• Science of Emotion and Sensibility
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• v.23 no.4
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• pp.105-116
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• 2020

The purpose of this study is to progress the limitations and disadvantages of existing safety clothing by applying high technology to current safety clothing that is produced and distributed only with fluorescent fabrics and retroreflective materials. Therefore, the industrial suspender-type safety belt and engineering technology are introduced, designed, and fabricated to help save a life in an emergency. First, the suspender-type safety belt to be developed is designed to emit light by LED attached to the film, and the body of the belt-wearer is recognized from a distance through retroreflection from the flashing LED. It aims to support people's safety by preventing accidents during roadside work, rescue activities, and sports activities at night. Second, with the development of advanced devices when the user is in an unconscious state due to distress or falls into an unconscious state due to distress or accident, the tilt sensor of the control unit attached to the belt automatically detects the angle of the human body and generates light and sound. It is intended to further enhance the utilization by mounting a sensing and signaling device that generates a distress signal and shaping it in the form of a belt attached to a vest that can be easily detached from the outside of the garment. When the wearer falls due to an accident, the tilt sensor of this belt detects the angle change and then the controller generates a high-frequency sound and repeated LED blinking signals at the same time. In the case of conventional safety vests, it is almost impossible to detect that the person is wearing a vest when there is no ambient light, but in case of the safety belts in this study, the sound and light signals of the safety belt enable us to find the wearer within 100 meters even when there is no ambient light.

• Nuclear Engineering and Technology
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• v.18 no.3
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• pp.200-208
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• 1986

The loss of coolant accident based on a double-ended cold leg break is analyzed with the discharge coefficient (Ca) of 0.4. This analysis covers the whole transient period from the start of depressurization to the complete refilling of the core by using RELAP4/MOD6-EM and RELAP4/ MOD6-HOT CHANNEL for the system thermal-hydraulics and the fuel performance during the blowdown phase respectively, and RELAP4/MOD6-FLOOD and TOODEE2 during the reflood phase. A simple analytical method has been developed to account for the lower plenum filling by approximating steam-water countercurrent flows and superheated wall effects at the downcomer during the refill period. Based on the informations. at the time of EOB (end-of-bypass), the refill duration time and the initial reflooding temperature were estimated and compared with the results from the RELAP4/MOD6, resulting in a good agreement. In addition, some parametric studies on the EOB were performed. The form loss coefficient between upper head and upper downcomer was found to be sensitive to the occurrence of the spurious EOB. Appropriate form loss coefficients should be taken into account to avoid the flow oscillations at the downcomer. The analyses with the six and three volume core nodalizations, respectively, show much similar trends in the system thermal-hydraulic performance, but the former case is recommended to obtain good results.

• Nuclear Engineering and Technology
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• v.26 no.1
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• pp.126-139
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• 1994

Best-estimate thermal-hydraulic codes, CATHARE2 V1.2 and RELAP5/MOD3, hate been assessed against the BETHSY 6.2 tc six-inch cold leg break loss-of-coolant accident (LOCA) test. Main objective is to analyze the overall capabilities of the two codes on physical phenomena of concern during the small break LOCA i.e. two-phase critical flow, depressurization, core water level de-pression, loop seal clearing, liquid holdup, etc. The calculation results show that the too codes predict well both in the occurrences and trends of major two-phase flow phenomena observed. Especially, the CATHARE2 calculations show better agreements with the experimental data. However, the two codes, in common, show some deviations in the predictions of loop seal clearing, collapsed core water level after the loop seal clearing, and accumulator injection behaviors. The discrepancies found from the comprision with the experimental data are larger in the RELAP5 results than in the CATHARE2. To analyze the deviations of the two code predictions in detail, several sensitivity calculations have been performed. In addition to the change of two-phase discharge coefficients for the break junction, fine nodalization and some corrections of the interphase drag term are made. For CATHARE2, the change of interphase drag force improves the mass distribution in the primary side. And the prediction of SG pressure is improved by the modification of boundary conditions. For RELAP5, any single input change doesn't improve the whole result and it is found that the interphase drag model has still large uncertainties.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.423-434
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• 2016

The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents.

• Analytical Science and Technology
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• v.17 no.6
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• pp.514-519
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• 2004

Soil samples collected from around the BOMARC Missile Site were measured for their activity concentrations and isotopic ratios of Pu and Am isotopes with particle sizes. The activity concentrations of Pu and Am in the BOMARC soil were remarkably higher than the fallout levels, and the activities decreased nearly exponentially with an increasing particle size of the soil due to a decreasing surface area. The activity ratios of Pu-238 / Pu-239, 240, Pu-241 / Pu-239, 240 and Am-241 / Pu-239, 240 observed in the BOMARC soil were much lower than those attributed to the nuclear reprocess plants and the Chernobyl fallout. Also, the atomic ratio of Pu-240 / Pu-239 in the BOMARC soil was remarkably lower than the fallout value influenced by the nuclear weapons testing and the Chernobyl accident. The atomic ratio of Pu-240 / Pu-239 was so close to the value of the weapons grade Pu released from the crash of a B52 plane in the Thule of the Greenland, such that the Pu isotopes detected in the BOMARC soil could have originated from the weapons grade plutonium.

