A study on the denitrification of reverse osmosis(RO) concentrated wastewater from sewage reuse treatment plant in P city was conducted using waste desulfurization agent obtained from desulfurization process. Sulfur-based autotrophic denitrifying carrier comprises the predetermined amount of waste iron sulfide (FeS, Fe2S3), mine drainage sludge and elemental sulfur showing mesoporisity with 9.9 nm (99 Å) of average pore size. Sulfur denitrifying bacteria and sulfur reducing bacteria were implanted into the pores of sulfur denitrifying carrier. Nitrate was not affected by empty bed contact time (EBCT). It is probably due to larger reducing capacity of the carrier than the concentration of nitrate in RO concentrated wastewater. Total nitrogen (T-N) removal efficiency exhibited about 90% after 4 days. Sulfate ion was surprisingly decreased with sulfur autotropic process due to the reduction of sulfate ion to HS- and S2- by sulfur reducing bacteria. Sulfide and hydrogen sulfide ions were then taken by Fe(OH)3, main component of mine drainage sludge, releasing OH-. Alkalinity was therefore maintained between 7.5 and 8.5 in pH by the released OH-. Also, it had the effect of suppressing the production of H2S, which causes bad odor.

In this study, we employed a small-scale experiment to demonstrate the introduction of a thin copper heat dissipation plate into a bentonite buffer layer of an engineered barrier system. This experiment designed for spent nuclear fuel disposal can effectively reduce the maximum temperature of the bentonite buffer layer, and ultimately, make it possible to reduce the area of the disposal site. For the experiment, a small-scale engineered barrier system with a copper heat dissipation plate was designed and manufactured. the thickness of the cylindrical buffer was about 2 cm, which was about 1/20 of KAERI Repository System (KRS). At a power supply of 250 W, the maximum buffer temperature reduced to a mere 1.8°C when the thin copper plate was introduced. However, the maximum surface temperature reduced to a remarkable 9.1°C, when a U-collar copper plate was introduced, which had a good contact with the other barrier layers. Consequently, we conclude that the introduction of the thin copper plate into the engineered barrier system for spent nuclear fuel disposal can effectively reduce the maximum buffer temperature in high-level radioactive waste disposal repositories.

Currently, Korea is planning to use various equipments and technologies for cutting, decontamination, compression, solidification, and packaging at decommissioning site of Kori unit 1 and Wolsung unit 1. In particular, Korea is considering to apply new technologies like inorganic acid decontamination, spent resion treatment technology not only to localize various decommissioning technology, but to meet the limit of 14,500 drums of the decommissioning waste per unit. However, before the techniques applied to decommissioning, it is necessary to demonstrate the effectiveness and the safety of the techniques. Because unlike the industrial fields, the failure of the decommissioning technique in nuclear power plants can cause the spread or leakage of radioactive materials. In the「Regulation on Technical Standards for Nuclear Reactor Facilities, Etc.」is stated that the licensee shall apply proven technology to decommissioning and if the licensee apply new technology, he must provide resonable evidence and prove its safety. In accordance with this approach, Nuclear Safety and Security Commission (NSSC) Notice No. 2021-24 can be applied to the decommissioning technology as one of a technical standard related with the demonstration of it. And it states 9 kinds of elements related to the radioactive waste management facilites and components like the management of radioactive effluents, prevention of contamination and overflow by radioactive materials, etc. But, they are mixed with the radioactive material considerations and industrial considerations, and these considerations are usually for the facilities, not equipments or techniques. On the other hand, in the IAEA Safety Standards Series No. WS-G-2.1 Section 6.15 to 6.20, it recommended to evaluate 12 considerations for the decontamination technique and 7 considerations for the dismantling techniques. The Decommissioning Guide in Germany recommends to consider 3 conditions for radiation protection of decontamination techniques and 4 conditions for dismantling techniques. Therefore, it is necessary to compare the safety requirements or recommendations related to the demonstration of decommissioning technology with the other countries to check there is something to learn from it.

