In 2022, new regulatory guidelines were announced in relation to the off-site dose calculation (ODC), and accordingly, measures to improve the off-site does calculation program (ODCP), kdose60, were reviewed. The main consideration is, first, that if multiple nuclear facilities are operated on the same site, the boundaries of the restricted areas shall be set as the overlapping outer boundaries of the restricted areas determined by calculation for each nuclear facility. Second, the external exposure caused by direct radiation from a number of nuclear facilities in the same site must be partially or fully applied depending on the facility and site characteristics. Third, the dose conversion coefficient should be evaluated by checking whether the effect of the daughter nuclides is properly reflected. Fourth, the soil contamination period is a factor to consider that radioactive substances deposited on the surface, such as particulate nuclides, affect residents over a long period of time. Fifth, due to the recent construction of Shin-Kori Units 5 and 6, there is a change in the site boundary of the Kori/Saeul site, so as the site boundary is expanded, it is required to add an exposure dose assessment point due to gas effluents and change the exposure dose assessment point according to crop intake. Therefore, through this study, the direction for improving the ODCP will be prepared by reviewing the recent revision of the regulatory guidelines.

The U.S. Nuclear Regulatory Commission (NRC) states that every environmental report prepared for the licensing stage of a Pressurized Water Reactor shall contain a statement concerning risk during the transportation of fuel and radioactive wastes to and from the reactor. Thus, the licensee should ensure that the radiological effect in accidents, as well as normal conditions in transport, do not exceed certain criteria or be small if cannot be numerically quantified. These are specified in 10 CFR Part 51 and applied in NUREG-1555 Supplement 1 Revision 1, which deals with Environmental Standard Review Plan. Corresponding regulations in Korea would be the Nuclear Safety and Security Commission Notice No. 2020-7. In Appendix 2 of the Notice, guides on the radiological environmental report for production and utilization facilities, spent nuclear fuel interim storage facilities, and radioactive waste disposal facilities. In this guide, unlike the regulations in the U.S., there are no obligations for radiological dose assessment for workers and public during the transportation. Therefore, overall regulations and their legal basis related to risk assessment during transportation conducted for the environmental report in the U.S. were analyzed in this study. On top of that, through the comparison with regulations in Korea, differences between the two systems were figured out. Finally, this study aims to find the points in terms of assessing transport risk to be revised in the current regulatory system in Korea.

In Korea, the NUREG-0017 methodology based on realistic model for reactor coolant concentrations are used to estimate the annual radioactive effluent releases for normal operation of nuclear power plant. The realistic model to estimate the radionuclide concentrations in reactor coolant is formulated as a standard, ANSI/ANS-18.1. This standard has provided a set of the reference radionuclide concentrations and adjustment factors for estimating the radioactivity in the principal fluid systems of target plant. Since ANSI/ANS-18.1 was first published in 1976, it was revised in 1984, 1999, 2016, and most recently in 2020. Therefore, this study analyzed revision history of assessment methodology of radioactive source term of light water reactors, which is ANSI/ANS-18.1. Assessment methodology of radioactive source term given ANSI/ANS-18.1 is by using radionuclide concentrations for reactor coolant and steam generator fluid of the reference plant and adjustment factors, which is modifying radioactive source term according to differences in design parameters between reference plant and target plant. There are three type of reference plant: PWR with u-tube steam generator, PWR with once-through steam generator, and BWR. This study analyzed for PWR with u-tube steam generator. Although the standard was revised, evaluation methodology and formula of adjustment factor have been retained, but some of items have been revised. First revision item is reduction of the number of radionuclides and decrease of radioactive concentration in reactor coolant. In the 1976 version of the standard, there were 71 target radionuclides, but the target nuclides have reduced to 57 in 1984 and 56 after 1999. In the case of radioactive concentration in reactor coolant, as the version of standard was updated, the radioactive concentration of 18 nuclides in 1984, 14 nuclides in 1999, and 25 radionuclides in 2016 was decreased. Most of the radionuclides with decrease radioactivity concentration were fission product, it is resulted from improvement of nuclear fuel performance. Second revision item is change of adjustment factors. After the revision in 2016, the adjustment factors for zinc addition plants using natural or depleted zinc are changed. This study analyzed revision history of evaluation methodology of radioactive source term of light water reactors. Furthermore, result of this study will be contributed to the improvement of understanding of assessment methodology and revision history for the radioactive source term.

