In recent years, people are increasingly interested in CO2 hydrogenation to produce value-added chemicals and fuels ( CH4, CH3OH, etc.). In the quest for an efficient treatment in CO2 methanation and methanolization, several technologies have been practiced, and DBD plasma technology gain attention due to its easily handling, mild operating conditions, strong activation ability, and high product selectivity. In addition, its reaction mechanism and the effect of packing materials and reaction parameters are still controversial. To address these problems efficiently, a summary of the reaction mechanism is presented. A discussion on plasma-catalyzed CO2 hydrogenation including packing materials, reaction parameters, and optimizing methods is addressed. In this review, the overall status and recent findings in DBD plasma-catalyzed CO2 hydrogenation are presented, and the possible directions of future development are discussed.

본 연구는 뷰티 분야에서의 디지털 기술 진보에 따른 변화의 해결방안으로 뷰티 교수자들의 신 기술 수용의도를 분석하여 뷰티산업을 이끌어 갈 예비 뷰티전문가를 교육하는데 있어서 전문적인 디지털 관련 기술 교과목을 적용하여 디지털 뷰티 전문가 배출의 가능성을 알아보고자 하였다. 교수자의 혁신의지, 주관적 규범, 자기효능감 등의 독립변수가 유용성 및 용이성 등 매개변수가 어떤 경로로 기술수용의도에 영향을 미치는 지를 분석하여 뷰티분야에서 신기술 관련 교육의 활성화 방안을 제시하는데 목적을 두고 있 다. 뷰티전공 교수자의 디지털 신기술 수용의도를 분석한 결과 신기술에 대한 혁신 의지와 주관적 규범, 자 기효능감은 매개변수인 용이성과 유용성에 정의 영향을 미치는 것으로 나타났다. 이러한 결과는 뷰티 분야 에서 신기술 적용이 필요하며, 이를 위한 교과목 개선이 필요하다고는 강하게 인식하고 있으나, 교수자 스 스로의 신기술 응용에 대한 두려움을 크게 느끼게 되므로 현실적으로 신규 교과목 적용이 어려워지고 있는 것으로 해석할 수 있다. 따라서 뷰티분야 교수자들이 쉽게 이해할 수 있는 디지털 기술 교육 관련 교재개발 이 우선시 되어야 하며, 관련 교육 프로그램 개발 및 활용이 확대되어야 할 것으로 사료된다. 또한 디지털 신기술 및 IT관련 학과와의 긴밀한 협업을 통해 뷰티전공자들을 대상으로 한 교육에 대한 이해와 교육 프 로그램 개발을 위한 상호연구가 지속적으로 이루어진다면 뷰티분야 맞춤형 디지털 신기술 교육프로그램 개발 많은 도움이 될 것이라 확신한다.

목적 : 본 연구는 시선추적 분야에 사용되는 적외선 동공 검출 기법을 기반으로 동공 크기 변화와 동공 간 거리 변화 분석을 통해 양안시기능을 타각적으로 측정하는 방법에 관해서 연구하고자 한다. 방법 : 남녀 23명(평균연령 27.27±8.84세)을 대상으로 원거리(3 m) 및 근거리(0.4 m)에서 적외선 필터를 사 용하여 만든 차폐렌즈(이하 차폐렌즈)로 커버테스트를 실시하여 편위량을 측정하였다. 또한, 근거리에서 눈앞에 각각 ±1.00 D의 렌즈를 부가한 상태에서 커버테스트를 실시하여 AC/A 비 측정을 하였다. 적외선 카메라를 사용 하여 위의 과정들을 모두 촬영하였고, OpenCV(Open Source Computer Vision)를 사용하여 영상 프레임별 동 공 간 거리와 동공 면적을 계산하였다. 결과 : 원거리 및 근거리 커버테스트 측정값의 T-검정 결과 원거리 및 근거리 커버테스트 모두 통계적으로 유 의한 차이(p<0.050)가 없는 것으로 나타났고, Bland-Altman 일치도 평가에서도 대부분의 값이 두 측정법 사이 95% 신뢰구간 내에 존재하였다. 경사 AC/A 비는 -1.00 D를 가입했을 때 p=0.015, +1.00 D를 가입했을 때 p=0.823으로 –1.00 D 렌즈를 가입했을 때 유의미한 차이가 발생하였고, 계산 AC/A 비는 유의한 차이가 나타나 지 않았다. AC/A 비의 일치도 평가 역시 두 검사법 간 평균 차이가 0.9 △로 나타났고, 산점도 역시 평균 차이 선 에 균일하게 배열하여 비교적 높은 일치도를 보였다. 결론 : 본 연구에서는 시선추적 기술에 활용되는 적외선 동공 검출 기법을 기반으로 양안시기능의 타각적 검사 법을 개발하였다. 동공 크기 변화와 동공 간 거리 변화 양상 분석을 통한 타각적 양안시기능 검사 방법이 실제 임 상에 적용된다면, 자각적 검사 방법의 한계를 극복할 수 있을 것으로 생각된다.

