• 콘크리트학회논문집
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• 제24권5호
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• pp.567-575
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• 2012

이 연구는 실험과 병행 화재에 노출된 철근콘크리트 구조물의 갤러킨 유한요소해석 방법을 제시하였다. 이 방법은 비선형 비정상 온도분포해석에 관한 것으로 2차원 삼각형 요소에 대한 해석기법을 구축하였다. 해석기법의 검증을 위하여 실규모 철근콘크리트 슬래브에 대한 내화실험을 실시하였으며, 실험 결과와의 비교를 통해 해석기법의 유효성을 확인하였다. 또한 콘크리트 부재의 내화성능에 대한 실험 결과를 분석하였다. 변수분석에서는 화재규모, 콘크리트의 온도의존성 열적특성값, 콘크리트의 함수율이 콘크리트의 내화성능에 미치는 영향을 평가하였다. 이 연구에서 구축된 수치해석모델은 다양한 화재규모와 대류, 복사 경계조건, 재료의 온도의존성 열적특성값을 자유롭게 고려할 수 있다. 또한 이 논문에서는 콘크리트 슬래브를 대상으로 표준화재곡선을 대상으로만 분석하였지만 관련된 철근콘크리트 기둥 골조 해석에 용이하게 사용될 수 있을 것으로 판단되었다.

• 한국전산유체공학회지
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• 제21권4호
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• pp.1-10
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• 2016

The analysis of characteristics of turbulent flow and thermal boundary layer for natural convection caused by fire along vertical wall is performed. The 4m-high vertical copper plate is heated and kept at a uniform surface temperature of $60^{\circ}C$ and the surrounding fluid (air) is kept at $16.5^{\circ}C$. The flow and temperature is solved by large eddy simulation(LES) of FDS code(Ver.6), in which the viscous-sublayer flow is calculated by Werner-Wengle wall function. The whole analyzed domain is assumed as turbulent region to apply wall function even through the laminar flow is transient to the turbulent flow between $10^9$<$Gr_z$<$10^{10}$ in experiments. The various grids from $7{\times}7{\times}128$ to $18{\times}18{\times}128$ are applied to investigate the sensitivity of wall function to $x^+$ value in LES simulation. The mean velocity and temperature profiles in the turbulent boundary layer are compared with experimental data by Tsuji & Nagano and the results from other LES simulation in which the viscous-sublayer flow is directly solved with many grids. The relationship between heat transfer rate($Nu_z$) and $Gr_zPr$ is investigated and calculated heat transfer rates are compared with theoretical equation and experimental data.

• Nuclear Engineering and Technology
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• 제52권8호
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• pp.1871-1880
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• 2020

Round robin analyses for vessel failure probabilities due to PTS events are proposed for plant-specific analyses of all types of reactors developed in Korea. Four organizations, that are responsible for regulation, operation, research and design of the nuclear power plant in Korea, participated in the round robin analysis. The vessel failure probabilities from the probabilistic fracture mechanics analyses are calculated to assure the structural integrity of the reactor pressure vessel during transients that are expected to initiate PTS events. The failure probabilities due to various parameters are compared with each other. All results are obtained based on several assumptions about material properties, flaw distribution data, and transient data such as pressure, temperature, and heat transfer coefficient. The realistic input data can be used to obtain more realistic failure probabilities. The various results presented in this study will be helpful not only for benchmark calculations, result comparisons, and verification of PFM codes developed but also as a contribution to knowledge management for the future generation.

• 대한기계학회논문집B
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• 제37권1호
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• pp.35-41
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• 2013

금형온도는 사출성형시 수지의 유동특성이나 성형품의 변형에 영향을 미치는 중요한 변수로서, 고온의 수지 주입과 냉각회로에 주입되는 냉각수의 영향을 받아 사출 사이클이 반복될수록 온도의 상승과 하강이 반복되는 주기적인 변화특성을 가지고 있다. 본 연구에서는 금형 냉각회로에 저온과 고온의 유체를 번갈아 주입하는 가변 금형온도 제어기법을 적용하여 성형전에는 금형온도를 높게 유지하고 성형후에는 낮게 유지함으로써 사출성형시 품질과 생산성을 동시에 높일 수 있는 연구를 수행하였다. 특히 열전달-유동해석을 연계한 다중사이클 사출성형 과도해석을 수행하여 수지와 금형, 냉각수간의 과도적인 온도변화를 수치적으로 고찰하였고, 기존 냉각방법과의 해석결과를 비교하여 제안된 가변 금형온도 제어기법의 가열 및 냉각과정에서의 효율성을 비교하였다.