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.

This study introduces the licensing process carried out by the regulatory body for construction and operation of the 2nd phase low level radioactive waste disposal facility in Gyeongju. Also, this study presents the experience and lessons learned from this regulatory review for preparing the license review for the next 3rd phase landfill disposal facility. Korea Radioactive Waste Agency (KORAD) submitted a license application to Nuclear Safety and Security commission (NSSC) on December 24, 2015 to obtain permit for construction and operation of the national engineered shallow land disposal facility at Wolsong, Gyeongju. NSSC and Korea Institute of Nuclear Safety (KINS) started the regulatory review process with an initial docket review of the KORAD application including Safety Analysis Report, Radiological Environmental Report and Safety Administration Rules. After reflecting the results of the docket review, the safety review of revised 10 application documents began on November 29, 2016. Total 856 queries and requests for additional information were elicited by thorough technical review until November 16, 2021. As the Gyeongju and Pohang earthquakes occurred in September 2016 and November 2017, respectively, the seismic design of the disposal facility for vault and underground gallery was enhanced from 0.2 g to 0.3 g and the site safety evaluation including groundwater characteristics was re-investigated due to earthquake-induced fault. Also, post-closure safety assessments related to normal/abnormal/human intrusion scenarios were re-performed for reflecting the results of site and design characteristics. Finally, NSSC decided to grant a license of the 2nd phase low level radioactive waste disposal facility under the Nuclear Safety Laws in July 2022. This study introduces important issues and major improvements in terms of safety during the review process and presents the lessons learned from the experience of regulatory review process.

The disposal criteria of the domestic LILW disposal facility specifies that fluidized substances such as the spent resin, the evaporator bottom should be solidified in a physically stable solid form, such as cementation and polymerization. And the solidified form applies requirements for compressive strength, immersion test, thermal circulation test, radiation irradiation test, leaching test, and free standing water measurement test. On the other hand, it is specified that immobilization iss applied to wastes with a total radioactivity concentration of more than 74,000 of radionuclides with a half-life exceeding 20 years among non-homogeneous wastes such as spent filters and DAW, but the test requirements are not applied. Nevertheless, it is necessary for waste generator to establish quality control standards for the manufacture of immobilized solid form through reviewing overseas cases and domestic regulations and technical standards. The test requirements for solidified solid form require measurement of structural stability (compressive strength, immersion, thermal cycling, irradiation test), leachability (leaching test), and free standing water measurement. A characteristic of the immobilized solid form is that it is not mixed with the waste and that the cement medium surrounds the waste. Therefore, the structural soundness is higher than that of the solidified solid mixed with waste. In addition, even when in contact with water, the cement medium blocks the contact between waste and water, thereby preventing the spread of radionuclides. Therefore, considering the characteristics of these immobilized solid form, compressive strength test and free standing water measurement are applied for structural soundness. For other tests, it is determined that application is unnecessary.

