해양 부유식 구조물의 계류시설에 대한 연구는 다양한 계류 시스템의 개발과 그 효율성에 대해 지 속적으로 연구되어 왔다. 계류 시스템은 구조물의 안전성, 내구성, 그리고 환경적인 책임을 모두 고려 해야하는 복합적인 설계 요소이다. 기존 계류바익은 해저지반의 특성에 크게 의존하며, 넓은 점유 면 적으로 인해 해양환경과 활동에 부정적인 영향을 줄 수 있다. 이러한 기존 계류시스템의 한계와 문제 점에 대응하기 위해, 본 연구에서는 새로운 계류시스템을 제안하고자 한다. 제안된 계류시스템은 계류 선이 해저지반에 고정되지 않고 중력식 기초 구조물을 거쳐 중간부력재에 연결된다. 중간 부력재는 상 시 과잉 부력상태로 상향력이 작용하며, 이로인해 계류선에 초기장력이 유도되며 전체 시스템에 강성 을 도입한다. 전체적으로 Semi-taut과 유사한 방식으로 계류 장력 변화에 따른 상단 플랫폼의 운동 제어와, 중간 부력재와의 상호연계 거동효과로 추가적인 동적 응답 저감이 기대된다. 본 연구에서는 제안된 신형식 계류시스템의 역학적 거동특성을 수치해석적으로 구현하고 파랑에 대한 Catenary system 대비 운동 저감 성능을 검증하였다.

목적 : 현재 주목받고 있는 스마트 글라스와 같은 증강현실 기술은 배터리, 발열, 착용감 문제 등으로 사용자의 불편을 야기시킨다. 본 연구는 이러한 문제들을 해결하고 사용자 편의성을 극대화할 수 있는 새로운 NED 광학계 를 설계하고자 한다. 방법 : Waveguide 타입과 비구면 구성을 적용한 허상광학계로 NED 광학계를 설계하였다. Code V 소프트웨어 를 사용하였고, 경량화, 높은 시야각, 낮은 왜곡 수차 및 우수한 이미지 품질을 설계 목표로 하여 비순차 광학계를 구성하였다. 결과 : 설계된 광학계는 배율 증가, 포커싱, 수차 감소에 초점을 맞춘 첫 번째 파트와 빛의 손실을 최소화하는 Waveguide 타입의 두 번째 파트로 구성하여 고효율성과 높은 해상도를 나타내었다. 사용자에게 선명하고 편안한 시각적 경험을 보장하고, 컴팩트한 디자인으로 휴대성과 사용자 편의성이 동시에 고려된 광학계를 설계하였다. 결론 : 본 연구에서 개발된 Waveguide 타입과 비구면 광학계는 빛의 손실을 최소화하여 망막까지 효율적으로 전달함으로써 NED 기술의 성능을 크게 향상시켰다. 이는 사용자에게 더 높은 선명도와 생동감 있는 시각적 경험 을 제공하며, 향후 고해상도 광학 응용 분야에 큰 기여를 할 것으로 판단된다.

이 연구는 OpenAI의 GPT 아키텍처를 활용하여 진로교육에 맞춤형 교육과정 설계 시스템을 개발하는 것을 목적으로 하였다. 현대 교육 환경에서 AI의 통합은 학습자 중심의 교육 방법론 개발에 필수적이며, 이 연구는 그러한 필요성에 응답하고자 하였다. GPT 아키텍처의 언어 처리 및 생성 능력과 국가직무능력표준(NCS) 기반 교육과정의 통합 가능성을 탐색하여 시스템의 바탕을 마련하였으며, 다양한 분야의 전문 가 검토를 통해 연구가 진행되었다. 검토에 참여한 전문가들은 컴퓨터공학, 교육공학, 진로교육, 대학 학 사지원팀 분야에서 활동하는 박사 및 석사 학위 소지자들로 구성되었으며, 이들의 피드백을 바탕으로 시 스템 설계와 개선이 이루어졌다. 연구의 산출물로는 AI 기술과 진로교육의 융합을 통한 교육 분야의 혁신 적 접근법이 적용된 시스템 개발을 위한 스토리보드가 제시되었다. GPT 아키텍처를 활용한 이 시스템은 학생들에게 개인화된 진로 교육과정을 제공하는 데 있어 중요한 역할을 할 수 있음을 보여주었다. 이 연구는 교육 기술의 발전과 학습자 중심의 맞춤형 교육 방법론 개발을 시도하는데 의미가 있으며, 미래 교육 기술 발전 방향에 대한 중요한 통찰을 제공한다. 또한, AI 기술을 활용한 교육과정 설계의 가능 성을 탐색하고, 교육 현장에서의 AI 적용을 위한 실질적인 기반을 마련하는 데 중점을 두었다. 이러한 접근은 학습자 개별의 요구를 충족시키는 동시에 교육의 효과성을 높이는 방향으로 교육 기술을 발전시 킬 수 있는 기회를 제공한다 .

