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In pressurized water reactor accident scenarios, the injection of water from the emergency core cooling system (ECCS) (ECC injection) might induce a pressurized thermal shock (PTS), affecting the reactor pressure vessel (RPV) integrity. Therefore, PTS is a vital research issue in reactor safety, and its analysis is essential for evaluating the integrity of RPVs, which determines the reactor life. The PTS analysis comprises a coupled analysis between thermal–hydraulic and structural analyses. The thermal–hydraulic approach is particularly crucial, and reliable computational fluid dynamic (CFD) simulations should play a vital role in the future because predicting the temperature gradient of the RPV wall requires data on the transient temperature distribution of the downcomer (DC). Since one-dimensional codes cannot predict the complex three-dimensional flow features during ECC injection, PTS is one reactor safety issue where CFD simulation can benefit from complement evaluations with thermal–hydraulic system analysis codes. This study reviewed from the viewpoint of the turbulence models most affecting PTS analysis based on papers published since 2010 on single- and two-phase flow CFD simulation for the experiment on PTS performed in the Rossendorf coolant mixing model (ROCOM), transient two-phase flow (TOPFLOW), upper plenum test facility (UPTF), and large-scale test facility (LSTF). The results revealed that in single-phase flow CFD simulation, where knowledge and experience are sufficient, various turbulence models have been considered, and many analyses using large eddy simulation (LES) have been reported. For two-phase flow analysis of air–water conditions, interface capturing/tracking methods were used in addition to two-fluid models. The standard k – ε and shear stress transport (SST) k – ω models were still in the validated phase, and various turbulence models have yet to be fully validated. In the two-phase flow analysis of steam–water conditions, many studies have used two-fluid models and Reynolds-averaged Navier-Stoke (RANS), and NEPTUNE_CFD, in particular, has been reported to show excellent prediction performance based on years of accumulated validation. [ABSTRACT FROM AUTHOR]

The drift-flux model (DFM) is of practical importance for analyzing gas–liquid flow in various applications, such as chemical reactors and nuclear power plants. This paper aims at developing an accurate flow-regime-independent DFM for estimating the void fractions of two-phase downward inclined flow. The effect of flow channel inclination on void fraction for downward two-phase flows was investigated comprehensively. First, 2868 data for void fractions of gas–liquid two-phase flows were summarized from 12 sources. The inclination angle ranged from 0° to 90°. Then, the existing drift-flux models were reviewed comprehensively. The existing models had poor predictive performance in terms of the whole experimental database. Therefore, a theoretically supported drift-flux model for downward inclined two-phase flow was established based on a deeper understanding of two-phase flow behavior. The influence of channel inclination angles on the distribution parameters and drift velocities was determined quantitatively. A total of 87.8% of estimated void fractions were within ±20% errors of experimental void fractions. The mean relative deviation, md, was −2.04%. [ABSTRACT FROM AUTHOR]

• New one-group drift-flux model (DFM) was developed for dispersed flows in annuli. • New two-group DFM was developed for dispersed flows in annuli. • Developed one-group and two-group DFMs were validated by extensive data. • Two-group DFM can be used for calculating two-group interfacial area concentration. • Two-group DFM can calculate two-group gas velocity using mixture momentum equation. Interfacial area concentration (IAC) plays a vital role in mass, momentum, and heat transfers between gas and liquid phases in two-phase flows. The IAC of group-one bubbles, including spherical and distorted bubbles, differs from the IAC of group-two bubbles, including cap, slug and churn-turbulent bubbles. The two-group interfacial area transport equation (IATE) can be used as a mechanistic model to obtain the IAC of each group. The equation demands the two-group gas velocity fields, leading to the need to solve an extra momentum equation, which is complicated. The two-group drift-flux model can prevent this complexity by providing two-group gas velocities without introducing the extra momentum equation. Despite the importance of gas-liquid flows in annuli, the drift-flux model for gas-liquid flows is not well-developed like it is for pipe flows. The main objective of this study is to develop a two-group drift-flux model for dispersed vertical flows in annuli. The two-group drift-flux model provides the flow characteristics of group-one and group-two bubbles, such as void fractions and velocities, which strongly affect the prediction of the available interfacial area concentration of the group-one and group-two bubbles for heat and mass transfer. Hence, after comprehensively investigating the available experimental data, this study first improves a one-group drift-flux model for dispersed gas and low-viscosity liquid upward flows in annuli and proposes a new two-group drift-flux model. Then, the models are evaluated using the respective group experimental data. The consistency between one-group and two-group drift-flux models is also investigated. This consistency helps to reduce the two gas momentum equations to one gas mixture momentum equation. The results of the comparison between model predictions and experimental data are satisfactory. [ABSTRACT FROM AUTHOR]

• Channel aspect ratio affected the distribution parameter in the drift-flux model. • Asymptotic distribution parameter was approximated at 1.2 for square channels. • Hydraulic diameter was used as characteristic length in a drift velocity correlation. • Drift-flux correlation predicted void fractions within ±13.9 % for adiabatic flow data. • Drift-flux correlation predicted void fractions within ±8.72 % for boiling flow data. Numerous industrial components and production processes encounter two-phase flows in rectangular channels. The drift-flux model, mathematically expressed in terms of the drift-flux parameters, such as the drift velocity and distribution parameter, is often utilized to solve two-phase flow problems. The constitutive equations of the drift-flux parameters have been limitedly developed and assessed for dispersed two-phase flows in rectangular channels. This study discusses the effects of geometrical parameters, such as the characteristic length and aspect ratio of a rectangular channel, on dispersed two-phase flow parameters. For the flow conditions of collected data in rectangular channels, it was observed that the hydraulic diameter was appropriate as the characteristic length scale for the drift velocity calculation since the drift flux correlation using the hydraulic diameter correlated better with experimental data. The asymptotic distribution parameters in the constitutive equation of the distribution parameter were approximated to be 1.35 and 1.20 for rectangular channel aspect ratios less than 0.25 and greater than 0.33, respectively. Due to the lack of data, the asymptotic distribution parameters were tentatively determined based on a linear interpolation scheme between 1.35 and 1.20 for the aspect ratios between 0.25 and 0.33. The constitutive equations of the drift-flux model were assessed using the collected data with various fluids, pressures, and flow conditions. The constitutive equations of the drift-flux model could predict the void fractions with negligible bias and random uncertainty of ±0.0358 and ±0.0434 (absolute value measure) and ±13.9 % and ±8.72 % (relative value measure) for adiabatic and boiling two-phase flows in rectangular channels, respectively. [ABSTRACT FROM AUTHOR]

• New method was developed to obtain subchannel average two-phase flow parameters. • New flow parameter mapping data was obtained for 8 × 8 rod bundle. • New two-group drift-flux model for 8 × 8 rod bundle was developed. • New two-group drift-flux model is useful for predicting interfacial area concentration. • Newly obtained data is useful for subchannel analysis code validation. Interfacial transfer terms in gas-liquid two-phase flows are formulated as the product of interfacial area concentration (IAC) and flux. The interfacial area transport equation (IATE) is essential for obtaining IAC in transient and developing two-phase flows. Two-group IATE was developed to account for the difference in the interfacial drag force between two groups of bubbles, where spherical and distorted bubbles are categorized as group one, while cap, slug, and churn turbulent bubbles are categorized as group two. The rigorous two-group approach requires two-group gas momentum equations, resulting in one additional momentum equation in the two-fluid model. The mixture gas momentum equation is considered to avoid the additional momentum equation. The two-group drift-flux model is necessary to calculate the two-group gas velocity from the mixture velocity. The present study develops an approximation methodology to acquire subchannel average and rod bundle average two-group two-phase flow parameters based on the validated power law assumption and a limited number of local data. The two-group model is developed for the distribution parameter and drift velocity for dispersed two-phase flows in rod bundles. The total performance of the newly developed two-group drift-flux model is evaluated by two-phase flow data in rod bundles. The results show that the developed two-group drift-flux model could well predict two-group gas velocities in rod bundles. This model is useful for subchannel thermal hydraulic analysis codes as a complement to the conventional one-group drift-flux model. [ABSTRACT FROM AUTHOR]

• A robust drift-flux model for upward high-viscosity two-phase flows was proposed. • The rigorous constitutive equations of the drift velocities were formulated based on drag law. • The dependence of drift-flux parameters on viscosities and flow regimes was identified. • The new theoretically-supported model achieved a superior predictive performance. The drift-flux model is of practical importance in predicting the void fraction of upward two-phase flows with high-viscosity liquid phase. This study aims to develop an accurate and robust drift-flux correlation for each type of upward high-viscosity two-phase flow to overcome the unsatisfactory predictive performance of the existing drift-flux correlations. First, an extensive literature review was conducted which collected more than 500 experimental void fractions of upward high-viscosity two-phase flows from 11 sources. The viscosity numbers for the collected database ranged from 0.0213 to 25.6. The existing drift-flux correlations of high-viscosity two-phase flows were critically reviewed. A comparison between the collected database and the existing drift-flux correlations indicated that none of the existing drift-flux correlations could predict the entire collected database with acceptable accuracy. The failure of the void fraction prediction using the existing drift-flux correlations drove the development of new drift-flux correlations based on the state-of-the-art understanding of the flow behaviors for upward high-viscosity two-phase flows. The rigorous constitutive equations of the drift velocities were formulated based on the drag law. The dependence of the distribution parameter on the physical properties and flow conditions was systematically analyzed. Finally, the new drift-flux correlation was developed for upward high-viscosity two-phase flows. This correlation was well-validated as demonstrating superior predictive performance to the existing correlations; for example, 92.2% of the collected void fractions were predicted within ± 20 % error without any systematic bias. The correlations are considered to have great practical potential in improving the analysis of high-viscosity two-phase flows. [ABSTRACT FROM AUTHOR]

