• Journal of the Korean Society for Marine Environment & Energy
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• v.18 no.4
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• pp.274-281
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• 2015

Thermal and nuclear power plants on shore commonly use the sea water for cooling facility. Discharged cooling water has the high kinematic energy potential due to amount of water flux. Numerical analysis was made to find the suitable combinations between the arrangement of tidal turbines and the overall dimensions of the discharged channel. Several parameters such as the turbine diameter to inlet size, and the axial distance to turbine size were investigated. Power coefficients for various test conditions were also compared to see the effect of inlet configurations such as single inlet and dual inlet. For the single inlet, the mean power coefficient appeared to be gradually decreased with increasing distance, and the maximum power was obtained when the turbine diameter was same as the inlet diameter. For the dual inlet, the tendency was similar so that the better result when the turbine diameter was same as the inlet diameter. It is expected that the present methodology can be extensively utilized to harness the high kinetic energy flow of the discharge channel of power plant.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.28 no.4
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• pp.361-368
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• 2015

Recently, the earthquake has been increasing a lot, damage of electric power facility has been serious as well. Nowadays, the importance of pipe support system such as Hanger, Brace, Snubber connecting the main structure have been emphasized. These devices can prevent pipe from damage so that reduce the vibration and shock acting on the pipe. For this reason, the FCD(Friction Concave Damper) was developed and has been expected to reduce the vibration on the pipe through the Friction Pendulum System. This paper was described the introduction of self-developed mechanical damper using the friction pendulum principle and the characteristic test was performed to verify the performance of the device. Additionally the test results have been compared with predicted F.A.P(FCD Analysis Program-self developed) results. As a result, reliability of design could be improved.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.35 no.5
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• pp.487-490
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• 2011

We used X-ray tomography to carry out an experimental study to visualize the effect of freeze and thaw cycles on the gas diffusion layer (GDL) in a polymer electrolyte membrane fuel cell (PEMFC). A PEMFC has freeze and thaw cycles if the fuel cell is operating at a below-freezing ambient temperature. The cycle permanently deforms the fuel-cell capillary structures and reduces the ability of the cell to generate electric power and also reduces its service life. The GDL is the thickest capillary layer in the fuel cell, so it experiences the most deformation. The X-ray tomography facility at the Pohang Accelerator Laboratory was used to observe the structural changes in GDLs induced by a freeze and thaw cycle. We discuss the effects of these structural changes on the power production and service life of PEMFCs.

• Korean Chemical Engineering Research
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• v.50 no.4
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• pp.595-603
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• 2012

International Thermonuclear Experimental Reactor (ITER) will be constructed in 2019 according to the JIA (Joint Implementation Agreement) of 7 countries. The ITER fusion fuel cycle consists of fusion vacuum vessel, tritium plant and fuelling system. The tritium plant provides the functions of storage, delivery, separation, removal and recovery of the deuterium and tritium used as fusion fuels for the ITER. The tritium plant systems supply deuterium and tritium from external sources and treat all tritiated fluids from ITER operation through Storage and Delivery System (SDS), Tokamak Exhaust Processing (TEP), Isotope Separation System (ISS), Water Detritiation System & Atmosphere Detritiation System (WDS & ADS) and Analysis System (ANS). In this paper, the functions and design requirements of the major systems in the tritium plant and the status of R&D are described. Korean party is developing the SDS for ITER tritium plant and partially attaining the WDS technology through the construction and operation experience of the Wolsong Tritium Removal Facility (WTRF). Now it is expected that researchers in other fields such as chemical engineering take part in the development of upcoming technologies for ISS and TEP.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.36 no.2
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• pp.113-120
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• 2023

This paper introduces a simulation method for the collision between roadblocks and vehicles using LS-DYNA. The need to evaluate the performance of anti-ram barriers to prepare for vehicle impact has increased since vehicle impact threats have been included as a design criterion for nuclear power plants. Anti-ram barriers are typically certified for their performance through collision experiments. However, because Koreas has no performance testing facilities for anti-ram barriers, their performance can only be verified through simulations. LS-DYNA is a specialized program for collision simulation. Various organizations, including NCAC, distributes numerical models that have been validated for their accuracy with collision tests. In this study, we constructed a finite element model of the most critical vehicle barrier module and simulated collision between roadblocks and vehicles. The calculated results were verified by applying the validation criteria for vehicle safety facility collision simulations of NCHRP 179.

• Proceedings of the Korean Institute of Information and Commucation Sciences Conference
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• 2017.10a
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• pp.702-704
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• 2017

In case of dismantling of nuclear power generation facility or radiation accident, the accurate information of gammaray source is essential for rapid decontamination. In order to more efficiently represent the position of the gamma ray to be removed, we create a spatial domain based on the real image. And we can perform decontamination of gamma-ray source more quickly by expressing the distribution of radiation source. The developed gamma ray imaging device overlaps with the visible image after gamma - ray detection and provides only two - dimensional image, but it does not show the distance information to the source. In this paper, we have developed a operation environment using the 3D visualization model for reporting effective decontamination operation.

