Republic of Korea (ROK) is operating the Integrated Environmental Radiation Monitoring Network (IERNet) in preparation for a radioactive emergency based on Article 105 of the Nuclear Safety Act (Monitoring of Nationwide Radioactive Environment). 215 radiation monitoring posts are monitoring a wide area, but their location is fixed, so they can’t cover areas where the post is not equipped around the Nuclear Power Plants (NPPs). For this, a mobile radiation monitoring system was developed using a drone or vehicle. However, there are disadvantages: it is performed only at a specific cycle, and an additional workforce is required. In this study, a radiation monitoring system using public transportation was developed to solve the above problems. Considering the range of dose rates from environmental radiation to high radiation doses in accidents, the detector was designed by combining NaI (TI) (in the low-dose area) and GM detector (in the high-dose area). Field test was conducted by installed on a city bus operated by Yeonggwang-gun to confirm the performance of the radiation monitoring system. As a result of the field test, it was confirmed that data is transmitted from the module to the server program in both directions. Based on this study, it will be possible to improve the radiation monitoring capability near nuclear facilities.

Radioactive waste generated in large quantities from NPP decommissioning has various physicochemical and radiological characteristics, and therefore treatment technologies suitable for those characteristics should be developed. Radioactively contaminated concrete waste is one of major decommissioning wastes. The disposal cost of radioactive concrete waste is considerable portion for the total budget of NPP decommissioning. In this study, we developed an integrated technology with thermomechanical and chemical methods for volume reduction of concrete waste and stabilization of secondary waste. The unit devices for the treatment process were also studied at bench-scale tests. The volume of radioactive concrete waste was effectively reduced by separating clean aggregate from the concrete. The separated aggregate satisfied the clearance criteria in the test using radionuclides. The treatment of secondary waste from the chemical separation step was optimally designed, and the stabilization method was found for the waste form to meet the final disposal criteria in the repository site. The final volume reduction rates of 56.4~75.4% were possible according to the application scenario of our processes under simulated conditions. The commercial-scale system designs for the thermomechanical and chemical processes were completed. Also, it was found that the disposal cost for the contaminated concrete waste at domestic NPP could be reduced by more than 20 billion won per each unit. Therefore, it is expected that the application of this technology will improve the utilization of the radioactive waste disposal space and significantly reduce the waste disposal cost.

Hydrogen isotopes (H, D, T) separation technologies have received great interest for treatments of tritiated liquid waste produced in Fukushima. In addition, the separated deuterium and tritium can be utilized in various industries such as semiconductors and nuclear fusion as expensive and rare resources. However, separating hydrogen isotopes in gas and liquid forms still requires energyintensive processes. To improve efficiency and performance of hydrogen isotope separation, we are developing water electrolysis, cryosorption, distillation, isotope exchange, and hydrophobic catalyst technologies. Furthermore, an analytical method is studied to evaluate the separation of hydrogen isotopes. This presentation introduces the current status of hydrogen isotope research in this research group.

A simulation model was developed for heavy water pre-enrichment and detritiation by the Combined Electrolysis and Catalytic Exchange (CECE) process. In the CECE process, heavy water enrichment and detritiation are based on the principle that concentrated in to water phase through an isotopic exchange reaction between water vapor and hydrogen gas produced by a water electrolysis. An operational analysis for a liquid phase catalytic exchange column was carried out by the model equations, composed of a material balance and combined equilibrium relationships for a scrubbing and catalyst bed, respectively. As a result of simulation, the optimum flow ratio of water to the rising hydrogen gas in contact with the down-coming water was predicted as the key variables in the separation performance analysis at a given feed flow rate and isotopic composition. From a graphical approach based on this model, the operating conditions can be determined within the range where the operating line does not meet the combined equilibrium line before reaching the specified target concentration.

