination process is to selectively recover and solidify radioactive metals so that radioactivity of the decontaminated carbon meets the release criteria. In this work, a preliminary version of mass flow diagram of the thermochemical decontamination process was established for representative wastes. Mass balance of each step was calculated based on physical and chemical properties of each constituent atoms. The mass flow diagram provides a platform to organize experimental results leading to key information of the process such as the final decontamination factor and radioactivity of each product.

Korea faces decommissioning the nation’s first commercial nuclear power plant, the Kori-1 and Wolseong-1 reactors. In addition, other nuclear power plants that will continue to operate will also face decommissioning over time, so it is essential to develop independent nuclear facility decommissioning and site remediation technologies. Among these various technologies, soil decontamination is an essential not only in the site remediation after the decommissioning of the highly radioactive nuclear facility, but also in the case of site contamination caused by an accident during operation of the nuclear facility. But the soil, which is a porous material, is difficult to decontaminate because radionuclides are adsorbed into the pores. Therefore, with the current decontamination technology, it is difficult to achieve the two goals of high decontamination efficiency and secondary waste reduction at the same time. In this study, a soil decontamination process with supercritical carbon dioxide as the main solvent was presented, which has better permeability than other solvents and is easy to maintain critical conditions and change physical properties. Through prior research, a polar chelating ligand was introduced as an additive for smooth extraction reaction between radionuclides present as ions in soil and nonpolar supercritical carbon dioxide. In addition, for the purpose of continuity of the process, a candidate group of auxiliary solvents capable of liquefying the ligand was selected. In this research evaluated the decontamination efficiency by adding the selected auxiliary solvent candidates to the supercritical carbon dioxide decontamination process, and ethanol with the best characteristics was selected as the final auxiliary solvent. In addition, based on the decontamination effect under a single condition of the auxiliary solvent found in the Blank Test process, the possibility of a pre-treatment leaching process using alcohol was tested in addition to the decontamination process using supercritical carbon dioxide. Finally, in addition to the existing Cs and Sr, the possibility of decontamination process was tested by adding U nuclides as a source of contamination. As a result of this research, it is expected that by minimizing secondary waste after the process, waste treatment cost could be reduced and the environmental aspect could be contributed, and a virtuous cycle structure could be established through reuse of the separated carbon dioxide solvent. In addition, adding its own extraction capacity of ethanol used for liquefaction of solid-phase ligands is expected to maximize decontamination efficiency in the process of increasing the size of the process in the future.

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.

A large amount of concrete radioactive waste is generated during the decommissioning of nuclear facilities, including nuclear power plants, and it is expected that the radioactive waste management expenses will be huge. In order to reduce the concrete radioactive waste, a decontamination or removal process is required, but working on concrete may present a risk of worker exposure in a high-radioactive space. Therefore, in this study, a remote control integrated decontamination equipment that can reduce concrete radioactive waste and ensure the safety of workers during the concrete decontamination process in a high-radioactive space was developed. The integrated decontamination equipment consists of remote movement, automatic surface contamination measurement, automatic surface decontamination and debris/dust removal systems. The remote movement system generates ‘mapping data’ of topographic features for the work space and ‘location data’ that coordinates the location of the integrated decontamination equipment through LiDAR (Light Detection and Ranging) sensor and SLAM (Simultaneous Localization And Mapping) technique. The user can check the location of the integrated decontamination equipment through ‘location data’ outside the work space, and can move it by remote control through wired/wireless communication. The automatic surface contamination measurement system uses a radiation detector and an automatic measurement algorithm to generate ‘surface measurement data’ based on the measurement distance interval and measurement time set by the user. ‘Surface measurement data’ is combined with ‘location data’ to create a visualized map of radioactive contamination, and users can intuitively realize the location and degree of contamination based on the map. The automatic surface decontamination system uses a laser and an automatic removal algorithm to decontaminate the concrete surface. Concrete debris and dust generated during this process were collected by the debris/dust removal system, minimizing waste generation and radiation exposure due to secondary pollution. The integrated decontamination equipment developed through this study was applied with technologies that reduced radioactive concrete waste and ensured the safety of workers. If technology verification and on-site applicability review are performed using concrete specimens simulating nuclear power plant or similar environments, it is reasoned to contribute to the domestic and overseas decommissioning industry.

