• Nuclear Engineering and Technology
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• v.36 no.5
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• pp.403-414
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• 2004

An experimental study of heat transfer characteristics near the critical pressure has been performed with an internally-heated vertical annular channel cooled by R-134a fluid. Two series of tests have been completed: (a) steady-state critical heat flux (CHF) tests, and (b) heat transfer tests for pressure reduction transients through the critical pressure. In the present experimental range, the steady-state CHF decreases with increase of the system pressure for fixed inlet mass flux and subcooling. The CHF falls sharply at about 3.8 MPa and shows a trend towards converging to zero as the pressure approaches the critical point of 4.059 MPa. The CHF phenomenon near the critical pressure does not lead to an abrupt temperature rise of the heated wall, because the CHF occurs at remarkably low power levels. In the pressure reduction transients, as soon as the pressure passes below the critical pressure from the supercritical pressure, the wall temperatures rise rapidly up to very high values due to the departure from nucleate boiling. The wall temperature reaches a maximum at the saturation point of the outlet temperature, and then tends to decrease gradually.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.61-71
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• 2022

Pool boiling heat transfer is widely applied in nuclear engineering fields. The influence of heating surface orientation on the pool boiling heat transfer has received extensive attention. In this study, the heating surface with different roughness was adopted to conduct pool boiling experiments at different inclination angles. Based on the boiling curves and bubble images, the effects of inclination angle on the pool boiling heat transfer and critical heat flux were analyzed. When the inclination angle was bigger than 90°, the bubble size increased with the increase of inclination angle. Both the bubble departure frequency and critical heat flux decreased as the inclination angle increased. The existing theoretical models about pool boiling heat transfer and critical heat flux were compared. From the perspective of bubble agitation model and Hot/Dry spot model, the experimental phenomena could be explained reasonably. The enlargement of bubble not only could enhance the agitation of nearby liquid but also would cause the bubble to stay longer on the heating surface. Consequently, the effect of inclination angle on the pool boiling heat transfer was not conspicuous. With the increase of inclination angle, the rewetting of heating surface became much more difficult. It has negative effect on the critical heat flux. This work provides experimental data basis for heat transfer and CHF performance of pool boiling.

• Proceedings of the KSME Conference
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• 2001.06e
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• pp.63-68
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• 2001

An experimental study was carried out to identify the various regimes of natural convective boiling and to determine the Critical Heat Flux(CHF) on a 70mm square surface which is inclined at $180^{\circ}$(upward), $90^{\circ}, \;45^{\circ}$. The heater block made of copper with cartridge heaters is submerged in a water tank with windows for visualization. As the heat flux increases from $100kW/m^2$ to $1.1MW/m^2$, the heat transfer regime migrates from the nucleate boiling to film boiling and results in a rapid heat up of the heater block. An explosive vapor generation on the heated surface, whose size and frequency are characterized by the heat flux, is visualized by using a digital camcorder with $512{\times}512$ pixel size at 30fps.

• Nuclear Engineering and Technology
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• v.31 no.2
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• pp.236-255
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• 1999

An improved mechanistic model was developed to predict a convective boiling critical heat flux (CHF) in the vertical round tubes with uniform heat fluxes. The CHF formula for subcooled and low quality boiling was derived from the local conservation equations of mass, energy and momentum, together with appropriate constitutive relations. The model is characterized by the momentum balance equation to determine the limiting transverse interchange of mass flux crossing the interface of wall bubbly layer and core by taking account of the convective shear effect due to the frictional drag on the wall-attached bubbles. Comparison between the present model predictions and experimental CHF data from several sources shows good agreement over a wide range of How conditions. The present model shows comparable prediction accuracy with the CHF look-up table of Groeneveld et al. Also the model correctly accounts for the effects of flow variables as well as geometry parameters.

• Proceedings of the KSME Conference
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• 2004.04a
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• pp.1470-1474
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• 2004

'Nanofluids' means suspension of common fluids with particles of the order of nanometers in size. The present research is an experimental study of critical heat flux (CHF) behavior in pool boiling of $water-TiO_2$ nanofluids under atmospheric pressure. CHF for pure water and $water-TiO_2$ nanofluids were respectively measured using disk-type copper block heater with 10mm diameter, and CHF of water with surfactant was also measured to consider the effect of surfactant used to disperse nanoparticle. The results show a large increase in CHF for $water-TiO_2$ nanofluids compared to pure water. After CHF occurred, heat flux in pool boiling for $water-TiO_2$ nanofluids was maintained in considerable value, but not for pure water.