• Transactions on Electrical and Electronic Materials
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• 제8권2호
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• pp.53-57
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• 2007

In this study, a novel slurry containing ceria as the abrasive particles was analyzed in terms of its frictional, thermal and kinetic attributes for interlayer dielectric (ILD) CMP application. The novel slurry was used to polish 200-mm blanket ILD wafers on an $IC1000_{TM}$ K-groove pad with in-situ conditioning. Polishing pressures ranged from 1 to 5 PSI and the sliding velocity ranged from 0.5 to 1.5 m/s. Shear force and pad temperature were measured in real time during the polishing process. The frictional analysis indicated that boundary lubrication was the dominant tribological mechanism. The measured average pad leading edge temperature increased from 26.4 to $38.4\;^{\circ}C$ with the increase in polishing power. The ILD removal rate also increased with the polishing power, ranging from 400 to 4000 A/min. The ILD removal rate deviated from Prestonian behavior at the highest $p{\times}V$ polishing condition and exhibited a strong correlation with the measured average pad leading edge temperature. A modified two-step Langmuir-Hinshelwood kinetic model was used to simulate the ILD removal rate. In this model, transient flash heating temperature is assumed to dominate the chemical reaction temperature. The model successfully captured the variable removal rate behavior at the highest $p{\times}V$ polishing condition and indicates that the polishing process was mechanical limited in the low $p{\times}V$ polishing region and became chemically and mechanically balanced with increasing polishing power.

• Nuclear Engineering and Technology
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• 제45권5호
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• pp.613-624
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• 2013

Multi-dimensional two-phase phenomena occur in many industrial applications, particularly in a nuclear reactor during steady operation or a transient period. Appropriate modeling of complicated behavior induced by a multi-dimensional flow is important for the reactor safety analysis results. SPACE, a safety analysis code for thermal hydraulic systems which is currently being developed, was designed to have the capacity of multi-dimensional two-phase thermo-dynamic phenomena induced in the various phases of a nuclear system. To validate the performance of SPACE, a two-dimensional two-phase flow test was performed with slab geometry of the test section having a scale of $1.43m{\times}1.43m{\times}0.11m$. The test section has three inlet and three outlet nozzles on the bottom and top gap walls, respectively, and two outlet nozzles installed directly on the surface of the slab. Various kinds of two-dimensional air/water flows were simulated by selecting combinations of the inlet and outlet nozzles. In this study, two-dimensional two-phase void fraction profiles were quantified by measuring the local gap impedance at 225 points. The flow conditions cover various flow regimes by controlling the flow rate at the inlet boundary. For each selected inlet and outlet nozzle combination, the water flow rate ranged from 2 to 20 kg/s, and the air flow rate ranged from 2.0 to 20 g/s, which corresponds to 0.4 to 4 m/s and 0.2 to 2.3 m/s of the superficial liquid and gas velocities based on the inlet port area, respectively.

• 한국압력기기공학회 논문집
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• 제17권1호
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• pp.18-27
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• 2021

This study aims to develop an improved evaluation technology for assessing CANDU-6 safety. For this purpose, the multiple steam generator tube rupture (mSGTR) followed by an unmitigated station blackout (SBO) in a CANDU-6 plant was selected as a hypothetical event scenario and the analysis model to evaluate the plant responses was envisioned into the MARS-KS input model. The model includes logic models for controlling the pressure and inventory of the primary heat transport system (PHTS) decreasing due to the u-tubes' rupture, as well as the main features of PHTS with a simplified model for the horizontal fuel channels, the secondary heat transport system including the shell side of steam generators, feedwater and main steam line, and moderator system. A steady state condition was successfully achieved to confirm the stable convergence of the key parameters. Until the turbine trip, the fuel channels were adequately cooled by forced circulation of coolant and supply of main feedwater. However, due to the continuous reduction of PHTS pressure and inventory, the reactor and turbine were shut down and the thermal-hydraulic behaviors between intact and broken loops got asymmetric. Furthermore, as the conditions of low-flow coolant and high void fraction in the broken loop persisted, leading to degradation of decay heat removal, it was evaluated that the peak cladding temperature (PCT) exceeded the limit criteria for ensuring nuclear fuel integrity. This study is expected to provide the technical bases to the accident management strategy for transient conditions with multiple events.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.3017-3029
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• 2023