The treatment of radioactive waste by melting has been mainly discussed with low-level waste (LLW). Considering that a large amount of waste in RV or RVI is intermediate-level waste (ILW), however, it is necessary to examine the possibility of treatment by melting of ILW. Different from LLW, melting of ILW with a high content of long-lived nuclides would lead to no free releasee, but has advantages in volume reduction, homogenization, and delay of release. In this paper, the possibility of melting as an alternative technology for the treatment of ILW in the future is reviewed by analyzing the benefits generated by melting ILW in the following aspects: 1) Similar to melting techniques of LLW, them of ILW are mostly based on well-known techniques, but it is necessary to review the feasibility of performing operations such as removal of solidified melt using remote equipment in abnormal situations such as loss of electricity. 2) It is necessary to specify radiation limits for the melting operation unless the ILW melting operation technique can guarantee that the risk of abnormal occurrence is very low. The main quantified radiation parameter is the ingot dose rate, which of 10 mSv/h is considered more reasonable. 3) Although the treatment of ILW by melting leads to a reduction in volume, the main characteristics of the waste still remain, and no waste can be disposed of for free release. Thus, the main potential benefits are improved long-term safety and reduced waste volume. 4) Reducing the surface-to-volume ratio of the molten material could reduce the amount of corrosive material per unit time and, consequently, increase long-term safety. Its effect on long-term safety is difficult to quantify precisely as it depends on several factors, such as the geometry of the original component or whether radionuclides were distributed on the surface of the original component or the induced radioactivity. 5) The volume reduction of ILW is estimated to be reduced by about 1/4 compared to the generated amount when assuming a disposal volume reduction factor of 3 and considering the dose reduction due to radioactive decay after long-term storage, however, due to the lack of knowledge about non-hazardous facility alternatives, it is difficult to evaluate cost-benefit. This is heavily influenced by both the final volume reduction and the potential to reduce the complexity of the repository’s technical barriers.

Gyeongju radioactive waste repository has been operated to dispose low and intermediate level radioactive waste in Korea since 2016. Currently, only deep geological disposal facility (1st) is in operation, surface disposal facility (2nd) is scheduled to operate from 2024. As a result, the annual amount of radioactive waste that can be disposed of at deep geological disposal facilities and surface disposal facilities is almost determined. According to this result, it was possible to derive the total annual disposal amount to dispose of all radioactive waste at the Gyeongju repository after landfill disposal facility (3rd) construction. To evaluate it, a predictive model has been designed and radioactive waste generation, storage, and disposal data were input. The predictive model is based on system dynamics, which is useful to analyze the correlation between input variables. As a result of analysis, radioactive waste generation amount and maximum annual radioactive waste disposal were predicted to reach 741,615 drum and 17,030 drum per year respectively. From these results, it seems that the expansion of radioactive waste acceptance system or temporary storage is necessary.

As the management procedure for self-disposal wastes stored in the radiation controlled area within the Korea Atomic Energy Research Institute (KAERI) have been established, and the types and quantities of wastes are increasing. In order to carry out the disposal of wastes with various generation histories, we expanded the processing range from surface contaminated waste, which was already in progress, to volumetric contaminated waste. In this paper, a case study of self-disposal of volumetric contaminated radioactive waste for which final disposal has been completed is described. In order to carry out of self-disposal of volumetric contaminated waste, it is important to collect representative samples and prove their representativenss. Based on this, a treatment plan was established after reviewing the history of the waste to be disposed of, and the treatment work was carried out as follows; waste collection, classification by size and shape, radiation (activity) measurement, sampling of representative samples, radioactivity concentration analysis, notification to regulatory bodies and question-and-answer, final disposal. The waste is judged have no potential for contamination because the polywood used to set the flat floor between the steel frame and floorboards in the experimental greenhouse didn’t come into contact with radioactive material. However, due to the conservative approach to the presence or absence of contamination, the treatment plan was established assuming volumetric contaminated waste. The type of waste is single wood, and the major contaminating radionuclides are Sr-85 and Cs-137. After the waste was collected and sorted by size and shape, it was weighed and a representative sampling amount and sampling method were set up. A direct method of surface contamination was performed on the entire area, and the representative sample was divided into three groups of homogenized population samples using the trisection method, with three points (upper/middle/below) were sampled at a 200:1 ratio, and radioactivity concentration analysis was conducted. After confirming that the concentration was below the allowable concentration for selfdisposal, the final disposal was completed after receiving approval after reporting to the regulatory body. As a result of radioactivity concentration analysis of representative samples, the maximum radioactivity concentration for each nuclide was Sr-85: < MDC (0.00178), Cs-137 : 0.00183 Bq/g (Sr-85 : 1 Bq/g, Cs-137 : 0.1 Bq/g), which meets the nuclide allowable concentration standard. It was confirmed that the total maximum fraction of 0.02 Bq/g satisfies the criteria (In the case of mixed nuclides, the sum of the fraction is less than 1). This paper introduces the establishment and implementation of self-disposal procedures based on the experience of self-disposal of radioactive waste with volumetric contaminants, and is going to utilize it as a basic material for self-disposal of radioactive waste with volumetric contaminants that will continue in the future and contribute to the reduction of radioactive wastes.

