Search

AbstractReflood thermal hydraulics remains a difficult and complex subject, and understanding the physical phenomena that occur during a reflood transient is important to nuclear safety. The Organisation for Economic Co-operation and Development/Nuclear Energy Agency (OECD/NEA) Rod Bundle Heat Transfer (RBHT) project was designed to provide unique experimental data for code assessment and model development. Participants, which came from 21 international organizations, used analysis codes including APROS, ATHLET, CATHARE, CTF, MARS, RELAP5, TRACE, and SPACE to simulate the tests performed in the RBHT facility.The experimental campaign carried out within the OECD/NEA RBHT project produced data for a total of 16 reflood tests conducted in two test series. An “open” test series consisted of 11 experiments, and a “blind” test series consisted of 5 experiments. In the blind tests, only the initial and boundary conditions were provided to participants prior to simulation of those experiments. Reflood rates ranged from 0.5 to 15 cm/s, thus producing data applicable to dispersed flow film boiling and inverted annular flow film boiling. Inlet subcooling ranged from 2.8 to 80 K. Tests with variable reflood rates and oscillatory reflood rates were included in the test matrix. This paper describes the project and presents a summary of major experimental and analytical findings. [ABSTRACT FROM AUTHOR]

The accurate prediction of the fluid flow mass and the heat transfer process as well as the system response during reflood transients has long been a critical and challenging issue for reactor system safety analyses. Accurate characterization of the flow and energy transport can also significantly facilitate the various system/component design and optimization tasks. In the current study based on the U.S. Nuclear Regulatory Commission/Pennsylvania State University Rod Bundle Heat Transfer (RBHT) reflood experimental data, a comprehensive uncertainty analysis framework is developed using DAKOTA. The developed framework is used to perform an in-depth reflood model validation and verification for the subchannel analysis code COBRA-TF. In the meantime, the artificial intelligence (AI)–based machine learning (ML) model for rod cladding temperature prediction during reflood is also developed and evaluated using the current framework. Key input parametric effects for reflood thermal-hydraulic prediction include the system pressure, inlet liquid temperature/enthalpy, inlet mass flow rate, and average bundle power input. The figure of merit under consideration is the peak cladding temperature variations. It is found in the current study that, while further model improvement is needed, COBRA-TF can predict the correct parametric trends when compared with the RBHT data. On the other hand, it is challenging for the pure AI-based ML models to correctly reflect the parametric trends. Suggestions for future ML model development are provided in the end. [ABSTRACT FROM AUTHOR]

The U.S. Nuclear Regulatory Commission has been conducting thermal-hydraulic research using the Rod Bundle Heat Transfer (RBHT) facility at the Pennsylvania State University since 2001. The facility has been used for five individual test programs: forced reflood, steam cooling, mixture level swell, dispersed droplet injection, and oscillatory reflood test series. While rod bundle thermal hydraulics has been extensively studied in the past, the RBHT data have provided new insights into rod bundle phenomena especially on the effects of spacer grids. This paper provides a summary of the RBHT test program and discusses some of the major findings from this research with the emphasis on reflood thermal hydraulics and the effect of spacer grids. Of particular interest are data that enable model and correlation development. Recent efforts have focused on the evaluation of RBHT data and development of improved models and correlations suitable for systems thermal-hydraulic codes such as TRACE and RELAP. Because of detailed instrumentation on and about spacer grids, RBHT data have enabled improved models for convective heat transfer enhancement and droplet breakup. New correlations for the inverted annular and the inverted slug film boiling regimes have also been developed as an initial step toward an improved model for dispersed droplet film boiling. [ABSTRACT FROM AUTHOR]