During the decommissioning of nuclear facilities, 3D digital model that precisely describes the work environment can expedite the accomplishment of the work. Thus, the workers’ exposure to radiation is minimized and the safety risk to the workers is reduced, while precluding inadvertent effects on the environment. However, it is common that the 3D model does not exist for legacy nuclear facilities as most of the initial design drawings are 2D drawings and even some of the 2D drawings are missing. Even in the case that all of the 2D drawings are intact, these initial design drawings need to be updated using asbuilt data because facilities get modified through years of operation. In those cases, 3D scanning can be a good option to quickly and accurately generate a structure’s actual 3D geometric information. 3D scanning is a technique used to capture the shape of an object in the form of point cloud. Point cloud is a collection of large number of points on the external surfaces of objects measured by 3D scanners. The conversion of point cloud to 3D digital model is a labor-intensive process as a human worker needs to recognize objects in the point cloud and convert the objects into 3D model, even though some of the conversion process can be automated by using commercial software packages. With the aim of full automation of scan-to-3D-model process, deep learning techniques that take point cloud as input and generate corresponding 3D model have been studies recently. This paper introduces an efficient scan simulation method. The simulator generates synthetic point cloud data used to train deep learning models for classifying reactor parts in robotic nuclear decommissioning system. The simulator is built by implementing a ray-casting mechanism using a python library called ‘Pycaster’. In order to improve the speed of simulation, multiprocessing is applied. This paper describes the ray casting simulation mechanism and compares the in-house scan simulator with an open source sensor simulation package called Blensor.

Most of the wastes generated when dismantling nuclear power plant were contaminated with lowlevel radioactive materials, therefore, applying a plasma melting system is a good option to dispose of the complex wastes safely. Melting system with plasma technology was developed to dispose single metal or composite objects. Its purpose is to secure final emissions satisfying final treatment conditions by controlling oxidization/ reduction reaction condition in detail during the melting process. A hollow plasma torch applied at plasma melting system could be operated with various plasmaforming gasses such as N2, Air, Ar, O2, and etc. The melting furnace was designed based on a double sealing structure to prevent risk factors; such as leaks, etc. in the reaction condition. The effect of the external air inflow on the melting conditions was minimized by carefully designing the object input device, torch mounting part, final object discharge part, etc.

Recently, it is being carried out the project to evaluate the properties of materials harvested from nuclear reactor after the decommissioning of Kori Unit 1. However, it is not sufficient adequate machining equipment and remote machining technique to perform the projects for evaluation of materials harvested from nuclear reactor. Thus, it is required to develop the remote machining technique in hotcell to evaluate the mechanical properties of nuclear reactor materials. The machining technique should be performed inside a hotcell to evaluate mechanical properties of materials harvested from nuclear reactor and is essential to prevent radiation exposure of workers. Also, it is essential to design the apparatus and develop the machining process so that it can be operated with a manipulator and minimize contamination in hotcell. In this research, development of remote specimen machining technique in hotcell such as machining apparatus, technique and process for compact tension specimens of material harvested from nuclear reactor are described. Remote machining technique will be useful in specimen machining to evaluate changes in mechanical properties of materials harvested in high-radioactive reactor. Also, it is expected that various types of specimens can be machining by applying the developed machining technique in the future.

In 2017, Kori unit 1 nuclear power plant was permanently shut down at the end of its life. Currently, Historical Site Assessment (HSA) for MARSSIM characteristics evaluation is being conducted according to the NUREG-1575 procedure, this is conducted through comprehensive details such as radiological characteristics preliminary investigation and on-site interview. Thus, the decommissioning of nuclear power plant must consider safety and economic feasibility of structures and sites. For this purpose, the establishment of optimal work plan is required which simulations in various fields. This study aims to establish procedure that can form a basis for a rational decommissioning plan using the virtual nuclear power plant model. The mapping procedure for 3D platform implementation consisted of three steps. First, scan the inside and outside of the nuclear power plant for decommissioning structure analysis, 3D modeling is performed based on the data. After that, a platform is designed to directly measure the radiation dose rate and mapped the derived to the program. Finally, mapping the radiation dose rate for each point in 3D using the radiation dose rate calculation factor according to the time change the measured value created on the 3D mapping platform. When the mapping is completed, it is possible to manage the exposure dose of workers according to the ALARA principle through the charge of radiation dose rate over time because of visualization of the color difference to the radiation dose rate at each point. For addition, the exposure dose evaluation considering the movement route and economic feasibility can be considered using developed program. As the interest in safety accidents for workers increases, the importance of minimum radiation dose and optimal work plan for workers is becoming increasingly important. Through this mapping procedure, it will be possible to contribute to the establishment of reasonable process for dismantling nuclear power plant in the future.

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.

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.

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 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.