Code-compliant seismic design should be essentially applied to realize the so-called emulative performance of precast concrete (PC) lateral force-resisting systems, and this study developed simple procedures to design precast industrial buildings with intermediate precast bearing wall systems considering both the effect of seismic and blast loads. Seismic design provisions specified in ACI 318 and ASCE 7 can be directly adopted, for which the so-called 1.5S y condition is addressed in PC wall-to-wall and wall-to-base connections. Various coupling options were considered and addressed in the seismic design of wall-to-wall connections for the longitudinal and transverse design directions to secure optimized performance and better economic feasibility. On the other hand, two possible methods were adopted in blast analysis: 1) Equivalent static analysis (ESA) based on the simplified graphic method and 2) Incremental dynamic time-history analysis (IDTHA). The ESA is physically austere to use in practice for a typical industrial PC-bearing wall system. Still, it showed an overestimating trend in terms of the lateral deformation. The coupling action between precast wall segments appears to be inevitably required due to substantially large blast loads compared to seismic loads with increasing blast risk levels. Even with the coupled-precast shear walls, the design outcome obtained from the ESA method might not be entirely satisfactory to the drift criteria presented by the ASCE Blast Design Manual. This drawback can be overcome by addressing the IDTHA method, where all the design criteria were fully satisfied with precast shear walls’ non-coupling and group-coupling strength, where each individual or grouped shear fence was designed to possess 1.5S y for the seismic design.

The aim of this study is to ensure the structural integrity of a canister to be used in a dry storage system currently being developed in Korea. Based on burnup and cooling periods, the canister is designed with 24 bundles of spent nuclear fuel stored inside it. It is a cylindrical structure with a height of 4,890 mm, an internal diameter of 1,708 mm, and an inner length of 4,590 mm. The canister lid is fixed with multiple seals and welds to maintain its confinement boundary to prevent the leakage of radioactive waste. The canister is evaluated under different loads that may be generated under normal, off-normal, and accident conditions, and combinations of these loads are compared against the allowable stress thresholds to assess its structural integrity in accordance with NUREG-2215. The evaluation result shows that the stress intensities applied on the canister under normal, off-normal, and accident conditions are below the allowable stress thresholds, thus confirming its structural integrity.

A transfer cask serves as the container for transporting and handling canisters loaded with spent nuclear fuels from light water reactors. This study focuses on a cylindrical transfer cask, standing at 5,300 mm with an external diameter of 2,170 mm, featuring impact limiters on the top and bottom sides. The base of the cask body has an openable/closable lid for loading canisters with storage modules. The transfer cask houses a canister containing spent nuclear fuels from lightweight reactors, serving as the confinement boundary while the cask itself lacks the confinement structure. The objective of this study was to conduct a structural analysis evaluation of the transfer cask, currently under development in Korea, ensuring its safety. This evaluation encompasses analyses of loads under normal, off-normal, and accident conditions, adhering to NUREG-2215. Structural integrity was assessed by comparing combined results for each load against stress limits. The results confirm that the transfer cask meets stress limits across normal, off-normal, and accident conditions, establishing its structural safety.