Accurate simulation of reflood transients of a reactor core is one of the top priorities for nuclear reactor thermal–hydraulic and safety analysis. An in-depth understanding of the two-phase flow and heat transfer process during reflooding can provide extremely useful information on the design and analysis of the various safety systems as well as on the development of numerical analysis tools. In the current paper, the status of the reflood transients in the past several decades is overviewed and discussed in detail, in the hope to provide a roadmap for future relevant research. Following an introduction of the current topic, the phenomenological aspects related to reflood are discussed first. Then, studies on the two-phase flow thermal–hydraulic non-equilibrium are reviewed and evaluated critically. The flow channel pressure drop as well as the spacer grid effect constitute important parts of reflood thermal–hydraulic and safety analysis, and various studies on these topics are discussed. Another important consideration during reflood transients is the liquid droplet behavior since it affects the rod bundle thermal response. In addition, the current paper also identifies the several limitations and the need for future research work on reflood thermal hydraulics. [ABSTRACT FROM AUTHOR]

• A new two-phase flow mass quality correlation was developed. • One-dimensional conservation equations were formulated for liquid and vapor phases. • Scaling analysis was performed to obtain an expression for the vapor drift velocity. • The new correlation predicts the mass quality in DFFB regime with improved accuracy. A new semi-empirical two-phase flow mass quality correlation was developed for the post-dryout dispersed flow film boiling (DFFB) regime in a 7 × 7 rod bundle geometry. Starting from the fundamental physics describing the two-phase flow in the dispersed droplet flow regime, one-dimensional conservation governing equations were formulated for the liquid and vapor phases by considering a detailed characterization of the liquid droplet phase, based on which the liquid phase velocity was related to the key flow parameters. A scaling analysis was then performed within a sub-channel of the rod bundle to obtain an expression for the vapor drift velocity, from which a suitable relationship for the two-phase flow quality was derived using the drift flux model. It was found that the vapor drift velocity in the DFFB regime is affected by many variables including: the droplet size, velocities of the liquid and vapor phases, void fraction, vapor superheat and fluid properties, all of which are important in characterizing the dispersed droplet flow. Based on the results of the RBHT reflood tests, which simulate the reflood transients of a LWR during an accident scenario, a new mass quality correlation was developed that is applicable to the conditions with system pressure ranging from 138 to 414 kPa, rod bundle peak power between 0.98 and 1.97 kW/m, inlet liquid subcooling temperature from 11 to 83 K, and inlet flooding rate ranging from 0.0191 to 0.0508 m/s. Comparisons with experimental data indicated that the new correlation is able to predict the actual mass quality in the post-dryout two-phase flow regime with significantly improved accuracy not only for the rod bundle geometry but also for the tube geometry. [ABSTRACT FROM AUTHOR]

Highlights • Two-phase flow thermal–hydraulic non-equilibrium is investigated in detail. • An advanced image processing technique is used for liquid droplet study. • A complete data reduction methodology is formulated and employed. • It is found that significant thermal–hydraulic non-equilibrium exists between the liquid and vapor phases. Abstract The current research aims at investigating the thermal–hydraulic non-equilibrium encountered in the dispersed flow film boiling (DFFB) regime under reflood conditions. A series of constant reflood tests have been performed at the 7 × 7 Rod Bundle Heat Transfer (RBHT) test facility and the experimental results have been analyzed in detail. The data not only is useful for the development of theoretical models for two-phase interfacial mass and heat transfer, void fraction, pressure drop, and rod bundle cooling prediction, but also can be used as a benchmark data base for evaluating the performance of various thermal–hydraulic analysis codes. In this paper, a complete data reduction methodology is employed to reveal and interpret important two-phase flow phenomena during reflood transients. The results show that for the region far away from the quench front location, significant thermal non-equilibrium may exist during reflood and thus poor heat removal capability of the flow mixture is expected. On the other hand, for the region immediately downstream of the quench front, the two-phase flow is found to be close to the thermal equilibrium state, indicating that both the interfacial heat transfer and the heat transfer between the rod surface and fluid mixture are very efficient. In addition, based on the liquid droplet measurement and force balance calculations, it is found that the slip ratio between the two separate phases is a strong function of the droplet size and reflood time. With a larger droplet size and a shorter distance from the quench location, the slip ratio is larger. The liquid carryover fraction at the exit of the test section is also determined as a function of the reflood time. Results indicate that only a small portion of the injected coolant mass is stored within the bundle. A significant amount of liquid mass into the test section is either evaporated or entrained in the vapor flow and eventually left the bundle. [ABSTRACT FROM AUTHOR]