• Journal of the Korean Institute of Gas
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• v.23 no.4
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• pp.1-9
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• 2019

Large-scale industrial boilers operating at high temperature and high pressure, have a large amount of water, and a large amount of energy is released at the time of explosion. Currently, most industrial boilers use gas fuel such as LNG and LPG, etc. and fuel exists in the same space as equipment, so there is a high possibility of secondary damage such as fire or explosion in the event of a boiler accident. Both special care and management are required to operate the very dangerous equipment that causes casualty 2.51 per accident. For boilers of a certain size or more, the Korea Energy Agency conducts inspections in accordance with the Energy Usage Rationalization Act, KS, and public notice of the Ministry of Industry, Trade and Resources. In this research, based on the results of the inspection, the hazard factorss are configured, and a questionnaire is conducted to the inspector, the equipment manager, the maintenance person, and the person in charge of the manufacturer. We analyzed the results by using AHP (Analytic Hierarchy Process). As a result of analysis, generally recognized hazard factorss are not good management, measurement failure, specification failure, water leak, leak analysis, but connection, welding, scale, and corrosion, etc. are relatively less important. It is judged that the adverse factors that are recognized to be highly important among all groups and careers are already well managed, but less important and adverse factors should be well managed to ensure that the safe usage of the boiler.

• Journal of Energy Engineering
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• v.27 no.4
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• pp.27-35
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• 2018

The passive containment cooling system (PCCS) has been designed to remove the released decay heat during the accident by means of the condensation heat transfer phenomenon to guarantee the safety of the nuclear power plant. The heat removal performance of the PCCS is mainly governed by the condensation heat transfer of the steam-air mixture. In this study, the heat removal performance of the PCCS was evaluated by using the MARS-KS code with a new empirical correlation for steam condensation in the presence of a noncondensable gas. A new empirical correlation implemented into the MARS-KS code was developed as a function of parameters that affect the condensation heat transfer coefficient, such as the pressure, the wall subcooling, the noncondensable gas mass fraction and the aspect ratio of the condenser tube. The empirical correlation was applied to the MARS-KS code to replace the default Colburn-Hougen model. The various thermal-hydraulic parameters during the operation of the PCCS follonwing a large-break loss-of-coolant-accident were analyzed. The transient pressure behavior inside the containment from the MARS-KS with the empirical correlation was compared with calculated with the Colburn-Hougen model.

• The Journal of the Institute of Internet, Broadcasting and Communication
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• v.23 no.2
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• pp.157-162
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• 2023

Among the failures of photovoltaic power generation facilities, failures caused by surges account for 20% of the total failure rate, and energy emissions of tens to hundreds [A] during power generation and electrical damage to inverters and connection boards lead to electrical safety accidents. In particular, in the case of lightning, an abnormal voltage is induced in an electric circuit to destroy insulation, and the current flowing at this time causes a fire and acts as a factor that accelerates the deterioration of parts. Due to this action, the problem of electrical safety of solar power generation devices spreading from outside the city center to the inside of the city center such as houses, apartments, and government offices is emerging. Since lightning strikes cause both field-based and conducted electrical interference, this effect increases with increasing cable length or conductor loops. In addition, surge damages not only solar modules, inverters and monitoring devices, but also building facilities, which can eventually cause operational shutdown due to fire of the photovoltaic power generation system and consequent financial loss. Therefore, in this paper, a lightning protection system for solar power generation devices is studied for the purpose of reducing property damage and human casualties due to the increase in fire and electrical safety accidents caused by lightning strikes in photovoltaic power generation systems.

• Proceedings of the KSME Conference
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• 2007.05b
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• pp.3491-3496
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• 2007

If a molten core is released from a reactor vessel into a reactor cavity during a severe accident, an important safety issue of coolability of the molten core from top-flooding and concrete ablation due to a molten core concrete interaction (MCCI) is still unresolved. The released molten core debris would attack the concrete wall and basemat of the reactor cavity, which will lead to inevitable concrete decompositions and possible radiological releases. In a OECD/MCCI project scheduled for 4 years from 2002. 1 to 2005. 12, a series of tests were performed to secure the data for cooling the molten core spread out at the reactor cavity and for the 2-D long-term core concrete interaction (CCI). The tests included not only separate effect tests such as a melt eruption, water ingression, and crust failure tests with a prototypic material but also 2-D CCI tests with a prototypic material under dry and flooded cavity conditions. The paper deals with the transient simulations on the CCI-2 test by using a severe accident analysis code, CORQUENCH, which was developed at Argonne National Laboratory (ANL). Similar simulations had been already per for me d by using MELCOR 1.8.5 code. Unlike the MELCOR 1.8.5, the CORQUENCH includes a melt eruption mode I and a newly developed water ingression model based on the water ingression tests under the OECD/MCCI project. In order to adjust the geometrical differences between the CCI-2 test (rectangular geometry) and the simulations (cylindrical geometry), the same scaling methodology as used in the MELCOR simulation was applied. For the direct comparison of the simulation results, the same inputs for the MELCOR simulation were used. The simulation results were compared with the previous results by using MELCOR 1.8.5.

• Journal of the Korean Society for Marine Environment & Energy
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• v.19 no.1
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• pp.52-61
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• 2016

To inject huge amount of $CO_2$ for CCS application, high pressure pipeline transport is accompanied. Rapid depressurization of $CO_2$ pipeline is required in case of transient processes such as accident and maintenance. In this study, numerical analysis on the depressurization of high pressure $CO_2$ pipeline was carried out. The prediction capability of the numerical model was evaluated by comparing the benchmark experiments. The numerical models well predicted the liquid-vapor two-phase depressurization. On the other hands, there were some limitations in predicting the temperature behavior during the supercritical, liquid phase and gaseous phase expansions.