• Proceedings of the KSME Conference
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• 2001.11b
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• pp.288-293
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• 2001

A critical heat flux (CHF) prediction method using an artificial neural network (ANN) was evaluated for application to the high-heat-flux (HHF) subcooled flow boiling. The developed ANN predictions were compared with the experimental database consisting of a total of 3069 CHF data points. Also, the prediction performance by the ANN was compared with those by mechanistic models and a look up table technique. The parameter ranges of the experimental data are: $0.33{\leq}D{\leq}37.5mm$, $0.002{\leq}L{\leq}4m$, $0.37{\leq}G{\leq}134Mg/m^2s$, $0.1{\leq}P{\leq}20MPa$, $50\leq{\Delta}h_{sub,in}\leq1660kJ/kg$, and $1.1{\leq}q_{CHF}\leq276MW/m^2$. $276MW/m^2$. It was found that 91.5% of the total data points were predicted within $a{\pm}20%$ error band, which showed the best prediction performance among the existing CHF prediction methods considered.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2771-2788
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• 2020

Experiments are conducted to study bubble flow behavior during the instability of subcooled boiling under uniform and non-uniform transverse heating. The non-uniform heat distribution introduces nonuniform bubble generation and condensation rates on the heated surface, which is different from the uniform heating. These bubble generation and condensation characteristics introduce a non-uniform local pressure distribution in the transverse direction, which creates an extra non-uniform pressure on the flowing bubbles. Therefore, different bubble flow behavior can be observed between uniform and non-uniform heating conditions. In the uniform heating, bubble velocity fluctuations are low, and the bubbles travel straight along the axial direction. In the non-uniform heating, more fluctuation in the bubble velocity occurs at low mass flow rate and high subcooled inlet temperatures, and reverse flow is observed. Additionally, the bubbles show a zigzag trajectory when they pass through the channel, which indicates the existence of cross flow in the transverse direction.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.26 no.11
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• pp.1513-1520
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• 2002

Effect of Convex Surface Curvature on the Onset of Nucleate Boiling of Subcooled Fluid Flow in Vertical Concentric Annuli An experimental study has been carried out to investigate the effect of the transverse convex surface curvature of core tubes on heat transfer in concentric annular tubes. Water is used as the working fluid. Three annuli having a different radius of the inner cores, Ri=3.18mm, 6.35mm, and 12.70mm with a fixed ratio of Ri/Ro=0.5 are used over a range of the Reynolds number between about 40,000 and 80,000. The inner cores are made of smooth stainless steel tubes and heated electrically to provide constant heat fluxes throughout the whole length of each test section. Experimental result shows that heat flux values on the onset of nucleate boiling of the smaller inner diameter model is much higher than that of the larger size test model.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.313-326
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• 2007

In this paper, an algorithm to reconstruct two orthogonal images into a three-dimensional image is developed in order to measure the bubble size and volume in a two-phase boiling flow. The central-active contour model originally proposed by P. $Szczypi\'{n}ski$ and P. Strumillo is modified to reduce the dependence on the initial reference point and to increase the contour stability. The modified model is then applied to the algorithm to extract the object boundary. This improved central contour model could be applied to obscure objects using a variable threshold value. The extracted boundaries from each image are merged into a three-dimensional image through the developed algorithm. It is shown that the object reconstructed using the developed algorithm is very similar or identical to the real object. Various values such as volume and surface area are calculated for the reconstructed images and the developed algorithm is qualitatively verified using real images from rubber clay experiments and quantitatively verified by simulation using imaginary images. Finally, the developed algorithm is applied to measure the size and volume of vapor bubbles condensing in a subcooled boiling flow.

• Proceedings of the Korean Nuclear Society Conference
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• 1998.05a
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• pp.449-456
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• 1998

A theoretical critical heat flux (CLE) model. based on lateral bubble coalescence on the heated wall, is proposed to predict the subcooled flow boiling CHF in a uniformly heated vertical tube. The model is based on the concept that a single layer of bubbles contacted to the heated wall events a bulk liquid from reaching the wall at near CHF condition. Comparisons between the model predictions and experimental data result in satisfactory agreement within less than 9.73 % root-mean-square error by the appropriate choice of the critical void fraction in the bubbly layer. The present model shows comparable performance with the CHF look-up table of Groeneveld et al.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.154-154
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• 2001

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.127-127
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• 1998

• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.98.1-98.1
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.568-569
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• 2005

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.151.2-151
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• 1999