Treatment methods such as interim storage and immobilization are being considered to dispose of intermediate level waste (ILW), but some wastes that have been treated in the past may require repackaging. Re-packaging means to cover repackaging of waste that has already been packaged in a waste container and re-packaging is required for the following reasons: loss of shielding or containment, damage to external handling features, package out-of-specification, insufficient records and external policy. The re-packaging includes various methods such as non-intrusive treatment, overpacking of waste package, external treatment of waste container, repair waste container, injection of stabiliser, disassemble waste package, high temperature process, and dissolve waste package. The purpose of this paper is to evaluate the re-packaging possibility for various wastes by identifying the main repackaging methods among the above various re-packaging methods. 1) Disposal outside of the waste container is a viable technique for most packages, as coating with a portable spray gun for low dose rate packages or remotely using a robotic arm for high dose rate packages. 2) Waste container repair is divided into welding repair and patching of waste container according to the degree of damage. Weld repair and patching are important techniques that can be used to add additional shielding, repair damaged areas, and improve the integrity of lifting gears that may not be compliant. 3) In general, disassembly of waste packages has been applied to loose drummed waste. Packages and waste forms are physically disassembled, reduced in size, and placed in different new packages. For practical solution, grouted waste is repackaged by cutting using proprietary equipment such as diamond saws, wire saws, core drilling and rupture techniques. 4) High-temperature process involves cutting the waste package and placing the pieces in a hot bath of inorganic liquid or molten metal, and the process is applicable to all waste types. However, treatment of all gases produced, compliance with waste types and acceptance criteria. Finally, dissolving waste packages, which is generally considered impractical due to the variety of chemicals and radionuclides present in ILW, is a process that is easier to perform on raw ILW than conditioned waste. An example of waste being re-packaged is when old drummed waste is recovered from an old storage facility and the waste needs to be repackaged into a form that meets modern standards for interim storage and disposal.

Molten Salt Reactor, which employs molten salt mixture as fuel, has many advantages in reactor size and operation compared to conventional nuclear reactor. In developing Molten Salt Reactor, Offgas system should be properly designed since the fission products in off-gas accelerates the corrosion in reactor structure materials and deteriorates the purity of liquid fuel. The design of off-gas system therefore requires the preliminary study of the behavior of evolved fission products in off-gas units and the development of off-gas model is crucial in developing such system. In this study, we corrected the off-gas illustrative model proposed by ORNL (Nuclear Engineering and Design, vol 385(15) 111529, 2021) by employing physically consistent concept of capture rate of fission product and holdup. For the application of the corrected off-gas model to Chloride-based 6 MW Molten Salt Reactor, major fission products were firstly determined from OpenMC based neutronics calculation and chain reaction related to the major fission products were defined. Based on these data, the holdup behavior of fission products in off-gas units (decay tank, caustic scrubber, Halide trap, H2O trap and charcoal bad) were investigated.