최근 미디어아트와 공연예술 분야에서 예술과 기술의 융합을 활용한 다원예술의 공연들이 늘어나고 있다. 공연 자와 관객들 사이의 실시간 커뮤니케이션이 이루어지는 공연예술은 관람객들이 전시의 형태로 이를 감상하는 미 디어아트와는 달리, 작가의 의도를 공연자의 퍼포먼스와 배경, 무대장치들을 통해 간접적으로 전달하는 형태를 취한다. 본 연구는 비언어적 커뮤니케이션의 형태로 관객들의 실시간 소통이 중요한 예술공연에서, 게임엔진과 실시간 인터렉션 기술들을 접목한 공연사례들을 분석해보자 한다. 게임과 영상 분야에서 주로 사용되었던 언리 얼 엔진과 비주얼 인터렉션 기술들은 실시간 리얼타임 비주얼 출력이라는 강력한 이점으로 인해 다양한 분야에 서 폭넓게 활용되고 있고, 예술공연 분야로도 점차 확대되고 있다. 따라서 미디어아트 공연분야에서 실시간 인터 렉션 기술이 접목된 다원예술 공연들을 살펴보고 어떠한 시도와 움직임들이 있는지 살펴보고 분석해보고자 한 다. 이를 통해 미디어아트 분야 뿐만 아니라, 공연예술 및 다원예술 분야에서도 미디어아트와 실시간 인터렉션 공연을 접목시킨 다양한 형태의 예술공연들이 늘어나고 관련 제작 및 방법론 연구에 도움이 되고자 한다.

The demand for energy is steadily rising because of rapid population growth and improvements in living standards. Consequently, extensive research is being conducted worldwide to enhance the energy supply. Transpiration power generation technology utilizes the vast availability of water, which encompasses more than 70% of the Earth's surface, offering the unique advantage of minimal temporal and spatial constraints over other forms of power generation. Various principles are involved in water-based energy harvesting. In this study, we focused on explaining the generation of energy through the streaming potential within the generator component. The generator was fabricated using sugar cubes, PDMS, carbon black, CTAB, and DI water. In addition, a straightforward and rapid manufacturing method for the generator was proposed. The PDMS generator developed in this study exhibits high performance with a voltage of 29.6 mV and a current of 8.29 μA and can generate power for over 40h. This study contributes to the future development of generators that can achieve high performance and long-term power generation.

After the Fukushima nuclear power plant accident in 2011, interest in technology for evaluating residents’ exposure to effluents generated from nuclear power plants at the time of the accident has increased. KHNP has developed the S-REDAP program and is using it to evaluate radiation dose and recommend resident protection measures in the event of a nuclear power plant emergency. Its main functions are source term evaluation, atmospheric diffusion evaluation, radiation dose evaluation, etc. Based on these evaluations, resident protection measures are evaluated. In Japan, evaluation is conducted through a program called SPEEDI-MP (System for Prediction of Environmental Emergency Dose Information Multi-model Package) created by JAEA (Japan Atomic Energy Agency). Similar to S-REDAP, the program also evaluates effluents emitted from nuclear facilities through source term evaluation and atmospheric diffusion factor evaluation. In JAEA, through a program using SPEEDI-MP, the source term evaluation was performed in collaboration with NSC (Nuclear Safety Commission) in the event of the Fukushima nuclear plant accident, and dose evaluation in Japan was performed 2 months as an atmospheric diffusion factor using meteorological data for 2 days. Through comparative analysis of evaluation data from Japan, improvements to the current program be derived.

