The dimensioning machine installed in the hot cell has been used for 20 years. It has been used for a long time so it was often malfunction due to aging and radiation. In addition, some parts of apparatus were discontinued and there were a lot of problems in maintenance and repair. In the old measuring system, the operator’s subjectivity was much involved. The process of control the focal length (distance between lens and specimen) by operator’s sense is a good example. The existing dimensioning machine was the Kh-7700 of Hirox Co., Ltd. As the equipment had been used for a longtime, additional utilities such as jigs, lighting module and servo motors have been customized and used. The same company’s apparatus was selected for the reasons that it did not need to manufacture a new utility so it could save the cost of radioactive waste disposal for existing utilities and its radiation resistance which has been used for 20 years in radiation environment. Lighting, standing, stage, controllers, cables and so on had been customized to provide remote control in hot cell. The installation was completed in March of this year in hot cell and has been successfully used until now. Through the improvement of dimensioning machine, an auto-focusing and multi-focusing were available. Therefore, they made it possible to produce high quality data and improve the accuracy of data. And this research could be a good example of how hot cell devices can be built and customized to achieve remote control.

This study presents distribution of naturally occurring radioactive materials in groundwater in Jeju island. Radon (222Rn) and potassium (40K) concentrations were performed by using RAD H2O of RAD7 and 940 Professional IC Vario, respectively. In addition, the activities of uranium and thorium nuclides were analyzed by ICP-MS. All of the groundwater samples were collected from 29 sites from August to October 2022. The radon concentrations in groundwater were in the range of 0 to 60 Bq L-1, and there was no groundwater exceeding the range of 148 Bq L-1 proposed by the US EPA. The distribution of uranium in groundwater varied from 0 to 0.6 μg L-1 and did not exceed 30 μg L-1, thresholds indicated by the US EPA.

Since radon was detected in mattresses of famous bed furniture brands in 2018, the nuclear safety and security commission (NSSC) announced the radiation safety management act in April 2021 to protect the public health and environment. This act stipulates the safety management of radiation that can be encountered in the natural environment such as the notification of radioactivity concentration of source materials, process by-products, the installation and operation of radioactive monitors. In this study, a model was established to predict radioactive exposure dose from radioactive materials such as radon and uranium detected in consumer products such as bed mattresses, pillows, shower, bracelets and masks in order to identify major radioactive substances that largely affect the exposure dose. A period of seven years from 2014 to 2020 was investigated for the source materials and exposure doses of consumer products containing naturally occurring radioactive materials (NORMs). We analyzed these using machine learning models such as classification and regression tree (CART), Random Forest and TreeNet. Index development and verification were performed to evaluate the predictive performance of the models. Overall, predictive performance was highest when Random Forest or TreeNet was used for each consumer product. Thoron had a great influence on the internal exposure dose of bedding, clothing and mats. Uranium had a great influence on the internal exposure dose of other consumer products except whetstones. When the number of data is very small or the missing value rate is high, it is difficult to expect accurate predictive performance even with machine learning techniques. If we significantly reduce the missing value rate of data or use the limit of detection value instead of missing values, we can build a model with more accurate predictive performance.

The segmentation of activated components including reactor vessel and reactor vessel internals requires many information. The primary information is material composition, trace materials in the composition, neutron flux during operation, etc. According to the EPRI report the primary basis of activity in a decommissioning source term is the activated metals from the reactor vessel and vessel internal components. The report indicates that over 95% of the radioactivity from decommissioning, except from spent nuclear fuel, consists of activated metals. These are from the reactor vessel, reactor internal structures and expendable components which are constructed primarily of various grades of stainless steel. Stainless steel contains appreciable levels of impurity cobalt. The common primary radionuclides of concern for the disposal environment from activated metals identified in US and international studies include C-14, Cl-36, Ni-59, Co-60, Ni-63, etc. The most common types of stainless steels used in reactor vessel construction and internal components include the Type 304(L), Type 316(L) and various grades of Inconel. The components of stainless steel are mainly Ni, Cr, Mo, Nb, etc., and when these elements are activated, they produce nuclides such as Nb-94, Tc-99, Sr-90, etc. In this study, the current status of activation analysis is reviewed to understand the effects of many variables. Also, the effect of trace materials is reviewed, including transformation of radioactive nuclides.

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.

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.

