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

• Proceedings of the Korean Nuclear Society Conference
• /
• 1997.10a
• /
• pp.552-557
• /
• 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
• /
• v.40 no.4
• /
• pp.275-281
• /
• 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.

• Journal of Mechanical Science and Technology
• /
• v.15 no.7
• /
• pp.931-940
• /
• 2001

A numerical method for simulating bubble motion during nucleate boiling is presented. The vapor-liquid interface is captured by a level set method which can easily handle breaking and merging of the interface and can calculate an interfacial curvature more accurately than the VOF method using a step function. The level set method is modified to include the effects of phase change at the interface and contact angle at the wall as well as to achieve mass conservation during the whole calculation procedure. Also, a simplified model to predict the heat flux in a thin liquid microlayer is developed. The method is applied for simulation of a sliding bubble on a vertical surface to further understand the physics of partial boiling. Based on the computed results, the effects of contact angle, wall superheat and phase change on a sliding bubble are quantified.

• Korean Journal of Air-Conditioning and Refrigeration Engineering
• /
• v.16 no.10
• /
• pp.942-951
• /
• 2004

This paper presents the results of a theoretical study of the effect of radiation during free convective laminar film boiling for methanol/water binary mixtures on an isothermal vertical wall at atmospheric pressure. With the well-known boundary layer theory as a basis, a theoretical model has been formulated into consideration for mass diffusion at liquid phase. The equations are numerically solved by a similarity method to investigate the effects of radiation emissivity on the surface with various parameters such as wall superheat and composition of more volatile component at liquid phase far from the wall. From the results, the distributions of the physical quantifies are investigated in both phases. New correlations are proposed to predict the heat transfer coefficient of binary mixtures. It is shown that the proposed correlations are in good agreement with numerical results and with Bromley's correlation within maximum $11\%$ errors. It is also found that as the wall superheat is increased, radiation effect becomes more important.

• The KSFM Journal of Fluid Machinery
• /
• v.14 no.6
• /
• pp.45-53
• /
• 2011

Multiphase lattice Boltzmann method (LBM) has been used to simulate the nucleate pool boiling directly. For the phase change model, the thermal model and the Stefan boundary condition were introduced to the isothermal LBM. The phase change model was validated by the bubble growth in a superheated liquid under no gravity. The bubble growth on and departure from a superheated wall has been simulated successfully. The preliminary results showed that the detail process of nucleate pool boiling was in good agreement with the experimental results.

• 한국전산유체공학회:학술대회논문집
• /
• 2004.03a
• /
• pp.49-55
• /
• 2004

This paper addresses a numerical simulation of the flow and heat transfer in a simplified model of helically coiled tube steam generator using a general purpose computational fluid dynamic analysis computer code. The steam generator model is comprised of a cylindrical shell and helically coiled tubes. A cold feed water entered the tubes is heated up, evaporates. and finally become a superheated steam with a large amount of heat transferred continuously from the hot compressed water at higher pressure flowing counter-currently through the shell side. For the calculation of tube side two-phase flow field formed by boiling, inhomogeneous two-fluid model is used. Both the internal and external turbulent flows are simulated using the standard k-e model. The conjugate heat transfer analysis method is employed to calculate the conduction in the tube wall with finite thickness and the convections in the internal and external fluids simultaneously so as to match the fluid-wall-fluid interface conditions properly. The numerical calculations are peformed for helically coiled tubes of steam generator at an integral type pressurized water reactor under normal operation. The effects of tube-side inlet flow velocity are discussed in details. The results of present numerical simulation are considered to be physically plausible based on the data and knowledge from previous experimental and numerical studies where available.

• Nuclear Engineering and Technology
• /
• v.56 no.2
• /
• pp.624-635
• /
• 2024

In two-fluid simulations of flow boiling, the modeling of the mean bubble diameter is a key parameter in the closure relations governing the intefacial transfer of mass, momentum, and energy. Monodispersed approach proved to be insufficient to describe the significant variation in bubble size during flow boiling in a heated pipe. A population balance model (PBM) has been employed to address these shortcomings. During nucleate boiling, vapor bubbles of a certain size are formed on the heated wall, detach and migrate into the bulk flow. These bubbles then grow, shrink or disintegrate by evaporation, condensation, breakage and aggregation. In this study, a parametric analysis of the PBM aggregation and breakage models has been performed to investigate their effect on the radial distribution of the mean bubble diameter and vapor volume fraction. The simulation results are compared with the DEBORA experiments (Garnier et al., 2001). In addition, the influence of PBM parameters on the local distribution of individual bubble size groups was also studied. The results have shown that the modeling of aggregation process has the largest influence on the results and is mainly dictated by the collisions due to flow turbulence.

• Nuclear Engineering and Technology
• /
• v.54 no.6
• /
• pp.2276-2296
• /
• 2022

This study validates the applicability of the CUPID code for simulating subcooled wall boiling under high-pressure conditions against number of DEBORA tests. In addition, a new numerical technique in which the interfacial momentum non-drag forces are calculated at the cell faces rather than the center is presented. This method reduced the numerical instability often triggered by calculating these terms at the cell center. Simulation results showed good agreement against the experimental data except for the bubble sizes in the bulk. Thus, a new model to calculate the Sauter mean diameter is proposed. Next, the effect of the relationship between the bubble departure diameter (D_dep) and the nucleation site density (N) on the performance of the Wall Heat Flux Partitioning (WHFP) model is investigated. Three correlations for D_dep and two for N are grouped into six combinations. Results by the different combinations show that despite the significant difference in the calculated D_dep, most combinations reasonably predict vapor distribution and liquid temperature. Analysis of the axial propagations of wall boiling parameters shows that the N term stabilizes the inconsistences in D_dep values by following a behavior reflective of D_dep to keep the total energy balance. Moreover, ratio of the heat flux components vary widely along the flow depending on the combinations. These results suggest that separate validation of D_dep correlations may be insufficient since its performance relies on the accompanying N correlations.

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