As part of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Rod Bundle Heat Transfer (RBHT) project, an experimental study was performed to investigate the entrained droplet sizes and velocities in a rod bundle under reflood conditions. Experimental results were obtained from the U.S. Nuclear Regulatory Commission/The Pennsylvania State University RBHT test facility using advanced dual laser measurement systems that allow for the simultaneous measurement of droplet behaviors at two axial locations during reflood transients. The RBHT facility is highly instrumented and contains a 7×7 electrically heated bundle with dimensions matching those in commercial pressurized water reactors. The combination of the measurement capabilities of the RBHT facility and the choice of appropriate experimental conditions allows for the measurement of unique droplet size and velocity distributions under different transient reflood conditions. [ABSTRACT FROM AUTHOR]

Heat transfer by mixed convection in a rod bundle occurs when convection is affected by both the buoyancy and inertial forces. Mixed convection can be assumed when the Richardson number ( Ri = Gr / Re 2 ) is on the order of unity, indicating that both forced and natural convection are important contributors to heat transfer. In the present study, data obtained from the Rod Bundle Heat Transfer (RBHT) facility was used to determine the heat transfer coefficient in the mixed convection regime, which was found to be significantly larger than those expected assuming purely forced convection based on the inlet flow rate. The inlet Reynolds ( Re ) number for the tests ranged from 500 to 1300, while the Grashof ( Gr ) number varied from 1.5 × 10 5 to 3.8 × 10 6 yielding 0.25 < Ri < 4.3. Using results from RBHT test along with the correlation from the FLECHT-SEASET test program for laminar forced convection, a new correlation ​is proposed for mixed convection in a rod bundle. The new correlation accounts for the enhancement of heat transfer relative to laminar forced convection. [ABSTRACT FROM AUTHOR]

A linear temporal stability of coaxial confined jets in a vertical tube involving heat and mass transfer at the interface is presented in this paper. A potential flow analysis that includes the effect of viscosity at the interface is performed in analyzing the stability of the system. Film boiling in a vertical tube gives rise to the flow configuration explored in this work. The effects of various non-dimensional parameters on the growth rate and the neutral curve are discussed. The heat transfer at the interface has been characterized by introducing a heat flux ratio between the conduction heat flux and the evaporation heat flux. Viscous forces and the heat and mass transfer at the interface are found to stabilize the flow both in the capillary instability region and Kelvin–Helmholtz instability region. Increasing heat and mass transfer at the interface stabilizes the flow to small as well as very large wave numbers. [ABSTRACT FROM AUTHOR]