Cellulose-based wastes can be degraded into short-chain organic acids at the cementitious radioactive waste repository. Isosaccharinic acid (ISA), one of the main degradation products, can form the chelate complex with metals and radionuclides, and these complexes have a potential that can accelerate to move the radionuclides to far-field from the repository. This study characterized the amount of generated ISA from typical cellulosic materials in the repository. Two different degradation experiments were conducted under alkaline conditions (saturated with Ca(OH)2 at pH 12.4): i) cellulosic material mixture under an opened condition (partially aerobic), and ii) cellulosic material under an anaerobic condition in a nitrogen-purged glove box. In the first case, three different types of cellulosic materials–paper, cotton, and wood– were mixed at the same ratio, and the experiments were carried out at three different temperatures (20°C, 40°C, and 60°C). It revealed that both the cellulose degradation rate and generated ISA concentration were high at high reaction temperatures, and various soluble degradation products such as formic acid and lactic acid were generated. The cellulose degradation in this work seems to still stay at a peeling-off process. In the second study, each type of cellulosic material was applied in its own batch experiments, and the amount of generated ISA was in the order of paper > wood > cotton. The above two experiments are supposed to be a long-term study until the generated ISA reaches an equilibrium state.

The number of nuclear power plants that are permanently shut down or decommissioned is increasing worldwide, and accordingly, research is being conducted on an appropriate method for disposing of radioactive waste generated during the decommissioning of nuclear power plants. In the case of waste liquid generated during the decommissioning of nuclear power plants, it is important not only to efficiently reduce waste but also to secure the suitability of disposal. One of the solidification treatment methods for radioactive waste is cement solidification, but since cement solidification has poor solidification properties and generates a large amount of waste, improvement activities have been pursued. This study aims to develop high-performance cement-based materials and solidification treatment technology for solidification of liquid radioactive waste generated during nuclear decommissioning in order to improve the problems of cement solidification treatment method. For the development of polymer cement, epoxy resin and polyamine/amide mixed type and general Portland cement were mixed in various ratios. The most appropriate mixing ratio was 4.5:2, which showed the highest compressive strength. A simulated waste liquid was prepared by referring to the preliminary decommissioning plan of Shin-Kori Units 5 and 6, and it was dried and made into granules. Polymer cement was injected into a drum filled with granules by vacuum pressure to prepare a waste form matrix. In the solidification process, granules made by drying the waste liquid were used, and the solidification agent was filled in between the granules, so the total volume of solid radwaste was reduced compared to the conventional cement solidification treatment method. As a result, the amount of waste decreased to about 1/3, and the volume reduction rate increased by about 2.2 times. The compressive strength of 3,243 psi was confirmed in the disposability performance test for the manufactured solid samples. The compressive strength after the thermal cycling test, irradiation test, microorganism test, and immersion test was 2,257 psi, 2,306 psi, 4,530 psi, and 2,263 psi, respectively, exceeding the acceptance criteria of 500 psi. The leaching index was 7~13, and no free standing water was generated.

Backfill is one of the main components of engineered barrier in a high-level waste repository. The material selection of the backfill determines the barrier performance of the backfill. Overseas, its related research has been carried out mainly in Sweden, Finland, Canada, and Japan. However, Korea has recently started backfill research, and it is urgent to select a potential material for establishing the concept of backfill material and conducting backfill research. This study reviews NEA report, potential materials for overseas backfill research, advantages and disadvantages of single and mixed backfill materials, cases of license applications in Finland and Sweden for the selection of potential materials for backfill in Korea’s high-level waste repository. The review results indicated that it is reasonable to carry out backfill research according to the following plan: Both single and mixed materials are considered as potential materials for backfill research; experiments and performance studies are conducted with these materials; and, based on the results, a potential material or candidate material for the backfill suitable for the HLW repository in Korea is determined. For this plan, the single material is tentatively selected, as in Sweden, as bentonite with a montmorillonite content of about 40-50%. Then, if the selection criteria for montmorillonite content are determined through experiments and performance studies, we determine the final potential backfill material. As for the mixed backfill material, the bentonite/crushed rock mixture seems to be more advantageous than the bentonite/sand mixture considering the disposing problem of crushed rock generated from tunnel excavation and economic feasibility through its recycling. It is thought that the bentonite used in the bentonite/crushed rock mixture should have a higher montmorillonite content than bentonite used as a single backfill material since the crushed rock acts as an inert material in the mixture. The results of this study can be used as basic data for selecting the backfill material to be applied to the high-level waste repository in Korea, and can be used as a guideline for selecting the potential material required for backfill experiments and performance studies to be carried out in the future.

Spent nuclear fuels in Korea are temporarily stored at the nuclear power plant site and it is expected that will become saturated from 2031. Deep geological disposal in engineered barrier system (EBS) is one of the most important options for disposing spent nuclear fuel. The disposal canister is the first barrier that prevents leakage of nuclides in the spent nuclear fuel to the environment. Therefore, the corrosion behavior of the canister materials are significant factors in determining the overall disposal period. Oxygen-free copper is the most widely used material for disposal canisters, and manufacturing methods include forging, cold spray, and electro-deposition. In this study, corrosion behavior of materials that have the potential to replace oxygen-free copper manufactured using various 3D printing method were analyzed. As a result of electrochemical analysis of various materials such as copper manufactured by the Atmospheric Plasma Spray (APS) process and Inconel 718 manufactured by the Direct Energy Deposition (DED) process, the possibility of replacing oxygen-free copper was confirmed.