항만 내 선박과 부두의 사고를 예방하기 위하여 통항 및 접안 안전성 평가를 통하여 안전한 부두가 건설되어 관리하고 있으나, 선 박의 접안 및 계류 과정에서 선박이 부두에 충돌하거나 로프로 인한 인명사고의 발생 등 예측할 수 없는 사고들이 종종 발생한다. 자동계류장 치는 선박의 신속하고 안전한 계류를 위한 자동화된 시스템으로 로봇 매니퓰레이터와 흡착 패드로 구성된 탈/부착 메커니즘을 가지고 있다. 본 논문은 자동계류장치의 흡착 패드의 위치 및 속도제어에 필요한 선체와의 변위 및 속도 측정 시스템을 다룬다. 자동계류장치에 적합한 측 정 시스템을 설계하기 위하여, 본 논문은 우선 센서의 성능 및 실외 환경적 특성 분석을 수행한다. 다음으로 이러한 분석 결과를 토대로 실외 부두환경에서 설치되는 자동계류장치에 적합한 변위 및 속도 측정시스템의 구성 및 설계 방법에 대해 기술한다. 또한 센서의 측정상태 감지 및 속도 추정을 위한 알고리즘을 제시한다. 제안된 방법은 다양한 속도 구간에서의 변위 및 속도 측정 실험을 통해 그 유용성을 검증한다.

KEPCO KPS is the contractor for the full system decontamination (FSD) of Kori Unit 1 and under preparation such as modification, lay out for equipment installation, setting up tie-in/out point for chemical injection and way to pressurize the system, of its successful performance. In this research, KPS introduced how KPS has designed and prepared for the FSD project and how will the chemical decontamination process be implemented. As described in the previous research, chemical decontamination process is planned to be conducted for three cycles and each cycle is consisted of oxidation, reduction, 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 IX by the heat. If the decontamination result does not meet the target DF and the dose rate, additional cycle can be conducted. Expected volume of process water for FSD is 200 m3. Three systems have been designated as decontamination targets: reactor coolant system (RCS), residual heat removal system (RHRS), chemical volume control system (CVCS). For the steady flow rate, existed plant equipment such as reactor coolant pump (RCP) will be operated and modifications on some components will be conducted. Due to the limited space for installation, decontamination equipment and other resources are distributed to three different places. KPS designed the layout of equipment installed inside the containment vessel. The layout contains the information of shielding for highly radiated equipment such as IX and filter skid.

Various radioactive metal wastes are generated during operation and decommissioning of nuclear facilities. Radioactive metal wastes with complex geometries or volumetric contamination can be difficult to decontaminate and disposal costs may increase. To solve these problems, the radioactive metal wastes can be treated by melting method. In this study, we designed a melting furnace system of air induction melting type, which is widely utilized due to its advantages of good thermal efficiency, uniform heating and guaranteed safety for radioactive material. By utilizing the melting furnace system, volatile radionuclides existed in the base material can be captured in the form of gas or dust by the filter. The radionuclides whose chemical properties can easily form metal oxides present as slag. For this reason, the specific radioactivity of the base material can be reduced. Radionuclides that are difficult to transport to slag and dust are uniformly distributed in the base material. A dedicated power supply and a transformer were necessary to be included in the melting furnace system since the induction furnace uses high-frequency currents. In addition, a hood is placed on top of the furnace to capture fumes generated during melting, and additional hoods were installed around the furnace to remove airborne dust. In particular, a dust collection unit consisting of a cyclone and a HEPA filter were constructed to effectively collect dust containing radionuclides. During the melting process, the slag is removed and accumulated separately, and the ingot production system was designed to produce the ingot using molten metal. The furnace was constructed for tilting the molten metal by moving the furnace using hydraulic system. The water cooling system and cooling tower were prepared to cool off the equipment with high temperature during melting is cooled off. The above process was specified in the operating procedure developed for this melting furnace system, and the operator shall operate and inspect according to the prescribed procedures. The radioactivity concentration in the sample taken in the step of tilting shall be analyzed whether they meet clearance level for self-disposal determined and publicly announced by the Commission. We can conduct self-disposal for the product of melting furnace system confirmed by the Commission as having the radioactivity concentration by nuclide not exceeding the value determined by the Commission.

