This study explores the impact of metal doping on the surface structure of spent nuclear fuels (SNFs), particularly uranium dioxide (UO2). SNFs undergo significant microstructural changes during irradiation, affecting their physical and chemical properties. Certain elements, including actinides and lanthanides, can integrate into the UO2 lattice, leading to non-stoichiometry based on their oxidation state and environmental conditions. These modifications are closely linked to phenomena like corrosion and oxidation of UO2, making it essential to thoroughly characterize SNFs influenced by specific element doping for disposal or interim storage decisions. The research employs X-ray diffraction (XRD), scanning electron microscopy (SEM), and Raman spectroscopy to investigate the surface structure of UO2 samples doped with elements such as Nd3+, Gd3+, Zr4+, Th4+, and ε-particles (Mo, Ru, Pd). To manufacture these samples, UO2 powders are mixed and pelletized with the respective dopant oxide powders. The resulting pellet samples are sintered under specific conditions. The XRD analysis reveals that the lattice parameters of (U,Nd)O2, (U,Gd)O2, (U,Zr)O2, and (U,Th)O2 linearly vary with increasing doping levels, suggesting the formation of solid solutions. SEM images show that the grain size decreases with higher doping levels in (U,Gd)O2, (U,Nd)O2, and (U,Zr)O2, while the change is less pronounced in (U,Th)O2. Raman spectroscopy uncovers that U0.9Gd0.1O2-x and U0.9Nd0.1O2-x exhibit defect structures related to oxygen vacancies, induced by trivalent elements replacing U4+, distorting the UO2 lattice. In contrast, U0.9Zr0.1O2 shows no oxygen vacancy-related defects but features a distinct peak, likely indicating the formation of a ZrO8-type complex within the UO2 lattice. ε-Particle doped uranium dioxide shows minimal deviations in surface properties compared to pure UO2. This structural characterization of metal-doped and ε-particle-doped UO2 enhances our understanding of spent nuclear fuel behavior, with implications for the characterization of radioactive materials. This research provides valuable insights into how specific element doping affects the properties of SNFs, which is crucial for managing and disposing of these materials safely.

Thermal cutting processes that can be applied to dismantling nuclear power plants include oxygen cutting, plasma cutting, and laser cutting. According to the global trend, research projects are being carried out in various countries to upgrade laser cutting, and many studies are also being conducted in Korea with plans to apply laser cutting processes when dismantling nuclear power plants. However, with the current technology level of the laser cutting process, the maximum thickness that can be cut is limited to 250 mm. Therefore, in this study, a laser-oxygen hybrid cutting process was implemented by adding a laser heat source to the oxygen cutting process that can cut carbon steel with a thickness of 250 mm or more (RV, beam, column, beam, etc.) when dismantling the nuclear power plant. This has the advantage of improving the cutting speed and reducing the cutting width Kerf compared to conventional oxygen cutting. In this research, the laser-oxygen hybrid cutting process consisted of laser cutting to which Raycus’ 8 kW Fiber Laser power source was applied and oxygen cutting to which hydrogen was applied with Fuel Gas. The oxygen torch was placed perpendicular to the test piece, and the laser head was irradiated by tilting 35° to 70°. The effects of cutting directions on quality and performance were studied, and cutting paths were selected by comparing cutting results. Thereafter, it was confirmed that there is an optimal laser output power according to the cutting thickness by studying the effect on the cutting surface quality by changing only the laser output power under the same cutting conditions. The results of this study are expected to be helpful in the remote cutting process using laser-oxygen hybrid cutting when dismantling domestic nuclear power plants in the future.