Lubricant oil waste contaminated with radioactive materials generated at nuclear facilities can be disposed of as industrial waste in accordance with self-disposal standards if only radioactive materials are removed. Lubricant oil used in nuclear facilities consists of oil of 75-85% and additives of 15-25%, and lubricant oil waste contains heavy metals, carbon, glycol, etc. In addition, lubricant oil waste from nuclear facilities contains metallic gamma-ray emission radionuclides including Co-60, Cs-137 and volatile beta-ray emission radionuclides such as C-14 and H-3, which are not present in lubricant oil waste from general industries and these radionuclides must be eliminated according to the Atomic Energy Act. In general industries, the wet treatment technologies such as acid-white soil treatment, ion purification, thin film distillation, high temperature pyrolysis, etc. are used as the refining technology of lubricant oil waste, but it is difficult to apply these technologies to nuclear industrial sites due to restrictions related with controlling the generation of secondary radioactive waste in sludge condition containing radionuclides of metal components, and limiting the concentration of volatile radioactive elements contained in refined oil to be below the legal threshold. In view of these characteristics, the refinement system capable of efficiently refining and treating lubricant oil waste contaminated with radioactive materials generated in nuclear facilities has been developed. The treatment process of this R&D system is as follows. First, the moisture in the radioactive lubricant oil waste pretreated through the preprocessing system is removed by the heated evaporating system, and the beta-emission radionuclides of H-3 and C-14 can be easily removed in this process. Second, the heated lubricant oil waste by the heated evaporating system is cooled through the heat exchanging system. Third, the particulate matters with gamma-ray emission radionuclides are removed through the electrostatic ionizing system. Forth, the lubricant oil waste is stored in the storage tank and the purified lubricant oil waste is discharged to the outside after sampling and checking from the upper, middle and lower positions of the lubricant oil waste stored in the storage tank. Using this R&D system, it is expected that the amount of radioactive waste can be reduced by efficiently refining and treating lubricant oil waste in the form of organic compounds contaminated with radioactive materials generated in nuclear facilities.

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.

High Integrity Container (HIC) made of polymer concrete was developed for the efficient treatment and safe disposal of radioactive spent resin and concentrate waste in consideration of the disposal requirements of domestic disposal sites. Permission for application of Polymer Concrete High Integrity Container (PC-HIC) to the domestic nuclear power plants (NPPs) has been completed or is under examination by the regulatory agency. In the case of 860 L PC-HIC for very-low-level-waste (VLLW) or low low-level-waste (LLW), the application of four representative NPPs has been approved, and the license for extended application to the rest NPPs is also almost completed. A licensing review is also underway to apply 510 L PC-HIC for intermediate and low-level-waste (ILLW) to representative nuclear power plants. In order to handle and efficiently store and manage PC-HICs and high-dose PCHIC packages, a gripper device that can be remotely operated and has excellent safety is essential, and the introduction of NPPs is urgent. The conventional gripper device developed by the PC-HIC manufacturer for lifting test to evaluate the structural integrity of PC-HIC requires a rather wide storage interval due to its design features, and does not have a passive safety design to handle heavy materials safely. In addition, work convenience needs to be reinforced for safety management of high radiation work. Therefore, we developed a conceptual design for a gripper device with a new concept to minimize the work space by reflecting on-site opinions on the handling and storage management conditions of radioactive waste in NPPs, and to enhances work safety with the passive safety design by the weight of the package and the function of checking the normal seating of the device and the normal operation of the grip by the detector/indicator, and to greatly improves the work efficiency and convenience with the wireless power supply function by rechargeable battery and the remote control function by camera and wireless monitoring & control system. Through design review by experts in mechanical system, power supply and instrumentation & control fields and further investigations on the usage conditions of PC-HICs, it is planned to facilitate preparations for the application of PC-HIC to domestic NPPs by securing the technical basis for a gripper device that can be used safely and efficiently and seeking ways to introduce it in a timely manner.