• Methodology to estimate subchannel average flow parameters was developed. • Subchannel average two-phase flow parameters were obtained in an 8 × 8 rod bundle. • Operating parameter effects on the flow characteristics were found. • The distribution parameter model for the rod bundle was validated independently. • The drift velocity model for the rod bundle was validated independently. Two-phase flows in rod bundles or tube bundles appear in heat transfer systems. The bundle-average (or one-dimensional) void fraction is a critical parameter in overall system design, performance evaluation, and safety assessment. Accurate bundle-average void fraction prediction requires a precise interfacial momentum transfer term formulated using the distribution parameter and drift velocity, two essential drift-flux parameters in the drift-flux model. The currently utilized modeling approach is based on the distribution parameter and drift velocity determined together through drift-flux plots. This results in possible compensation errors between the distribution parameter and drift velocity in the model validation process. The independent validation of the constitutive equations for the distribution parameter and drift velocity has been challenging due to experimental difficulty in measuring detailed two-phase flow structures in rod bundles. The present study proposes an approximation methodology to obtain subchannel average two-phase flow parameters using local two-phase flow data measured at two local points: the subchannel center and minimum gap center in a rod bundle. The distributions of subchannel average two-phase flow parameters are invaluable in benchmarking subchannel analysis codes. Comprehensive subchannel average void fraction and gas velocity mappings are given to discuss the effects of gas and liquid velocities, pressure, spacer grid, and developing length on the two-phase flow characteristics in the rod bundle. The bundle-average distribution parameter values calculated by subchannel average two-phase flow parameters are used for independent validation of the constitutive equations of the distribution parameter. The bundle-average drift velocity values back-calculated using the distribution parameter values are also used for independent validation of the constitutive equation of the drift velocity. [ABSTRACT FROM AUTHOR]

• A new two-group drift-flux model (DFM) was developed for large-diameter pipes. • Group-one drift velocity was proposed by considering group-two effects. • Group-one and group-two distribution parameters were determined. • The one-group DFM could be reproduced by the two-group DFM. • The model was applicable to adiabatic two-phase flows. Interfacial heat and mass transfer are prevalent in industrial processes. The interfacial transfer rate can be obtained by the product of their fluxes and interfacial area concentration (IAC) calculated by the interfacial area transport equation (IATE). Bubbles show different behavior according to their sizes. Hence, bubbles are classified into two groups. Consequently, two-group IATE is required causing to use of two gas momentum equations leading to more complexity. The present study suggests a new reliable two-group drift-flux modeling to reduce the two gas momentum equations to one gas mixture momentum equation for gas-liquid flow in large-diameter pipes. The model is developed based on the drift-flux model concept and experimental data. Group-one and group-two distribution parameters and drift velocities are validated through experimental data. The results show that the proposed two-group drift-flux model can support the concept of drift velocity from the bubbly to beyond the bubbly flow and consistency between the one-group and two-group drift-flux models. Moreover, steam-water data are used to validate the applicability of the model in steam-water flows condition. The developed two-group drift-flux model is indispensable for reducing the two gas momentum equations to one gas mixture momentum equation when two-group IATE is implemented into thermal-hydraulic codes to improve the prediction accuracy of IAC. [ABSTRACT FROM AUTHOR]

• Effects of inlet boundary conditions on interfacial area concentration changes are investigated. • Some non-uniform inlet boundaries significantly affect interfacial area concentration changes. • Predictive capability of the IATE calculation is evaluated under inlet boundary conditions. • IATE's predictive capability deteriorates for several non-uniform inlet flow boundary conditions. Bubbly flows appear in many heat and mass transfer equipment. The interfacial area concentration is a critical parameter in determining the performance of the heat and mass transfer systems. The one-dimensional one-group interfacial area transport equation (IATE) was developed based on area-averaged bubbly flow data collected under uniform inlet flow boundary conditions. The applicability of the IATE to the flow systems with non-uniform inlet flow boundary conditions was not comprehensively examined. This study investigated the effect of inlet liquid and gas velocity profiles on measured interfacial area concentration changes along the flow direction. The data used in this study were collected for vertical upward air-water two-phase flow in a rectangular channel with a gap length of 10 mm and width of 200 mm. The test conditions were the superficial gas velocity ranging from 0.0870 to 0.288 m/s and the superficial liquid velocity ranging from 0.515 to 2.53 m/s. The interfacial area concentrations with uniform and non-uniform inlet flow boundary conditions measured in a large vertical rectangular channel were compared under the same area-averaged flow rate conditions. Then, the predictive capability of the IATE to the bubbly flow test data with uniform and non-uniform inlet flow boundary conditions was assessed. Some non-uniform inlet flow boundary conditions, including channel-center peaking of liquid flow (CPF), channel-center peaking of gas flow (CPG), single-side-wall peaking of liquid flow (SPF), and single-side-wall peaking of gas flow (SPG), significantly affected the interfacial area concentration changes along the axial location at relatively high superficial gas or liquid velocity conditions. The IATE's predictive capability deteriorated for several non-uniform inlet flow boundary conditions, including double-side-wall peaking of gas flow (DPG) and single-side-wall peaking of liquid flow (SPF), specifically at relatively high superficial liquid velocity conditions. The maximum percentage error of the interfacial area concentration prediction by the IATE for these non-uniform inlet flow boundary conditions reached 75%. [ABSTRACT FROM AUTHOR]

• Interfacial drag forces for interior, edge, and corner subchannels were formulated in a rod bundle. • Subchannel-average void fraction covariance was empirically modeled for saturated boiling flow. • Subchannel-average void fraction covariance was analytically modeled for subcooled boiling flow. • The covariance models were validated by steam-water data up to 8.6 MPa. • The relative velocity covariances were calculated from the void fraction covariances. Accurate simulation of boiling two-phase flows in a rod bundle is indispensable for the robust, economical, and safe design of various heat transfer systems using the rod bundle configuration. Subchannel analysis codes are used for this purpose. Interfacial drag force modeling significantly affects the prediction accuracy of the void fraction. The void fraction and relative velocity covariances constitute the interfacial drag force. However, the covariances are currently not considered in existing subchannel codes due to a lack of reliable constitutive equations to calculate the void fraction and relative velocity covariances. This study aims to model the subchannel-average void fraction and relative velocity covariances for subcooled and saturated boiling flows in three types of subchannels in a rod bundle. The considered subchannels are interior, edge, and corner subchannels. The subchannel-average void fraction and relative velocity covariances for saturated boiling flow are modeled by the data obtained from local void fraction data collected for saturated boiling flow in an 8 × 8 rod bundle under pressures from 1.0 to 8.6 MPa. The subchannel-average void fraction and relative velocity covariances for subcooled boiling flow are modeled based on the bubble-layer thickness model. The modeled subchannel-average void fraction and relative velocity covariances are well validated with the experimental data. The modeled subchannel-average void fraction and relative velocity covariances are expected to be implemented in subchannel analysis codes to improve the void fraction prediction accuracy in each subchannel type. [ABSTRACT FROM AUTHOR]

• The channel geometry affects the velocity profile, average film thickness, wave height, base film thickness, and wave frequency, while it has little effect on wave velocity. • The system pressure affects the magnitude of average film thickness, base film thickness, and wave characteristics. • For boiling flow, the heating state should be considered as an influencing factor. • Wave velocities in the wispy annular flow regime and other annular flow regimes behave differently. • An accurate description of the structure of the gas-liquid interface is necessary for a better understanding of annular flow and modeling of liquid film characteristics and wave characteristics. A comprehensive understanding of annular flow characteristics is needed to improve the transport efficiency of petroleum and natural gas pipelines, and the efficiency and safety of heat transfer equipment, such as condensers, evaporators, steam generators, and nuclear reactors. For the annular flow characteristics, liquid film characteristics and wave characteristics are of essential importance for heat and mass transfer in annular flows. This paper reviews previous studies on these parameters, including prediction models and correlations for liquid film characteristics and wave characteristics of annular flows. Then, the paper examines the predictive capability of the models and correlations using existing databases. The data comparison considers the following aspects: the channel geometry effect, pressure effect, and heating state. The channel geometry affects the velocity profile, average film thickness, wave height, base film thickness, and wave frequency, while it has little effect on wave velocity. The pressure effect on these parameters is evident as it changes the gas inertia. For wave velocity in boiling flows, the predictions of current adiabatic flow models do not agree well with the current database values indicating the importance of heating state. [ABSTRACT FROM AUTHOR]

• A drift-flux correlation for a large annulus is developed. • A threshold length scale between medium and large annuli is proposed. • Applicability of drift-flux correlation to an eccentric annulus is confirmed. • Distribution parameter in churn flow is independent of channel size. In view of the practical importance of predicting void fraction in two-phase flow, a correlation based on the drift-flux model is often utilized in many industrial fields. To use the drift-flux type correlation, two drift-flux parameters such as distribution parameter and void fraction weighted mean drift velocity should be modeled accurately. Since the drift-flux parameters are susceptible to channel size, they are modeled differently depending on whether the channel size is characterized as small (or mini), medium or large. Although drift-flux type correlations with modeled distribution parameter and drift velocity for medium size channels with various geometries have been well-developed, limited drift-flux correlations are available for large size channels. Two-phase flow characteristics in a large size channel are quite different from those in a medium size channel at low flow conditions. Large slug bubbles observed in a medium size channel do not form in a large size channel due to the presence of surface instability. As large slug bubbles disintegrate into smaller cap bubbles, the drift velocity is increased. Complex two-phase flow characteristics in a large size channel also affect the distribution parameter. This study successfully models the distribution parameter and drift velocity for two-phase flow in a vertical concentric annulus. The drift-flux correlation with modeled distribution parameter and drift velocity is validated by existing data taken in large size annuli. The applicability of the drift-flux correlation to two-phase flow in a vertical eccentric annulus is also examined by existing data. [ABSTRACT FROM AUTHOR]

