• Journal of Advanced Marine Engineering and Technology
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• v.39 no.7
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• pp.696-701
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• 2015

Two-phase flow boiling experiments were conducted in 15 micro-channels with a depth of 0.2 mm, width of 0.45 mm, and length of 60 mm. FC-72 was used as the working fluid, and the mass fluxes ranged from 200 to $400kg/m^2s$. Tests were performed over a heat flux range of $5-40kW/m^2$ and vapor quality range of 0.1-0.9. The heat transfer coefficient sharply decreased at a lower heat flux and then was kept approximately constant as the heat flux is increased. Based on the measured heat transfer data, the flow pattern was simply classified into bubbly, slug, churn, and wavy/annular flows using the existing method. In addition, these classified results were compared to the transition criterion to wavy/annular regime. However, it was found that the existing transition criterion did not satisfactorily predict the transition criterion to annular regime for the present data.

• Journal of Advanced Marine Engineering and Technology
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• v.18 no.3
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• pp.26-33
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• 1994

A new reliable method to prediet the axial vapor fraction distribution from the measured probability density of the liquid bulk temperature is suggested in this paper. And also the actual quality of the subcooled boiling flow is easily calculated from the liquid bulk temperature. When the heat generating rate is reached to the CHF value, the sharp wall temperature increasing by the wall temperature fluctuation is occurred under the CHF condition. This paper presents the simple wall temperature fluctuation model of transition boiling by the repeating process of overheating and quenching, when the coalescent bubble passes slowly near the wall. Experiments for the subcooled R-113 flow are carride-out in the range of(0.9399~4.461)${\times}10^6$kg/$m^2$hr mass velocity and 10~3$0^{\circ}C$ intel subcooling condition.

• Proceedings of the KSME Conference
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• 2002.08a
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• pp.403-406
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• 2002

A number of the experiments on the phenomenon in which the thermal energy was transferred from a hot sphere to the surrounding water through the film boiling process had been conducted. As the sphere only carried the thermal energy associated with its initially high temperature but did not contain any other thermal source, the film boiling was only driven by the decreasing temperature of the sphere and, thus, was time dependent. The results from the experiments showed that the temperature of the sphere was slowly decreased in the beginning. This corresponded to the period in which the sphere was penetrating the water surface. Later, when the sphere was fully submerged and the transition film boiling was observed over the whole surface, the temperature of the sphere was decreased relatively much faster. In the last stage, the temperature of the sphere was again slowly decreased. This was considered caused by the relatively low temperature of the sphere, which reduced and later ceased the film boiling process. In addition, the estimation of the departure rate of the steam bubbles from the film layer was also correlated for the experiments.

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.155-155
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• 1999

Pool boiling critical heat flux (CHF) have been investigated using plate type test sections with different widths (3 cm & 4 cm) and lengths (10 cm, IS cm & 20 cm) under various incli- nation angles. As the inclination angle increases from $0^{\circ}$ (horizontally facing downward plate) to $30^{\circ}$, CHF sharply increases. After that angle, CHF gradually increases with the increase of the inclination angle. There must be a transition angle between $0^{\circ}$ and $30^{\circ}$, at which the CHF increase rate remarkably changes. According to the comparison of present and previous ex- periments, the transition angle may be affected by heater size and increase with the increase of heater size. The size effect of heated surface on CHF is noticeable in the L15 & L20 series and W4 series; however, it seems to be difficult to find the size effect in L10 series and W3 series.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4776-4797
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• 2022

In the one-dimensional thermal-hydraulic (TH) codes, a subcooled boiling model to predict the void fraction profiles in a vertical channel consists of wall heat flux partitioning, the vapor condensation rate, the bubbly-to-slug flow transition criterion, and drift-flux models. Model performance has been investigated in detail, and necessary refinements have been incorporated into the Safety and Performance Analysis Code (SPACE) developed by the Korean nuclear industry for the safety analysis of pressurized water reactors (PWRs). The necessary refinements to models related to pumping factor, net vapor generation (NVG), vapor condensation, and drift-flux velocity were investigated in this study. In particular, a new NVG empirical correlation was also developed using artificial neural network (ANN) techniques. Simulations of a series of subcooled flow boiling experiments at pressures ranging from 1 to 149.9 bar were performed with the refined SPACE code, and reasonable agreement with the experimental data for the void fraction in the vertical channel was obtained. From the root-mean-square (RMS) error analysis for the predicted void fraction in the subcooled boiling region, the results with the refined SPACE code produce the best predictions for the entire pressure range compared to those using the original SPACE and RELAP5 codes.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.3 no.1
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• pp.51-60
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• 1991

The aim of this study is to investigate the heat transfer characteristics in the transient cooling process of a high temperature wall. The slow transient cooling experiment was carried out with a copper block of high thermal capacity. The results of these experiments are as follows. 1. Temperature histories measured by the thermocouple, which is 0.99, 2.00, 2.99mm from the heat transfer surface showed monotonous during the cooling process. These variation are the curves of typical temperature histories in film-boiling, transition-boiling, and nucleate-boiling regions. 2. The temperature histories were measured by thermocouple installed in the copper block. The variations of the surface heat fluxes and surface temperature were computed from the numerical solution method TDMA from the measured temperature histories for radial position one dimensional heat transfer inverse problem. The boiling curves were found by the computed temperature histories. 3. The rewetting point which starts to change from film boiling to nucleate boiling is not connected with the mass velocity and it were found that the temperature of rewetting point indicated about $100^{\circ}C$. 4. The heat flux of rewetting point was about $10^5Kcal/m^2h$, at that time, the heat transfer coeficient indicated about $1000Kcal/m^2h^{\circ}C$ irrelevent to mass velocity. 5. The wall superheat decreases as the pressure increases. But I found that rewetting point appeared under higher condition in the wall temperature.

