• Proceedings of the KSME Conference
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• 2004.04a
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• pp.1933-1938
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• 2004

Nucleate boiling experiments with constant wall temperature of heating surface were performed using R113 for almost saturated pool boiling conditions. A microscale heater array and Wheatstone bridge circuits were used to maintain a constant wall temperature condition and to measure the heat flow rate with high temporal and spatial resolutions. Bubble images during the bubble growth were taken as 5000 frames a sec using a high-speed CCD camera synchronized with the heat flow rate measurements. The geometry of the bubble during growth time could be obtained from the captured bubble images. The bubble growth behavior was analyzed using the new dimensionless parameters for each growth regions to permit comparisons with previous results at the same scale. We found that the new dimensionless parameters can describe the whole growth region as initial and later respectively. The comparisons showed good agreement in the initial and thermal growth regions. The required heat flow rate for the volume change of the observed bubble was estimated to be larger than the instantaneous heat flow rate measured at the wall. Heat, which is different from the instantaneous heat supplied through the heating wall, can be estimated as being transferred through the interface between bubble and liquid even with saturated pool conditions. This phenomenon under a saturated pool condition needs to be analyzed and the data from this study can supply the good experimental data with the precise boundary condition (constant wall temperature).

• Transactions of the Korean Society of Mechanical Engineers B
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• v.25 no.9
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• pp.1237-1244
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• 2001

An experimental study was carried out to identify the various regimes of natural convective pool boiling and to determine the boiling heat transfer curve and Critical Heat Flux(CHF) on a vertical square surface having a 70mm width and a 70mm height. The heater made of copper block with embedded cartridge heaters is submerged in a water tank at atmospheric pressure. As the heat flux increases from 100kW/㎡ to 1.2MW/㎡, the heat transfer regime migrates from the nucleate boiling to the film boiling. The boiling heat transfer data are fitted by Rohsenow type correlation. An explosive vapor generation on the heated surface, whose size and frequency are characterized by the heat flux, is visualized using a high speed digital imaging system.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.23 no.11
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• pp.718-725
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• 2011

Characteristics of flow boiling heat transfer in microchannels were investigated experimentally. The microchannels consisted of 9 parallel trapezoidal channels with each channel having 205 ${\mu}m$ of bottom width, 800 ${\mu}m$ of depth, $3.6^{\circ}$ of sidewall angle, and 7 cm of length. Tests were performed with R113 over a mass velocity range of 150~920 $kg/m^2s$, heat flux of 10~100 $kW/m^2$ and inlet pressures of 105~195 kPa. Flow boiling heat transfer coefficient in microchannels was found to be dominated by heat-flux. However the effect of mass velocity was not significant. Contrary to macrochannel trends, the heat transfer coefficient was shown to decrease with increasing thermodynamic equilibrium quality. A new correlation suitable for predicting flow boiling heat transfer coefficient was developed based on the laminar single-phase heat transfer coefficient and the nucleate boiling dominant equation. Comparison with the experimental data showed good agreement.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1417-1428
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• 2020

The design of a nuclear reactor core requires basic thermal-hydraulic information concerning the heat transfer regime at which onset of nucleate boiling (ONB) will occur, the pressure drop and flow rate through the reactor core, the temperature and power distributions in the reactor core, the departure from nucleate boiling (DNB), the condition for onset of flow instability (OFI), in addition to, the critical velocity beyond which the fuel elements will collapse. These values depend on coolant velocity, fuel element geometry, inlet temperature, flow direction and water column above the top of the reactor core. Enough safety margins to ONB, DNB and OFI must-emphasized. A heat transfer package is used for calculating convection heat transfer coefficient in single phase turbulent, transition and laminar regimes. The main objective of this paper is to study the possibility of power upgrading of WWR-S research reactor from 2 to 10 MW_th. This study presents a one-dimensional mathematical model (axial direction) for steady-state thermal-hydraulic design and analysis of the upgraded WWR-S reactor in which two types of plate fuel elements are employed. FOR-CONV computer program is developed for the needs of the power upgrading of WWR-S reactor up to 10 MW_th.

