• 대한기계학회:학술대회논문집
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• 대한기계학회 2003년도 추계학술대회
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• pp.1133-1138
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• 2003

The pressure tubes, which contain high temperature heavy water and fuel, are within the core of a CANDU nuclear reactor, and are thus subjected to high stresses, temperature gradient, and neutron flux. Further, it is well known that pressure tubes of cold-worked Zr-2.5Nb materials result in hydrogen diffusion, which create fully-hydrided regions (frequently called Blister). Thus a proper investigation of hydrogen diffusion within zirconium-alloy nuclear components, such as CANDU pressure tube and fuel channels is essential to predict the structural integrity of these components. In this respect, this paper presents numerical investigation of hydrogen diffusion to quantify the hydrogen concentration for blister growth of CANDU pressure tube. For this purpose, coupled temperature-hydrogen diffusion analyses are performed by means of two-dimensional finite element analysis. Comparison of predicted temperature field and blister with published test data shows good agreement.

• 설비공학논문집
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• 제20권2호
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• pp.87-96
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• 2008

Flow condensation heat transfer coefficients(HTCs) of R123 and R245fa are measured in a horizontal plain tube. The main test section in the experimental flow loop is made of a plain copper tube of 9.52 mm outside diameter and 530 mm length. The refrigerant is cooled by passing cold water through an annulus surrounding the test section. Tests are performed at a fixed saturation temperature of $50\;{\pm}\;0.2\;^{\circ}C$ with mass fluxes of 50, 100, $150\;kg/m^2s$ and heat flux of $7.3{\sim}7.7\;kW/m^2$. Heat transfer data are obtained in the vapor quality range of $10{\sim}90%$. Test results show that the flow condensation HTCs of R245fa are overall 7.9% higher than those of R123 at all mass fluxes. The pressure drop of R245fa is smaller than that of R123 at the same heat flux. In conclusion, R245fa is a good candidate to replace ozone depleting R123 currently used in chillers from the view point heat transfer and environmental properties.

• Nuclear Engineering and Technology
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• 제49권1호
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• pp.92-102
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• 2017

The analysis of steam generators as an interface between primary and secondary circuits in light water nuclear power plants is crucial in terms of safety and design issues. VVER-1000 nuclear power plants use horizontal steam generators which demand a detailed thermal hydraulics investigation in order to predict their behavior during normal and transient operational conditions. Two phase flow field simulation on adjacent tube bundles is important in obtaining logical numerical results. However, the complexity of the tube bundles, due to geometry and arrangement, makes it complicated. Employment of porous media is suggested to simplify numerical modeling. This study presents the use of porous media to simulate the tube bundles within a general-purpose computational fluid dynamics code. Solved governing equations are generalized phase continuity, momentum, and energy equations. Boundary conditions, as one of the main challenges in this numerical analysis, are optimized. The model has been verified and tuned by simple two-dimensional geometry. It is shown that the obtained vapor volume fraction near the cold and hot collectors predict the experimental results more accurately than in previous studies.

• 설비공학논문집
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• 제17권7호
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• pp.649-658
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• 2005

Flow condensation heat transfer coefficients (HTCs) of R22, R410, Propane (R290) were measured inside a horizontal 9 hole aluminum multi-channel flat tube. The main test section in the refrigerant loop was made of a 0.53m long multi-channel flat tube of hydraulic diameter of 1.4 mm. Refrigerant was cooled by passing cold water through an annulus surrounding the test section. Data were obtained in qualities of $0.1\~0.9$ at mass flux of $200\~400kg/m^2s$ and heat flux of $7.3\~7.7kW/m^2$ at the saturation temperature of $40^{\circ}C$. All popular heat transfer correlations in single-phase subcooled liquid flow and flow condensation originally developed for large single tubes predicted the present data of the multi channel flat tube within $25\%$ deviation when effective heat transfer area was used in determining experimental data. This suggests that there is little change in flow characteristics and patterns when the tube diameter is reduced down to 1.4 mm diameter range. Hence, a modified correlation based on the present data was proposed which could be applied to small diameter tubes with effective heat transfer area. The correlation showed a mean deviation of less than $20\%$ for all data.

• 대한설비공학회:학술대회논문집
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• 대한설비공학회 2005년도 동계학술발표대회 논문집
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• pp.249-255
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• 2005

Flow condensation heat transfer coefficients(HTCs) of R22, R4IO, Propane(R290) were measured inside a horizontal 9 hole aluminum multi-channel flat tube. The main test section in the refrigerant loop was made of a 0.53 m long multi-channel flat tube of hydraulic diameter of 1.4 mm. Refrigerant was cooled by passing cold water through an annulus surrounding the test section. Data were obtained in qualities of 0.1 ${\sim}$ 0.9 at mass flux of $200{\sim}400$ $kg/m^2s$ and heat flux of $7.3{\sim}7.7$ $kW/m^2$ at the saturation temperature of $4^{\circ}C$. All popular heat transfer correlations in single-phase subcooled liquid flow and flow condensation originally developed for large single tubes predicted the present data of the multi channel flat tube within 25% deviation when effective heat transfer area was used in determining experimental data. This suggests that there is little change in flow characteristics and patterns when the tube diameter is reduced down to 1.4 mm diameter range. Hence, a modified correlation based on the present data was proposed which could be applied to small diameter tubes with effective heat transfer area. The correlation showed a mean deviation of less than 20% for all data.