After the major radioactivation structures (RPV, Core, SG, etc.) due to neutron irradiation from the nuclear fuel in the reactor are permanently shut down, numerous nuclides that emit alpha-rays, beta-rays, gamma-rays, etc. exist within the radioactive structures. In this study, nuclides were selected to evaluate the source term for worker exposure management (external exposure) at the time of decommissioning. The selection of nuclides was derived by sequentially considering the four steps. In the first stage, the classification of isotopes of major nuclides generated from the radiation of fission products, neutron-radiated products, coolant-induced corrosion products, and other impurities was considered as a step to select evaluation nuclides in major primary system structures. As a second step, in order to select the major radionuclides to be considered at the time of decommissioning, it is necessary to select the nuclides considering their half-life. Considering this, nuclides that were less than 5 years after permanent suspension were excluded. As a third step, since the purpose of reducing worker exposure during decommissioning is significant, nuclides that emit gamma rays when decaying were selected. As a final step, it is a material made by radiation from the fuel rod of the reactor and is often a fission product found in the event of a Severe accident at a nuclear power plant, and is excluded from the nuclide for evaluation at the time of decommissioning is excluded. The final selected Co-60 is a nuclide that emits high-energy gamma rays and was classified as a major nuclide that affects the reduction of radiation exposure to decommissioning workers. In the future, based on the nuclide selection results derived from this study, we plan to study the evaluation of worker radiation exposure from crud to decommissioning workers by deriving evaluation results of crud and radioactive source terms within the reactor core.

For the release of the nuclear power plant site after the decommissioning, a reliable exposure dose assessment considering the environmental impact of residual radionuclides is essentially required. In this study, the Derived Concentration Guideline Level (DCGL) for the hypothetically contaminated surface soil at the Wolsong nuclear power plant (NPP) unit 1 site was preliminarily calculated by using the RESRAD-OFFSITE computational code and compared with the other case studies. Moreover, radiation exposure dose for local residents and relevant exposure pathways were quantitatively analyzed based on the calculation model established through this work. For the target site modeling, the source term was determined by referring to the previous case studies regarding the nuclear power plant decommissioning, quantification analysis data of pressure tubes of Wolsong NPP unit 1, and radionuclide data estimated by using the MCNP/ORIGEN-2 code. In total, 14 different radioisotopes such as Ag-108m, C-14, Co-60, Cs-134/137, Fe-55, H-3, Nb-93m/94, Ni-63, Sb-125, Sn-121m, Sr-90, and Zr-93 were considered as target radionuclides. In addition, the geological structure model of the Wolsong NPP site was established based on the final safety analysis report of Wolsong NPP unit 1. The distribution coefficients (Kd) were taken from the JAEA-SDB to estimate the migration/retardation behavior of various radionuclides under the groundwater condition of the Wolsong NPP site. In the present work, the DCGL values were calculated according to the site release criterion of 0.1 mSv/yr, which indicates the radiation protection standard for the site release. Moreover, the exposure pathway and sensitivity analyses were conducted to assess the sensitive input parameters remarkably influencing the calculation result. For the evaluation of exposure dose for local residents, a site layout centered around Wolsong NPP unit 4, located in the closest proximity to the residents’ habitation area, was alternatively established and all potential exposure pathways were considered as a comprehensive resident farmer scenario. The results obtained from this study are expected to serve as a preliminary case study for the DCGL values regarding the surface soil at the Wolsong NPP unit 1 site and for evaluating the radiation exposure dose to local residents resulting from the residual radioactivity at the site after the decommissioning.