This paper evaluates the performance of the U.S. Nuclear Regulatory Commission's (NRC's) thermal hydraulic code TRAC/RELAP Advanced Computational Engine (TRACE) against experimental reflood data from the NRC/Pennsylvania State University (NRC/PSU) Rod Bundle Heat Transfer (RBHT) test facility, as an integral step in verification of code accuracy. This paper is an extension of the NURETH-20 conference paper by the first author (Garrett et al., 2023) that has been recommended for consideration and submission to Nuclear Engineering and Design. An international study on reflood thermal-hydraulics, sponsored by the Nuclear Energy Agency (NEA) Working Group on Accident Management and Analysis (WGAMA), was conducted with data collected in the NRC/PSU RBHT test facility, located at the Pennsylvania State University. A series of 16 benchmark tests were conducted, with conditions covering a carefully selected range of oscillatory, variable stepped and constant rate reflood injection velocities. These unique conditions are useful for code validation and model improvement. These 16 tests were segmented into 11 open tests, followed by five blind tests. This paper covers the five blind tests as the 11 open tests were covered by Garrett et al. at the NURETH-19 conference (Garrett et al., 2021). For TRACE code benchmarking, a numerical model with the same dimensions as the RBHT facility was used. The initial and boundary conditions for this model were taken from experimental measurements. Many of the test conditions were chosen to examine sensitivities to important parameters, such as reflood liquid subcooling, reflood rate, and system pressure. The wide range of test conditions served to test the code and provide insight to its strengths and potential areas of improvement. These novel experiments were vital in this effort. Simulations were made for five reflood tests and comparisons between predicted and measured results were made for the transient cladding temperatures, vapor temperature, bundle liquid mass fraction, carryover fraction, and steam exhaust fraction. The comparison presented in this paper has provided useful insight into code improvements. Studies to more accurately model reflood phenomena are currently underway as a result of the work presented in this paper. • TRACE reflood validation study against RBHT reflood experimental data. • Validation against RBHT data suggest TRACE's droplet sizes are too large. • Novel transient reflood data under controlled forced flow conditions. • Further research from the first author and others supports these findings too. • Novel RBHT reflood entrained droplet measurements pursued from these findings. [ABSTRACT FROM AUTHOR]

This work investigates the separate effects of liquid subcooling, substrate material, and surface micro-structure on the film boiling characteristics. A quenching facility was constructed to conduct vertical quenching experiments using rods with various substrate materials and surface morphologies. The surface morphology is characterized using field emission scanning electron microscopy (FESEM). Stainless steel, zirconium, and Inconel-600 rods are used with a diameter of 9.5 mm that simulates the size of fuel rods in commercial nuclear reactors. Other Inconel-600 rods with different porosity percentages are used to study the effect of fouling on the quenching behavior. The surface temperature and wall heat flux at the surface are deducted from the temperatures measured by the thermocouples embedded inside the rods using an inverse heat conduction code, from which the heat transfer coefficient and Nusselt number are calculated. The data are used to develop a generalized heat transfer correlations that includes the effects of liquid subcooling and substrate materials. It predicted the data within ± 40 % error band. The results also suggest that the heat transfer coefficient increases gradually as the sample cools down and in higher subcooled pools. Moreover, the variation in the substrate material shows a significant effect on the heat transfer characteristics. However, the surface micro-structure impact on the film boiling regime is negligible. • Parametric study of the effects of liquid subcooling, substrate material, transient heat transfer, and surface microstructure on the heat transfer characteristics were experimentally investigated. • Surface characterization analyses were performed to obtain detailed information needed to identify the effects of surface microstructure. • The variation in the heat transfer coefficient with the heated surface temperature was studied. • New heat transfer correlations were developed to be used in film boiling regime. [ABSTRACT FROM AUTHOR]

• We review the history of the International Topical Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH) conference series. • We provide a summary of progress in the thermal-hydraulic field in the last 4 decades. The year 2020 marks the 40th anniversary of the first International Topical Meeting on Nuclear Reactor Thermal-hydraulics (NURETH-1). Hosted by the thermal-hydraulics division (THD) of the American Nuclear Society, the NURETH series is the premier topical meeting exclusively dedicated to advances in nuclear reactor thermal-hydraulics. In this article, which opens a special issue dedicated to the 40th anniversary of NURETH-1, we provide a brief history of the NURETH series. We dedicate the bulk of the manuscript to a summary of the progress in thermal-hydraulics in the past 40 years, highlighting, key contributions presented in the NURETH series of conferences. We emphasize that, given the size and complexity of the field examined, this cannot be considered a comprehensive review. However, We hope the reader will find this article useful to reflect on the advances in the past 40 years and the current state of the art. [ABSTRACT FROM AUTHOR]