The nuclear power plant (NPP) decommissioning market is expected to expand not only domestically but also overseas. Proven technologies must be applied to decommission NPP. This is based on Article 41-2, Paragraph 2 of the domestic ‘Enforcement Decree Of The Nuclear Safety Act’. Proven technology refers to technology that has verified that it can be applied in the field through demonstration. In other words, in order to carry out NPP decommissioning, verification must be done. Demonstration refers to reducing technological uncertainty and directly verifying services implemented in the field. From a technology commercialization perspective, demonstration requires an approach based on technology readiness level (TRL) from a technology perspective and market readiness level (MRL) from a market perspective. The characteristics of demonstration also differ depending on the characteristics of each field. The demonstration in the field of nuclear energy is the demonstration of demand matching. This is to confirm the feasibility of the technology in the company’s required environment. In order to perform demonstration, a scenario must be derived by reflecting demonstration design considerations. After evaluating the derived scenario, an actual assessment is conducted using lab-based demonstration/virtual environment demonstration/real environment demonstration. What must be preceded by an actual assessment is confirming the consumer’s requirements. In this study, the necessary environment and requirements of consumer’s to perform NPP decommissioning were reviewed. The domestic decommissioning procedure requirements management system presents decommissioning procedures, potential worker accidents, and worker requirements. In the case of foreign countries, it was confirmed that complex wide need, cost benefit, risk reduction, waste generation, operation, reliability and maintenance (RAM) improvement and quantitative measures were evaluated for the technology to be demonstrated. Also the requirements for demonstrating decommissioning need to a detailed review of actual decommissioning cases. Therefore, a comparison must be made between the requirements based on actual NPP decommissioning cases and the requirements derived from this research process. Afterwards, the empirical research approach proposed by the Ministry of Trade, Industry and Energy was applied. The empirical research approach proposed by the Ministry of Trade, Industry and Energy is to secure a track record over a certain period of time and performance under conditions similar to the actual environment in the final research stage at the TRL level 6 to 8. Through this, it will be possible to confirm the suitability of overseas technology for domestic application.

For safe and successful decommissioning, it is one of the most important procedures that establishing the goal and complying with regulations of which final status of decommissioned site and building. The dose criteria for cyclotron facilities should be established and applied to reuse the site and building, since building and component of a cyclotron facility have been activated by incident secondary neutrons from radioactive isotope processes (e.g. 18O(p,n)18F, etc.). Furthermore, appropriate approaches should be applied to demonstrate compliance with the dose criteria for reliability of reuse. It is of noted that U.S. NRC (Nuclear Regulatory Commission) has confirmed that the residual radioactivity which distinguishable from background radiation results in a TEDE (Total Effective Dose Equivalent) does not exceed 25 mrem (0.25 mSv) per year as radiological criteria for unrestricted use of not only nuclear power plants but also cyclotron facilities referred to 10 CFR Part 20.1402. In addition, U.S. NRC noted the two approaches (i.e. dose assessment methods and, DCGL and final status surveys) which can be applied for demonstrating compliance with the dose criteria of 10 CFR Part 20 and recommended DCGL and FSS approach based on advantages and disadvantages of the two approaches. In order to using DCGL and FSS approach, U.S. NRC suggested screening approach; using DandD Version 2 which assesses TEDE under ICRP 28 and site-specific approach; using all models or computational codes which approved by NRC staff. There are several foreign cases that release of cyclotron facilities after decommissioning (i.e. U.S. and Japan). U.S., for examples, there are two DCGL approach cases and one dose modeling case based on 25 mrem per year same as reactor facilities. The dose modeling case, however, which may not be really used in Korea because of its low applicability. On the other hand, Japan case did not establish any radiological criteria for site and building reuse such as DCGL and just confirm “no more contamination” which is all residual radioactivity is lower than MDC based on real survey. Japan case also may not be used in Korea since criteria of “no more contamination” is not clear and hard to apply for all sites. Considering regulations and criteria for site release and reuse in Korea, this study aims to suggest radiological criteria and the demonstration approach of compliance for decommissioning of cyclotron facilities based on Nuclear Safety Acts and NSSC notices.