Abstract The study of two-phase flow in rectangular channels is of great importance due to their use in marine reactors, chemical reactors, test reactors, and neutron spallation sources. Due to the practical importance of the drift-flux correlation for two-phase flow analysis in industrial systems, distribution parameter and drift velocity have been studied for vertical upward two-phase flow in a large size rectangular channel. Due to the presence of liquid recirculation in the bubbly flow regime in large size rectangular channels, the distribution parameter, drift velocity, and flow regime transition criteria are different from those found in small or medium size rectangular channels. The use of hydraulic equivalent diameter as the characteristic length scale in channels with significant asymmetry and its effect on interfacial area concentration prediction are also discussed. Under these considerations a new drift-flux correlation has been developed for upward two-phase flow in large size rectangular channels. Comparison of the newly developed correlation with several datasets shows reasonable agreement. [ABSTRACT FROM AUTHOR]

Highlights • Constitutive equations for a vertical rod bundle are reviewed. • Dependence of distribution parameter on flow conditions is discussed. • 1D momentum equation by considering void fraction distribution is discussed. • Effect of channel size on interfacial area concentration is discussed. Abstract In view of the quality assurance of two-phase flow simulations, CSAU (Code Scalability, Applicability, and Uncertainty) methodology and code V & V (Verification and Validation) have been proposed. The estimation of simulation uncertainty is indispensable in using best-estimate computational codes. A key of successful two-phase flow simulations is to use the state-of-the-art constitutive equations to close the mathematical system used in two-phase flow analyses. The advanced constitutive equations should be developed based on “physics” behind phenomena and should consider scaling parameters which enable their application beyond test conditions used for a code validation. Two-phase flow simulations in a rod bundle is important in various industrial apparatuses such as heat exchangers and nuclear reactors. Constitutive equations for two-phase flows in a vertical rod bundles have been advanced in recent five years. In view of this, this paper provides a comprehensive review of most advanced constitutive equations for two-phase flow analyses in a vertical rod bundle. The constitutive equations of two-phase flow parameters reviewed in this paper are flow regime map, void fraction, void fraction covariance and relative velocity covariance, interfacial area concentration and wall friction. In addition, an exact formulation of one-dimensional momentum equation in two-fluid model considering void fraction distribution is discussed. [ABSTRACT FROM AUTHOR]

Highlights • Interfacial area transport equation and its available constitutive models are reviewed. • Latest interfacial area concentration experimental data in bubbly flows are collected. • 8 bubble coalescence and breakup models are evaluated with the collected databases. • New bubble coalescence and breakup model has been developed for bubbly flows. • New bubble coalescence and breakup model can predict the collected databases well. Abstract The performance of thermal-hydraulics analysis codes heavily depends on the reliable models developed for interfacial transfer terms in the two-fluid model. This requires the accurate modeling of the interfacial area concentration (IAC). The interfacial area transport equation (IATE) developed from the population balance equation is one of the promising ways to be able to dynamically describe the IAC evolutions in the two-phase flows. A lot of efforts have been made to develop the IATE and its bubble coalescence and breakup models in the past two decades. The one-group IATE with its bubble coalescence and breakup models serves as an important and promising model to predict the IAC behaviors in the bubbly flows now. This paper made an extensive survey on the recent theoretical and experimental researches on the IATE and collected the available 8 sets of bubble coalescence and breakup models and the latest experimental databases measured with the up-to-date measuring techniques. The 8 sets of bubble coalescence and breakup models are originally developed to model the bubble evolution in bubbly flows and the small bubble evolution in the flows beyond bubbly flow. The collected data covers the bubbly flows taken the vertical pipes with the diameter ranging from 0.0508 m to 0.200 m and the height-to-diameter ratio ranging from 4 to 113. To confirm the reliability of the 8 sets of bubble coalescence and breakup models, this paper systematically compared and evaluated them with the collected experimental data in the 1D one-group IATE numerical calculations. Although the bubble coalescence and breakup models of Sun et al. (2004, Modeling of bubble coalescence and disintegration in confined upward two-phase flow. Nucl. Eng. Des. 230, 3–26) and Hibiki and Ishii (2000, One-group interfacial area transport of bubbly flows in vertical round tubes. Int. J. Heat Mass Transfer 43, 2711–2726) were found to show the best and second-best predicting performance for the collected latest experimental databases of bubbly flows with data-number-averaging mean absolute relative errors of 33.5% and 61.7% respectively, their relative errors are not satisfactory. Hibiki and Ishii (2000, One-group interfacial area transport of bubbly flows in vertical round tubes. Int. J. Heat Mass Transfer 43, 2711–2726) model only consisting of the contributions of the bubble random collision coalescence and turbulent impact breakup mechanisms can reasonably predict the IAC experimental data in the bubbly flows with high-liquid flow-rate flow conditions and significantly overpredict the IAC experimental data in the bubbly flows with low-liquid flow-rate flow conditions. To further improve the performance of the 1D one-group IATE, this paper has developed a bubble coalescence and breakup model by greatly improving the model of Hibiki and Ishii (2000, One-group interfacial area transport of bubbly flows in vertical round tubes. Int. J. Heat Mass Transfer 43, 2711–2726) with the addition of the contribution of the wake entrainment bubble coalescence mechanism which prevails in low-liquid flow-rate bubbly flows and optimized the adjustable coefficients of the 3 major constitutive models with the collected latest experimental databases of bubbly flows. The data-number-averaging mean absolute relative error of the newly-proposed bubble coalescence and breakup model is accordingly improved to be 17.1% in the predictions of the collected latest experimental databases of bubbly flows. [ABSTRACT FROM AUTHOR]

Highlights • A critical review addresses issues in one-group interfacial area transport models. • A critical review addresses future directions of interfacial area modeling. • 1D interfacial area models of turbulent diffusion & lateral migration are derived. • A preliminary interfacial area model for high-pressure conditions is developed. Abstract This paper provides the state-of-the-art critical review on the modeled source and sink terms of the one-group interfacial area transport equation. The reviewed source and sink terms include models developed by Wu et al., Hibiki and Ishii, Hibiki et al., Yao and Morel, Park et al., Hibiki et al., Kataoka et al., Nguyen et al., Shen and Hibiki and Hazuku et al. The present critical review assesses major issues in modeling the one-group source and sink terms. Important conclusions are that the existing source and sink terms are not well-validated for developing and transient flows and the applicability of the source and sink terms to high pressure conditions such as prototypic nuclear reactor conditions has not been well-discussed. In view of these, coefficients of the source and sink terms in Hibiki and Ishii’s model have been modified for their application to prototypic nuclear reactor conditions. In addition, approximated area-averaged source and sink terms due to turbulent diffusion and lateral migration are derived by area-averaging local source and sink terms due to turbulent diffusion and lateral migration developed by Kataoka et al. The role of the area-averaged source and sink terms due to turbulent diffusion and lateral migration on the interfacial area transport should be tested in a future study. This paper also overviews important model validation challenges to be addressed in a future study. The challenging model validation should address the model performance for various test sections and flow conditions. They are (1) the effect of channel size (small (or micro/mini)-to-large channels) on the interfacial area transport, (2) the scalability of the interfacial area transport equation to prototypic nuclear reactor conditions, (3) the effect of covariance due to phase distribution on the interfacial area transport, (4) the effect of inlet conditions on the interfacial area transport, (5) the applicability of the interfacial area transport equation to transient and developing flow conditions, and (6) the applicability of the interfacial area transport equation to various flow channels including a rod bundle. [ABSTRACT FROM AUTHOR]

Highlights • Heat transfer correlations of two-component two-phase flow in pipes were reviewed. • Heat transfer data of two-component two-phase flow in pipes were reviewed. • None of existing correlations could predict collected databases successfully. • The dependence of heat transfer coefficient on flow parameters was explained. • A heat transfer correlation was developed for two-component two-phase flow. Abstract The prediction of heat transfer coefficient for gas–liquid two-phase flows in vertical pipes is important for designing many industrial and engineering apparatuses such as nuclear power plants, solar collectors and petroleum pipelines. The mechanism of two-component two-phase heat transfer is more complicated than that of single-phase heat transfer. This study aims at developing a robust correlation of heat transfer coefficient for two-component two-phase slug flows in vertical pipes. The theoretical framework for the correlation development is inspired by considering classic Reynolds and Chilton–Colburn analogies. The framework demonstrates that the functional dependence of the two-component two-phase heat transfer coefficient on a void fraction or a two-phase multiplier is expressed by the Reynolds and Chilton–Colburn analogies. The correlation for the heat transfer coefficient of two-component two-phase slug flows in vertical pipes has been developed and validated by more than 200 existing data. The newly developed correlation predicts 95.1% of the collected two-phase heat transfer data within ± 30% error and the mean absolute relative deviation of the correlation is estimated to be 14.2%. The extended applicability of the newly developed correlation to other flow regimes such as bubbly, churn and annular flow regimes has been verified by the existing experimental data. The possible application of the newly developed correlation to large pipes is also discussed. In summary, it is expected that the newly developed correlation can predict the heat transfer coefficient of two-component two-phase flows under various conditions such as vertical upward and downward flows, developing and fully developed flows, laminar and turbulent flows and all two-phase flow regimes. The application of the correlation to vertical large pipes is also considered to be reasonable. [ABSTRACT FROM AUTHOR]