In the pressurized water nuclear reactors (PWRs), the upper and bottom head penetration nozzles, the geometric asymmetry of the welded part increases from the center to the outer part, increasing the possibility of defects. For this reason, it is important to perform early detection and management through analysis of defects occurring in the welded parts of upper and bottom penetration nozzles of reactor vessel. However, it is very difficult to operate boat sampling of the welding area because the spacing of the penetration nozzle of the bottom head of the reactor is very narrow. In addition, it is more difficult to collect welded specimens of bottom penetration nozzles by electrical discharge machining in a boric acid water environment of nuclear reactor. In this work, to overcoming these technical difficulties, we developed a boat sampling robot system, which is composed of the specimen collection electrode head, borate-mediated discharge electrode and control system. Also, we performed basic performance tests and summarize the results.

To develop technology for extracting energy resources from seawater, we first investigated the research experiences of domestic experts. The survey items included the types of adsorbents that can adsorb dissolved resources in seawater, the subjects of experiments, and the scope of research. We divided the types of adsorbents into organic and inorganic categories and compared their adsorption performance. We also examined how adsorption experiments were conducted using simulated solutions and confirmed whether there were any experiences of conducting experiments in actual seawater. A total of 14 domestic research papers on extracting dissolved resources from seawater were reviewed, excluding topics such as removing dissolved resources from seawater and seawater desalination. This review provides an understanding of domestic research trends and will be helpful in setting directions for future research and development.

Decommissioning plan of nuclear facilities require the radiological characterizations and the establishment of a decommissioning process that can ensure the safety and efficiency of the decommissioning workers. By utilizing the rapidly developed ICT technology, we have developed a technology that can acquire, analyze, and deliver information from the decommissioning work area to ensure the safety of decommissioning workers, optimize the decommissioning process, and actively respond to various decommissioning situations. The established a surveillance system that monitors nuclide inventory and radiation dose distribution at dismantling work area in real time and wireless transmits data for evaluation. Developed an evaluation program based on an evaluation model for optimizing the dismantling process by linking real-time measurement information. We developed a technology that can detect the location of dismantling workers in real time using stereovision cameras and artificial intelligence technology. The developed technology can be used for safety evaluation of dismantling workers and process optimization evaluation by linking the radionuclides inventory and dose distribution in dismantling work space of decommissioning nuclear power plant in the future.

The concrete structure of a nuclear power plant is a major safety structure that performs shielding functions to block radioactive materials and radiation, heat removal, and isolation functions. Therefore, concrete structures of nuclear power plants must prove structural safety from immediately after construction to dismantling, and a representative method for this is to investigate compressive strength. The compressive strength and specimen standards of concrete structures are specified in ASTM C 42/C 42M, and samples must be obtained through core drilling in order to collect samples according to this standard. However, commercial equipment requires anchor installation work causes radiation dust generation. Even commercial products have developed equipment that does not require anchor installation work, but it can only be applied to flat walls and cannot be applied to curved walls such as bioshields. To solve this problem, a method of fixing to the scaffolding pipe was designed. The equipment developed based on this method fundamentally blocks the generation of radioactive dust. The vertical position can be adjusted using guide shafts and jack screws, and the horizontal position can be adjusted using scaffolding clamps. In addition, the distance between the installation location and the wall can be adjusted by adjusting the scaffolding clamp location of the device. Lastly, it can be rotated to the left and right, so that even on a curved wall, the sampling position can be performed perpendicular to the wall. Core drills that take specimens for measuring compressive strength use the wet type. Core drilling by wet type in radioactively contaminated concrete leads to the disposal of sludge as radioactive waste. Water supplied during core drilling is scattered in all directions by the rotation of the core drill bit, which causes radiation exposure to workers, so measures must be taken to ensure that the water does not splash and gather in one place. Nileplant Co., Ltd. has developed a sludge collection device that can be used with a core drilling device. It can be inserted into a 4-inch core drill bit to meet the specimen regulations of ASTM C 42/C 42M, and nylon resin was used as a material to withstand friction with water, and the wall of the drainage part was thickened to increase durability. Based on these results, it is expected to be able to work more quickly and safely when collecting core drilling samples of radioactively contaminated concrete or radiation and concrete.