Domestic commercial low- and intermediate-level radioactive waste storage containers are manufactured using 1.2 mm thick cold-rolled steel sheets, and the outer surface is coated with a thin layer of primer of 10~36 μm. However, the outer surface of the primer of the container may be damaged due to physical friction, such as acceleration, resonance, and vibration during transportation. As a result, exposed steel surfaces undergo accelerated corrosion, reducing the overall durability of the container. The integrity of storage containers is directly related to the safety of workers. Therefore, the development of storage containers with enhanced durability is necessary. This paper provides an analysis of mechanical properties related to the durability of WC (tungsten carbide)-based coating materials for developing low- and intermediate-level radioactive waste storage containers. Three different WC-based coating specimens with varied composition ratios were prepared using HVOF (high-velocity oxy-fuel) technique. These different specimens (namely WC-85, WC-73, and WC-66) were uniformly deposited on cold-rolled steel surfaces ensuring a constant thickness of 250 μm. In this work, the mechanical properties of the three different WCbased coaitng materials evaluated from the viewpoints of microstructure, hardness, adheision force between substrate and coating material, and wear resistance. The cross-sectional SEM-EDS (Scanning Electron Microscope-Energy Dispersive X-ray Spectroscopy) images revealed that elements W (tungsten), C (carbon), Ni (nickel), and Cr (chromium) were uniformly distributed within the each coating layers which was approximately 250 μm thick. The average hardness values of HWC-85 and HWC-73 were found to be 1,091 Hv (Vickers Hardness) and 1,083 Hv, respectively, while the HWC-66 exhibited relatively lower hardness value of 883 Hv. This indicates that a higher WC content results in increased hardness. Adhesion force between and substrates and coating materials exceeded 60 MPa for all specimens, however, there were no significant differences observed based on the tungsten carbide content. Furthermore, a taber-type abrasion tester was used for conducting abrasion resistance tests under specific conditions including an H-18 load weight at 1,000 g with rotational speed set at 60 RPM. The abrasion resistance tests were performed under ambient temperatures (RT: 23±2°C) as well as relative humidity levels (RH: 50±10%). Currently, the ongoing abrasion resistance tests will include some results in this study.

Activated carbon (AC), extensively used across various industrial sectors, serves as a sponge for different types of gases due to its porous carbon material. These gases are attracted to the carbon substrate via van der Waals forces. In nuclear power plants, AC is commonly used to adsorb radioactive gases such as 86Kr and 134Xe, as well as radioiodine sources like 131I and 133I from gaseous effluents. Even if the adsorbed radioactive gases and radioiodine decay into non-radioactive elements, the spent AC still contains radioactive species with long half-lives, such as 3H (Tritium, T) and 14C (radiocarbon). Minimizing and separating waste that contains long-lived nuclides (e.g., 14C) are pivotal components of an efficient waste management approach. A challenging aspect of effectively managing disposed AC is to minimize long-lived radioactive substances by eliminating them. This paper explores and summarizes the technology used to remove pollutants (3H, 14C) trapped within the pores of Activated carbon through thermochemical vacuum and surface oxidation processes. By recycling and reusing spent Activated carbon, we anticipate a reduction in the volume of radioactive waste, leading to decreased disposal costs. Furthermore, this paper will contribute as a valuable reference in future studies, enhancing the understanding of vacuum thermal desorption and surface oxidation of used Activated carbon.

For the performance and safety assessments of deep geological disposal, developing scenarios, which represent possible long-term changes in the surface environment, is required. These scenarios are formulated using a list of FEPs (Features, Events, and Processes) that describes characteristics of disposal system components. In this study, using international FEP (IFEP) list from OECD/NEA, the individual FEPs related to uplift-subsidence and erosion-deposition were analyzed, and the correlation between each FEP was evaluated. From the IFEP list, the elements related to uplift-subsidence and erosion-deposition processes that cause long-term changes in the surface environment were identified. Uplift-subsidence, erosion - deposition, and the long-term change factors caused by them were analyzed and a correlation diagram was produced according to their interactions. Basis for the integrated analysis of long-term changes in the surface environment and the construction of long-term change scenarios were established considering the evaluation of the factors that cause uplift-subsidence and erosiondeposition, and their correlation with the hydrology-hydrogeology, topography and local climate of the affected surface. The results of this study will be used for systematically formulating scenarios of long-term changes in the surface environment due to uplift-subsidence and erosion-deposition based on natural phenomena. And, it may be necessary to modify and supplement the correlation of domestic FEPs based on the correlation diagram of IFEPs in order to analyze long-term changes in the surface environment in an integrated manner.