Despite its advantages such as safety, unnecessary pretreatment, and decontamination of waste with complex geometry, conventional ultrasonic decontamination technology has been only used to remove loose contaminants, oil and grease, not fixed contaminants due to the limitations in increasing the intensity in the high frequency range. Thus, ultrasound has been used as an auxiliary method to accelerate chemical decontamination of radioactive wastes or chemicals were added to the solution to increase the decontamination efficiency. The recently developed high-intensity focused ultrasound (HIFU) decontamination technology overcomes these limitations by combining multiple frequencies of ultrasonic waves in a specific arrangement, making it possible to remove most fixed contaminants, including radioactive micro particles less than 1 micrometer within half an hour. KEPCO NF and EnesG developed mobile HIFU decontamination equipment and successfully demonstrated the decontamination effect on various radionuclides found in nuclear power plants by treating radioactive metal waste to the level below free release criteria. The mobile HIFU decontamination equipment used in the demonstration can be operated anywhere where water is supplied, including controlled area in nuclear power plants, and is expected to be used widely for decontamination and free release of metal radioactive wastes.

For the disposal of radioactive waste from nuclear facilities, assessing their radioactivity inventories is essential. As a result, countries with nuclear facilities are implementing assessment schemes tailored to their respective policies and available resources for radioactive waste management. This paper specifically describes the assessment scheme for radioactivity inventory applied to metal waste generated during the dismantling of the Japan Power Demonstration Reactor (JPDR), a 1.25 MW BWR. The distinctive aspect of the Japanese approach lies in the fact that, for a pair of a key nuclide and a difficult-to-measure (DTM) nuclide that lack a significant correlation in their concentrations, the mean activity concentration method was used. In this method, an arithmetic average of all measurements of the DTM nuclide from representative drums, including MDAs (Minimum Detectable Activities), was assigned to the concentration of the DTM nuclide for all drums, regardless of the concentration of its paired key nuclide. Conversely, for a specific pair of a key nuclide and a DTM nuclide with a significant correlation, the scaling factor method was applied, as is common in many other countries. This Japanese case can serve as a valuable reference for Korea, which does not have the option of using the mean activity concentration method in its assessment scheme.

Properties of bentonite, mainly used as buffer and/or backfill materials, will evolve with time due to thermo-hydro-mechanical-chemical (THMC) processes, which could deteriorate the long-term integrity of the engineered barrier system. In particular, degradation of the backfill in the evolution processes makes it impossible to sufficiently perform the safety functions assigned to prevent groundwater infiltration and retard radionuclide transport. To phenomenologically understand the performance degradation to be caused by evolution, it is essential to conduct the demonstration test for backfill material under the deep geological disposal environment. Accordingly, in this paper, we suggest types of tests and items to be measured for identifying the performance evolution of backfill for the Deep Geological Repository (DGR) in Korea, based on the review results on the performance assessment methodology conducted for the operating license application in Finland. Some of insights derived from reviewing the Finnish case are as follows: 1) The THMC evolution characteristics of backfill material are mainly originated from hydro-mechanical and/or hydrochemical processes driven by the groundwater behavior. 2) These evolutions could occur immediately upon installation of backfill materials and vary depending on characteristics of backfill and groundwater. 3) Through the demonstration experiments with various scales, the hydro-mechanical evolution (e.g. advection and mechanical erosion) of the backfill due to changes in hydraulic behavior could be identified. 4) The hydro-chemical evolution (e.g. alteration and microbial activity) could be identified by analyzing the fully-saturated backfill after completing the experiment. Given the findings, it is judged that the following studies should be first conducted for the candidate backfill materials of the domestic DGR. a) Lab-scale experiment: Measurement for dry density and swelling pressure due to saturation of various backfill materials, time required to reach full saturation, and change in hydraulic conductivity with injection pressure. b) Pilot-scale experiment: Measurement for the mass loss due to erosion; Investigation on the fracture (piping channel) forming and resealing in the saturation process; Identification of the hydro-mechanical evolution with the test scale. c) Post-experiment dismantling analysis for saturated backfill: Measurement of dry density, and contents of organic and harmful substances; Investigation of water content distribution and homogenization of density differences; Identification of the hydro-chemical evolution with groundwater conditions. The results of this study could be directly used to establishing the experimental plan for verifying performance of backfill materials of DGR in Korea, provided that the domestic data such as facility design and site characteristics (including information on groundwater) are acquired.