Film boiling is an important phenomenon in the evaluation of an emergency core cooling system following a hypothetical loss of coolant accident in a nuclear reactor. This study investigates the effects of liquid subcooling, surface oxidation, and surface materials on the minimum film-boiling temperature . Quenching experiments were performed using stainless steel and zirconium (Zr) test samples. The samples were heated to a temperature well above then plunged vertically in various degrees of liquid subcooling pools. A visualization study using a high-speed camera was conducted to capture the quenching behavior. Additionally, surface characterization analyses including X-ray diffraction, scanning electron microscopy, and energy dispersive X-ray spectroscopy were performed to quantify the surface conditions. Results indicate that liquid subcooling has a strong influence on . The visualization study shows a very thin vapor formation around the test sample for higher subcooling pools which explains the enhancement in the heat transfer. It is observed from the surface characterization analyses that the variations in the surface condition of the stainless steel and Zr causes the vapor bubbles to depart differently in the nucleate boiling regime. Furthermore, the effect of surface oxidation is clearly noticeable in the Zr test sample compared to the stainless steel test sample due to the oxidation kinematic of each substrate material. It is found that the substrate thermophysical properties have a significant impact on . Comparing the bare substrates shows that for the same degrees of liquid subcooling pool, the value of for the Zr sample is ∼30°C to 60°C higher compared to the stainless steel sample. Moreover, increasing the degrees of liquid subcooling contributes to a significant increase in that varies between ∼50°C and 70°C for both samples. [ABSTRACT FROM AUTHOR]

Droplet size and distribution across a spacer grid in a heated rod bundle during reflood stage of a postulated design basis accident loss of coolant accident (LOCA) are studied experimentally using the Rod Bundle Heat Transfer (RBHT) test facility. Effects of spacer grid conditions, quench front location during reflood, and inlet water subcooling on droplet field are investigated. Experimental results show that the droplet sizes decrease when they pass through a dry spacer grid due to cutting and shattering effects at the lower edges of the spacer grid. On the other hand, for a wet grid, the droplet sizes downstream of the grid could be larger than those at the upstream locations, as the droplet generation mechanism, for a wet grid, is due to the formation of liquid ligaments at the trailing edge of the spacer grid and the subsequent aerodynamic breakup of these ligaments into relatively larger droplets induced by instabilities. It is found that, as the quench front propagates upward, sizes of incoming droplets increase correspondingly as a result of the droplet-vapor thermal-hydraulic interactions along the flow channels. The experimental results also indicate that larger droplet diameters occur when the inlet water subcooling is higher. Results of the present study can be utilized to develop models for droplet field behavior under accident scenarios. These models can be incorporated into nuclear reactor thermal-hydraulic safety analysis codes such as COBRA-TF and TRACE. [ABSTRACT FROM AUTHOR]

A correlation is developed for the liquid carryover fraction during both constant and oscillatory reflood with application to nuclear power accidents. The correlation, semi-empirical in nature, is based on extensive experiments in a 7 × 7 rod bundle array at the Nuclear Regulatory Commission/Pennsylvania State University Rod Bundle Heat Transfer (NRC/PSU RBHT) facility. Liquid carryover and entrainment can significantly impact quench behavior and maximum cladding temperatures during a postulated accident; therefore, accurate predictions of these phenomena are critical. The parameters associated with the stability at the liquid–vapor interface, and hence the liquid entrainment and subsequent carryover are established. These parameters provide the basic form of the correlation with experimental data from the NRC/PSU RBHT facility to determine the model coefficients. Using a weighted least-squares method that minimizes the error between modeled and experimental carryover fraction, we implemented a genetic algorithm to identify the parameters of this correlation. The proposed model is compared with experimental data and predictions from the NRC's TRACE code and lies within 20% error margin. • A liquid carryover fraction correlation is developed for both constant and oscillatory reflood. • First correlation to include effects of an oscillating flow and moving quench front. • Correlation parameters identified using a genetic algorithm and weighted least-squares method. • Correlation is compared with experimental data and predictions from NRC's TRACE code. • Correlation predicts within a 20% error margin. [ABSTRACT FROM AUTHOR]