• Nuclear Engineering and Technology
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• v.32 no.1
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• pp.17-33
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• 2000

A mechanistic model based on wall-attached bubble coalescence, previously developed by the authors, was extended to predict a vow high critical heat flux (CHF）in highly subcooled flow boiling, especially for high mass flux and small tube diameter conditions. In order to take into account the enhanced condensation due to high subcooling and high mass velocity in small diameter tubes, a mechanistic approach was adopted to evaluate the non-equilibrium flow quality and void fraction in the subcooled water flow boiling, with preserving the structure of the previous CHF model. Comparison of the model predictions against highly subcooled water CHF data showed relatively good agreement over a wide range of parameters. The significance of the proposed CHF model lies in its generality in applying over the entire subcooled flow boiling regime including the operating conditions of fission and fusion reactors.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
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• v.24 no.9
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• pp.685-692
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• 2012

In this work, nucleate pool boiling heat transfer coefficients(HTCs) of 5 refrigerants of different vapor pressure are measured on horizontal Thermoexcel-E square surface of 9.53 mm length. Tested refrigerants are R32, R22, R134a, R152a and R245fa. HTCs are taken from 10 $kW/m^2$ to critical heat fluxes for all refrigerant at $7^{\circ}C$. Wall and fluid temperatures are measured directly by thermocouples located underneath the test surface and in the liquid pool. Test results show that critical heat fluxes(CHFs) of Thermoexcel-E enhanced surface are greatly improved as compared to that of a plain surface in all tested refrigerants. CHFs of all refrigerants on the Thermoexcel-E surface are increased up to 100% as compared to that of the plain surface. The improvement of Thermoexcel-E surface in CHF, however, is lower than that of the low fin surface. HTCs on Thermoexcel-E surface increase with heat flux. But after certain heat flux, HTCs began to decrease due to the difficulty in bubble removal caused by the inherent complex nature of this surface. Therefore, at heat fluxes close to the critical one, sudden decrease in HTCs needs to be considered in thermal design with Thermoexcel-E surface.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2474-2479
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• 2024

In-Vessel Retention (IVR) is a key technology to retain the molten core in the reactor vessel during severe accidents of Pressurized-water reactors (PWRs). In order to gain the safety margin of IVR, it is crucial to enhance the critical heat flux (CHF) of the reactor vessel, which is submerged in a water pool. To enhance the CHF, we have designed and additive-manufactured porous grid plates with a 3-D printer for design flexibility. We measured the CHF for the porous grid plate on the boiling heat transfer surface and found that the CHF was enhanced by 50 % more than that of the bare surface. The CHF enhanced more with a narrower grid pitch and a lower grid height. The visual observation study revealed that the vapor film was formed at the bottom of the grid plate.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1068-1078
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• 2018

Experiments were performed to investigate bubble behaviors and pool boiling Critical Heat Flux (CHF) on a thin flat rectangular copper heater fabricated on Printed Circuit Board (PCB), at various inclination angles. The surface inclination angles were $0^{\circ}$, $45^{\circ}$, $90^{\circ}$, $135^{\circ}$, and $180^{\circ}$. Results showed the Onset of Nucleate Boiling (ONB) heat flux increased with increasing heater orientation from $0^{\circ}$ to $90^{\circ}$, while early ONB occurred when the heater faced downwards ($135^{\circ}$ and $180^{\circ}$). The nucleate boiling was observed to be unstable at low heat flux (1-21% of CHF) and changed into typical boiling when the heat flux was above 21% of CHF. The result shows the CHF decreased with increasing heater orientation from $0^{\circ}$ to $180^{\circ}$. In addition, the bubble departure diameter at the heater facing upwards ($0^{\circ}$, $45^{\circ}$, and $90^{\circ}$) was more prominent compared to that of the heater facing downward ($135^{\circ}$). The nucleation site density also observed increased with increasing heat flux. Moreover, the departed bubbles with larger size were observed to require a longer time to re-heat and activate new nucleation sites. These results proved that the ONB, CHF, and bubble dynamics were strongly dependent on the heater surface orientation.

• Transactions of the Korean Society of Mechanical Engineers
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• v.18 no.11
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• pp.2986-2998
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• 1994