Recent progress in nuclear thermal-hydraulics simulations has been largely focused on coupling with other computational packages, improved closure models for subcooled boiling and for bubbly flows, and the development of higher-fidelity simulation capabilities (Kulesza et al., 2016). While high-fidelity 3D simulation is important for model validation, scientific understanding, and some design calculations, it can be prohibitively expensive for system design applications or applications involving large geometries. Thus, there is also a need for practical, simplified approaches for those applications. The two-fluid model strikes a balance between detail and computational resources, but requires the accurate specification of several key constitutive models. These include (1) interfacial forces, (2) interfacial area concentration, (3) two-phase turbulence, and (4) wall and bulk boiling and condensation. In many modern CFD packages, uncertainties in the local interfacial area concentration can have strong effects on the ability to predict the other key parameters. This paper demonstrates that the drag force in 3D CFD can be formulated in much the same way as in 1D system analysis codes and that this approach can be used to formulate a model for interfacial area concentration. The method is also applied to two-group approaches to consider the difference in transport properties for different bubble size classes. This approach may open a method to calculate the interfacial forces without the need for interfacial area transport equations. This reduces the number of differential equations and avoids the modeling challenges associated with bubble breakup and coalescence kernels and the need to specify the inlet interfacial area concentration a priori . The new method is expected to decouple the effects of interfacial area uncertainty and calibrated coefficients, and should provide reasonable local bubble diameters for both group-1 and group-2 bubbles. The approaches proposed in this study are applicable to two-phase flow simulations in rather simple geometries such as upward two-phase flow in vertical channels. In view of many applications for upward two-phase flow in vertical channels, including nuclear reactor systems, the proposed methods are considered useful. [ABSTRACT FROM AUTHOR]

Utilization of one-dimensional system analysis code such as TRACE, RELAP5, TRAC-BF1 to evaluate gas-liquid two-phase flow behaviors in nuclear power plants is crucial for the plant-level safety assessment. In the two-fluid model, interfacial momentum transfer between two phases is expressed under interfacial drag term in the momentum equation. For the rigorous and accurate expression of interfacial drag term and drift flux parameter, covariance due to the area averaging of void fraction distribution must be considered. In the present paper, an effect of the covariance on void fraction prediction in pipe flow was numerically assessed by implementing Hibiki and Ozaki's model into the interfacial drag term in the one-dimensional two-fluid model. For the low flow rate with high void fraction conditions, it was found that the inclusion of covariance model slightly underestimated void fraction value than that calculated by the drift-flux model. This underestimation comes from the momentum source term in the two-fluid model, which was derived under the assumption of uniform void fraction distribution. Therefore, in this paper, momentum source term was rederived with consideration of void fraction covariance and a complete set of momentum equation and constitutive formulations for the one-dimensional two-fluid model is presented. [ABSTRACT FROM AUTHOR]

Steam submerged jet condensation has been widely used in many industry applications, especially in nuclear engineering. Since the heat and mass transfer varies with the condensation patterns as well as steam-water interface shapes, the identification and classification of the condensation regime is indispensable to conduct thermal-hydraulic analyses and to select adequate constitutive equations such as heat transfer coefficient. Different definitions and criteria have been used to classify the steam jet condensation regime. Due to the different transition criteria of condensation regime and multi-parameters’ effect, existing condensation regime maps do not agree each other. In this paper, unified criteria and definitions for each regimes have been proposed based on the dynamic behavior and geometrical shape of steam-water interface. The re-classified condensation regimes are chugging regime, hemispherical bubble oscillation regime, condensation oscillation regime, stable condensation regime and steam escape regime. The existing analytical model, empirical correlations and empirical map developed under different test section designs and test conditions failed to predict the existing experimental condensation regime transition boundaries. The discrepancy in the transition boundaries among existing experimental data is significant due partly to the subjective classification and complicated dependence on multiple parameters. Further efforts on both analytical and experimental researches are encouraged in the future. [ABSTRACT FROM AUTHOR]

Convection, wherein forced and natural convections are prominent, is known as mixed convection. Specifically, when a forced convection flow is downward, this flow is called opposing flow. The objectives of this study are to gain a comprehensive understanding of opposing flow mixed convection heat transfer and to establish the prediction methodology by evaluating existing correlations and models. Several heat transfer correlations have been reported related to single-phase opposing flow; however, these correlations are based on experiments conducted in various channel geometries, working fluids, and thermal flow parameter ranges. Because the definition of nondimensional parameters and their validated range confirmed by experiments differ for each correlation reported in previous studies, establishing a guideline for deciding which correlation should be selected based on its range of applicability and extrapolation performance is important. This study reviewed the existing heat transfer correlations for turbulent opposing flow mixed convection and the single-phase heat transfer correlations implemented in the thermal–hydraulic system codes. Furthermore, the authors evaluated the predictive performance of each correlation by comparing them with the experimental data obtained under various experimental conditions. The Jackson and Fewster, Churchill, and Swanson and Catton correlations can accurately predict all the experimental data. The effect of the difference in the thermal boundary conditions, i.e., uniform heat flux and uniform wall temperature, on the turbulent mixed convection heat transfer coefficient is not substantial. The authors confirmed that heat transfer correlations using the hydraulic-equivalent diameter as a characteristic length can be used for predictions regardless of channel-geometry differences. Furthermore, correlations described based on nondimensional dominant parameters can be used for predictions regardless of the differences in working fluids. The authors investigated the extrapolation performance of the mixed convection heat transfer correlations for a wide range of nondimensional parameters and observed that the Jackson and Fewster, Churchill, and Aicher and Martin correlations exhibit excellent extrapolation performance with respect to natural and forced convection flows, indicating that they can be applied beyond the parameter range validated experimentally. [ABSTRACT FROM AUTHOR]

Downward two-phase flow is observed in light water reactor accident scenarios such as loss of coolant accident (LOCA) and loss of heat sink accident (LOHS) due to loss of feed water or a secondary pipe break, and so, it is vital to have a thorough understanding of the flow mechanisms and regimes. With this point of view, flow regime transition criteria for vertical downward flow for a range of pipe diameters of 24–101.6 mm has been developed. Several models looked at the radial distribution of the bubbles and the wake effect of leading bubbles while others looked into the Kelvin-Helmholtz instability seen at the gas-liquid interface. The newly developed criteria have been compared to flow regime maps obtained via subjective and objective means, consisting of air-water data at atmospheric conditions as well as at an elevated pressure of 0.2 MPa. The new model is also compared to flow regime maps developed with different inlet conditions. Overall, the present model showed good agreements with the available data, with the exception of several 50.8 mm ID flow regime maps of different inlet conditions as well as a self-organizing neural network. This study also highlights the need for a more objective and consistent flow regime map data for large diameter pipes, the identification of cap-bubbly and churn-turbulent flows in these maps, and the deviations observed between a supervised and self-organizing neural network (SONN). [ABSTRACT FROM AUTHOR]

A drift-flux correlation has been often used to predict void fraction of gas-liquid two-phase flow in a horizontal channel due to its simplicity and practicality. The drift-flux correlation includes two important drift-flux parameters, namely, the distribution parameter and void-fraction-weighted-mean drift velocity. In this study, an extensive literature survey for horizontal two-phase flow is conducted to establish void fraction database and to acquire existing drift-flux correlations. A total of 566 data is collected from 12 data sources and 4 flow-regime-dependent and 1 flow-regime-independent drift-flux correlations are identified. The predictive capability of the existing drift-flux correlations is assessed using the collected data. It is pointed out that the drift velocity determined by a regression analysis may include a significant error due to a compensation error between distribution parameter and drift velocity. In this study, a simple flow-regime-independent drift-flux correlation is developed. In the modeling approach, the void-fraction-weighted mean drift velocity is approximated to be 0 m/s, whereas the distribution parameter is given as a simple function of the ratio of non-dimensional superficial gas velocity to non-dimensional mixture volumetric flux. The newly developed correlation shows an excellent predictive capability of void fraction for horizontal two-phase flow. Mean absolute error (or bias), standard deviation (random error), mean relative deviation and mean absolute relative deviation of the correlation are 0.0487, 0.0985, 0.0758 and 0.206, respectively. The prediction accuracy of the correlation is similar to the correlation of Chexal et al. (1991), which was formulated based on the drift-flux parameters by means of many cascading constitutive relationships with numerous empirical parameters. [ABSTRACT FROM AUTHOR]

Eulerian two-fluid model (TFM) has been the workhorse for several applications involving vertical bubbly flows due to its computational efficiency especially when applied to large-scale systems. The constituent phases are treated as interpenetrating continuous media, and the stress terms are usually modeled using Large Eddy Simulation (LES) and Reynolds Averaged Navier Stokes (RANS) approaches. Turbulence in the liquid phase plays an important role in determining the void fraction distribution. Besides, turbulence parameters are used in the closure models for interfacial terms which would determine heat and mass transfer or species composition in a given system. Hence, it is necessary to model the turbulence field accurately. LES-TFM approach has produced reasonably accurate results, albeit the sub-grid scale modeling of interfacial terms remains to be validated. There is a lack of a universal approach to model turbulent bubbly flows using RANS-TFM, and the research in developing the transport equations and closure terms is extensive. At present, the choice of one model over the other is mostly ad hoc, and a systematic analysis is required to determine their applicability. In the current review, the different BIT models and their applications have been summarized. Further, some of the shortcomings in the existing approaches are identified and recommendations for future work are made based on the analysis. [ABSTRACT FROM AUTHOR]