In nuclear power plant, there were many contaminated tanks dispose of radioactive fluid waste. These tanks are made of stainless-steel, and corrosion can occur when tanks are exposed to radioactive fluid waste containing moisture for a long time. Therefore, those sludge waste including radionuclide should be collected, solidified, and disposed of. If sludge can be melted, sludge can be easily solidified. However, melting points of sludge components (Fe2O3, NiO, Cr2O3) are very high as 1565, 1955, and 2435 , respectively. Therefore, melting sludge is difficult. If a solidification auxiliary material such as cement or asphalt is used to help solidify, solidification can easily occur, but cement and asphalt are vulnerable to heat. Vitrification using glass material can be solidification method, but the waste loading ratio of glass material is higher than 50%. High waste loading ratio is weakness in terms of volume reduction of waste. In this study, ferro frit powder (Na2O, K2O, CaO, Al2O3, B2O3, SiO2, ZnO) is used as solidification auxiliary material. When ferro frit powder mixed with sludge material are melted in sludge material, melted ferro frit powder can stick sludge material and can solidify sludge material without melting. Sludge can be solidified by using ferro frit powder with a smaller waste loading ratio than the vitrification method. However, since the waste loading ratio of the solidification auxiliary material is small, if ferro frit powder is not uniformly distributed between sludge powder, solidification may not be performed properly. Although the mixing ratio between sludge and ferro frit in solidified sludge is same, when the distribution of ferro frit powder in sludge is non-homogeneous, the difference in chemical and physical characteristics as compressive strength and leaching resistance can be observed in solidified sludge. As the ferro frit mixing ratio in the site where ferro frit exists was relatively high, the melting point of the mixed powder (sludge+ferro frit) decreased, and the mixed powder could not maintain its shape and melted. In the case of the area where ferro frit does not exist, since only the stainless-steel oxide sludge exists, sludge was not melted, and the shape was maintained. However, it was confirmed that the leaching resistance was lowered by visually observing the color change of the leachate within a short period of time (about 2 hours) when solidified sludge was immersed in the leachate.

3D modeling is a technology for representing real objects in a virtual 3D space or modeling and reproducing the physical environment. 2D drawings for viewing the existing building structure have limitations that make it difficult to understand the structure. By implementing this 3D modeling, specific visualization became possible. 3D technology is being applied in a wide range of preevaluation work required for nuclear decommissioning. In Slovakia, 3D modeling was applied to determine the optimal cutting strategy for the primary circuit before dismantling the VVER type Bohunice V1. In Japan, the Decommissioning Engineering Support System (DEXUS) program has been developed that incorporates VRDose, a decommissioning engineering support system based on 3D CAD models. Through this, the cutting length of the work object and the quantity of containers for packaging waste are calculated, exposure dose calculations in various dismantling scenarios, and cost estimation are performed. Korea also used 3D technology to evaluate the decommissioning waste volume for Kori Unit 1 and to evaluate the optimal scenario of the decommissioning process procedure for the research reactor Unit 1. 3D technology is currently being used in various pre-decommissioning evaluations for VVER, PWR, and research reactors. Overseas, a program that matches various decommissioning pre-evaluation tasks with cost estimation has also been developed. However, most 3D technologies are mainly used as a support system for dose evaluation and amount of decommissioning waste calculation. In this study, 3D modeling was performed on the PHWR structure, and physical and radiological information about the structure was provided. The information on the structure can present the unit cost for the work object by confirming the standard of the applied unit cost factor (UCF). The UCF presents the unit cost for repeated decommissioning operations. The decommissioning cost of the work object can be obtained by multiplying the UCF by the number of repetitions of the work. If the results of this study are combined with the process evaluation and waste quantity estimation performed in previous studies, it is judged that it will be helpful in developing an integrated NPP decommissioning program that requires preliminary evaluation of various tasks. In addition, it is judged that a clear cost estimation of the object to be evaluated will be possible by matching the 3D work object with the UCF.

The decommissioning process of Kori Nuclear Power Plant No.1, which was permanently suspended in 2017, various studies and attention on the decommissioning of nuclear power plants and waste management are being focused. In particular, decommissioning of high-risk facilities should take into account both safety and economic aspects. Small defects in the decommissioning process may lead to major disasters, and the resulting economic losses will cause enormous damage at the national level. In order to prevent such damage, various decommissioning process simulations within a virtual environment should be performed, and process errors and results should be collected and analyzed through simulation to derive the optimal decommissioning scenario as possible. The platform introduced in this paper builds a virtual environment based on drawing and modeling data of Kori Nuclear Power Plant No.1 and automatically creates an optimized cutting path for dismantling the facility and internal structure, and simulates a cutting process similar to reality using Robot Arm. In addition, it is possible to derive and analyze a cutting process scenario by processing process results such as time required for work and cutting distance collected through simulation.