• Journal of the Korean Chemical Society
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• v.59 no.5
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• pp.454-465
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• 2015

The goal of this study is to investigate pre-service chemistry teachers’ understanding of the boiling process of a liquid mixture. We surveyed 65 students in the chemistry program of the College of Education about the boiling point of a 50%(by mole) ethanol aqueous solution and the temperature changes during heating. We then interviewed 9 of these students. According to the survey results, the percent of the pre-service teachers who thought that the boiling point of the ethanol solution would be ‘between the boiling points of ethanol and water (78-100 ℃)’ and ‘the same as that of ethanol’ were 52.3% and 35.4%, respectively. The majority of those who stated the former explained that the boiling point of the ethanol solution increased due to the effects of attraction or blocking by water molecules. Most of those who believed the latter explained that physical properties such as the boiling point would not be changed by the addition of water. With regard to the temperature change during heating, 69.2% of the teachers thought that the temperature would increase gradually while boiling, which some thought resulted from the increasing amount of water in the solution as the ethanol boiled off. Others thought that two temperature plateaus would be observed as each component of the liquid mixture underwent phase transition at its specific boiling point. When asked about the particle model of the gas phase during the boiling and evaporation process, some students drew both ethanol and water during evaporation but only ethanol when boiling. We discussed several alternative concepts of pre-service chemistry teachers about the boiling process of liquid mixtures and ways to improve their understanding.

• Journal of the Korean Society of Visualization
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• v.18 no.3
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• pp.23-34
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• 2020

Two-phase flow and heat transfer characteristics during the reflood phase of a single heated rod in the KHU reflood experimental facility were examined. Two-phase flow behavior during the reflooding experiment was carefully visualized along with transient temperature measurement at a point inside the heated rod. By numerically solving one-dimensional inverse heat conduction equation using the measured temperature data, time-resolved wall heat flux and temperature histories at the interface of the heated rod and coolant were obtained. Once water coolant was injected into the test section from the bottom to reflood the heated rod of >700℃, vast vapor bubbles and droplets were generated near the reflood front and dispersed flow film boiling consisted of continuous vapor flow and tiny liquid droplets appeared in the upper part. Following the dispersed flow film boiling, inverted annular/slug/churn flow film boiling regimes were sequentially observed and the wall temperature gradually decreased. When so-called minimum film boiling temperature reached, the stable vapor film between the heated rod and coolant was suddenly collapsed, resulting in the quenching transition from film boiling into nucleate boiling. The moving speed of the quench front measured in the present study showed a good agreement with prediction by a correlation in literature. The obtained results revealed that typical two-phase flow and heat transfer behaviors during the reflood phase of overheated fuel rods in light water nuclear reactors are well reproduced in the KHU facility. Thus, the verified reflood experimental facility can be used to explore the effects of other affecting parameters, such as CRUD, on the reflood heat transfer behaviors in practical nuclear reactors.

• Transactions of the Korean Society of Mechanical Engineers
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• v.11 no.3
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• pp.425-436
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• 1987

Accurate predictions of heat transfer coefficient of vertical laminar film-boiling are very important in many engineering applications. There are many predictions, however they are not exact as yet, since they have used the assumption of constant thermodynamic properties in the analysis. In this paper, heat transfer of vertical film boiling was analysized by Runnge Kutta method using veriable thermodynamic properties. 1/4 interval method was exployed for the prediction of unknown wall boundary condition. Numerical computations were performed with varying the wall temperature and the free stream velocity of liquid. Results show that assumption of constant thermodynamic properties induced considerable error in predicting the heat transfer coefficient, friction factor, film thickness, and critical length for transition to turbulent flow. Comparision of the predicted heat transfer coefficient of present analysis with that from Bromley's correlation shows that the use of general latent heat in Bromely equation instead of modified latent heat is more desireable since it makes the coefficient of Bromley equation into constant.

• Nuclear Engineering and Technology
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• v.54 no.3
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• pp.1085-1097
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• 2022

A printed circuit steam generator (PCSG) is being considered as the component for pressurized water reactor (PWR) type small modular reactor (SMR) that can further reduce the physical size of the system. Since a steam generator in many PWR-type SMR generates superheated steam, it is expected that dryout front oscillation can potentially cause thermal fatigue failure due to cyclic thermal stresses induced by the transition in boiling regimes between convective evaporation and film boiling. To investigate the fatigue issue of a PCSG, a reference PCSG is designed in this study first using an in-house PCSG design tool. For the stress analysis, a finite element method analysis model is developed to obtain the temperature and stress fields of the designed PCSG. Fatigue estimation is performed based on ASME Boiler and pressure vessel code to identify the major parameters influencing the fatigue life time originating from the dryout front oscillation. As a result of this study, the limit on the temperature difference between the hot side and cold side fluids is obtained. Moreover, it is found that the heat transfer coefficient of convective evaporation and film boiling regimes play an essential role in the fatigue life cycle as well as the temperature difference.