An experimental investigation of horizontal forced vibration effect on void fraction variation of subcooled boiling flow was carried out in this study. In order to simulate the fuel assembly subchannel of a boiling water reactor (BWR), an annular test section with inner and our diameters of 19.1 and 38.1 mm was utilized for subcooled boiling tests under atmospheric pressure. The annular test section was attached to an eccentric-cam vibrator, which was driven by a low-speed motor and can produce horizontal forced vibrations with frequency up to 20 Hz and maximum displacement of 22.2 mm. The inlet liquid velocity and subcooling were set as v f , in = 0.25–1.00 m/s and Δ T Sub = 5–20 °C. Different heat fluxes of q ″ = 0.058–0.193 MW/m 2 were loaded through the center heater rod, and the void fraction and fluid temperature were measured during the tests under stationary (no vibration) and vibration conditions. Test results show that in the subcooled boiling region, the void fraction and fluid temperature can vary under horizontal forced vibrations, and the variation trends were presented in N Zu - N Sub and v f , in -〈 α 〉 plots. These variations can be explained by the potential changes of thermal boundary layers (TBL) and the heat transfer enhancement under vibrations. In addition, no significant change of void fraction and fluid temperature was found in or near the saturated boiling conditions under vibrations. [ABSTRACT FROM AUTHOR]

High fidelity 3D computational simulation for gas–liquid two-phase flow is getting important to simulate various thermal–hydraulic phenomena in nuclear related components. Single-phase computational simulation techniques have reached to a certain level of reasonable prediction accuracy for the purposes of designs and performance analyses of industrial equipment. However, two-phase computational simulation techniques have not reached to the level of reliable prediction due mainly to the difficulty in the modeling of interfacial transfer terms. Accurate modeling of the interfacial forces including the interfacial area modeling is one of the keys to predict the distribution of gas phase in various two-phase flow systems successfully. This paper is aiming at reviewing the interfacial force modeling including recent advance of the interfacial area transport equation. This paper discusses the frame-work of bubble-wall collision force which is potentially used in place of the wall lubrication force applicable for laminar flow. This paper also discusses the frame-work of the bubble collision force considering the effect of the bubble coalescence on the bubble collision frequency. The review of existing interfacial force models including newly developed bubble-wall collision force and recent advance of the interfacial area transport equation offers the most advanced knowledge of constitutive equations necessary for improving the predictive capability of the two-phase flow computational simulation codes. [ABSTRACT FROM AUTHOR]

Drift-flux parameters have been often used to formulate one-dimensional interfacial drag force in dispersed two-phase flow, which is one of key parameters to predict void fraction using one-dimensional thermal-hydraulic codes. This approach is called “Andersen approach”, which has been widely used in one-dimensional nuclear thermal-hydraulic system analysis codes such as TRACE, RELAP5 and TRAC-BF1. However, the current formulation of one-dimensional interfacial drag force ignores important void fraction covariance and relative velocity covariance when local interfacial drag force is converted to one-dimensional interfacial drag force. The impact of neglecting void fraction covariance and relative velocity covariance on one-dimensional interfacial drag force and relative velocity has been discussed in detail. In view of the importance of the drift-flux parameters, void fraction covariance and relative velocity covariance on one-dimensional formulation of the interfacial drag force, three constitutive equations have been developed for upward boiling two-phase flow in a vertical pipe. The validity of the modeled void fraction covariance and relative velocity covariance for subcooled and bulk boiling flow in a vertical pipe has been verified by boiling R12 data taken in a vertical pipe with the diameter of 19.2 mm under the pressure simulating prototypic nuclear reactor thermal-hydraulic conditions. The correlation of void fraction covariance agrees with the boiling flow data in the vertical pipe with the mean absolute error, standard deviation, mean relative deviation and mean absolute relative deviation being 0.828, 3.43, 10.3% and 33.5%, respectively. The correlation of relative velocity covariance agrees with the boiling flow data in the vertical pipe with the mean absolute error, standard deviation, mean relative deviation and mean absolute relative deviation being −0.00394, 0.0663, −0.184% and 5.11%, respectively. Due to the great importance of the void fraction covariance and relative velocity covariance on one-dimensional interfacial drag force formulation, it is highly recommended to include the void fraction covariance and relative velocity covariance in the one-dimensional formulation of the interfacial drag force used in nuclear thermal-hydraulic system analysis codes. [ABSTRACT FROM AUTHOR]

The void fraction prediction is critical in nuclear safety analyses. Available one-dimensional computational codes heavily depend on the two-fluid model. The interfacial drag force in the momentum equation of the two-fluid model is currently formulated using the drift-flux model. Thus, the drift-flux model is critical for advanced two-phase flow analysis computational codes. The drift-flux model is characterized by two essential parameters: the distribution parameter and the drift velocity. The distribution parameter expresses the covariance of the area-averaged product of a void fraction and mixture volumetric flux. Flow channel geometry is a significant factor affecting the distribution parameter. The drift velocity expresses the relative velocity between gas and liquid phases. A bubble shape regime affects the drift velocity. The bubble shape regime is susceptible to flow regime dependence on the bubble Reynolds number. Bubbles in a spherical or distorted particle regime (group-1 bubbles) dominate in bubbly flow at a low void fraction, whereas bubbles in a slug or cap bubble regime (group-2 bubbles) play a critical role in beyond-bubbly flow. Whether slug bubbles or cap bubbles appear depends on the channel size. Thus, flow channel size significantly affects the drift velocity. The drift velocity dynamically changes along the two-phase flow, transforming from bubbly to beyond-bubbly flows. The present study first formulates the evolution process of the drift velocity using the two-bubble-group approach. A two-group-based formulation is simplified to model the drift velocity behavior at the bubbly-to-beyond-bubbly flow transition. A scheme is proposed to explicitly calculate the distribution parameter and drift velocity using operating parameters. A drift-flux correlation with the new drift velocity model is evaluated by 750 data collected from 8 different sources. Tested fluid systems are nitrogen-water, air-water, and steam-water systems. Flow channel diameter ranges from 0.0508 to 0.305 m. Test section height (L)-to-channel diameter (D) ratio L / D ranges from 9.41 to 130. Superficial gas velocity is changed from 0.0100 to 11.2 m/s, whereas superficial liquid velocity is changed from 0 (pool condition) to 2.6 m/s. Operating pressure covers from 0.1 to 4.6 MPa. A systematic evaluation demonstrates the validity of the newly developed drift velocity model for a wide range of test conditions. This model is simple enough to be implemented into existing one-dimensional computational codes without any substantial code architecture change. • A new drift-flux correlation was developed for medium-to-large-size channels. • Drift velocity was modeled by the two-group bubble approach. • The applicable range covered pressures from 0.1 to 4.6 MPa. • The correlation was validated for channel sizes up to 0.305 m. • The correlation was applicable to adiabatic and boiling two-phase flows. [ABSTRACT FROM AUTHOR]

The two-fluid model is comprised of the mass, momentum, and energy equations for liquid and gas phases separately. The terms of interfacial transfer govern mass, momentum, and energy transfer across the gas-liquid interface. The interfacial area concentration is critical to formulating the interfacial transfer terms. The one-dimensional Interfacial Area Transport Equation (IATE) has excellent potential in dynamically predicting the interfacial area concentration. The sink and source terms modeling for the interfacial area concentration is vital in the successful IATE development. The current IATE in bubbly flows considers the sink terms caused by bubble coalescence owing to random bubble collision and wake entrainment. The source term is caused by bubble breakup owing to turbulent impact. The turbulent diffusion term has not been considered in the current practice of one-dimensional IATE. This study developed the equation to predict the turbulent diffusion terms in the one-dimensional IATE for bubbly flow in vertical round channels. The predictive capability of the developed equation was validated with the collected database. The non-dimensionalized turbulent diffusion terms predicted by the developed equation agreed with those directly calculated by the experimental data. The developed turbulent diffusion term was also implemented in the one-dimensional IATE. The interfacial area concentrations predicted by the one-dimensional IATE with the turbulent diffusion term had a satisfactory agreement with the experimental results by the statistical parameters of the mean relative deviation of 3.76% and the mean absolute relative deviation of 10.1%. • The interfacial area transport equation (IATE) research was conducted. • Turbulent diffusion term for one-dimensional IATE in bubbly flows was developed. • The developed equation of turbulent diffusion term was validated with experimental data. • The IATE with turbulent diffusion term was validated with experimental data. [ABSTRACT FROM AUTHOR]

In order to investigate the potential seismic vibrations effect on two-phase flow in an annular channel, experimental tests with air-water two-phase flow under horizontal vibrations were carried out. A low-speed eccentric-cam vibration module capable of operating at motor speed of 45–1200 rpm ( f  = 0.75–20 Hz) was attached to an annular channel, which was scaled down from a prototypic BWR fuel sub-channel with inner and outer diameters of 19.1 mm and 38.1 mm, respectively. The two-phase flow was operated in the ranges of 〈 j f 〉 = 0.25–1.00 m/s and 〈 j g 〉 = 0.03–1.46 m/s with 27 flow conditions, and the vibration amplitudes controlled by cam eccentricity ( E ) were designed for the range of 0.8–22.2 mm. Ring-type impedance void meters were utilized to detect the area-averaged time-averaged void fraction under stationary and vibration conditions. A systematic experimental database was built and analyzed with effective maps in terms of flow conditions (〈 j g 〉-〈 j f 〉) and vibration conditions ( E - f and f - a ), and the potential effects were expressed by regions on the maps. In the 〈 j g 〉-〈 j f 〉 maps, the void fraction was found to potentially decrease under vibrations in bubbly flow regime and relatively lower liquid flow conditions, which may be explained by the increase of distribution parameter. Whereas and the void fraction may increase at the region closed to bubbly-to-slug transition boundary under vibrations, which may be explained by the changes of drift velocity due to flow regime change from bubbly to slug flows. No significant change in void fraction was found in slug flow regime under the present test conditions. [ABSTRACT FROM AUTHOR]