The high-level radioactive waste repository must ensure its performance for a long period of time enough to sufficiently reduce the potential risk of the waste, and for this purpose, multibarrier systems consisting of engineered and natural barrier systems are applied. If waste nuclides leak, the dominating mechanisms facilitating their movement toward human habitats include advection, dispersion and diffusion along groundwater flows. Therefore, it is of great importance to accurately assess the hydrogeological and geochemical characteristics of the host rock because it acts as a flow medium. Normally, borehole investigations were used to evaluate the characteristics and the use of multi-packer system is more efficient and economical compared to standpipes, as it divides a single borehole into multiple sections by installing multiple packers. For effective analyses and groundwater sampling, the entire system is designed by preselecting sections where groundwater flow is clearly remarkable. The selection is based on the analyses of various borehole and rock core logging data. Generally, sections with a high frequency of joints and evident water flow are chosen. Analyzing the logging data, which can be considered continuous, gives several local points where the results exhibit significant local changes. These clear deviations can be considered outliers within the data set, and machine learning algorithms have been frequently applied to classify them. The algorithms applied in this study include DBSCAN (density based spatial clustering of application with noise), OCSVM (one class support vector method), KNN (K nearest neighbor), and isolation forest, of which are widely used in many applications. This paper aims to evaluate the applicability of the aforementioned four algorithms to the design of multi-packer system. The data used for this evaluation were obtained from DB-2 borehole logging data, which is a deep borehole locates near KURT.

Even though a huge amount of spent nuclear fuels are accumulated at each nuclear power plant site in Korea, our government has not yet started to select a final disposal site, which might require more than several km2 surface area. According to the second national plan for the management of high-level radioactive waste, the reference geological disposal concept followed the Finnish concept based on KBS-3 type. However, the second national plan also mentioned that it was necessary to develop the technical alternatives. Considering the limited area of the Korean peninsula, the authors had developed an alternative disposal concepts for spent nuclear fuels in order to enhance the disposal density since 2021. Among ten disposal concepts shown in the literature published in 2000’s, we narrowed them to four concepts by international experiences and expert judgements. Assuming 10,000 t of CANDU spent nuclear fuels (SNF), we designed the engineered barriers for each alternative disposal concept. That is, using a KURT geological conditions, the engineered barrier systems (EBS) for the following four alternative concepts were proposed: ① mined deep borehole matrix, ② sub-seabed disposal, ③ deep borehole disposal, and ④ multi-level dispoal. The quantitative data of each design such as foot prints, safety factors, economical factors are produced from the conceptual designs of the engineered barriers. Five evaluation criteria (public acceptance, safety, cost, technology readiness level, environmental friendliness) were chosen for the comparison of alternatives, and supporting indicators that can be evaluated quantitatively were derived. The AHP with domestic experts was applied to the comparison of alternatives. The twolevel disposal was proposed as the most appropriate alternative for the enhancement of disposal efficiency by the experts. If perspectives changes, the other alternatives would be preferred. Three kinds of the two-level disposal of CANDU SNF were compared. It was decided to dispose of all the CANDU spent nuclear fuels into the disposal holes in the lower-level disposal tunnels because total footprint of the disposal system for CANDU SNF was much smaller than that for PWR SNF. Currently, we reviewed the performance criteria related to the disposal canister and the buffer and designed the EBS for CANDU SNF. With the design, safety assessment and cost estimates for the alternative disposal system will be carried out next year.

This paper presents a study on the design and implementation of a secure contactless system leveraging Quick Response (QR) codes as a core component. The main goal of this system is to bridge the gap between strong security and improved user experience within the realm of digital interaction. The system's versatility can be expanded with broad compatibility with a variety of applications. Utility can be expanded to areas such as contactless payments, electronic ticketing, secure identity verification, and convenient access to medical records. The international standardization of QR codes ensures seamless cross-platform compatibility, strengthening their role in the digital ecosystem. We actually create and develop a non-contact security QR code system and check the expandability of the system. This study highlights the pivotal role of QR codes within the realm of secure contactless systems. Through its effective balance of digital security and user convenience, QR codes are emerging as an important element in the continued development of a secure and user-friendly digital environment. The potential for future research lies in exploring more complex use cases and further advancements that improve both security and user-centered design.

This research introduced a command-filtered backstepping control of mirror system to maintain laser communication between satellite and ground station. This requires a 2 degree of freedom gimbal mirror system using DC motors for target acquisition, pointing, and tracking (APT) system. This APT system is used for laser communication between satellite and ground stations. To track these desired angles, we have to control DC motors using introduced command-filtered backstepping controller (CFBSC) with disturbance. Command filtered backstepping controller has second order filter instead differentiation for simple and fast calculation. Introduced command-filtered backstepping control gives a smooth control signal for intermediate states. Simulation results verify that CFBSC outperforms SMC in terms of tracking error and disturbance rejection.

