• Proceedings of the KIPE Conference
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• 2015.11a
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• pp.205-206
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• 2015

케이블 및 버스바는 전원을 공급하는 부대 장치로 산업 전반에 걸쳐 널리 사용되고 있다. 대부분의 경우 일정 주기로 점검 및 교체 등의 유지 보수를 진행하나 ITER IVC 버스바 및 지중케이블 등의 특수한 경우에서는 접근성 및 작업성의 이유로 유지 보수를 진행 할 수 없다. 이러한 경우, 절연물 내부에 전기 트리가 발생했을 때에는 대책이 없다. 전기 트리는 부분적인 고전계에 의한 진성파괴, 전하의 주입 또는 추출에 의한 파괴 및 전기-기계적 응력(맥스웰 응력)에 의한 파괴, 미소 부분 방전에 의한 파괴의 가장 큰 고장요인이 되는 현상이며, 다양한 사고 사례가 알려져 있다. 본 연구에서는 유한요소해석법을 이용하여 전기 트리를 가지는 사각 버스바에 대한 전계해석을 진행하였고, 전기 트리를 방지 할 수 있는 사각 버스바의 절연 설계에 대한 방안을 제시한다. 이 설계 방안은 향후 전기 트리를 고려한 사각 버스바의 절연 설계 기초 데이터로 활용 될 것이다.

• Structural Engineering and Mechanics
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• v.47 no.1
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• pp.75-98
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• 2013

This study focuses on shear strengthening performance of simply supported reinforced concrete (RC) T-beams bonded by glass fibre reinforced polymer (GFRP) strips in different configuration, orientations and transverse steel reinforcement in different spacing. Eighteen RC T-beams of 2.5 m span are tested. Nine beams are used as control beam. The stirrups are provided in three different spacing such as without stirrups and with stirrups at a spacing of 200 mm and 300 mm. Another nine beams are used as strengthened beams. GFRP strips are bonded in shear zone in U-shape and side shape with two types of orientation of the strip at $45^{\circ}$ and $90^{\circ}$ to the longitudinal axis of the beam for each type of stirrup spacing. The experimental result indicates that the beam strengthened with GFRP strips at $45^{\circ}$ orientation to the longitudinal axis of the beam are much more effective than $90^{\circ}$ orientation. Also as transverse steel increases, the effectiveness of the GFRP strips decreases.

• Proceedings of the SAREK Conference
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• 2006.06a
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• pp.596-601
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• 2006

Because of insufficiency of energy resources and pollution of environment, it is necessary to develop alternative energy sources. Nuclear fission energy is used widely for source of electric Power but being restricted due to radioactivity problem. Nuclear fission is highlighted as the new generation of nuclear energy and researched worldwide because of low risk of radiation effect. The representatives of fusion research is China's EAST, KSTAR of Korea and ITER of world. Korea Superconducting Tokamak Advanced Research(KSTAR) project is on progress for the completion in August, 2007. In this study, the research of utility system for KSTAR be carried out. The utility system of KSTAR is consist of water cooling & heating system, $N_2$ gas system, DI water system, service water system and instrument air & auto control system. The progress of KSTAR utility system is under commissioning state after construction completion. The optimal operation scenario will be verified during commissioning and adopted to the KSTAR operation.

• Nuclear Engineering and Technology
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• v.40 no.1
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• pp.87-92
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• 2008

Tokamak reactor system analysis code was developed at KAERI (Korea Atomic Energy Research Institute) and is used here for the conceptual development of a DEMO reactor. In the system analysis code, prospects of the development of plasma physics and the relevant technology are included in a simple mathematical model, i.e., the overall plant power balance equation and the plasma power balance equation. This system analysis code provides satisfactory results for developing the concept of a DEMO reactor and for identifying the necessary R&D areas, both in the physics and technology areas for the realization of the concept. With this system analysis code, the performance of a DEMO reactor with a limited extension of the plasma physics and technology adopted in the ITER design. The main requirements for the DEMO reactor were selected as: 1) demonstrate tritium self-sufficiency, 2) generate net electricity, and 3) achieve a steady-state operation. It was shown that to access an operational region for higher performance, the main restrictions are presented by the divertor heat load and the steady-state operation requirements.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.100-108
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• 2023