The engineered barrier system (EBS), composed of spent nuclear fuel, canister, buffer and backfill material, and near-field rock, plays a crucial role in the deep geological repository for high-level radioactive waste. Understanding the interactions between components in a thermo-hydro-mechanical -chemical (THMC) environment is necessary for ensuring the long-term performance of a disposal facility. Alongside the research project at KAERI, a comprehensive experimental facility has been established to elucidate the comprehensive performance of EBS components. The EBS performance demonstration laboratory, which installed in a 1,000 m2, consists of nine experimental modules pertaining to rock mechanics, gas migration, THMC characteristics, buffer-rock interaction, buffer & backfill development, canister corrosion, canister welding, canister performance, and structure monitoring & diagnostics. This facility is still conducting research on the engineering properties and complex interactions of EBS components under coupled THMC condition. It is expected to serve as an important laboratory for the development of the key technologies for assessing the long-term stability of engineered barriers

In this research, KPS manufactured Full System Decontamination (FSD) equipment, which is consisted of Oxidizing Agent Manufacturing System (OAMS), Chemical Injection System (CIS), RadWaste Treatment System (RWTS), Chemical Waste Decomposition & Treatment System (CWDS) and conducted demonstration test to prepare Decontamination and Decommissioning (D&D) project of Kori nuclear power plant in Korea. Each equipment of FSD was modularized due to the limited size of equipment hatch of Kori nuclear power plant. To simulate the expected circumstances in nuclear power plant such as usage of heater or position of each equipment, additional equipment was used. The chemical concentration and flow rate of process water for FSD were used as mentioned in the previous study by KHNP CRI. FSD was conducted for three cycles and each cycle was consisted of oxidation, reduction, chemical decomposition and purification. Oxidation and reduction process were conducted at 90°C. Chemical decomposition and purification process were conducted at 40°C due to the damage of UV lamp and IX by the heat. Total volume of process water for FSD demonstration test was 2.5 m2. KPS conducted decontamination performance review by calculating thickness reduction and weight loss of installed specimen. Operational review was conducted as if FSD test was conducted in the field based on the result of demonstration test. One of the most prioritized features is the workers’ safety. Also, the appropriate position of equipment needs to be considered to meet the required specification of component.

In the pilot scale test, the two scale-up factors (Electric energy per order EEO, Electric energy per mass EEM) were conducted to design the Chemical Waste Decomposition & Treatment System (CWDS). The CWDS consist of two kind UV lamp reactors to improve the decomposition rate of oxalic acid, which are low pressure amalgam UV lamp and medium pressure UV lamp. The two reactors were connected in series, and the hydrogen peroxide is mixed through a line mixer at the front of the reactor and injected into the reactors. The CWDS was connected with the full system decontamination equipment to purify the residual oxalic acid after chemical decontamination process. The full system decontamination equipment were included Oxidizing Agent Manufacturing System (OAMS), Chemical Injection System (CIS), RadWaste Treatment System (RWTS) to operate the Oxidation/Reduction decontamination process and purify the process water. After decontamination process, the waste water will be cooled down into the 40°C and passed through the UV reactor at 110 gpm with hydrogen peroxide injection. The concentration of waste water is expected oxalic acid 1,700 ~ 2,000 ppm, Iron 5 ~ 20 ppm. As a result of the CBD test in the laboratory with simulated waste liquid, the amount of Low pressure amalgam lamp UV dose required to decompose 95% of oxalic acid in 2 m2 waste water was up to 1,800 mJ/cm2. The amount of medium pressure lamp UV dose was up to 450 mJ/cm2 at the same condition. We conducted demonstration test using 2 m2 waste water after the oxidation/reduction decontamination process, the decomposition rate 95% was obtained by low pressure amalgam UV lamp and medium pressure UV lamp reactor each.