In the late in-vessel phase of a nuclear reactor severe accident, the internal heat transfer and crust evolution during the debris bed melting process have important effects on the thermal load distribution along the vessel wall, and further affect the reactor pressure vessel (RPV) failure mode and the state of melt during leakage. This study coupled the phase change model and large eddy simulation to investigate the variations of the temperature, melt liquid fraction, crust and heat flux distributions during the debris bed melting process in the hypothetical severe accident of HPR1000. The results indicated that the heat flow towards the vessel wall and upper surface were similar at the beginning stage of debris melting, but the upward heat flow increased significantly as the development of the molten pool. The maximum heat flux towards the vessel wall reached 0.4 MW/m². The thickness of lower crust decreased as the debris melting. It was much thicker at the bottom region with the azimuthal angle below 20° and decreased rapidly at the azimuthal angle around 20-50°. The maximum and minimum thicknesses were 2 and 90 mm, respectively. By contrast, the distribution of upper crust was uniform and reached stable state much earlier than the lower crust, with the thickness of about 10 mm. Moreover, the sensitivity analysis of initial condition indicated that as the decrease of time interval from reactor scram to debris bed dried-out, the maximum debris temperature and melt fraction became larger, the lower crust thickness became thinner, but the upper crust had no significant change. The sensitivity analysis of in-vessel retention (IVR) strategies indicated that the passive and active external reactor vessel cooling (ERVC) had little effect on the internal heat transfer and crust evolution. In the case not considering the internal reactor vessel cooling (IRVC), the upper crust was not obvious.

• 한국재료학회지
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• 제34권2호
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• pp.111-115
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• 2024

Silicon carbide (SiC) has emerged as a promising material for next-generation power semiconductor materials, due to its high thermal conductivity and high critical electric field (~3 MV/cm) with a wide bandgap of 3.3 eV. This permits SiC devices to operate at lower on-resistance and higher breakdown voltage. However, to improve device performance, advanced research is still needed to reduce point defects in the SiC epitaxial layer. This work investigated the electrical characteristics and defect properties using DLTS analysis. Four deep level defects generated by the implantation process and during epitaxial layer growth were detected. Trap parameters such as energy level, capture-cross section, trap density were obtained from an Arrhenius plot. To investigate the impact of defects on the device, a 2D TCAD simulation was conducted using the same device structure, and the extracted defect parameters were added to confirm electrical characteristics. The degradation of device performance such as an increase in on-resistance by adding trap parameters was confirmed.

• 생물환경조절학회지
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• 제18권4호
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• pp.341-347
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• 2009

본 연구는 수출과채류 시설원예 재배농가의 에너지 절감을 위한 보온자재의 개발을 목표로 수행되었으며 본 보에서는 시설 온실에 사용되어지는 단일 보온자재와 조합형 다겹보온자재에 대해 보온 특성을 조사하기 위하여 실험과 수치해석을 수행하였다. 실험해석의 경우, 실험모듈을 통해 내부 열원의 보온 효과를 조사하기 위해 내 외부의 온도는 K형 열전대와 데이터 획득 장치로 측정하였고, 측정된 온도를 통해 보온자재의 보온특성을 구명하였다. 수치해석은 상용코드인 CFX-11을 이용하였고 다겹보온자재의 내부 공기층은 고려하지 않고 해석하였으며 해석에서 필요한 다겹보온자재의 물성치인 열전도도는 과도 열선법에 의해 측정되는 QTM-500을 사용하였다. 실험 결과, 조합형 보온자재가 단겹 보온자재에 비해 약 45~55%까지 보온율이 높았고, 조합조건에 따라 보온 효과가 달라지기 때문에 보온성이 우수한 조합 조건을 얻을 수 있을 것으로 예상된다. 수치해석 결과와 실험 결과와의 보온율을 비교해 보면 수치해석의 결과가 실험 결과의 보온율에 비해 다소 저하하는 경향을 나타내었는데 이는 다겹보온자재 내부의 공기층을 무시하여 나타난 오차로서 향후 보온자재 내부의 공기층을 고려할 경우 보다 정확한 수치해석 결과를 얻을 수 있을 것이다.