Organic waste generated by small and medium-sized (S&M-sized) metal decontamination in NPP decommissioning. To lower the concentration of these organic substances for a level acceptable at the disposal site, the project of “Development of Treatment Process of Organic Decontamination Liquid Wastes from Decommissioning of Nuclear Power Plants” is being carried out. The conditioning and treatment process of organic liquid waste was designed. Also, the literature was investigated to make simulated organic liquid waste, and the composition of these waste was analyzed and compared. As the decontamination agent, organic acids such as EDTA, oxalic acid, citric acid are used. The sum of the concentrations of these organic materials was set to a maximum value of 1,000 ppm. The major metal ions of the decontamination liquid waste estimated are 59Fe, 51Cr, 54Mn, 63Ni, and the concentrations are respectively 527, 163, 161, 159 ppm. Additional major metal ions are 60Co, 58Co, 137Cs. 58Co is replaced by 60Co because it has the same chemical properties as 60Co. Unlike the HLW, the contamination level of S&M-sized metal in primary system was quite low, so 60Co is set to 2,000 Bq/g. Considering the contribution of fission and gamma ray dose constant, 137Cs was estimated to 360 Bq/g. Also, suspended solids of decontamination liquid waste were set at 500 ppm. Under these assumptions, the simulated organic liquid waste was made, and then organic substances and metal ions were analyzed with TOC analyzer and ICP-OES. The TOC analysis value was expected to 392 ppm in consideration of the equivalent organic quantity. the test result was 302 ppm. Some of organics appears to have been decomposed by acid. The values of metal ions (Fe3+, Cr3+, Mn2+, Ni2+) analyzed by ICP-OES are 139, 4, 152, 158 ppm, respectively. A large amount of Cr3+ and Fe3+ were expected to exist as ions, but they existed in the form of suspended solid. Mn2+ and Ni2+ came out similar to the expected values. The designed conditioning and treatment process is largely divided into pretreatment, conditioning, and decomposition processes. After collecting in the primary liquid waste storage tank, large particulate impurities and suspensions are removed through a pretreatment process. In the conditioning process, treated liquid waste passes through UF/RO membrane system, and pure water is discharged to the environment after monitoring. Concentrated water is decomposed in the electrochemical catalyst decomposition process, then this water secondarily passes through the RO membrane system and then discharged to the environment after monitoring. Through an additional experiment, the conditioning and treatment process will be verified.

In a nuclear power plant, the activated corrosion products are deposited on the reactor coolant system. The activated corrosion products must be removed to reduce the radiation exposure to workers before maintaining or decommissioning of the nuclear power plant. In order to remove the remove the activated duplex oxide layer generated on the reactor coolant system in the pressurized water reactor (PWR), the Cyclic SP (Sulfuric acid/Permanganate)-HyBRID (Hydrazine Based Reductive metal Ion Decontamination) process developed by KAERI (Korea Atomic Energy Research Institute) can be used. After applying the Cyclic SP-HyBRID process, a sulfate-rich waste powder containing the radionuclide is generated, and the radioactive powder has to be stabilized for final disposal. In the previous study, it was confirmed that the low-temperature sintering method can be applied to immobilize the sulfate-rich waste powder. Thus, immobilization of the Cyclic SP-HyBRID process waste powder was carried out by the low-temperature sintering method using a low melting point glass, and the physicochemical and radiological characteristics of a waste form were evaluated in this study. As a result, the compressive strength of the waste form increased with increasing sintering temperature and sintering time. It is considered that the result was caused by the difference in the band gap between the bismuth borate and zinc borate, which are the products during the sintering process. It was verified that the physical stability was maintained after the 107 Gy of irradiation test. In addition, it was confirmed that the radioactive metal hydroxides contained in the waste powder converted to metal oxide forms, which have the lower solubility, at the sintering temperature. Finally, the waste form was evaluated as a low-level radioactive waste from the concentration of radionuclides contained in the waste form.