• Solar Energy
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• v.1 no.1
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• pp.30-36
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• 1981

Pool Boiling heat transfer from controlled arrays of artificial nucleation sites was studied experimentally. Distilled water were boiled from artificial sites of uniform size, shape and spacing, drilled in superfinished copper horizontal surfaces at site density of 16, 25, 36, 49, 64, 81, 100 per $2.25cm^2$. The results confirm the boiling heat transfer from artificial sites can be improved by increasing the site density N/A or temperature difference ${\Delta}T$ or both. Following experimental correlation were developed for predicting the heat transfer rate from the heating surface which has artificial sites. $$q/A = C(T_s - T_{sat})^{1.811}(N/A)^{0.41}$$

• Proceedings of the KSME Conference
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• 2001.11b
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• pp.330-336
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• 2001

Water heat transfer experiments were carried out in a uniformly heated annulus with a wide range of pressure conditions. The local heat transfer coefficients for saturated water flow boiling have been measured just before the occurrence of the critical heat flux (CHF) along the length of the heated section. The trends of the measured heat transfer coefficients were quite different from the conventional understanding for the heat transfer of saturated flow boiling. This discrepancy was explained from the nucleate boiling in the liquid film of annular flow under high heat flux conditions.

• Nuclear Engineering and Technology
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• v.41 no.10
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• pp.1285-1292
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• 2009

The effects of the narrowed upside gap on nucleate pool boiling heat transfer in a vertical annulus were investigated experimentally. For the study, a stainless steel tube with a diameter of 25.4 mm and saturated water that kept an atmospheric condition were used. The ratio between the gaps measured at the upper and the lower regions of the annulus ranged from 0.18 to 1. Two different lengths of the modified gap also were investigated. The change in heat transfer due to the modified gap became evident as the gap ratio decreased and the length of the gap increased. As the gap ratio became less than 0.51, a significant decrease in heat transfer was observed compared to the plain annulus. The longer gap size resulted in an additional decrease in heat transfer. The major cause for the tendency was attributed to the formation of lumped bubbles around the upper region of the annulus followed by the increased flow friction between the fluid and the surface around the modified gap.

• Nuclear Engineering and Technology
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• v.35 no.3
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• pp.191-205
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• 2003

Effects of the width and location of a flow disturbing circular plate, installed at a vertical tube surface, on nucleate pool boiling heat transfer of water at atmospheric pressure have been investigated experimentally. Through the tests, changes in the degree of intensity of liquid agitation have been analyzed. The plate changes the fluid flow around the tube as well as heat transfer coefficients on the tube surface. It is identified that the plate width changes the rate of the circulating flow whereas its location changes the growth of the active agitating flow. Moreover, the flow chugging was observed at the downside of the plate.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.546-551
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• 1997

This paper presents a prediction of critical heat flux (CHF) in bubbly flow regime using dry-spot model proposed recently by authors for pool and flow boiling CHF and existing correlations for forced convective heat transfer coefficient, active site density and bubble departure diameter in nucleate boiling region. Without any empirical constants always present in earlier models, comparisons of the model predictions with experimental data for upward flow of water in vertical, uniformly-heated round tubes are performed and show a good agreement. The parametric trends of CHF have been explored with respect to variations in pressure, tube diameter and length, mass flux and inlet subcooling.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.32 no.5
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• pp.327-334
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• 2008

Effects of tube inclination on pool boiling heat transfer have been studied for the saturated water at atmospheric pressure. For the analysis, seven inclination angles varying from the horizontal to the vertical and two tube diameters(25.4 and 30.0 mm) are tested. According to the results, inclination angles result in much change on heat transfer. For the same wall superheat(about $5.3^{\circ}C$) the ratio between two heat fluxes for the $45^{\circ}$ inclined and the vertical has the value of more than five when the tube diameter is 25.4mm. As the inclination angle is increasing from the horizontal to the vertical direction heat transfer is gradually increasing because of the increase in liquid agitation. However the detailed tendency depends on the ratio between the tube length and the diameter.