The demand of accurate prediction for two-phase flow behavior in a nuclear reactor core composed of fuel rod bundles or a heat exchanger composed of heat transfer tube bundles requires comprehensive understanding of flow regime, void fraction, heat transfer and pressure drop. In comparison with the great success in developing the two-phase flow regime transition criteria for simple geometries such as pipes, annulus and rectangular channels, limited researches have been performed for developing the flow regime transition criteria of upward two-phase flow in vertical rod bundles. In the vertical rod bundles, slug bubbles spanning the bundle casing cannot exist due to their surface instability and the two-phase flow characteristics in the vertical rod bundles are different from those in pipes, annulus or rectangular channels whose channel size are smaller than the length scale of the surface instability. This study has proposed a new flow regime transition criteria model based on the analysis on the underlying physics of the upward two-phase flow behavior in the vertical rod bundles. A reliable drift-flux correlation to predict void fraction in the vertical rod bundle developed recently has been used in modeling the flow regime transition criteria. This study has classified the flow regime into 6 distinct flow regime such as bubbly, finely dispersed bubbly, cap-bubbly, cap-turbulent, churn and annular flows. The newly developed flow regime transition criteria have been compared with existing 4 flow regime maps obtained in vertical rod bundles. The fluid systems include air-water and steam-water. A fairly good agreement with some discrepancies has been obtained between the newly developed transition criteria and the measured flow regime maps. [ABSTRACT FROM AUTHOR]

A steam generator thermal-hydraulic code based on homogeneous flow model has been useful based on its numerical stability and simpler formulation. One of key parameters for a steam generator thermal-hydraulic analysis is void fraction which determines two-phase mixture density and affects two-phase mixture velocity. These parameters are important for a heat transfer tube vibration analysis. A void fraction-quality correlation is very important to accurately convert the quality into the void fraction. The void fraction-quality correlation should preferably be applicable to parallel and cross flows in rod or tube bundles since two-phase flow in the steam generator encounters flow configuration change from the parallel flow along the tube bundle in the riser section of the steam generator to the cross flow in the U-bend section of the steam generator. A set of correlations depending on flow configuration such as parallel and cross flows, rod or tube array pattern and mass flux is developed based on legacy Smith correlation. The correlation agrees with the parallel and cross flow data with the mean absolute error (or bias) of 0.117% and the standard deviation (random error) of 2.26% and with the mean absolute error (or bias) of 0.760% and the standard deviation (random error) of 6.21%, respectively. The correlations are further simplified to a single correlation applicable for parallel and cross flow in rod or tube bundles. The Smith correlation with a modified constant entrainment parameter e being 0.5 is recommended for predicting void fraction in the steam generators. The Smith correlation with e = 0.5 is expected to be applicable for parallel and cross flows with various rod or tube array patterns including normal square, parallel triangular and normal triangular arrays. [ABSTRACT FROM AUTHOR]

In relation to void fraction prediction of cross-flow in horizontal tube bundle of shell-tube heat exchangers, a drift-flux correlation has been developed to meet the demand on the study of two-phase flow gas and liquid velocities, two-phase pressure drop, heat transfer, flow patterns and flow induced vibrations in the shell side. Two critical parameters such as distribution parameter and drift velocity have been modeled. The distribution parameter is obtained by constant asymptotic values and taking into account the differences in channel geometry. The drift velocity is modelled depending on the density ratio and the non-dimensional viscosity number. The relationship between the channel averaged and gap mass velocity has been discussed in order to obtain the superficial gas and liquid velocities in the drift-flux correlation. The newly developed drift-flux correlation agrees well with cross-flow experimental databases of air-water, R-11 and R-113 in parallel triangular, normal square and normal triangular arrays with the mean absolute error of 1.06% and the standard deviation of 4.47%. In comparison with other existing correlations, the newly developed drift-flux correlation is superior to other studies due to the improved accuracy. In order to extend the applicability of the newly developed drift-flux correlation to void fraction of unity, an interpolation scheme has been developed. The newly developed drift-flux correlation is able to calculate the void fraction of cross-flow over a full range with different sub-channel configurations in shell-tube type heat exchangers. [ABSTRACT FROM AUTHOR]

Several passive safety systems are included in the design of the Economic Simplified Boiling Water Reactor (ESBWR) safety system. To investigate the capability of the ESBWR passive safety system, the integral facility designed for ESBWR was applied to perform a series of experiments simulating Loss of Coolant Accidents (LOCAs) break events of the ESBWR. This paper is intended to summarize the LOCAs test results performed at the Purdue University Multi-Dimensional Integral Test Assembly for ESBWR application (PUMA-E) facility, namely, the Main Steam Line Break (MSLB), Bottom Drain Line Break (BDLB), Feed Water Line Break (FWLB) and Gravity Drain Line Break (GDLB). Furthermore, this study focuses on the comparison of important parameters in the main components such as Reactor Pressure Vessel (RPV), Containment, and Suppression Pool (SP) of the integral facility for those types of the LOCAs tests in three different phases of LOCAs, namely, the blowdown phase, Gravity Driven Coolant System (GDCS) period and long term cooling PCCS period. The various responses of passive safety function by GDCS, Isolation Coolant System (ICS), and Passive Containment Coolant System (PCCS) at the three different phases of LOCAs to the four types of the LOCAs tests are discussed and analyzed. [ABSTRACT FROM AUTHOR]

1521 frictional pressure drop data points were collected from 12 literatures. The database included adiabatic and diabatic systems composed of 10 working fluids. The diameter range was from 0.1 to 3 mm, and the pressure drop ranged from 1.26 kPa/m to 2 MPa/m. 17 existing single channel two-phase pressure drop correlations including homogenous flow model and separated flow model were evaluated with the database. The comparison results showed that the correlations of Pamitran et al., Hwang and Kim, Mishima and Hibiki and Zhang et al. gave better predictions than the others. However, the performance evaluation results also showed the mean absolute errors higher than 40%. Based on the relatively large prediction error by the existing correlations, a new correlation was proposed by classifying the flow conditions into four regimes (namely, 1: gas laminar-liquid laminar, 2: gas laminar-liquid turbulent, 3: gas turbulent-liquid laminar, 4: gas turbulent-liquid turbulent). The newly developed correlation is expressed by a function of the two-phase Reynolds number, Re tp , the two-phase viscosity number, N μ t p , and the gas quality, x , and was able to predict the two-phase frictional pressure drop with the mean absolute percentage error of 17.4%. The correlation demonstrated an excellent performance for predicting the two-phase frictional pressure drop in mini/micro single channels. [ABSTRACT FROM AUTHOR]

The prediction capability of the two-fluid model for gas–liquid dispersed two-phase flow depends on the accuracy of the closure relations for the interfacial forces. In previous studies of two-phase flow Computational Fluid Dynamics (CFD), interfacial force models for a single isolated bubble has been extended to disperse two-phase flow assuming the effect in a swarm of bubbles is similar. Limited studies have been performed investigating the effect of the bubble concentration on the lateral phase distribution. Bubbles, while moving through the liquid phase, may undergo turbulence-driven random collision with neighboring bubbles without significant coalescence. The rate of these collisions depends upon the bubble approach velocity and bubble spacing. The bubble collision frequency is expected to be higher in locations with higher bubble concentrations, i.e., volume fraction. This turbulence-driven random collision causes the diffusion of the bubbles from high concentration to low concentration. Based on experimental observations, a phenomenological model has been developed for a “turbulence-induced bubble collision force” for use in the two-fluid model. For testing the validity of the model, two-phase flow data measured at Purdue University are utilized. The geometry is a 10 mm × 200 mm cross section channel. Experimentally, non-uniform inlet boundary conditions are applied with different sparger combinations to vary the volume fraction distribution across the wider dimension. Examining uniform and non-uniform inlet data allows for the influence of the volume fraction to be studied as a separate effect. The turbulence-induced bubble collision force has been implemented in ANSYS CFX. The assessment results show agreement with the measured data, correctly capturing the redistribution of volume fraction downstream with uniform and non-uniform inlet profiles. In particular, for the non-uniform data, the transverse redistribution of volume fraction at downstream locations is captured. This signifies the importance of bubble–bubble collision phenomena in correctly predicting volume fraction distributions. [ABSTRACT FROM AUTHOR]

In view of the practical importance of the interfacial area transport equation in flow analysis based on the two-fluid model, especially for flow analysis in a reactor core which has a relatively small channel configuration, the dominant mechanism in the interfacial area transport in small diameter pipes was firstly identified by evaluating the experimental data obtained for pressurized bubbly flows under microgravity, where the source and sink terms of the interfacial area concentration due to bubble expansion and wake entrainment collision are negligible. The sink term of the interfacial area concentration of bubbly flow in small diameter pipe was modeled in terms of the shear collision mechanism. The one-dimensional interfacial area transport equation with the developed sink term was evaluated by using 74 datasets obtained for adiabatic gas-liquid bubbly flows, including upward flows and microgravity flows in pipes with inside diameters of 9.0, 5.0 and 3.0 mm. A detailed discussion was provided on the interfacial area transport mechanism in small-diameter pipe. The overall standard deviation of the absolute error between predicted and measured interfacial area concentrations was 11.8%. The present model will reproduce the interfacial area transport of bubbly two-phase flows in small-diameter pipes with inside diameters of 3.0–9.0 mm, at superficial liquid and gas velocities over the ranges 〈 j f 〉 = 0.073–0.949 m/s, and 〈 j g 〉 = 0.00434–0.0500 m/s, and also over a wide range of axial distances, z / D = 5–188. [ABSTRACT FROM AUTHOR]