The intermediate level spent resins waste generated from water purification for the the moderator and primary heat transport system during operaioin of heavy water reactor (HWR). Especially, moderator resins contain high level activity largely because of their C-14 content. So spent resins are considered as a problematirc solid waste and require special treatment to meet the waste acceptance criteria for a disposal site. Various methods have been studied for the treatment of spent resins which include thermal, destructive, and stripping methods. In the case of solidification methods, cement, bitument or organic polymers were suggested. In the 1990s, acid stripping using nitric acid and thermal treatment methods were actively investigated in Canada to remove C-14 nuclide from waste resin. In Japan, thermal distructive method was studied in the 1990s. Since 2005, KAERI developed acid stripping method using phosphate salt. However, acid stripping method are not suitable due to large amounts of 2nd waste containing acid solution with various nuclides. To solve this probelm, KAERI has been suggested the microwave treatment method for C-14 selective removal from waste resin in the 2010s. Pilot scale demonstration tests using radioactive waste resin generated from Wolsung unit 1 and unit 2 were successfully conducted and 95% of C-14 was selectively removed from the radioactive waste resin. In recent years, price of C-14 source is dramatically increased due to market growth of C-14 utilization and exclusive supply chain depending on China and Russia. High purity of C-14 were captured in HWR waste resin. Interest of C-14 recovery research from HWR waste resin is currently increased in Canada. In this study, microwave method is suggested to treat HWR waste resin with C-14 recovery process. Additionally, status of waste resin management and research trends of HWR waste resin treatment are introduced.

Laser scabbling has the potential to be a valuable technique capable of effectively decontaminating highly radioactive concrete surface at nuclear decommissioning sites. Laser scabbling tool using an optical fiber has a merits of remote operation at a long range, which provides further safety for workers at nuclear decommissioning sites. Furthermore, there is no reaction force and low secondary waste generation, which reduces waste disposal costs. In this study, an integrated decontamination system with laser scabbling tool was employed to test the removal performance of the concrete surface. The integrated decontamination system consisted of a fiber laser, remote controllable mobile cart, and a debris collector device. The mobile cart controlled the translation speed and position of the optical head coupled with 20 m long process fiber. A 5 kW high-powered laser beam emitted from the optical head impacted the concrete block with dimensions of 300 mm × 300 mm × 80 mm to induce explosive spalling on its surface. The concrete debris generated from the spalling process were collected along the flexible tube connected with collector device. We used a three-dimensional scanner device to measure the removed volume and depth profile.

Air conditioning facilities in nuclear power plants use pre-filters, HEPA filters, activated carbon filters, and bag filters to remove radionuclides and other harmful substances in the atmosphere. Spent filters generate more than 100 drums per year per a nuclear power plant and are stored in temporary radioactive waste storage. Plasma torch melting technology is a method that can dramatically reduce volume by burning and melting combustible, non-flammable, and mixed wastes using plasma jet heat sources of 1,600°C or higher and arc Joule heat using electric energy, which is clean energy. KHNP CRI & KPS are developing and improving waste treatment technology using MW-class plasma torch melting facilities to stably treat and reduce the volume of radioactive waste. This study aims to develop an operation process to reduce the volume of bag filter waste generated from the air conditioning system of nuclear power plants using plasma torch melting technology, and to stably treat and dispose of it. It is expected to secure stability and reduce treatment costs of regularly generated filter waste treatment, and contribute to the export of radioactive waste treatment technology by upgrading plasma torch melting technology in the future.

