• 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.

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.977-986
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• 2019

In the thermal-hydraulic system codes, such as MARS and RELAP5/MOD3, the Savannah River Laboratory (SRL) model has been adopted as a subcooled boiling model. It, however, has been shown that the SRL model cannot take into account appropriately the effects of inlet liquid velocity and hydraulic diameter on axial void fraction development. To overcome the problems, Ha et al. (2018) proposed a modified SRL model, which is applicable to low-pressure and low-Pe conditions (P < 9.83 bar and $Pe{\leq}70,000$) only. In this work, the authors extended the modified SRL model by proposing a new net vapor generation (NVG) model and a wall evaporation model so that the new subcooled boiling model can cover a wide range of thermal-hydraulic conditions with pressures ranging from 1.1 to 69 bar, heat fluxes of $97-1186kW/m^2$, Pe of 3600 to 329,000, and hydraulic diameters of 5-25.5 mm. The new model was implemented in the MARS code and has been assessed using various subcooled boiling experimental data. The results of the new model showed better agreements with measured void fraction data, especially at low-pressure conditions.

• 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.

• 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 Korean Nuclear Society Conference
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• 1997.10a
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• pp.552-557
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• 1997

Based on the bubble coalescence adjacent to the heated wall as a flow structure for CHF condition, Chang and Lee developed a mechanistic critical heat flux (CHF) model for subcooled flow boiling. In this paper, improvements of Chang-Lee model are implemented with more solid theoretical bases for subcooled and low-quality flow boiling in tubes. Nedderman-Shearer's equations for the skin friction factor and universal velocity profile models are employed. Slip effect of movable bubbly layer is implemented to improve the predictability of low mass flow. Also, mechanistic subcooled flow boiling model is used to predict the flow quality and void fraction. The performance of the present model is verified using the KAIST CHF database of water in uniformly heated tubes. It is found that the present model can give a satisfactory agreement with experimental data within less than 9% RMS error.

• Journal of Advanced Marine Engineering and Technology
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• v.17 no.2
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• pp.36-43
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• 1993

Real quality and axial void fraction distribution of subcooled refrigerant flow is very important to predict the heat transfer rate and pressure drop in the design of refrigerating system. In the subcooled boiling region, the liquid bulk temperature is still below the corresponding saturation temperature. But beyond the net vapor generation point, bubble detachment is occured actively from the vapor layer formed on the wall. A reliable method to predict the vapor fraction from the liquid bulk temperature is suggested in this paper. And also the actual quality of the subcooled R-113 flow is calculated in the range of 261-1239kg/$m^2$s mass velocity and 10-30K subcooling.

• 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.

• 한국전산유체공학회:학술대회논문집
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• 2010.05a
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• pp.320-325
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• 2010

A multiphase CFD analysis is performed to investigate the effect of near-wall grid for simulating a subcooled boiling flow in vertical tube. The multiphase flow model used in this CFD analysis is the two-fluid model in which liquid(water) and vapor(steam) are considered as continuous and dispersed fluids, respectively. A wall boiling model is also used to simulate the subcooled boiling heat transfer at the heated wall boundary. The diameter and heated length of tube are 0.0154 m and 2 m, respectively. The system pressure in tube is 4.5 MPa and the inlet subcooling is 60 K. The near-wall grid size in the non-dimensional wall unit ($y_{w}^{+}$) was examined from 64 to 172 at the outlet boundary. The CFD calculations predicted the void distributions as well as the liquid and wall temperatures in tube. The predicted axial variations of the void fraction and the wall temperature are compared with the measured ones. The CFD prediction of the wall temperature is shown to slightly depend on the near-wall grid size but the axial void prediction has somewhat large dependency. The CFD prediction was found to show a better agreement with the measured one for the large near-wall grid, e.g., $y_{w}^{+}$ > 100.

• Journal of Advanced Marine Engineering and Technology
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• v.40 no.4
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• pp.275-281
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• 2016

Subcooled boiling under low pressure was numerically investigated using computational fluid dynamics(CFD). The wall boiling model was used for simulating the subcooled boiling; this model requires sub-models consisting of bubble departure diameter, nucleation site density and bubble departure frequency. The CFD code CFX provides the default models based on experimental data. Because these models are mostly developed under high pressure conditions, it would not be predicted well in low pressure conditions. Thus in this study, CFD validation for subcooled boiling under low pressure was analyzed. The numerical results were compared with experimental data from published paper. Simulations were performed with mass flux ranging from 250 to $750kg/m^2s$, heat flux ranging from 0.37 to $0.77MW/m^2$ and constant outlet pressure of 0.11 MPa. Employing the empirical correlation developed under low pressures could increase the accuracy of numerical analysis.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.40 no.7
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• pp.457-470
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• 2016

Most CFD codes, that mainly adopt the heat partitioning model as the wall boiling model, have shown low accuracies in predicting the two-phase flow parameters of subcooled boiling phenomena under low pressure conditions. In this study, a number of subcooled boiling experiments in vertical channels were analyzed using a thermal-hydraulic component code, CUPID. The prediction of the void fraction distribution using the CUPID code agreed well with experimental data at high-pressure conditions; whereas at low-pressure conditions, the predicted void fraction deviated considerably from measured ones. Sensitivity tests were performed on the submodels for major parameters in the heat partitioning model to find the optimized sets of empirical correlations suitable for low-pressure subcooled flow boiling. The effect of the K-factor on the void fraction distribution was also evaluated.