Reactor pressure vessels and steam generators generated in the process of dismantling nuclear power plants or replaced steam have various shape and occupy a considerable amount of the disposal site when disposed of in original shape. For the development of domestic technologies related to the disposal of large wastes, it is necessary to secure technologies for reducing large radioactive metal wastes, including technologies such as decontamination, cutting, melting, and residual radioactivity evaluation. Cases of disposal of steam generators in Europe and the United States were investigated. Except for u-tubes, steam generators are less contaminated or easily decontaminated, so it is possible to reduce the volume of waste subject to final disposal by exempting a significant amount through appropriate treatment. Korea Hydro & Nuclear Power Co. is currently temporarily storing 24 steam generators at 41.6 billion won. This paper presents a method to exempt more parts of the steam generator and reduce the volume of waste by properly combining mechanical cutting thermal cutting and melting to dispose of the steam generator. Currently the decontamination and dismantling industries of nuclear facilities are gradually expanding around the world. Therefore, it is necessary to localize the treatment technology for metal waste generated during maintenance and dismantling. The result of this study can be used to establish waste reduction and disposal method for dismantling steam generators.

Radioactively contaminated metal components from a nuclear power plant must be decontaminated to reduce the risk of radiation exposure to workers, which can be cleaned using a foam decontamination used to reduce the amount of wastewater significantly. Metal components with a fixed radioactive contamination can be effectively decontaminated using a foam consist of 0.5wt% nonionic surfactant, 0.5 M H2SO4, and 0.2 M Ce(SO4)2. However, strongly acidic wastewater is generated from the decontamination method, which contains a high concentration of the nonionic surfactant and ionic materials with radioactive nuclides. This wastewater must be treated as a stable form. In this study, an integrated process of precipitation and low pressure distillation was evaluated for the treatment of wastewater. It was confirmed that the surfactant and ionic materials were effectively removed from the wastewater through the integrated process.

As the number of nuclear power plants whose design life has expired worldwide increases, the attempts are continuing to complete the project of nuclear back-end cycle, the last task of the nuclear industry. Decontamination is essential in the process of dismantling nuclear facilities and restoration sites to remove all or some of the regulatory controls from an authorized facility. Among radioactive wastes, particularly contaminated soil is characterized by difficult physical decontamination because radionuclides are adsorbed between soil particles, that is, pores. Therefore, chemical decontamination is mainly used, which has the disadvantage of generating a lot of secondary waste. In order to overcome these disadvantages, an eco-friendly soil decontamination process is being developed that can drastically reduce the amount of secondary waste generated by using supercritical carbon dioxide. Supercritical carbon dioxide can easily control its physical properties and has both liquid and gas properties. However, since supercritical carbon dioxide is non-polar, additives are needed to extract polar metal ions, which are the goal of decontamination. Therefore, ligand with both CO2-philic and metal binding regions was selected. In previous studies, the decontamination efficiency of soil was evaluated by reacting contaminated soil with solid ligand and co-ligand at once. When solid ligands were used, the decontamination efficiency was lower than expected, which was expected because chemical substances were somewhat difficult to exchange in the closed process. In this study, in order to increase the efficiency of the decontamination process, the need for a process of liquefying ligand and continuously flowing it has been raised. Therefore, a co-solvent that dissolves well at the same time in SCCO2, ligand, and co-ligand was selected. In the selection process, a total of eight substances were selected by dividing into six polar substances and two non-polar substances through various criteria such as economic feasibility, eco-friendliness, and harmlessness. Thereafter, ethanol was finally selected through solubility evaluation for SCCO2 and additives. It is expected that a more effective decontamination process can be constructed when the additive is liquefied using a solvent selected from the results of this study.

In a nuclear facility, the base metal can be radiologically contaminated during the operation. They must be decontaminated to reduce the radiation exposure to workers before decommissioning of the nuclear facility. In order to decontaminate the nuclear facility, it is possible to apply a perfluorocarbon (PFC) based emulsion consisted of surfactant and decontamination reagent. The PFC has high resistance for the radiation decomposition, and PFC based emulsion can be easily stabilized using the ultrasonication method. During decontamination process, a dispersion stability of the emulsion affects to the decontamination performance because the decontamination reagents dispersed in the emulsion contact contaminated surface. In this study, the dispersion stability the PFC based emulsion was evaluated following the composition of the emulsion and dispersion condition such as temperature, ultrasonication time. It was confirmed that the concentration of surfactant is highly related to the dispersion stability from the result of Turbiscan analysis using the multiple light scattering method. It was also verified that the droplet size of the decontamination reagent in the stable emulsion was smaller than that in the unstable emulsion. This phenomena can be explained by the relationship between the interfacial tension and droplet size. Finally, the recovering test of the PFC from the spent PFC-based decontamination emulsion was conducted using distillation method. The distillation test was performed using vacuum distillation unit, and the distillation temperature was 80°C. From the distillation test, about 95 % of PFC was recovered by distillation. From this result, it is considered that PFC-based decontamination emulsion reduces the volume of the secondary waste.