In this study, the process of compressing/packaging the spent filters of Kori Unit 1, which was conceptually presented in the previous study, is advanced so that disposal suitability for each step can be secure efficiently. In particular, the differences between the previous study and this study are that the disposable filters are screened using an In-Situ Object Counting System (ISOCS), and the method of collecting representative samples for development of scaling factor is specified. The process of compressing/packaging the spent filters consists of 7 stages as follows. 1) Collecting: The spent filters temporarily stored in the filter room are collected by dose and type remotely using a robot system to minimize the radiation exposure of workers according to a pre-established packaging plan. 2) Screening: The gamma activity concentration of the spent filters received by the robot system is measured by ISOCS. The spent filters below the low-level waste concentration limit and the surface dose are transferred into the compression system, while the others are returned in the filter room again. 3) Sampling: The external perforator drilling/cutting the filter was developed for sampling required for the new scaling factors. Since the sampling is collected remotely, the risk of exposure to workers can be reduced. The newly developed scaling factor will be used to verify the disposal suitability of the packages. 4) Compression: According to the pre-established plan, the spent filter collected by dose and type, is supplied to the compression system considering the dose and radionuclide inventory. Whether to additionally store the compressed filter in the drum is determined by checking the accumulated dose. 5) Immobilization: Immobilization with a safety material is necessary when inhomogeneous wastes, like spent filters, have the total radionuclide concentration with a half-life of more than 20 years is 74,000 Bq/g or more and for filling rate or non-dispersible treatment of particulates. 6) Packaging and Analysis: Waste information is labelled onto the package after the measurements of surface dose rate and surface contamination. Finally, using the drum assay system, the gamma radionuclide concentration is measured to identify at least 95% of the total radioactivity concentration of the package. 7) Temporary Storage and Delivery: The packages are moved to temporary storage in the plant prior to disposal. After establishing the plan for delivery and applying for a takeover request to KORAD, if the acceptance inspection is passed, the packages are transported to the disposal facility.

With the aging of nuclear power plants (NPPs) in 37 countries around the world, 207 out of 437 NPPs have been permanently shutdown as of August 2022 according to the IAEA. In Korea, the decommissioning of NPPs is emerging as a challenge due to the permanent shutdown of Kori Unit 1 and Wolsong Unit 1. However, there are no cases of decommissioning activities for Heavy Water Reactor (HWR) such as Wolsong Unit 1 although most of the decommissioning technologies for Light Water Reactor (LWR) such as Kori Unit 1 have been developed and there are cases of overseas decommissioning activities. This study shows the development of a decommissioning waste amount/cost/process linkage program for decommissioning Pressurized Heavy Water Reactor (PHWR), i.e. CANDU NPPs. The proposed program is an integrated management program that can derive optimal processes from an economic and safety perspective when decommissioning PHWR based on 3D modeling of the structures and digital mock-up system that links the characteristic data of PHWR, equipment and construction methods. This program can be used to simulate the nuclear decommissioning activities in a virtual space in three dimensions, and to evaluate the decommissioning operation characteristics, waste amount, cost, and exposure dose to worker. In order to verify the results, our methods for calculating optimal decommissioning quantity, which are closely related to radiological impact on workers and cost reduction during decommissioning, were compared with the methods of the foreign specialized institution (NAGRA). The optimal decommissioning quantity can be calculated by classifying the radioactivity level through MCNP modeling of waste, investigating domestic disposal containers, and selecting cutting sizes, so that costs can be reduced according to the final disposal waste reduction. As the target waste to be decommissioning for comparative study with NAGRA, the calandria in PHWR was modeled using MCNP. For packaging waste container, NAGRA selected three (P2A, P3, MOSAIK), and we selected two (P2A, P3) and compared them. It is intended to develop an integrated management program to derive the optimal process for decommissioning PHWR by linking the optimal decommissioning quantity calculation methodology with the detailed studies on exposure dose to worker, decommissioning order, difficulty of work, and cost evaluation. As a result, it is considered that it can be used not only for PHWR but also for other types of NPPs decommissioning in the future to derive optimal results such as worker safety and cost reduction.

For the disposal of radioactive waste generated from nuclear power plants, characterization of radioactive waste is essential. For characterization, samples of radioactive waste are directly collected or an indirect method is used through X-ray, etc. Through indirect analysis, which is a non-destructive method, the density, filling height, homogeneity and inter structure of the waste container can be analyzed. Currently, foreign institutions are in the process of developing a technology to perform characterization of radioactive waste through indirect analysis. In particular, research on improving internal image accuracy through image analysis techniques, improving measurement methods and enhancing portability for field application is ongoing. Through the review of such technology development trends, it will be utilized in the development of domestic radioactive waste disposal technolgy.