3D imaging equipment is essential for automated robotic operations that cut radiologically contaminated structure and transfer segmented pieces in nuclear facility dismantling site. Automated dismantling operations using programmed robotic arms can make conventional nuclear facility dismantling operations much more efficient and safer, so dismantling technologies using robotic arms are being actively researched. Resolving the position uncertainty of the target structure is very important in automated robot work, and in general industries, the problem of position uncertainty is solved through the method of teaching the robot in the field, but at the nuclear facility dismantling site, the teaching method by workers is impossible due to activated target structures. Therefore, 3D imaging equipment is a key technology for a remote dismantling system using automated robotic arms at nuclear facility dismantling site where teaching methods are impossible. 3D imaging equipment available in radioactive and underwater environments is required to be developed for a remote dismantling system using robotic arms because most commercial 3D scanners are available in air and certain 3D scanners available in radioactive and underwater environments cannot satisfy requirements of the remote dismantling system such as measurement range and radiation resistance performance. The 3D imaging equipment in this study is developed based on an industrial 3D scanner available in air for efficient development. To protect the industrial 3D scanner against water and radiation, a housing is designed by using mirrors, windows and shieldings. To correct measurement errors caused by refraction, refraction model for the developed 3D imaging equipment is defined and parameter studies for uncertain variables are performed. The 3D imaging equipment based on the industrial 3D scanner has been successfully developed to satisfy the requirements of the remote dismantling system. The 3D imaging equipment can survive up to a cumulative dose of 1 kGy and can measure a 3D point cloud in the air and in water with an error of less than 1 mm. To achieve the requirements, a proper industrial 3D scanner is selected, a housing and shielding for water and radiation protection is designed, refraction correction are performed. The developed 3D imaging equipment is expected to contribute to the wider application of automated robotic operations in radioactive or underwater environments.

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.

The design life of the radioactive waste carrier, the CHEONG JEONG NURI, is in the year 2034, when the decommissioning of Kori Unit 1 is expected. As only IP-2 type transport containers (7.5- tons, 1.6 m (W) × 3.4 m (L) × 1.2 m (H)) can be loaded onto the CHEONG-JEONG-NURI, the radioactive decommissioning waste (RDW) transport containers neither of 35-tons maximum weight nor ISO type can be accommodated. Accordingly, either a new vessel (NV) to replace the CHEONGJEONG- NURI or a change in the loading dock design of the CHEONG-JEONG-NURI is required. In this study, the necessity of building a NV capable of accommodating the issued containers above is analyzed focusing, (1) the estimated building and operating costs of the NV, and (2) the economic feasibility of the NV ‘s RDW transportation scenarios. Among bulk carriers, the CHEONG-JEONG-NURI was designed as handy-size ship type. It is operated reflecting various design requirements to satisfy the domestic/international legal requirements. To estimate the cost of the NV, the same vessel type and design criteria of the CHEONG-JEONGNURI were considered. The shipping price information of the Korea Ocean Business Corporation, as of August 2022, the building cost of bulk carrier Handysize (building NV type) is about USD 30 million. Considering domestic/overseas variables, such as future labor costs, international inflation, interest rate hike, etc., the building costs are expected to continuously rise. Furthermore, vessel operation costs of crew labor, vessel, fuel, and insurance are incurred separately. Due to the increase in oil price, and wages of special positions, such as general seafarers and radiation safety managers, the NV’s operating cost is expected to be about KRW 3.8 billion every year, which is about KRW 1.1 billion higher than that of the CHEONG-JEONG-NURI. The expected total cost of building and operating the NV is about KRW 65 billion. Assuming the repayment period of the NV building cost is the same as that of the CHEONG-JEONG-NURI building cost reimbursement agency and analyzing the economic feasibility of the transport scenario of the NV built by adding up about KRW 3.8 billion of the operating cost, cost about KRW 880 million per voyage of the NV built is expected, which being KRW 620 million more than the current cost (KRW 260 million) per trip of the CHEONG-JEONG-NURI. Therefore, transporting the RDW to the disposal facility through sustainable use of the CHEONGJEONG- NURI (considering design life extension and design change) is evaluated as more appropriate than building NV.