Decontamination of spent nuclear fuel from decommissioned nuclear reactors is crucial to reduce the volume of intermediate-level waste. Fuel cladding hulls are one of the important parts due to high radioactivity. Their decontamination could possibly enable reclassification as low-level waste. Fuel cladding hulls used in research reactors and being developed for conventional light water reactors are Al-Mg and Fe-Cr-Al alloys, respectively. Therefore, the recovery of these component metals after decontamination is necessary to reduce the volume of highly radioactive waste. Electrochemical approach is often chosen due to its simplicity and effectiveness. Non-aqueous solvents, such as molten salts (MSs) and ionic liquids (ILs), are preferred to aqueous solvents due to the absence of hydrogen evolution. However, MSs and ILs are limited by high temperature and high synthesis cost, along with toxicity issues. Deep eutectic solvents (DESs) are synthesized from a hydrogen bond acceptor (HBA) and donor (HBD) and exhibit outstanding metal salt solubility, wide electrochemical window, good biocompatibility, and economic production process. These characteristics make DES an attractive candidate solvent for economic, green, and efficient electrodeposition compared with aqueous solvents such acids or nonaqueous solvents such as MSs or ILs. In this research, the feasibility of electrodeposition of Al-Mg and Fe-Cr-Al alloys in ChCl:EG, the most common DES synthesized from choline chloride (ChCl) and ethylene glycol (EG), will be tested. A standard three-electrode electrochemical cell with an Au plated working electrode and Al wires for counter and reference electrodes is utilized. Two electrolyte solutions (Al-Mg and Fe-Cr-Al) are prepared by dissolving 100 mM of each anhydrous metal chloride salts (AlCl3, MgCl2, CrCl3, and FeCl2) in ChCl:EG. Cyclic voltammogram (CV) is measured at 5, 10, 15, and 20 mV·s−1 to observe the redox reactions occurring in the solutions. Electrodeposition of each alloy is performed via chronoamperometry at observed reduction potentials from CV measurements. The deposited surfaces and cross-sections are examined by scanning electron microscopy and energy dispersive spectroscopy (SEM-EDS) to analyze the surface morphology, cross-section composition, and thickness. Authors anticipate that the presence of different metals will greatly affect the possibility of electrodeposition. It is expected that although all metals are distributed throughout the surface, the morphology, in terms of particle size and shape, would differ depending on metals. Different metals will be deposited by layers of an approximate thickness of a few μm each. This research will illustrate a potential for recovery and electrodeposition of other precious radioactive metals from DES.

The permanent shutdown of NPPs (nuclear power plants) has been growing steadily around the world. Also, permanent shutdown of old NPPs has been determined following to Kori-1 and Wolsong-1 in the Korea. Among issues of decommissioning of NPPs, especially, the management of radioactive waste is the most sensitive issue. According to IAEA, a large volume of radioactive concrete waste would be generated from decommissioning of nuclear facilities. Also, EC (European commission) expected that about 500 million tons of concrete will be produced in Europe by 2060 due to decommissioning of NPPs. It is known that the radioactive concrete consists of surface contaminated concrete and activated concrete. So, if contaminants from radioactive concretes can be removed using specific technologies, volume reduction of concrete can be achieved. Since there is no experience of decommissioning of NPPs in Korea, it is important to analyze previous cases. In this study, decontamination of radioactive concrete is analyzed through previous studies. Decontamination technologies of concrete are composed of mechanical methods, chemical methods, and thermal methods. Mechanical methods are physical technologies separating contaminants from concrete using scabbling, milling, and vacuum cleaning. In chemical methods, contaminants were removed from concrete using an oxidizing agent/reducing agent, acid/base. Thermal methods are removal technologies using lasers, microwaves, and pulsed power discharge. Some methods still have practical use cases, and further research is needed on the issue of generation of secondary waste. Review on the experience of decontamination of concrete show that waste of concrete generated during decommissioning of NPPs are expected to have effect of large volume reduction. However, many studies are needed because secondary waste and decontamination cost is sensitive issue of concrete generated during decommissioning of NPPs. In order to successful decommissioning of NPPs in Korea, various research of decontamination of concrete are need