In this study, a neural network model inspired by a one-dimensional convolution U-net is developed to automatically accelerate edge localized mode (ELM) detection from big diagnostic data of fusion devices and increase the detection accuracy regardless of the hyperparameter setting. This model recognizes the input signal patterns and overcomes the problems of existing detection algorithms, such as the prominence algorithm and those of differential methods with high sensitivity for the threshold and signal intensity. To train the model, 10 sets of discharge radiation data from the KSTAR are used and sliced into 11091 inputs of length 12 ms, of which 20% are used for validation. According to the receiver operating characteristic curves, our model shows a positive prediction rate and a true prediction rate of approximately 90% each, which is comparable to the best detection performance afforded by other algorithms using their optimized hyperparameters. The accurate and automatic ELM-burst detection methodology used in our model can be beneficial for determining plasma properties, such as the ELM frequency from big data measured in multiple experiments using machines from the KSTAR device and ITER. Additionally, it is applicable to feature detection in the time-series data of other engineering fields.

• Advances in concrete construction
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• v.9 no.3
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• pp.267-278
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• 2020

This paper reports on the result of an experimental investigation carried out to study the compressive strength and sorptivity properties of blended cement concrete exposed to 5% and 10% MgSO₄ solution using fly ash (FA) and silpozz. Usually in sulphate environment the minimum grade of concrete is M30 and the mix design is done for target mean strength of 39 MPa. Silpozz is manufactured by burning of agro-waste rice husk in designed furnace in between 600° to 700℃ which is one of the main agricultural residues obtained from the outer covering of rice grains during the milling process. There are four mix series taken with control mix. The control mix made 0% replacement of FA and silpozz with Ordinary Portland Cement (OPC). The first mix series made 0% FA and 10-30% replacement of silpozz with OPC. The second mix series made with 10% FA and 10-40% replacement of silpozz with OPC. The third mix series made 20% FA and 10-30% replacement of silpozz with OPC and the fourth mix series made 30% FA and 10-20% silpozz replaced with OPC. The samples (cubes) are prepared and cured in normal water and 5% and 10% MgSO₄ solution for 7, 28 and 90 days. The studied parameters are compressive strength and strength deterioration factor (SDF) for 7, 28 and 90 days. The water absorption and sorptivity tests have been done after 28 days of normal water and magnesium sulphate solution curing. The investigation reflects that the blended cement concrete incorporating FA and silpozz showing better resistance against MgSO₄ solution when compared to normal water curing (NWC) samples.

• Nuclear Engineering and Technology
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• v.52 no.8
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• pp.1784-1791
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• 2020

The 14 MeV neutron generator facility is being developed by the Institute for Plasma Research India to conduct the lab scale experiments related to Indian breeding blanket system for ITER and DEMO. It will also be utilized for material testing, shielding experiments and development of fusion diagnostics. Occupational radiation exposure control is necessary for the all kind of nuclear facilities to get the operational licensing from governing authorities and nuclear regulatory bodies. In the same way, the radiation exposure for the 14 MeV neutron generator facility at the occupational worker area and accessible zones for general workers should be under the permissible limit of AERB India. The generator is designed for the yield of 10¹² n/s. The shielding assessment has been made to estimate the radiation dose during the operational time of the neutron generator. The facility has many utilities and constraints like ventilation ducts, accessible doors, accessibility of neutron generator components and to conduct the experiments which make the shielding assessment challenging to provide proper safety for occupational workers and the general public. The neutron and gamma dose rates have been estimated using the MCNP radiation transport code and ENDF -VII nuclear data libraries. The ICRP-74 fluence to dose conversion coefficients has been used for the assessment. The annual radiation exposure has been assessed by considering 500 h per year operational time. The provision of local shield near to neutron generator has been also evaluated to reduce the annual radiation doses. The comprehensive results of radiation shielding capability of neutron generator building and local shield design have been presented in the paper along with detailed maps of radiation field.

