In this study, a survey focusing on the status of clothing interest, inconveniences resulting from clothing, preferred design items, etc. was conducted on 364 elderly women to suggest aesthetically and functionally appropriate indoor wear design for at home elderly women aged 60 years or older. The survey results showed that in general, the respondents’ interest in clothing was high, and more respondents in their 70s or older had difficulty in the action of opening and closing. With respect to considerations when purchasing clothes, color was considered more important than design as respondent’s age increased, and size was regarded as the most important factor especially among those in their 80s. The preferred top styles were T-shirts and blouses among those in their 60s and 70s, and T-shirts and shirts among those in their 80s. The preferred sleeve lengths were “below the elbow” and “above the wrist” in all age groups. The preferred sleeve hem type was “tightening” in all age groups. The most preferred bottom styles were “straight-leg pants” and “elastic waistband.” This study suggests the design items of indoor wear, including top, bottom, and overgarment for warmth, appropriate for elderly women at home based on the survey results. The study results are expected to serve as basic data necessary for the revitalization of the clothing industry for elderly women.

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.

중국의 고건축물(古建築物)은 인류문명중의 걸작으로 고대인 의 출중한 지혜와 정밀한 기예 및 장인정신의 결정체이다. 이 들 건축물의 기세는 성대하며 건축양식의 풍채 또한 독특하며 정교하고 섬세하였다. 특히 중국 강남 지역의 고건축물은 원림 (園林), 수곽(水廓, 水鄕) 등과 청와(靑瓦)와 백벽(白壁)으로 이 루어진 독특한 풍경을 지니고 있다. 또한 그 청와와 백벽으로 이루어진 고건축물 중에서도 강남지역의 독특한 예술매력을 보 여주는 휘파(徽派) 건축이 가장 대표적이다. 오늘날 현대 장신구 디자인에서는 전통문화에 대한 전수와 계승, 그리고 정감 및 예술에 대한 사상의 구현을 더욱 중요시 하고 있다. 이에 본고에서는 강남지역의 청와와 백벽 건축물의 요소를 현대 장신구의 디자인에 유입함을 연구하였다. 휘파(徽 派)건축의 수묵(水墨)양식 건축은 청신하고 아담하며 고풍스러 운 느낌이다. 이는 현대 장신구의 수수하고 진중하며 개성적인 모습으로 전개되어 전통적인 장신구와의 귀족적적이며 화려하 고 고루함 등과 상대적으로 차별이 되었다. 이들 휘파(徽派) 건 축물과 현대 디자인은 예술형식의 융합이며 전통에 대한 추억 이다. 그러므로 본고는 현대 예술 디자인의 창조적인 탐색이다. 이를 바탕으로 장신구 디자인을 탐구하는 동학들에게 새로운 디자인의 창의와 폭넓은 사고에 도움이 되길 바란다.

The production process of ultra-pure water (UPW) involves dozens of unit processes such as reverse osmosis (RO), pretreatments, membrane degasifier, and several ion exchange processes. Recently, continuous electrodeionization (CEDI) has replaced the 2-bed and 3-tower (2B3T) ion exchange process. As a result, the majority of wastewater in UPW production now comes from the RO concentrate. The important of RO in UPW production is to produce high-quality water with a low ion concentration (around 1 mg/L) for CEDI feed water. Minimizing RO concentrate is essential to reduce the wastewater produced in the UPW production process. This can be achieved by maximizing the recovery of the RO system. However, increasing the recovery is limited by the water quality of the RO permeate. To ensure high-quality permeate water, the RO system is designed with a two-pass configuration. The recovery of each pass in the RO system is limited (e.g., < 85%) due to the expected increase in permeate water concentration at higher RO feed water concentrations. Interestingly, tests using 4-inch RO modules with low concentration feed water (≤ 35 mg/L as NaCl) revealed that the permeate concentration remains almost constant regardless of the feed water concentration. This implies that the recovery of the first RO pass can be increased as long as the average feed/concentrate concentration of the second RO pass is less than 35 mg/L. According to this design criterion for the RO system, the recovery of the first and second RO pass, with a feed water concentration of 250 mg/L as NaCl, can be increased up to 94.8% and 96.0%, respectively. Compared to the conventional RO system design (e.g., 70% and 80% for the first and second RO pass) for UPW production, this maximum recovery design reduces the volume of RO feed and concentrate by up to 38.4% and 89.2%, respectively.