Decontamination and Dismantlement (D&D) are of great interest to owner of decommissioning as a large number of old nuclear facilities around the world are either shutdown or soon to be decommissioned. D&D are key steps in the decommissioning of nuclear power plants (NPPs). These activities typically generate a significant volume of radioactively contaminated waste. However, as much as 90% or more of this waste is lightly contaminated metal and concrete that could potentially be cleared for recycle or beneficial reuse, rather than disposed of as radioactive waste. The objective of this study is to provide reference for the application of current technologies to cost-effectively reduce the volume of radioactive waste associated with decommissioning, through review of experiences with decontamination of NPPs materials for unrestricted release, recycle or reuse, Also, highlights the importance of ongoing efforts to harmonize regulations and standards for radioactive waste management globally to enable reuse and recycle of valuable materials generated during decommissioning. The presented results in the balance of this study are organized to align with the sequence of operations for executing reuse or recycle of material for a decommissioning project. Concrete from buildings has most commonly been used for backfill of voids onsite, while metal has most commonly been melted or cleared into the conventional scrap recycling industry. Copper and lead, commonly found in cables and shielding, have high residual value and are thus highly desirable for recycling. Steel and stainless steel, while not inherently valuable, are present in many large components, such that decontamination for recycling can be cost-effective compared to disposal as radioactive waste. The decontamination techniques range from simple, inexpensive methods to complex, aggressive methods, each with advantages in various scenarios and limitations in others. Treatment often involves the sequential application of two or more decontamination techniques (e.g., chemical decontamination followed by abrasive blasting). Strategies for the characterization of materials for recycling include analyzing material in place before dismantlement, analyzing removed samples before or after dismantlement, and evaluating bulk material removed after dismantlement. If clearance and recycling are permitted, metals can be released to the conventional scrap recycling market, and concrete rubble can be used as backfill material onsite. In general, successful reuse/recycle projects require consideration of reuse/recycling objectives and implementation of associated planning activities early in the decommissioning process. The practicality of reuse/recycle depends on a number of high level (country and region-specific) and component level (material and case specific) factors. Since this information is useful to those responsible for planning or implementing the decommissioning of nuclear facilities, it is expected that it will be of great help especially to those in charge of decommissioning plan and managers in charge of decommissioning projects.

Laser scabbling experiments were conducted with the aim of developing concrete decontamination technology. Laser scabbling system contains a 6 kW fiber laser (IPG YLS-6000, λ=1,070 nm) and optical head, which are connected with process fiber (core dia.: 600 μm, length: 20 m). Optical head consists of two lenses (f = 160 mm and 100 mm) to collimate and focus laser beam. The focused laser beam is passed through the small diameter of nozzle (throat dia.: 3 mm) to prevent the laser-produced debris into head. And then, the focused beam is directed toward concrete block as continuously diverging. The diverged laser beam was incident on the high-strength concrete with 300 mm (length) × 300 mm (height) × 80 mm (width) to induce explosive spalling on the concrete surface. The optical head was moved by X-Y-Z manipulate coupled with a computerized numerical control while scabbling. We have observed not only active spalling on the concrete surface but energetic scattering of laserproduced debris when scabbling on high-strength concretes. It indicates the need for a device capable of collecting the laser-produced debris. We newly designed and manufactured dust collector coupled with cylindrical tube to prevent scattering of laser-produced debris into ambient environment. The collecting system was evaluated by estimating the collecting efficiency for laser-produced debris while scabbling. The collecting efficiency was calculated on the basis of the information on the mass loss of concrete block after laser scabbling and the mass of collected debris in a container. The collecting efficiency was found to be over 85%.