• Journal of Convergence for Information Technology
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• v.11 no.12
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• pp.30-38
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• 2021

In this paper, we propose a method for extracting thermal data from thermal image and improving detection of heating equipment using the data. The main goal is to read the data in bytes from the thermal image file to extract the thermal data and the real image, and to apply the composite image obtained by synthesizing the image and data to the deep learning model to improve the detection accuracy of the heating facility. Data of KHNP was used for evaluation data, and Faster-RCNN is used as a learning model to compare and evaluate deep learning detection performance according to each data group. The proposed method improved on average by 0.17 compared to the existing method in average precision evaluation.As a result, this study attempted to combine national data-based thermal image data and deep learning detection to improve effective data utilization.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.19 no.2
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• pp.155-162
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• 2023

In nuclear power plants, the flow rate information is a major indicator of the performance of rotating equipment such as pumps, and is a very important one required for facility operation and maintenance. To measure a flow rate, various types of methods have been developed and used. Among them, the differential pressure type using orifice and the direct doppler type using ultrasonic waves are the most commonly used. However, these flow rate measurement methods have limitations in installation, conditions and status of the measuring part, etc. To solve this problem, we have studied a new technique for measuring flow rate from scratch. In this paper, we have devised a technique to estimate the flow rate using an average moving velocity of large-scale eddy in turbulence that occurs in the piping flow field. The velocity of the large-scale eddy can be measured using the pressure fluctuation signals on the inner surface of the pipe. To estimate the flow rate, at first a cross-correlation function is applied to the two pressure fluctuation signals located at different positions in the down stream for calculating the time delay between the moving eddies. In order to validate the proposed flow rate estimation method, CFD analyses for the internal turbulence flow in pipe are conducted with a fixed flow condition, where the pressure fluctuation signals on the pipe inner surface are simulated. And then the average flow velocity of the large scale eddy is to be estimated. The estimated flow velocity is turned out to be similar to the fixed (known) flow rate.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.2_spc
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• pp.327-336
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• 2020

The high-flux advanced neutron application reactor (HANARO) is a research reactor with thermal power of 30 MW applied in various research and development using neutrons generated from uranium fission chain reaction. A degasifier tank is installed in the ancillary facility of HANARO. This facility generates gas pollutants produced owing to internal environmental factors. The degasifier tank is designed to maintain the gas contaminants below acceptable levels and is monitored using an analyzer in the gas sampling panel. If condensate water is generated and flows into the analyzer of the gas sampling panel, corrosion occurs inside the analyzer's measurement chamber, which causes failure. Condensate water is generated because of the temperature difference between the degasifier tank and analyzer when the gas flows into the analyzer. A heating system is installed between the degasifier tank and gas sampling panel to suppress condensate water generation and effectively remove the condensate water inside the system. In this study, we investigated the efficiency of the heating system. In addition, the variations in the pipe temperature and the amount of average condensate water were modeled using a wall condensation model based on the changes in the fluid inlet temperature, outside air temperature, and heating cable-setting temperature.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.17 no.3
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• pp.339-346
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• 2019

Based on the $8^{th}$ Basic Plan for Electric Power Demand and Supply, an estimation has been made for inventories and characteristics of spent fuel (SF) to be generated from existing and planned nuclear power plants. The characteristics under consideration in this study are dimensions, fuel array, $^{235}U$ enrichment, discharge burnup, and cooling time for each fuel assembly. These are essentially needed for designing a disposal facility for SFs. It appears that the anticipated quantity by the end of 2082 is about 62,500 assemblies for PWR SFs. The inventories of Westinghouse-type and Korean-type SFs were revealed to be 60% and 40%, respectively as of the end of 2018. The proportion of SFs with initial $^{235}U$ enrichment below 4.5 weight percent (wt%) was shown to be approximately 90% in total as of the end of 2018. As of 2077, more than 97% of SFs generated from Westinghouse-type nuclear reactors were shown to have cooling time of over 50 years. As of 2125, more than 98% of SFs generated from Korean-type nuclear reactors were shown to have cooling time of over 45 years. Based on these results, for the efficient design of a disposal system, it is reasonable to adopt two types of reference spent fuel. SF of KSFA with $^{235}U$ enrichment of 4.5 wt%, discharge burnup of 55 GWd/tU, and cooling time of 50 years was determined as reference fuel for Westinghouse-type SFs; SF of PLUS7 with $^{235}U$ enrichment of 4.5 wt%, discharge burnup of 55 GWd/tU, and cooling time of 45 years was determined as reference fuel for Korean-type SFs.