Construction industry is considered as one of the most high-risk industries for work-related injuries and fatalities, accounting for more than half of fatalities in Korea. Advanced countries have recognized the critical role of designers in preventing construction accidents and have established regulations on design for safety. In line with this, the Korean government have also implemented regulations that require owners and designers to review the safety of design outcomes. However, it has been observed that designers face challenges in identifying hazards and integrating design solutions at the design stage mainly due to their shortage of required knowledge and skills. This study aimed to examine design solutions that can be applied to prevent fall accidents in the construction industry, and to establish a relationship between these solutions and fatal fall accidents occurred over the past three years in Korea. This study also analyzed the relationships of four variables (construction type, cost, work type, and fall location) with design solutions. The results indicated that all four variables have significant relationships with design solutions, with fall location showing the strongest relationship. The design solutions and their relationships with fatal fall accidents identified in this study can be utilized in identifying hazard and integrating design solutions to ensure design for safety.

본 연구는 고속도로용 RC 교각 기둥구조에 대하여 축방향 기존 철근을 중공철근으로 대체하는 설계방안을 제시하였 다. 동일직경 기준으로 기존 이형철근을 중공철근으로 대체할 수 있는 합리적인 설계방안을 제시하였으며, 기존 축방향 배근량 을 감소하는 방안을 제안하였다. 본 연구에서 제안한 설계방안을 검증하기 위하여 3차원 유한요소 구조해석을 수행하였으며, 압 축하중에 의한 변수 수치해석을 통하여 본 연구에서 제안한 방안의 타당성을 제시하였다. 향후 다양한 변수 수치해석 및 실물 시험을 통하여 본 연구에서 제시한 설계방안에 대한 추가 검증이 필요하다.

Reinforcement learning (RL) is widely applied to various engineering fields. Especially, RL has shown successful performance for control problems, such as vehicles, robotics, and active structural control system. However, little research on application of RL to optimal structural design has conducted to date. In this study, the possibility of application of RL to structural design of reinforced concrete (RC) beam was investigated. The example of RC beam structural design problem introduced in previous study was used for comparative study. Deep q-network (DQN) is a famous RL algorithm presenting good performance in the discrete action space and thus it was used in this study. The action of DQN agent is required to represent design variables of RC beam. However, the number of design variables of RC beam is too many to represent by the action of conventional DQN. To solve this problem, multi-agent DQN was used in this study. For more effective reinforcement learning process, DDQN (Double Q-Learning) that is an advanced version of a conventional DQN was employed. The multi-agent of DDQN was trained for optimal structural design of RC beam to satisfy American Concrete Institute (318) without any hand-labeled dataset. Five agents of DDQN provides actions for beam with, beam depth, main rebar size, number of main rebar, and shear stirrup size, respectively. Five agents of DDQN were trained for 10,000 episodes and the performance of the multi-agent of DDQN was evaluated with 100 test design cases. This study shows that the multi-agent DDQN algorithm can provide successfully structural design results of RC beam.

Korean innovative SMR has been implemented developing with improved safety/economy and i- SMR technology development project to secure a competitive edge in SMR. For nuclear power plants, according to the revision of the Nuclear Safety Act (2013.6), it is mandatory to be reflected in the aging management program of nuclear power plants, and the aging management and regulation of major nuclear power plants are being strengthened. For i-SMR, chemistry environment and management strategy is essential to mitigate corrosion and radiation fields, since it has compacted and integrated module designs. Since 1994, zinc injection into the reactor coolant system (RCS) has been applied more than 100 PWRs in the world to mitigate primary water stress corrosion cracking (PWSCC) and to reduce outof- core radiation fields. In domestic NPPs, 7 have been applying zinc injection and had up to 90% radiation field reductions. For this reason, SMR needs to apply zinc injection for chemistry strategy. Zinc target concentration will be 5~40 ppb at i-SMR, based on Ni-Fe-Cr materials as same as PWRs. Zinc injection location is in volume and purification control system between the volume control tank and charging P/P where the pressure is moderate. Zinc injection skid can consist of two micro-controllable pump (one for operation and one for stand-by) and one injection tank (batching tank for zinc solution). Zn, Ni, Si, Fe, and activated corrosion products should be monitored to identify zinc injection controls and trends. Flux mapping for core performance monitoring should be evaluated. The application of zinc will be essential and effective and bring sustainable reliability for corrosion control and mitigation strategy to meet the risk-free i-SMR development.

In this study, in relation to the demolition of the building as a research reactor, in order to establish a basic design for preparation for relocation and installation of the TRIGA Mark-II, the present conditions such as actual measurements and structural safety were investigated, as well as technologies and cases related to the relocation and installation of cultural properties. Based on this, the basic design for the relocation and installation of cultural assets was established by reviewing the disassembly and transport design of the TRIGA Mark-II and the basic plan for the relocation site. Although the structural safety of the current self-weight of the structure is judged to be reasonable, when lifting the structure, it is necessary to consider a method of lifting the foundation by reinforcing the foundation so that the tensile force can be minimized in the structure. As for the technology to be applied before TRIGA Mark-II, the technology before non-transplacement was confirmed as the most reasonable method in terms of preserving the original form, securing safety, and securing economic feasibility. Among the non-replacement technologies, the methods that can be applied before reactor 1 can be largely classified into three types. The three methods to be reviewed can be largely classified into the traditional rail movement method, the movement method using transport equipment, and the crane movement method. Each required period was calculated from the basic design results, and the modular trailer method was judged to be the most efficient. From the basic design results, the required period for each stage according to the mobile construction method was calculated. Depending on the calculation result, the modular trailer method is judged to be the most efficient. However, the final construction method should be selected according to the detailed design results. Overall, the results obtained through this study suggest that it is possible to create a memorial hall without the previous installation of TRIGA Mark-II if the structure foundation is composed independently of the building foundation after conducting a detailed characteristic investigation on the foundation of the TRIGA Mark-II structure.