Safe management of spent nuclear fuel (SNF) is a key issue to determine sustainability of current light water reactor (LWR) fleet. However, none of the countries are actually conducting permanent disposal of SNFs yet. Instead, most countries are pursuing interim storage of spent nuclear fuels in dry cask storage system (DCSS). These dry casks are usually made of stainlesssteels for resistibility against cracking and corrosion, which can be occurred over a long-term storage period. Nevertheless, some corrosion called Chloride-Induced Stress Corrosion Cracking (CISCC) can arise in certain conditions, exacerbating the lifetime of dry casks. CISCC can occur if the three conditions are satisfied simultaneously: (i) residual tensile stress, (ii) material sensitization, and (iii) chloride-rich environment. A residual tensile stress is developed by the two processes. One is the bending process of stainless-steel plates into a cylindrical shape, and the other is the welding process, which can incur solidification-induced stress. These stresses provide a driving force of pit-to-crack transition. Around the fusion weld areas, chromium is precipitated at the grain boundary as a carbide form while it depletes chromium around it, leading to material susceptible to pitting corrosion. It is called sensitization. Finally, coastal regions, where nuclear power plants usually operate, tend to have a higher relative humidity and more chloride concentration compared to inland areas. This high humidity and chloride ion concentration initiate pitting corrosion on the surface of stainless-steels. To prevent initiation of CISCC, at least one of the three conditions should be removed. For this, several surface engineering techniques are under investigation. One of the most promising approaches is surface peening method, which is the process that impacts the surface of materials with media (e.g., small pins, balls, laser pulse). By this impact, plastic deformation on the surface occurs with compressive stress that counteracts with pre-existing residual tensile stress, so this approach can prevent pit-to-crack transition of stainless-steels. Also, cold spray deposition can prevent CISCC. Cold spray deposition is a method of spraying fine metal powder to a substrate by accelerating them to supersonic velocity with propellant gas. As a result, a thin coating composed of the feedstock powders can protect the substrate from outer corrosive environments. In addition, the impact of the feedstock powder on the substrate during the process provides compressive stress, similar to the peening method.

This study presents a method for analyzing the surface temperatures of specific facilities, such as the 5 MWe reactor within the Yongbyon Nuclear Complex, to explore its potential utility in monitoring suspected nuclear-related activities in North Korea using thermal infrared (TIR) satellite imagery (Landsat series). TIR band data is utilized to derive surface temperatures in the specified areas, and the temperatures are analyzed on a monthly basis to examine any patterns within these regions. This research provides a pattern-of-life on temperature variation for the target areas through multiple TIR image datasets, offering additional information to analyze facilities’ operational status in remote and inaccessible regions.

프로바이오틱스 제품에 대한 수요가 지속적으로 증가하 고 있으며, Lactobacillus 균주가 가장 대중적인 프로바이 오틱스로 널리 사용되고 있다. 프로바이오틱스는 기준에 적합한 균수의 확보가 중요하며 제조원가나 시간 등을 낮 추기 위해 배양법의 개발이 필요하므로 Lactobacillus 생 산을 위한 배양 조건이 최적화되었다. 반응표면방법론에 의한 통계적 최적화에서 반응 변수에 영향을 미치는 독립 변수의 최적 조건은 Lactobacillus acidophilus의 경우 22.55 시간(배양시간), 25oC(배양온도), 3.41%(프리바이오틱스 농 도); Lactiplantibacillus plantarum의 경우 24시간, 30.86oC, 2.00%; Lacticaseibacillus rhamnosus의 경우 66.67시간, 35oC, 3.41%이었다. Lactobacillus의 최적 배양조건은 예측 한 결과와 실제 결과가 밀접하게 일치하는 것을 확인하였 다. 이러한 데이터는 수율 높은 Lactobacillus를 생산하는 데 중요한 포인트를 제공할 것이다.

Injection molding is a process of shaping resin materials by heating them to a temperature above their melting point and then using a mold. The resin material is injected into and cooled within the mold cavity, solidifying into the desired shape. The core and cavity components that make up the mold cavity are crucial elements for the precision molding in injection molding. In the case of precision mold production, the application of 5-axis machining technology is required to ensure high machining quality for complex shapes, and among these factors, the tool angle is a critical machining condition that determines the surface roughness of the workpiece. In this study, we aim to measure the surface roughness of the machined surface of KP4A specimens during machining processes with variations in the tool angle and analyze the correlation between the tool angle and surface roughness.