본 연구는 아동․청소년의 태권도시범단 활동을 통해 교우관계 문제, 자아존중감에 미치 는 효과에 대해 알아보았다. 본 연구는 태권도시범단이 아동․청소년의 교유관계문제를 어 떻게 해결하는지, 그리고 자아존중감은 어떻게 달라지는지 연구해 보는 것이 필요하다고 보았다. 연구의 목적은 대인관계문제와 낮은 자아존중감으로 힘들어 하는 아동․청소년을 돕는 기초자료가 되고자 한다. 연구대상은 U시와 G시 2지역에 소재한 태권도장에서 초등 학생과 중학생 수련생을 대상으로 총 50부의 자료를 분석하였다. 실험 기간은 2022년 2월 부터 5월까지 진행하고 실험집단 25명과 비교집단 25명을 각각 사전-사후 비교 분석하였다. 교우관계 문제와 자아존중감에 미치는 효과를 검증하기 위해 자료 분석은 SPSS 2".0을 이 용하여 대응t-검증을 실시하였다. 연구결과, 실험집단은 태권도시범단 활동 후 교우관계 문 제와 자아존중감에 유의미한 효과가 나타났다. 이 결과, 아동․청소년의 태권도시범단 활동 이 교우관계 문제를 개선하고 자아존중감이 상승했다는 결과를 알수 있었다.

Deep geological disposal (DGD) of spent nuclear fuels (SNF) at 500 m–1 km depth has been the mainly researched as SNF disposal method, but with the recent drilling technology development, interest in deep borehole disposal (DBD) at 5 km depth is increasing. In DBD, up to 40SNF canisters are disposed of in a borehole with a diameter of about 50 cm, and SNF is disposed of at a depth of 2–5 km underground. DBD has the advantage of minimizing the disposal area and safely isolating highlevel waste from the ecosystem. Recently, due to an increasing necessity of developing an efficient alternative disposal system compared to DGD domestically, technological development for DBD has begun. In this paper, the research status of canister operation technology and plans for DBD demonstration tests, which subjects are being studied in the project of developing a safety-enhancing high-efficiency disposal system, are introduced. The canister operation technology for DBD can be divided into connection device development and operation technology. The developing connection device, emplacing and retrieving canisters in borehole, adopted the concept of a wedge thus making replacement equipment at the surface unnecessary. The new connection device has the advantage of being well applied with emplacement facilities only by simple mechanical operation. The technology of operating a connection device in DBD can be divided into drill pipe, coiled tubing, free-drop, and wireline. The drill pipe is a proven method in the oil industry, but requiring huge surface equipment. The coiled tubing method uses a flexible tube and shares disadvantages as the drill pipe. The free-drop is a convenient method of dropping canister into a borehole, but has a weakness in irretrievability in an accident. Finally, the wireline method can be operational on a small scale using hydraulic cranes, but the number of operated canisters at once is limited. The test facility through which the connection device is to be tested consists of dummy canister, borehole, lifting part, monitoring part, and connecting device. The canister weight is determined according to the emplacement operation unit. The lifting part will be composed following wireline consisting of a crane, a wire and a winding system. The monitoring part will consist of an external monitoring system for hoists and trolleys, and an internal monitoring system for the connection device’s location, pressure, and speed. In this project, a demonstration test will be conducted in a borehole with 1km depth, 10 cm diameter provided by KAERI to verify operation in the actual drilling environment after design improvement of the connecting device. If a problem is found through the demonstration test, the problem will be improved, and an improved connection device will be tested to an extended borehole with a 2 km depth, 40 cm diameter.