Downward two-phase flow is observed in light water reactor accident scenarios such as loss of coolant accident (LOCA) and loss of heat sink accident (LOHS) due to loss of feed water or a secondary pipe break. Hence, a comprehensive literature review has been performed for the co-current downward two-phase flow with information on the flow regime transitions and flow characteristics for each regime in the downward flow. The review compares the experimental data of the flow regime map and the current available transition models. Objectivity of the data varies on the method utilized as a certain degree of subjectivity is still present in the most objective method. Nevertheless, experimental data through subjective methods such as direct visualization or analysis of a wire mesh sensor (WMS) data were still studied in this review. Despite the wide range of flow regime data for numerous pipe sizes, a consensus was not reached for the effect of pipe sizes on flow regime transition. However, it is known that a larger pipe results in greater degree of coalescence at lower gas flow rates ( Hibiki et al., 2004 ). The introduction of a flow straightener at the inlet led to less coring and fluid rotation and inevitably, reduced bubble coalescence. This also resulted in the disappearance of the kinematic shock wave phenomenon, contrary to an inlet without a flow straightener. The effect of flow inlet, flow location, pipe diameter and bubble interfacial forces on the radial distribution as well as bubble coalescence and breakup rate are studied. Moreover, the interfacial area concentration and the bubble coalescence and breakup mechanisms are shown to vary in the axial direction as well as with flow rate, flow area and pressure drop. The liquid velocity field, bubble shape and shear stress are studied for a stationary slug bubble with downward liquid flow. Furthermore, the relationship between the plug and foam flow shape profiles, relative velocity, void fraction and gas slug velocity at an elevated pressure of 0.2 MPa studied by Sekoguchi et al. (1996) are also analyzed, together with the five plug flow sub-regime groups located in the low slip and high slip velocity regions. For the annular flow, the relationship between liquid film thickness, entrainment mechanisms, film velocity and shear stress are studied as well. Alike to plug flow, five sub-regimes in the annular flow are also examined along with the bubble and droplet entrainment mechanisms. The paper also discusses the pressure drop for bubbly, slug, foam, falling film and annular flow regimes, with a particular focus on the most accurate interfacial friction factor correlation for annular flow and its applicability for a wide range of pipe diameters. The flow instability of a system such as static and dynamic instability in the presence of a downcomer, for both single and parallel heated channels are examined too. Finally, the most accurate and versatile drift-flux correlation applicable to all downward flow regimes is highlighted and compared to drift-flux type correlations as it will be a stepping stone to attain a more accurate co-current downward flow transition model. Further experimental effort is essential to achieve a strong foothold in the understanding of co-current downward two-phase flow, as it is vital for nuclear engineering applications. [ABSTRACT FROM AUTHOR]

In the initial event on a Loss of Coolant Accident (LOCA) of a typical Boiling Water Reactor (BWR), mostly non-condensable gas is forced through blowdown pipes submerged downwardly in the Suppression Pool (SP). The void penetration length in the SP is considered as a key parameter to design an alignment of Emergency Core Cooling System (ECCS) pump suction pipe in a pool for avoiding the problem of the gas intrusion to the pump. Therefore, it is very important to predict accurately the void penetration length in the SP during a LOCA. However, the existing models or correlations to predict the void penetration length in a water pool have been developed using test data with small diameter submerged tubes less than or equal to 3.5 cm. In this study, new correlations to predict the axial void penetration length during the initial and the quasi-steady phase air injection, respectively, have been developed particularly for a large diameter submerged pipe cases using the previous experimental results with large diameter submerged pipe of 7.6 and 10.2 cm. The developed correlations consider a bubble plume developed in a liquid medium as a lumped model and use the field equations with necessary assumptions to determine the equation of void penetration length. The predictions of the newly developed correlation are tested by comparing the computed results with available experimental data from the previous studies. The comparisons show a good agreement between them. [ABSTRACT FROM AUTHOR]

• Experiments to investigate the flow features in a vertical narrow rectangular channel are performed. • A database of two-group bubble local parameters in the two-phase flow is presented. • Abrupt and gradual axial two-group bubble evolutions at flow regime transition happen. • The two-group interfacial area transport equation (IATE) is evaluated using the narrow rectangular channel data. • The IATE evaluation results show that the cap-bubbly and slug flow prediction needs improvement. This study has experimentally investigated flow characteristics of two group bubbles in upward two-phase flows in a vertical narrow rectangular channel with 0.993 mm in gap, 40 mm in width, 2000 mm in height, and 1.94 mm in equivalent hydraulic diameter, D h. The local measurements are performed at seven axial positions (z / D h =52, 104, 155, 207, 415, 622, 829 (z : the axial distance from the inlet) using a high-speed camera. The database of the locally evolving parameters has been established by processing the images from the high-speed camera measurements. The measured parameters include the void fraction, interfacial area concentration (IAC), Sauter mean diameter and number density of group-1 and -2 bubbles, and gas-phase velocity of all bubbles. A total of 14 flow conditions are set at four different superficial liquid velocities ranging from 0.214 m/s to 2.08 m/s and different superficial gas velocities ranging from 0.0755 m/s to 1.70 m/s at the inlet. The observed flow regimes cover the bubbly, cap-bubbly, and slug flows. The measured void fraction ranges from 3.92 % to 42.6 %. The axially developing features of the two-phase flows are analyzed for the three flow regimes. Abrupt and gradual axial two-group bubble evolutions with intensive bubble coalescence at the flow regime transition are, respectively, observed at low and high superficial liquid velocity conditions. The newly-obtained experimental database has been used to evaluate the existing two-group interfacial area transport equation (IATE) and its corresponding source and sink models for the two-phase flow in the vertical narrow rectangular channel. Group-1 bubble IAC mean relative error of 21.1% and group-2 bubble IAC mean relative error of 33.4% have been obtained in the evaluations. The existing two-group IATE cannot satisfactorily predict the intensive bubble coalescence and the intragroup transport of the two-group bubbles at the transition from bubbly to slug flow in the vertical narrow rectangular channel. Therefore, it is necessary to improve the existing models of the intergroup and intragroup bubble interactions in the existing two-group IATE in the future. [ABSTRACT FROM AUTHOR]

• Void fraction data for the tight-lattice rod bundle at various operation conditions were collected. • The dependency of drift-flux parameters on physical properties and bundle geometries were assessed. • The existing drift-flux correlations developed for the reactor safety analysis for rod bundles were reviewed. In view of the energy system design involving two-phase flow in various industrial apparatus, the void fraction is one of the most critical parameters, which characterizes gas fraction in a two-phase mixture. Accurate prediction of a void fraction is indispensable to estimate two-phase mixture levels and represent the flow behaviors in a given flow channel. A drift-flux type correlation has been utilized in existing one-dimensional thermal-hydraulic system analysis codes to calculate various two-phase flow parameters in a rod bundle. Due to the demand for higher performance of equipment, a tight-lattice configuration of the rod array has been given increased attention. The tight-lattice rod-bundle geometry has a much smaller rod diameter-to-rod pitch length than a typical rod-bundle utilized in equipment, such as compact heat exchangers, advanced steam generators, and next-generation nuclear reactors. As a result, it is expected that the flow characteristics may differ from conventional rod-bundle geometries. Therefore, developing the reliable void fraction prediction model for accurately simulating the two-phase flow behavior in a tight-lattice rod-bundle will be essential from the safe, economical, and robust design aspect. The present study extensively reviews existing drift-flux correlations applicable to rod-bundle geometries. Their performances were assessed against available experimental data collected from test facilities equipped with tight-lattice configuration, and applicability towards tight-lattice bundle geometry was considered. [ABSTRACT FROM AUTHOR]

Natural circulation driven nuclear reactors are prone to flow instability during the startup transients. This paper intends to provide the state-of-the-art reviews on the theoretical analysis and experimental studies on flow instability in three types of natural circulation driven reactors, ranging from conventional nuclear reactors to small modular reactors. Brief overviews of three categories of startup flow instability, i.e., density wave oscillations, flashing instability, and Geysering instability, are provided. A critical review is conducted for the scaling analysis and design of small scaled test facility. The review of obtaining quasi-steady state stability maps in the dimensionless stability plane through frequency domain analysis and experimental tests provides the state-of-the-art methodology of analyzing the flow instability. Experimental startup instability during different initial startup procedures is reviewed. Although extensive efforts have been made to study the flow instability, further work is required to improve the scaling ability of experimental investigation and the accuracy of code analysis. Some discussions for future research directions are given. [ABSTRACT FROM AUTHOR]

In relation to recent V & V standard stream, accurate benchmark of two-phase flow CFD (Computational Fluid Dynamics) using interfacial area transport equation is indispensable to validate the interfacial area transport equation, interfacial force models and two-phase flow turbulence models. This is particularly true for future detailed three-dimensional nuclear reactor core analysis. This paper intends to provide the state-of-the-art review on the two-phase flow CFD using the interfacial area transport equation. This paper discussed the basics of two-group interfacial area transport equation and component sink and source terms of interfacial area concentration such as bubble interaction terms and wall nucleation term. Brief overviews of interfacial force models and two-phase flow turbulence models are also provided. Due to the importance of databases for benchmarking two-phase flow CFD using the interfacial area transport equation, databases utilized for the CFD simulations have been summarized. A critical review has been conducted for vertical upward two-phase flow CFD simulations for adiabatic bubbly flows, boiling bubbly flows and adiabatic flows in cap-bubbly, cap-turbulent and churn-turbulent flow regimes. The review of extensive existing two-phase flow CFD using the interfacial area transport equation provides the state-of-the-art knowledge of the advancement and the current status of the two-phase flow CFD and its limitation at present. Although extensive efforts have been made for benchmarking the two-phase flow CFD using the interfacial area transport equation in the past decade, further efforts to improve the prediction accuracy of the two-phase flow CFD should be made for general use. Some discussions for future research directions are given. [ABSTRACT FROM AUTHOR]