• 교육시설 논문지
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• 제18권3호
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• pp.35-41
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• 2011

Performance of a small-capacity solar absorption cooling system was investigated experimentally. Ten sets of evacuative-tube solar-heat collectors and a 5 kW single-stage absorption cooler were combined to produce a hybrid cooling system. The performance of the cooling system was measured using a tim-coil unit installed in a small plastic storage. It was found from the test on a sunny day of May that when the temperature of the hot water supplied from the solar collectors to the generator of the absorption cooler reached $60^{\circ}C$, the absorption cooler started cooling and the cold water temperature measured from the fan-coil unit reached $18^{\circ}C$. The COP, which is defined as the ratio of the cooling power to the total electrical power input was higher than 1.0.

• Nuclear Engineering and Technology
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• 제51권3호
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• pp.723-730
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• 2019

Acquiring instrumentation signals generated from nuclear power plants (NPPs) is essential to maintain nuclear reactor integrity or to mitigate an abnormal state under normal operating conditions or severe accident circumstances. However, various safety-critical instrumentation signals from NPPs cannot be accurately measured on account of instrument degradation or failure under severe accident circumstances. Reactor vessel (RV) water level, which is an accident monitoring variable directly related to reactor cooling and prevention of core exposure, was predicted by applying a few signals to deep neural networks (DNNs) during severe accidents in NPPs. Signal data were obtained by simulating the postulated loss-of-coolant accidents at hot- and cold-legs, and steam generator tube rupture using modular accident analysis program code as actual NPP accidents rarely happen. To optimize the DNN model for RV water level prediction, a genetic algorithm was used to select the numbers of hidden layers and nodes. The proposed DNN model had a small root mean square error for RV water level prediction, and performed better than the cascaded fuzzy neural network model of the previous study. Consequently, the DNN model is considered to perform well enough to provide supporting information on the RV water level to operators.

• Nuclear Engineering and Technology
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• 제26권2호
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• pp.257-264
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• 1994

원자력발전소 중대사고시 용융된 노심과 잔류냉각수가 증기폭발을 일으켜 원자로 격납용기의 건전성을 위협할 수 있다. 본 연구에서는 증기폭발을 모사할 수 있는 실험 장치를 제작하고, 물과 프레온을 사용하여 증기폭발실험을 수행하였다. 이때 고속카메라를 사용하여 폭발현상을 관측하였고, 동압측정기와 압력분출관을 이용하여 생성되는 폭발압력과 기계적인 에너지를 계측하였다. 이를 토대로 증기폭발의 중요인자들(물의 온도, 물의 주입속도, 물의 주입 시간, 그리고 냉매의 깊이)에 대한 민감도 분석을 수행하였다. 그리고, 압력용기 바닥의 구조물이 용융/냉각재의 반응에 미치는 영향을 살펴보기위하여 실험용기 내부에 그리드를 설치하여 폭발실험을 실시하였다. 물/프레온의 폭발실험에서 계측된 기계적에너지를 이용한 에너지효율은 0.5∼l.6%인 것으로 계산되었다.

• Journal of Advanced Marine Engineering and Technology
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• 제40권3호
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• pp.152-156
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• 2016

폐수 열원 히트펌프 시스템은 목욕탕, 공장 또는 하천수 등을 히트펌프의 열원으로 사용한다. 증발기는 폐수로부터의 오염을 완화시키기 위하여 휜이 없는 나관식(bare tube) 열교환기가 주로 사용되고 있다. 대부분의 열 저항은 관외를 흐르는 폐수측에 생성되는 파울링에 의한 것이며, 히트펌프의 성능을 급격하게 감소시킨다. 따라서 본 연구에서는 나관식 열교환기 하부에서 버블 유동층을 형성시키고 세정 볼을 사용하여 관 외측의 파울링을 저감시키는 장치를 개발하였다. 실험 조건으로, 냉수온도 $20^{\circ}C$, 폐수 온도 $40^{\circ}C$, 관 내측 유량 100L/h, 수조 내 물의 양 50L, 열교환기 면적 $0.161m^2$ 이다. 실험결과 세정 볼 없이 버블유동층만 형성한 경우에는 버블유동층이 없는 경우에 비해 파울링에 의해 생성된 열 저항이 약 56% 감소되었으며, 세정 볼의 개수 밀도가 8,000(Number of $ball/m^3$) 일 때, 파울링에 의해 생성된 열 저항은 버블유동층 및 세정 볼이 없을 때와 비교하여 약 86% 감소되었다.

• 한국가스학회지
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• 제2권4호
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• pp.25-33
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• 1998

본 연구는 LNG냉열활용을 위해 초저온 열교환시스템을 제작하고 액화질소와 에틸렌-글리콜 수용액을 작동유체로 사용하여 증발 열전달 특성실험을 수행하였다. 초저온 열교환기는 2중관식 열교환기이며, 내부관 지름이 8, 15 mm이며 길이는 6m이다. wire-coil inserts를 사용하여 열전달촉진 성능평가를 수행하였다. 액화질소와 에틸렌-글리콜 수용액의 출입구 온도, 벽면온도, 유량, 압력을 측정하였고, 이를 증발 열전달계수와 누셀트수를 계산하는데 사용하였다. 열전달상관식을 누셀트수, 프란틀수와 등가레이놀즈수의 멱법칙관계로 제안하였고, 그 결과 열전달촉진관이 평활관보다 2.5 ${\~}$ 5.5배정도 열전달이 증가되었다. 이 상관식을 이용하여 LNG냉열이용을 위한 초저온 2중관 열교환기를 설계하였고, 그 결과로 제시하였다.