A vitrification facility control area is formed to control and monitor the vitrification facility process, and the control system is designed to manage the vitrification facility more safely and effectively. The control system is largely composed of a process control system and an off-gas monitoring system. The process control system is operated so that operation variables can be maintained in a normal state even in normal and transient conditions, and is designed so that the vitrification facility can be stably maintained in the event of an abnormality in the facility. The process control system consists of Programmable Logic Controller (PLC) and Local Control Panel (LCP), which controls and monitors each unit device. In addition, operation variables are provided to the operator so that the operator can manage operation variables during process control in a centralized manner for the operation of the vitrification facility. The off-gas monitoring system is operated to monitor whether the off-gas discharged to the environment is stably maintained within the standard level, and the off-gas is monitored through an independent monitoring system.

After melting glass at a high temperature of about 1,100 degrees in the Cold Crucible Induction Melter (CCIM) of the vitrification facility, radioactive waste is fed into the CCIM to vitrify radioactive waste. Accordingly, since the metal sector of the CCIM contacts the high-temperature molten glass, cooling water is supplied to continuously cool the metal sector. The cooling system is divided into primary and secondary cooling water systems. The primary cooling water flows inside the metal sector of the CCIM to maintain the metal sector within normal temperature, thereby forming a glass layer between the metal sector and the high-temperature melting glass. The secondary cooling system is a system that cools the primary cooling water that cools the metal sector, and removes heat generated from the primary cooling system. In addition, it is designed to stably supply cooling water to the secondary cooling water system through an emergency cooling water system so that cooling water can be stably supplied to the secondary cooling water system in the event of secondary cooling water loss. Therefore, it is designed to maintain the facility stably in the event of loss of cooling water for the CCIM of the vitrification facility.

Since the first operation of the Gori No. 1 nuclear power plant in Korea was started to operate in 1978, currently 24 nuclear power plants have been being operated, out of which 21 plants are PWR types and the rest are CANDU types. About 30% of total electricity consumed in Korea is from all these nuclear power plants. The accumulated spent nuclear fuels (SNFs) generated from each site are temporarily being stored as wet or dry storage type at each plant site. These SNFs with their high radiotoxicity, heat generating, and long-lived radioactivity are currently the only type of high-level radioactive waste (HLW) in Korea, which urgently requires to be disposed of in deep geological repository. Studies on disposal of HLW in various kind of geological repositories have been carried out in such countries as Sweden, Finland, United States, and etc. with their own management policies in consideration of their situations. In Korea long-term R&D research program for safe management of SNF has also been conducted during last couple of decades since around 1997, during which several various type of disposal concepts for disposal of SNFs in deep geological formations have been investigated and developed. The first concept developed was KAERI Reference Disposal System (KRS) which is actually very much similar to Swedish KBS-3, a famous concept of direct disposal of SNF in stable crystalline rock at a depth of around 500 m which has been regarded as one of the most plausible method worldwide to direct disposal of SNF. The world first Finnish repository will be also this type. Since the characteristics of SNF discharged from domestic nuclear reactors have been changed and improved, and burnup has sometimes increased, a more advanced deep geological repository system has been needed, KRS-HB (KRS with High Burnup SNF) has been developed and in consideration of the dimensions of SNFs and the cooling period at the time point of the disposal time, KRS+, a rather improved disposal concept has also been subsequently developed which is especially focused on the efficient disposal area. Recently research has concentrated on rather advanced disposal technology focused on a safer and more economical repository system in recent view of the rapidly growing amount of accumulated SNF. Especially in Korea the rock mass and the footprint area for the repository extremely limited for disposal site. Some preliminary studies to achieve rather higher efficiency repository concept for disposal of SNF recently have already been emphasized. Among many possible ones for consideration of design for high-efficiency repository system, a double-layered system has been focused which is expected to maximize disposal capacity within the minimum footprint disposal area. Based on such disposal strategy a rather newly designed performance assessment methodology might be required to show long-term safety of the repository. Through the study some prerequisites for such methodological development will be roughly checked and investigated, which covers FEP identification and pathway and scenario analyses as well as preliminary conceptual modeling for the nuclide release and transport in near-field, far-field, and even biosphere in and around the conceptual repository system.