Today, the domestic and international nuclear power industry is experiencing an acceleration in the scale of the nuclear facility decommissioning market. This phenomenon is also due to policy changes in some countries, but the main reason is the rapid increase in the proportion of old nuclear power plants in the world, mainly in countries that introduced nuclear power plants in the early stages. Decontamination is essential in the process of decommissioning nuclear facilities. Among various decontamination targets, radionuclides are adsorbed between pores in the soil, making physical decontamination quite difficult. Therefore, various chemical decontamination technologies are used for contaminated soil decontamination, and the current decontamination technologies have a problem of generating a large amount of secondary wastes. In this study, soil decontamination technology using supercritical carbon dioxide is proposed and aimed to make it into a process. This technology applies cleaning technology using supercritical fluids to decontamination of radioactive waste, it has important technical characteristics that do not fundamentally generate secondary wastes during radioactive waste treatment. Supercritical carbon dioxide is harmless and is a very useful fluid with advantages such as high dissolution, high diffusion coefficient, and low surface tension. However, since carbon dioxide, a non-polar material, shows limitations in removing polar and ionic metal wastes, a chelating ligand was introduced as an additive. In this study, a ligand material that can be dissolved in supercritical carbon dioxide and has high binding ability with polar metal ions was selected. In addition, in order to increase the decontamination efficiency, an experiment was conducted by adding an auxiliary ligand material and ultrasonic waves as additives. In this study, the possibility of liquefaction of chelating ligands and auxiliary ligands was tested for process continuity and efficiency, and the decontamination efficiency was compared by applying it to the actual soil classified according to the particle size. The decontamination efficiency was derived by measuring the concentration of target nuclides in the soil before and after decontamination through ICP-MS. As a result of the experiment, it was confirmed that the liquefaction of the additive had a positive effect on the decontamination efficiency, and a difference in the decontamination efficiency was confirmed according to the actual particle size of the soil. Through this study, it is expected that economic value can be created in addition to the social value of the technology by ensuring the continuity of the decontamination process using supercritical carbon dioxide.

The purpose of full system decontamination before decommissioning a nuclear power plant is to reduce radiation exposure of decommissioning workers and to reduce decommissioning waste. In general, full system decontamination removes the CRUD nuclides deposited on the inner surface of the reactor coolant system, chemical and volume control system, residual heat removal system, pressurizer, steam generator tube, etc. by chemical decontamination method. The full system decontamination process applied to Maine Yankee and Connecticut Yankee in the USA, Stade, Obrigheim, Unterweser, Nekawestheim Unit 1 in Germany, Mihama Unit 1 and 2 in Japan, Jose Cabrera Unit 1 in Spain, and Barseback Unit 1 and 2 in Sweden are HP/CORD UV, NP/CORD UV, and DfD. In this study, the quantity of 60Co radioactivity removal, metal removal, ion exchange resin and filter generation according to reactor power, surface area and volume of the full system decontamination flow path, and the decontamination process were compared and analyzed. In addition, the quantity of 60Co radioactivity removal by each nuclear power plant was compared and analyzed with the evaluation results of the 60CO radioactivity inventory of the Kori Unit 1 full system decontamination loops conducted by SAE-AN Enertech Corporation.

Activated corrosion products deposited on the reactor coolant system in a nuclear power plant should be removed to reduce the radiation exposure to workers. Chemical decontamination processes using organic acids have been widely applied to remove the activated corrosion products. However, they are highly corrosive to the base metal and generate a considerable amount of ion exchange resin waste, which is hard to be treated. In order to resolve this problem, KAERI has been developed a chemical decontamination process using chelate-free inorganic acid, HyBRID (Hydrazine Based Reductive metal Ion Decontamination) process. Especially, the Cyclic SP (Sulfuric acid/Permanganate)- HyBRID process was suggested as the decontamination process for applying to the remove the double oxide layer generated on the reactor coolant system in the pressurized water reactor (PWR). During the Cyclic SP-HyBRID process, the process is continuously applied without discharging or recharging of the decontamination process solution from the primary circuit. Thus, it is necessary to include the removal processes of the decontamination reagents middle of the Cyclic SP-HyBRID process, e.g., ‘Mn removal step’ for removing the permanganate ions and ‘hydrazine decomposition step’ for decomposition of the remaining hydrazine. During these removal processes, the metal ions can also be removed from the process solution. In this study, the behaviors of metals were investigated during the Cyclic SP-HyBRID process. The concentration changes of metal ions in the process solution were analyzed using atomic absorption (AA) spectroscopy. The metal precipitates generated during the process were characterized using X-ray diffraction (XRD) and Fourier Transform Infrared (FT-IR) spectroscopy. From the results of the analysis, it was observed that the metal ions dissolved in the process solution were converted into metal hydroxides and precipitated at the Mn removal process. It was confirmed by equilibrium calculation result that the OH− ions generated at the Mn removal can react with the metal ions and form the metal hydroxides. It is considered that this removal behaviors of the metals can contribute the decontamination performance.