• Proceedings of the KSME Conference
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• 2001.06d
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• pp.195-200
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• 2001

An experimental study on critical heat flux (CHF) has been performed for water flow in a uniformly heated vertical 3 by 3 rod bundle under low flow and a wide range of pressure conditions. The objective of this study is to investigate the parametric trends of CHF with 3 by 3 rod bundle test section where three unheated rods exist. The general trends of the CHF are coincident with previous understandings. At low flow and system pressure above 3 MPa, some critical qualities are larger than 1.0 due to counter-current flow in test sections. Since there is a supply of water to the heated section from unheated section, the maximum CHFs at system pressure between 2 and 4 MPa are not shown.

• Journal of the Korean Society for Geothermal and Hydrothermal Energy
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• v.16 no.2
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• pp.1-6
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• 2020

This work experimentally explored the influence of nano-fouling on CHF, flow boiling heat transfer coefficient, contact angle, and surface roughness. In this study, the flow velocity conditions are established at 0.5, 1.0, and 1.5 m/s. Also, the nanoparticles of oxidized MWCNT were deposited on a heat transfer surface for 0, 120, 180, and 240 sec. As the results, it was found that CHF and superheated temperature were increased in case of nano fouling on the heat transfer surface in oxidized MWCNT fluid. Also, the contact angle and surface roughness decreased when flow velocity and nano coating increased.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.75-81
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• 1996

COBRA-TF is a two fluid, three field subchannel code. Three fields are continuous vapor, continuous liquid and droplet. Some assessments are conducted to validate the related models and to estimate a code ability through dryout and post-CHF experiment in a tube and DNB test in rod bundles. It turned out form dryout and post-CHF experiment that the predicted dryout locations and wall temperature profiles are in close agreement with the experiments. On the other hand, DNB prediction of COBRA-TF are performed for two kinds of rod bundles along with EPRI CHF correlation. To estimate its performance COBRA-IV of homogeneous model is also run for the same data. The results say that COBRA-TF/EPRI is better in DNB prediction than COBRA-IV/EPRI. In addition the thermal-hydraulic behaviors due to the different two-phase flow models are presented at the condition of CHF.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.316-321
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• 1997

The mixing vane effect on the Critical Heat Flux (CHF) is discussed with focus on the vortex now effect. In the subchannel approach, this effect is not quantified by the calculation model, but directly taken into account by the CHF correlation itself through data analysis. The vortex now effect is identified the two Westinghouse correlations, and then the CHF margin issue given rise to by the Vantage-5H design change is evaluated and discussed. It is noted that deficiency about CHF dependency on the vortex flow effect could induce an error in the Departure from Nucleate Boiling Ratio (DNBR) sensitivity Calculation.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.05a
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• pp.292-297
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• 1997

An alternating conditional expectation (ACE) algorithm, a kind of non-parametric regression method, is proposed to generate empirical correlations automatically. The ACE algorithm yields an optimal relationship between a dependent variable and multiple independent variables without any preprocessing and initial assumption on the functional forms. This algorithm is applied to a collection of 12,879 CHF data points for forced convective boiling hi vertical tubes to develop a new critical heat flux (CHF) correlation. The meat root mean square, and maximum errors of our new correlation are -0.558%, 12.5%, and 122.6%, respectively. Our CHF correlation represents the entire set of CHF data with an overall accuracy equivalent to or better than that of three existing correlations.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.24 no.4
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• pp.485-494
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• 2000

The purpose of this paper is to investigate a critical heat flux(CHF) during forced convective subcooled and saturated boiling in free water jet system impinged on a rectangular heated surface. The surface is supplied with subcooled or saturated water through a rectangular jet. Experimental parameters studied are a width of heated surface, a height of supplementary water and a degree of subcooling. Incipient boiling point is observed in the temperature of 6${\~}8^{\circ}C$ of superheat of test specimen. CHF depends on jet velocity for various boiling-involved coolant system. CHF also is proportional to the nozzle exit velocity to the power of n, where n is 0.55 and 0.8 for subcooled and saturated boiling, respectively. CHF is enhanced with a higher jet velocity, higher degree of subcooling and smaller width of a heated surface.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05b
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• pp.349-355
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• 1996

In this study, the mechanistic critical heat flux (CHF) model and correlation for water are derived based on flow excursion (or Ledinegg instability) criterion and the simplified two-phase homogeneous model. The relationship between CHF for the water and the principal parameters such as mass flux heat of vaporization, heated length-to-diameter ratio, vapor-liquid density ratio and inlet subcooling is derived on the developed correlation. The developed CHF correlation predicts very well at the applicable ranges, 1 < P < 40 bar, 1, 300 < G 27, 00 kg/$m^2$s and inlet quality is less than -0.1. The overall mean ratio of predicted to experimental CHF value is 0.988 with standard deviation of 0.046.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.11a
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• pp.224-229
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• 1996

An experimental study on critical heat flux (CHF) and two-phase flow visualization has been performed for water flow in internally-heated, vertical, concentric annuli under near atmospheric pressure. Tests have been done under stable forced- circulation, upward and downward flow conditions with three test sections of relatively large gap widths (heated length = 0.6 m. inner diameter = 19 mm, outer diameter = 29, 35 and 51 mm). The outer wall of the test section was made up of the transparent Pyrex tube to allow the observation of flow patterns near the CHF occurrence. The CHF mechanism was changed in the order of flooding, chum-to-annular flow transition, and local dryout under a large bubble in churn flow as the flow rate was increased from zero to higher values. Observed parametric trends are consistent with the previous understanding except that the CHF for downward flow is considerably lower than that for upward 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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• 1998.05a
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• pp.443-448
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• 1998

With using artificial neural networks (ANNs), an analytical study related to the heated length effect on critical heat flux(CHF) has been carried out to make an improvement of the CHF prediction accuracy based on local condition correlations or table. It has been carried out to suggest a feasible criterion of the threshold length-to-diameter (L/D) value in which heated length could affect CHF. And within the criterion, a L/D correction factor has been developed through conventional regression. In order to validate the developed L/D correction factor, CHF experiment for various heated lengths have been carried out under low and intermediate pressure conditions. The developed threshold L/D correlation provides a new feasible criterion of L/D threshold value. The developed correction factor gives a reasonable accuracy fur the original database, showing the error of -2.18% for average and 27.75% for RMS, and promising results for new experimental data.