• Journal of the KSME
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• v.24 no.2
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• pp.92-98
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• 1984

열전달영역은 임계열류속점(CHF)을 기준으로 pre-CHF 영역과 post-CHF 영역의 두가지로 대 별된다. Post-CHF 영역에 해당하는 열전달에는 천이비등열전달과 막비등열전달이 있으며 천 이비등은 CHF점과 최소 막비등점 사이에서 일어나는 현상으로 핵비등과 막비등이 조합된 열 전달 기구에 해당하고 막비등은 가열표면이 안정된 증기막에 의해 덮여 있는 상태의 열전달 기 구에 속한다. 전고에서는 pre-CHF와 CHF 열전달 영역의 특성을 살펴보았고 본고에서는 천이 비등과 최소막비등온도 및 막비등에서의 열전달 상관식의 특성을 살펴 보고자 한다.

• Journal of the KSME
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• v.23 no.6
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• pp.419-426
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• 1983

본 고에서는 첫째, Pre-CHF영역에서의 열전달 및 비등급기에 여러 상관식이 제안되었으나 열수력현상의 복잡성 때문에 완전한 이론해는 구해져 있지 않다. 둘째, 각 상관식은 그 예측에 있어서 정량적으로 많은 차이가 있다. 셋째, 상관식의 적용범위가 한정되어 있기 때문에 사용 상관식의 선택을 조건에 따라 신중히 하여야 한다.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.41 no.1
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• pp.1-8
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• 2017

A new empirical correlation was developed for application to the tandem tubes for saturated water at atmospheric pressure. The correlation was obtained by using experimental data and the least square method to calculate the bundle effect. A statistical analysis was performed to identify the suitability of the correlation. The correlation predicted the experimental data within ${\pm}8%$. The applicable ranges of the correlation correspond to a tube pitch of 28.5~114 mm, an elevation angle of $0^{\circ}{\sim}90^{\circ}$, an inclination angle of $0^{\circ}{\sim}90^{\circ}$, and heat fluxes of $0{\sim}120kW/m^2$ of the lower and upper tubes.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.8
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• pp.527-533
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• 2016

A new empirical correlation was developed to identify the effect of an inclination angle on pool boiling heat transfer coefficient of a tube submerged in the saturated water at atmospheric pressure. Through the experiments and the survey of published results 431 data points were obtained and the nonlinear least square method was used as a regression technique. The heat flux of the tube($0{\sim}120kW/m^2$), inclination angle($0^{\circ}{\sim}90^{\circ}$), and the length divided by the diameter of a tube(18~42.52) were selected as major parameters. The newly developed correlation well predicts the experimental data within ${\pm}18%$, with some exceptions.

• The Magazine of the Society of Air-Conditioning and Refrigerating Engineers of Korea
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• v.45 no.12
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• pp.88-92
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• 2016

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.150.2-150
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• 1999

• Journal of Advanced Marine Engineering and Technology
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• v.35 no.6
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• pp.750-756
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• 2011

An experimental study was carried out to measure the heat transfer coefficient in flow boiling to deionized water in a microchannel having a hydraulic diameter of $500{\mu}m$. Tests were performed in the ranges of heat fluxes from 100 to 400 kW/$m^2$, vapor qualities from 0 to 0.2 and mass fluxes of 200, 400 and 600 kg/$m^2s$. From the experimental results, it is found that the measured heat transfer coefficients is independent of mass flux and quality, and is somewhat dependent of heat flux. Measured data of heat transfer are compared to a few available correlations proposed for macroscale. The previous correlations for macroscale overpredicted the flow boiling heat transfer coefficient for the test conditions considered in this work.

• Journal of Advanced Marine Engineering and Technology
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• v.39 no.7
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• pp.702-708
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• 2015

Two-phase flow boiling experiments were conducted using FC-72 as the working fluid. The micro-channels consisted of 15 channels with a depth of 0.2 mm, width of 0.45 mm, and length of 60 mm. Tests were performed over a mass flux range of $200-400kg/m^2s$, heat flux range of $5.6-49.0kW/m^2$, and vapor quality range of 0.02-0.93. Based on the results of the experiment, the heat transfer mechanism by nucleate boiling was dominant at a lower vapor quality (x<0.2), whereas that in the region of a vapor quality greater than 0.2 was complexly influenced by nucleate boiling and forced convection boiling. The nucleate boiling and forced convection boiling could be expressed as functions of the boiling number and convection number, respectively. In addition, the heat transfer coefficient obtained by the experiment was compared with the heat transfer coefficient by the existing correlation.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.212-212
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• 1999

• Journal of Advanced Marine Engineering and Technology
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• v.40 no.4
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• pp.275-281
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• 2016

Subcooled boiling under low pressure was numerically investigated using computational fluid dynamics(CFD). The wall boiling model was used for simulating the subcooled boiling; this model requires sub-models consisting of bubble departure diameter, nucleation site density and bubble departure frequency. The CFD code CFX provides the default models based on experimental data. Because these models are mostly developed under high pressure conditions, it would not be predicted well in low pressure conditions. Thus in this study, CFD validation for subcooled boiling under low pressure was analyzed. The numerical results were compared with experimental data from published paper. Simulations were performed with mass flux ranging from 250 to $750kg/m^2s$, heat flux ranging from 0.37 to $0.77MW/m^2$ and constant outlet pressure of 0.11 MPa. Employing the empirical correlation developed under low pressures could increase the accuracy of numerical analysis.