본 연구에서는 Monopile 방식 풍력발전기 강구조물의 부식을 방지하기 위하여 S355 steel의 표면 거칠기에 따른 용사 코팅 상태에 관한 연구를 수행했다. 일차적으로는 시편별 서로 다른 표면거칠기를 부여하기 위해 밀링머신에 페이스 커터를 결합하여 시편별로 다른 조건의 Ra값 기준 표면거칠기를 부여했다. 실험 조건으로는 시편 가공 시 4가지의 회전속도(60, 400, 1200, 2000 rpm), feed rate 150(mm/min) 조건을 선정했다. 2차로는 와이어 용융 방식의 아크 용사 코팅을 실시했다. 코팅 조건으로는 분사 거리 200mm, 전압 24V, 전류 120A, 분사 압력 5bar, 와이어 삽입 속도 30g/mm, 와이어 직경 2mm이다. 용사 코팅 후 FE-SEM으로 표면을 관찰한 결과 모든 시편의 S355 면과 코팅층(아연-알루미늄) 사이에 유격이 발생하지 않고 성공적으로 안착이 되었음을 확인할 수 있었다.

본 연구에서는 미국 해양 대기청(NOAA)의 NOAA-20 위성에 장착된 차세대 고해상도 복사계인 VIIRS로부터 산출된 적외 해수면온도의 자료를 수집하고, 실측 자료와의 일치점을 생산하여 한반도 주변 해역에서의 정확도를 검증 하였다. 2020년 5월부터 2023년 6월까지 최근 3년간의 자료를 사용하였고, 총 75,700개의 일치점을 생산하였다. NOAA-20/VIIRS 해수면온도는 표층 뜰개 부이 관측 해수면온도와 비교해보았을 때 약 0.52K의 평균 제곱근 오차와 – 0.12 K의 평균 편차를 보였고, 이는 전구 해역을 대상으로 한 기존의 정확도 검증 연구 결과값을 상회하는 수치였다. NOAA-20 해수면온도의 오차 특성 분석 결과 겨울과 봄에는 음의 편차가, 여름철에는 양의 편차를 보이는 계절적 특 성이 나타났으며, 15-16시에 최대 평균 제곱근오차, 최대 양의 편차 및 22-24시에 최소 평균제곱근오차, 최소 편차를 가지는 일간 변화를 보였다. 이외에도 NOAA-20 해수면온도의 오차는 풍속, 위성 천정각, 연안으로부터의 거리, 해수면 온도의 공간 구배 크기에 영향을 받아 변동하는 특성이 나타났다. 전반적으로 위성 해수면온도의 편차값은 14ms1 이 상의 풍속 범위에서 풍속이 커질수록 양의 방향으로 증가하는 경향을 보였으며, 5 m s1 이하의 낮은 풍속 범위에서는 풍속이 약해질수록 낮/밤 자료에 따라 각각 양의 방향, 음의 방향으로 편차가 증가하였다. 위성 천정각이 커질수록 해 수면온도의 오차 범위는 급격하게 증가하였으며, 연안에 근접할수록 (<300 km) 위성 해수면온도의 오차가 증가하는 것 을 확인할 수 있었다. 해수면온도의 공간 구배는 그 크기가 커질수록 위성 해수면온도의 평균 제곱근 오차를 증폭시키 는 경향이 나타났다. 국지적인 해역에서의 위성 해수면온도 정확도 및 오차 특성은 전구 해역에서의 전반적인 특성과는 다르게 나타날 수 있다는 점을 고려할 때 본 연구는 향후 한반도 주변해에서 VIIRS 해수면온도를 활용하기 위한 선행 연구로 해수면온도 오차의 변동 특성 및 분포에 대한 깊은 이해가 필요함을 시사한다.