After the Fukushima accident, significant amount of radioactively contaminated waste has been generated with 50~250 m3/day and stored in tanks of the Fukushima Daiichi nuclear power plant site. The contaminated water is treated by various treatment facility such as KURION, SARRY, Reverse Osmosis, and ALPS to remove 62 radioactive nuclides except H-3. For the contaminated water treatment process, massive secondary wastes such as sludge, spent adsorbent, and so on as by-product are being generated by the facilities. In Japan, to treat the secondary wastes, melting technologies such as GeoMelt, In-can vitrification and Cold Crucible Induction Melting vitrification are considered as a candidate technologies. In this study, the technologies were reviewed, and the advantage and disadvantage of each technology were evaluated as the candidate technologies for treatment of the secondary wastes.

KHNP’s vitrification technology introduced a commercialized vitrification facility to the Hanul nuclear power site after a commercialization test through a lab test and a pilot plant at KHNP-CRI. France’s ANADEC (consortium with CEA, Orano, ECM Technologies and Andra) conducted a feasibility evaluation from FY2018 to FY2021 to apply In-Can vitrification, which was developed to treat Fukushima Effluent Treatment Waste (FETW) such as carbonate slurry and ferric slurry generated from ALPS (Advanced Liquid Processing System-Multi Radionuclides Removal) facilities for waste treatment in Fukushima, Japan. For commercialization, the following method was used. First, through the Laboratory scale studies, the possibility of high waste loading (60wt% in dry mass) of slurry on borosilicate matrix was tested. In addition, the volatility of radionuclide was evaluated through radionuclides surrogates with a Bench-scale mockup and glass discharge (100 kg) was evaluated through In-Can vitrification process verification. The feeding system was improved through a pilot scale test, and finally, glass discharge (300 kg) was evaluated after large amount of waste was treated through an industrial prototype (Fullscale) at the CEA Marcoule site (France).

The acceptance criteria for low and intermediate level radioactive waste disposal facilities in Korea to regulate that homogeneous waste, such as concentrated waste and spent resin, should be solidified. In addition, solidification requirements such as compressive strength and leaching test must be satisfied for the solidified radioactive waste solidified sample. It is necessary to develop technologies such as the development of a solidification process for radioactive waste to be solidified and the characteristics of a solidification support. Radioactive waste solidification methods include cement solidification, geopolymer solidification, and vitrification. In general, low-temperature solidification methods such as cement solidification and geopolymer solidification have the advantage of being inexpensive and having simple process equipment. As a high-temperature solidification method, there is typically a vitrification. Glass solidification is generally widely used as a stabilization method for liquid high-level waste, and when applied to low- and intermediate-level radioactive waste, the volume reduction effect due to melting of combustible waste can be obtained. In this study, the advantages and disadvantages of the solidification process technology for radioactive waste and the criteria for accepting the solidified material from domestic and foreign disposal facilities were analyzed.

Plasma melting technology is a high-temperature flame of about 1,600°C or higher generated using electrical arc phenomena such as lightning, and radioactive waste generated during the operation and dismantling of nuclear power plants is not classified according to physical characteristics. It is a technology that can meet waste disposal requirements through treatment and reduction. Plasma torch melting technology was used for volume reduction and stable treatment of HVAC filters generated from nuclear power plants HVAC (Heating Ventilation and Air Conditioning). filter was treated by placing 1 to 3 EA in a drum and injecting it into a plasma melting furnace at 1,500°C, and the facility was operated without abnormal stop. A total of 132.5 kg of filter was treated, and the high-temperature melt was normally discharged four times. It was confirmed that the plasma torch melting facility operates stably at 500 LPM of nitrogen and 370-450 A of current during filter treatment. Through this study, the possibility of plasma treatment of filters generated at nuclear power plants has been confirmed, and it is expected that stable disposal will be possible in the future.