The decommissioning of nuclear-related facilities at the end of their design life generates various types of radioactive waste. Therefore, the research on appropriate disposal methods according to the form of radioactive waste is needed. This study is about the solidification of uranium contaminated soils that may occur on the site of nuclear facilities. A large amount of radioactively contaminated soil waste was generated during the decommissioning of the uranium conversion plant in KAERI, and research on the proper disposal of this waste has been actively conducted. Numerous minerals in the soil can become glass-ceramic through the phase change of minerals during the sintering process. This method is effective in reducing the volume of waste and the glassceramic waste form has excellent mechanical strength and leaching resistance. In this study, the optimum temperature and time conditions were established for the production of glass-ceramic sintered body of soil. The compressive strength and leachability of the sintered body made by applying the optimal conditions to simulated waste was confirmed. The basic physicochemical properties of simulated soil waste were identified by measuring the pH, moisture content, density, and organic matter content. The elemental compositions in the soil was confirmed by XRF. Soils were classified by particle size, and each sample was compressed with a pressure of 150 MPa or more to prepare a green body. Based on the TG-DSC analysis, an appropriate heating temperature was set (>1,000°C), and the green body was maintained in a muffle furnace for 2~6 hours. The optimal sintering conditions were selected by measuring the compressive strength and volume reduction efficiency of the sintered body for each condition. The difference between the green body and sintered body was observed by XRD and SEM. In the experiments for evaluation of additives, the selected chemical substances were mixed with the soil sample in a rotator. Based on the results of TG-DSC, sintered body was made at 850°C, and the compressive strength and volume reduction were compared. Based on the results, the most effective additive was determined, and the appropriate ratio of the additive was found by adjusting the range of 1~5 wt%. This study was confirmed that the sintered soil waste showed sufficient stability to meet the disposal criteria and effective volume reduction for final disposal.

When decommissioning a nuclear power plant, the structure must be made to a disposable size. In general, the cutting process is essential when dismantling a nuclear power plant. Mainly, thermal cutting method is used to cutting metal structures. The aerosols generated during thermal cutting have a size distribution of less than 1 μm. The contaminated structures are able to generate radioactive aerosols in the decommissioning. Radioactive aerosols of 1 μm or less are deposited in the respiratory tract by workers’ breathing, causing the possibility of internal exposure. Therefore, workers must be protected from the risk of exposure to radioactive aerosols. Prior knowledge of aerosols generated during metal cutting is important to ensure worker safety. In this study, the physical and chemical properties of the aerosol were evaluated by measuring the number and mass concentrations of aerosols generated when cutting SUS304 and SA508 using the laser cutting method. High-resolution aerosol measuring equipment (HR-ELPI+, DEKATI) was used to measure the concentration of aerosols. The HR-ELPI+ is an impactor-type aerosol measuring equipment that measures the aerosol number concentration distribution in the aerodynamic diameter range of 6 nm to 10 um in real-time. And analyze the mass concentration of the aerosol according to the diameter range through the impactor. ICP-MS was used for elemental mass concentration analysis in the aerosol. Analytical elements were Fe, Cr, Ni and Mn. For the evaluation of physical and chemical properties, the MMAD of each element and CMAD were calculated in the aerosol distribution. Under the same cutting conditions, it was confirmed that the number concentration of aerosols generated from both materials had a uni-modal distribution with a peak around 0.1 um. CMAD was calculated to be 0.072 um for both SUS304 and SA508. The trend of the CMAD calculation results is the same even when the cutting conditions are changed. In the case of MMAD, it was confirmed that SUS304 had an MMAD of around 0.1 μm in size for only Fe, Cr and Mn. And SA508, Fe, Cr, Ni and Mn were all confirmed to have MMAD around 0.1 μm in size. The results of this study show that a lot of aerosols in the range of less than 1 μm, especially around 0.1 μm in size, are generated when metal is cut using laser cutting. Therefore, in order to protect the internal exposure of workers to laser metal cutting when decommissioning NPPs, it is necessary to protect from nano-sized aerosols beyond micron size.

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.