대조차 개방형 조간대 퇴적물에 대한 태풍 효과를 한국 서해안 고창 동호리 조간대에서 연구하였다. 2010년 태 풍 곤파스 전·후와 2018년 솔릭 전·후에 나타난 표층 퇴적물 조직, 집적률, 퇴적상 변화를 관측하였다. 두 개의 태풍 곤파스와 솔릭은 각각 2010년 9월 1일과 2일, 2018년 8월 23일과 24일 사이에 한반도 남서부 연안에 상륙하여 내륙을 관통하였다. 태풍 곤파스 전·후와 솔릭 전·후에 고창 동호리 조간대의 측선을 따라 30m 간격으로 표층 퇴적물을 채취 하고 집적률을 측정하였다. 동호리 조간대를 평균고조면, 평균해수면, 평균저조면을 기준으로 고조대, 중조대, 저조대로 세분하여 연구하였다. 두 태풍 모두 태풍 후에 고조대는 퇴적되었고 중조대와 저조대는 침식되었다. 고조대에서 저조대 방향으로 갈수록 집적률이 감소하였다. 고조대의 표층 퇴적물 조직은 두 태풍 모두 태풍 후에 평균 입도는 세립해지고 분급은 양호해졌다. 중조대의 표층 퇴적물 조직은 변화가 거의 없었으며, 저조대는 일부 구간에서 태풍 후에 띠 형태의 퇴적체가 나타났다.

This study used optical and scanning electron microscopy to analyze the surface oxidation phenomenon that accompanies a γ'-precipitate free zone in a directional solidified CM247LC high temperature creep specimen. Surface oxidation occurs on nickel-based superalloy gas turbine blades due to high temperature during use. Among the superalloy components, Al and Cr are greatly affected by diffusion and movement, and Al is a major component of the surface oxidation products. This out-diffusion of Al was accompanied by γ' (Ni3Al) deficiency in the matrix, and formed a γ'-precipitate free zone at the boundary of the surface oxide layer. Among the components of CM247LC, Cr and Al related to surface oxidation consist of 8 % and 5.6 %, respectively. When Al, the main component of the γ' precipitation phase, diffused out to the surface, a high content of Cr was observed in these PFZs. This is because the PFZ is made of a high Cr γ phase. Surface oxidation of DS CM247LC was observed in high temperature creep specimens, and γ'-rafting occurred due to stress applied to the creep specimens. However, the stress states applied to the grip and gauge length of the creep specimen were different, and accordingly, different γ'-rafting patterns were observed. Such surface oxidation and PFZ and γ'-rafting are shown to affect CM247LC creep lifetime. Mapping the microstructure and composition of major components such as Al and Cr and their role in surface oxidation, revealed in this study, will be utilized in the development of alloys to improve creep life.

자율운항선박 기술은 점차 발전하고 있다. 하지만 완전 자율운항선박이 등장하기 전까지는 원격운영센터에서 원격운영자가 제 어하는 형태를 가지게 될 것이다. 그러나 현재 그들의 면허체계는 국내외적으로 정해지지 않았다. 역량이 검증되지 않은 원격운영자의 등 장은 항행 안전에 위험이 될 것이다. 본 논문에서는 문헌 연구를 통해 원격운영자의 면허체계를 위해 고려해야 할 평가기준과 국내 해기 사 면허체계 내에서 수립할 수 있는 방안을 모색하고 AHP를 활용하여 분석하였다. 그 결과 원격운영자의 면허체계를 위해서는 우선적으 로 법률제정이 필요하고 선박직원법 제4조 해기사 면허의 직종에 원격운영자의 직종을 추가하는 방안이 가장 선호되었다. 이에 따라 원 격운영센터의 조직구성도 기존 선박의 선교인적관리 조직과 유사하게 형성할 수 있을 것이다. 본 연구는 자율운항선박의 인적 관리 측면 에서 원격운영자의 효율적인 양성과 항행 안전에 이바지할 수 있을 것이다.