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.

In order to use nuclear energy stably, high level radioactive waste including spent nuclear fuel that is inevitably discharged from nuclear power plants after electricity generation must be managed safely and isolated from the human living area for a long period of time. In consideration of the accumulated amount of spent nuclear fuel anticipated according to the national policy for HLW management, the area required for the deep geological repository facility is expected to be very large. Therefore, it is essential to conduct various studies to optimize the area required for the disposal of spent nuclear fuel in cases where the nationally available land is extremely limited, such as in Korea. In this study, as part of such research, the strategies and the requirements for the preliminary design of a high efficiency repository concept of spent nuclear fuel were established. For PWR spent nuclear fuel, seven assemblies of spent nuclear fuel can be accommodated in a disposal canister, and high burnup of spent nuclear fuel was taken into consideration, and the source terms such as the amount and time of discharge and disposal were based on the 2nd national basic plan. By evaluating the characteristics, the amount of decay heat that can be accommodated in the disposal canister was optimized through the combination of seven assemblies of spent nuclear fuel. The cooling period of the radiation source for the safety assessment of the repository system was set at 55 years, and the operation of the repository would start from 2070 and then the disposal schedule would be conducted according to the disposal scenario based on the national basic plan. With these disposal strategies described above, the main requirements for setting up the conceptual design of the high efficiency repository system to be carried out in this study were described below. • A combination of seven spent nuclear fuels with high heat and spent nuclear fuels with low heat was loaded into a disposal canister, and the thermal limit per disposal canister was 1,600 W. • In order to maintain the long-term performance of the repository, the maximum temperature design limit in the buffer material was set to 130°C. • In the deep disposal environment, the safety factor [yield strength/maximum stress] required to maintain the structural stability of the disposal canister should be maintained at 2.0 or higher so that integrity of the canister can be maintained even under long-term hydrostatic pressure and buffer swelling pressure in the deep disposal environment. • The repository should have a maximum exposure dose of 10 mSv/yr or less, which is the legal limit in case of a single event such as an earthquake, and the risk level considering natural phenomena and human intrusion, which is less than the legal limit of 10-6/yr. These strategies and requirements can be used to develop the high-efficiency geological disposal concept for spent nuclear fuels as an alternative disposal concept.

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.

An important goal of dismantling process is the disassembling of a spent nuclear fuel assembly for the subsequent extraction process. In order to design the rod extractor and cutter, the major requirements were considered, and the modularization design was carried out considering remote operation and maintenance. In order to design the rod extractor and cutter, these systems were analyzed and designed, also the concept on the rod extraction and cutting were considered by using the solid works tool. The main module consists of five sub-modules, and the function of each is as follows. The clamping module is an assembly fixing module using a cylinder so that the nuclear fuel assembly can be fixed after being placed. The Pusher module pushes the fuel rods by 2 inches out of the assembly to grip the fuel rods. The extraction module extracts the fuel rods of the nuclear fuel assembly and moves them to the consolidation module. The consolidation module collects and consolidates the extracted fuel rods before moving them to the cutting device. And the support module is a base platform on which the modules of the main device can be placed. The modules of level 2 can be disassembled or assembled freely without mutual interference. For the design of fuel rods cutter, the following main requirements were considered. The fuel rod cut section should not be deformed for subsequent processing, and the horizontally mounted fuel rods must be cut at regular intervals. The cutter should have the provision for aligning with the fuel rod, and the feeder and transport clamp should be designed to transfer the fuel rods to the cutting area. The main module consists of 6 sub-modules, and function of each is as follows. The cutting module is a device that cuts the fuel rods to the appropriate depth for notching. The impacting module is a device that impacts the fuel rods and moves them to the collection module. The transfer module is a device that moves the fuel rods to the cutting module when the aligned fuel rods enter the clamp module. The clamping module is a device to clamp the fuel rods before moving them to the cutting module. The collection module is a container where the rod-cuts are collected, and the support module is a base platform on which the modules of the main device can be placed. The module of level 3 can be disassembled or assembled after the cutting module of level 2 is installed, and the modules of level 2 can be disassembled or assembled freely without mutual interference.