• Journal of Hydrogen and New Energy
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• v.26 no.2
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• pp.120-126
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• 2015

Recently, the ever-increasing use of fossil fuels for rapid industrial development and population significantly caused an environment pollution and global warming such as climate change. So research and development of sustainable and eco-friendly energy have been performed. Especially the interest in nuclear fusion fuel was significantly increased from the developed countries. The system of fusion energy production was tritium separation, storage and delivery, and purification. Republic of Korea is in charge of Storage and Delivery System (SDS) in the International Thermonuclear Experimental Reactor (ITER). Welding part of the SDS bottles for storing the tritium is known to be susceptible to hydrogen embrittlement. In this study, conducted a study for the relaxation of the stability and hydrogen embrittlement of the weld area. The hydrogen heat treatment was processed through the Pressure-Composition-Temperature (PCT) device according to the time variation. Also mechanical properties such as impact test and hardness test according to using the alkaline cleaning liquid for hydrogen embrittlement relief and the fracture was observed by scanning electron microscopy (SEM) after the mechanical properties evaluation.

• Journal of Hydrogen and New Energy
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• v.26 no.2
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• pp.114-119
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• 2015

Nowdays, fossil fuels have been used as an important resource in development of industry. But it is limited and caused climate change such as pollution and global warming. So nuclear fusion research is being issued with tritium to develop eco-friendly and sustainable energy. Republic of Korea is in charge of Storage and Delivery System (SDS) in the International Thermonuclear Experimental Reactor (ITER), weld present in the SDS bottles are easily exposed to the hydrogen embrittlement of special characteristics of the hydrogen in hydrogen atmosphere, When the hydrogen embrittlement is rapidly progresses, the cracking is generated in the weld zone. Due to this cracking, the risk of leakage of tritium into the atmosphere occurs. In this study, hydrogen heat treatment was processed through the Pressure-Composition-Temperature (PCT) device according to the time variation. Also mechanical properties such as rupture strength test, three point bend test and hardness test in accordance with the respective time have been conducted and the fracture was observed by scanning electron microscopy(SEM) after the mechanical properties evaluation.

• Journal of Hydrogen and New Energy
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• v.24 no.2
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• pp.128-135
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• 2013

The nuclear fuel cycle plant is composed of various subsystems such as a fuel storage and delivery system (SDS), a tokamak exhaust processing system, a hydrogen isotope separation system, and a tritium plant analytical system. Korea is sharing in the construction of the International Thermonuclear Experimental Reactor (ITER) fuel cycle plant with the EU, Japan, and the US, and is responsible for the development and supply of the SDS. Hydrogen isotopes are the main fuel for nuclear fusion reactors. Metal hydrides offer a safe and convenient method for hydrogen isotope storage. The storage of hydrogen isotopes is carried out by absorption and desorption in a metal hydride bed. These reactions require heat removal and supply respectively. Accordingly, the rapid storage and delivery of hydrogen isotopes are enabled by a rapid cooling and heating of the metal hydride bed. In this study, we designed and manufactured a vertical-type hydrogen isotope storage bed, which is used to enhance the cooling performance. We present the experimental details of the cooling performances of the bed using various cooling parameters. We also present the modeling results to estimate the heat transport phenomena. We compared the cooling performance of the bed by testing different cooling modes, such as an isolation mode, a natural convection mode, and an outer jacket helium circulation mode. We found that helium circulation mode is the most effective which was confirmed in our model calculations. Thus we can expect a more efficient bed design by employing a forced helium circulation method for new beds.