This study performed a survey on existing correlations for interfacial area concentration (IAC) prediction and collected an IAC experimental database of two-phase flows taken under various flow conditions in large diameter pipes. Although some of these existing correlations were developed by partly using the IAC databases taken in the low-void-fraction two-phase flows in large diameter pipes, no correlation can satisfactorily predict the IAC in the two-phase flows changing from bubbly, cap bubbly to churn flow in the collected database of large diameter pipes. So this study presented a systematic way to predict the IAC for the bubbly-to-churn flows in large diameter pipes by categorizing bubbles into two groups (group 1: spherical or distorted bubble, group 2: cap bubble). A correlation was developed to predict the group 1 void fraction by using the void fraction for all bubble. The group 1 bubble IAC and bubble diameter were modeled by using the key parameters such as group 1 void fraction and bubble Reynolds number based on the analysis of Hibiki and Ishii (2001, 2002) using one-dimensional bubble number density and interfacial area transport equations. The correlations of IAC and bubble diameter for group 2 cap bubbles were developed by taking into account the characteristics of the representative bubbles among the group 2 bubbles and the comparison between a newly-derived drift velocity correlation for large diameter pipes and the existing drift velocity correlation of Kataoka and Ishii (1987) for large diameter pipes. The predictions from the newly-developed two-group IAC correlation were compared with the collected experimental data in gas–liquid bubbly to churn flow regimes in large diameter pipes and their mean absolute relative deviations were obtained to be 28.1%, 54.4% and 29.6% for group 1, group 2 and all bubbles respectively. [ABSTRACT FROM AUTHOR]

In view of an important role of a one-dimensional drift-flux correlation in nuclear thermal-hydraulic system analysis codes, several drift-flux correlations such as Lellouche–Zolotar, Chexal–Lellouche, TRAC-BF1 and Ozaki correlations have been reevaluated by rod bundle test data taken in FRIGG and NUPEC test facilities. The mean absolute error of void fraction representing a correlation bias of the Lellouche–Zolotar, Chexal–Lellouche, TRAC-BF1 and Ozaki correlations are, respectively, −1.0, 0.5, −6.3 and −3.3% for the FRIGG test data and 2.0, 2.3, −0.4 and −0.7% for the NUPEC test data. The effects of unheated rods, axial and radial power distributions, large unheated center rod and geometry of a shroud or casing on void fraction are identified. The presence of unheated rods with similar size of other heated rods tends to increase a distribution parameter in a drift-flux correlation, whereas the presence of a large unheated center rod tends to decrease the distribution parameter. The axial and radial power distributions do not have significant effect on void fraction within the tested axial and radial power distribution range. The Ozaki correlation is recommended for predicting void fraction in a BWR core but it is suggested to reduce the distribution parameter in the Ozaki correlation if a large unheated center rod exists in the core. It is indicated that drift-flux correlations developed based on bounded rod bundle test facility data may overestimate the distribution parameter for a PWR core. [ABSTRACT FROM AUTHOR]

• An extensive literature survey on inverted flow regime is carried out. • Existing interfacial heat transfer correlations interfacial frictional drag models were reviewed. • The inverted annular flow correlations utilized in modern reactor safety analysis codes are reviewed. Understanding the inverted annular flow (IAF) boiling regime is crucial for various engineering disciplines. The IAF is a post-dryout, or post-CHF film boiling regime in forced convective flow in tubes or channels. The flow structure of the IAF is characterized by the liquid core present in the channel center, which is separated from the surrounding heating source by a vapor film envelope. It has been studied extensively as one of the film boiling regimes since 1950s. In the core reflooding phase of the light water reactors, in the event of the loss-of-coolant-accident (LOCA), one of the dominant heat transfer modes is film boiling. During that process, various flow regimes are generated in the flow channel, including the IAF. The IAF appears immediately downstream of the pre-CHF flow regime and initiates at the quench front. It gradually progresses to an inverted slug and dispersed flow as it travels downstream. For the improved safety of engineering systems, including modern and advanced nuclear reactors, proper utilization of heat transfer and drag coefficients of IAF for the postulated LOCA is crucial. The present study extensively reviews the existing correlations proposed for IAF heat transfer and drag. In addition, the flow regime transition criteria applicable for IAF, based on the liquid jet-breakup concept, are reviewed. IAF correlations utilized in modern reactor safety analysis codes will also be revisited. [ABSTRACT FROM AUTHOR]

The bubble departure characteristics are studied in flow boiling considering the available experimental data in the literature. The current bubble departure diameter and departure frequency models are reviewed and compared with the collected database, revealing the current modeling shortcomings in forced convective flows. Based on an energy balance approach at the heated surface, the important dimensionless groups are identified and correlated for the bubble departure diameter and departure frequency. These dimensionless groups are the Jakob number, Boiling number, density ratio, and Prandtl number. The bubble departure diameter data also suggests some effect of the flow geometry in mini-channels. The newly proposed semi-empirical models for the bubble departure diameter and departure frequency are shown based on the available database to be accurate to within ±22% and ±35% respectively. This work also identifies important future experimental considerations for the bubble departure characteristics including data at elevated pressures, larger Prandtl numbers, and around the boundary between mini and conventional channels. [ABSTRACT FROM AUTHOR]

The concurrent upward two-phase flow of air and water in a long vertical large diameter pipe with an inner diameter ( D ) of 200 mm and a height ( z ) of 26 m ( z / D = 130) was investigated experimentally at low superficial liquid velocities from 0.05009 to 0.3121 m/s and the superficial gas velocities from 0.01779 to 0.5069 m/s. The resultant void fractions range from 0.03579 to 0.4059. According to the observations using a high speed video camera, the flow regimes of bubbly, developing cap bubbly and fully-developed cap bubbly flows prevailed in the flows. The developing cap bubbly flow appeared as a flow regime transition from bubbly to fully-developed cap bubble flow in the vertical large diameter pipe. The developing cap bubbly flow changes gradually and lasts for a long time period and a wide axial region in the flow direction, in contrast to a sudden transition from bubbly to slug flows in a small diameter pipe. The analysis in this study showed that the flow regime transition depends not only on the void fraction but also on the axial distance in the flow and the pipe diameter. The axial flow development brings about the transition to happen in a lower void fraction flow and the increase of pipe diameter causes the transition to happen in a higher void fraction flow. The measured void fraction showed an N -shaped axial changing manner that the void fraction increases monotonously with axial position in the bubbly flow, decreases non-monotonously with axial position in the developing cap bubbly flow, and increases monotonously again with axial position in the fully-developed cap bubbly flow. The temporary void fraction decrease phenomenon in the transition region from bubbly to cap bubbly flow can be attributed to the formation of medium to large cap bubbles and their gradual growth into the maximum size of cap bubble and/or cluster of large cap bubbles in the developing cap bubbly flow. In order to predict the N -shaped axial void fraction changing behaviors in the flow regime transition from bubbly to cap bubbly flow, the existing 12 drift flux correlation sets for large diameter pipes are reviewed and their predictabilities are studied against the present experimental data. Although some drift flux correlation sets, such as those of Clark and Flemmer (1986) and Hibiki and Ishii (2003) , can predict the present experimental data with reasonable average relative deviations, no drift flux correlation set for distribution parameter and drift velocity can give a reliable prediction for the observed N -shaped axial void fraction changing behaviors in the region from bubbly to cap bubbly flow in a vertical large diameter pipe. [ABSTRACT FROM AUTHOR]

• The channel size effect on drift-flux parameters is detailed. • Critical size such as 30 times Laplace length does not apply to non-circular channels. • The bundle distribution parameter depends on bundle casing geometry and rod array. • The channel size effect is more pronounced at low pressure and low flow conditions. • The channel size has less impact for beyond-bubbly flows. Gas-liquid two-phase flows are utilized in various heat and mass transfer processes that appear in numerous industrial apparatus. Typical examples are chemical reactors, heat exchangers, nuclear reactors, deaerators, condensers, etc. The numerical simulation of gas-liquid two-phase flows, which is crucial for efficient, safe, and optimized apparatus design, requires an accurate model developed based on the physics of the two-phase fluid dynamics. The two-fluid model is considered the most accurate two-phase conservation equations used in computer simulation codes to predict the thermal-hydraulic behavior of two-phase flows. The drift-flux parameters, such as the distribution parameter and drift velocity in the drift-flux model, are utilized in formulating the interfacial drag force in the two-fluid model-based simulation codes. The drift-flux model is an insightful model considering the difference between gas and liquid velocities. The effect of phase and velocity distributions on the void fraction is considered through the two drift-flux parameters. The distribution parameter and drift velocity are critical parameters in the two-phase flow formulation through the two-fluid model, which are the backbone of thermal-hydraulic analysis codes. The constitutive equations for the distribution parameter and drift velocity developed for medium-size channels reach a mature level but do not apply to industrial-size channels or large-size channels. The complicated two-phase flow dynamics in large-size channels affect the distribution parameter and drift velocity significantly. The distribution parameter in a large-size channel increases at low pressure and low liquid flow conditions due to induced secondary flow in the large-size channel. The drift velocity also increases due to cap bubbles formed by the surface instability of large bubbles. Multi-dimensional two-phase flow dynamics observed in large-size channels complicate the distribution parameter and drift velocity modeling. Thus, the drift-flux modeling in large-size channels is a much more subtle task than that in medium-size channels. This fact has driven the research to establish the drift-flux type correlations by modeling the distribution parameter and drift velocity in large-size channels. The current paper aims to provide state-of-the-art knowledge of the current status of the recent development of the drift-flux type correlations in large-size channels. The discussed items cover the formulation of the one-dimensional drift-flux model, typical drift-flux correlations developed for medium-size channels, distribution parameters for subcooled and saturated boiling flows, unique two-phase flow dynamics in large-size channels, critical size at the boundary between medium and large-size channels, and existing drift-flux correlations for large-size channels. The flow channel geometries discussed in the current paper are circular, annulus, rectangular, square, vertical rod bundle, and horizontal tube bundle. The future research direction is also discussed. [ABSTRACT FROM AUTHOR]