There is no heat transfer correlation available in the Inverted Slug Film Boiling (ISFB) regime. In this study, a new correlation for film boiling in a vertical channel is proposed using a semi-empirical model. This correlation is applicable in the Inverted Annular Film Boiling (IAFB) heat transfer regime as well as ISFB heat transfer regime. A semi-empirical model is developed to establish the functional dependence of the Nusselt number on various controlling parameters. Test data obtained from the transient reflood tests carried out in a full length 7 × 7 vertical Rod Bundle Heat Transfer (RBHT) facility are used to develop the correlation. The predicted Nusselt number agrees within 15% with the RBHT data set and other data set available in the literature. This heat transfer correlation is applicable up to a void fraction of 90%, i.e., both in the IAFB and ISFB regimes. The proposed correlation is also applicable to a tube geometry using a modified coefficient in the correlation with an error less than 10%. [ABSTRACT FROM AUTHOR]

This study develops a method to characterize the elbow-region for vertical-upward-to-horizontal air–water bubbly two-phase flows through a 90° elbow with a non-dimensional centerline radius of curvature, R C / D , of three. The dissipation of the elbow-effects in the horizontal region downstream of the elbow is characterized by the change in the local void fraction distribution. To quantify the dissipation characteristics, the elbow-strength, S , is defined as the inverse of the second moment of the local void fraction distribution. It is found that the elbow-strength decreases exponentially with increasing development length in the elbow-region. The slope of decrease of the elbow-strength in the elbow-region is determined by the dissipation parameter, β . The dissipation parameter determines the development length required for the elbow-effects on the two-phase flow parameters to become negligible. It is found that the dissipation parameter is a linear function of the mixture Reynolds number for the flow conditions investigated in the present study. [ABSTRACT FROM AUTHOR]

This study investigates the geometric effects of a 90-degree vertical-upward elbow on local two-phase flow-parameters in an air–water system, and develops an experimental database for interfacial area transport modeling. The experimental facility is constructed from 5.08 cm inner diameter acrylic pipes and includes vertical and horizontal sections interconnected by a 90-degree vertical glass elbow. The elbow has a radius of curvature of 15.24 cm and is installed at L / D = 63 from the inlet. A four-sensor conductivity probe is used to measure time-averaged local two-phase flow parameters including: void fraction, bubble velocity, interfacial area concentration, and bubble frequency at ten axial locations along the test section. It is observed that the bubbles moving through the vertical-upward elbow are entrained by the secondary flow leading to a bimodal distribution in bubbly flow conditions. For the flow conditions investigated within the study, this bimodal distribution occurs regardless of the bubble distribution upstream of the elbow. It is found that the change in bubble distribution downstream of the elbow is strongly correlated to the dissipation of the elbow effects. Furthermore, the dissipation characteristics as well as the length of dissipation region for the vertical-upward elbow are found to be a strong function of the liquid-phase flow rate. [ABSTRACT FROM AUTHOR]

Abstract: The conventional thermal-hydraulic nuclear reactor system analysis codes utilize a two-field, two-fluid formulation for two-phase flows. To provide closure, static flow regime transition criteria and algebraic relations are employed to estimate the interfacial area concentration (a i ). To better reflect the dynamic evolution of two-phase flow, an experimental version of TRACE is developed to estimate the a i using the one-group interfacial area transport equation (IATE), which includes mechanistic models for bubble coalescence and disintegration. These models account for: (1) bubble break-up due to impact of a turbulent eddy, (2) bubble coalescence due to random collision driven by turbulent eddies, and (3) bubble coalescence due to the acceleration of a bubble in the wake region of a preceding bubble. To assess the impact of including bubble interaction models in TRACE, code predictions of experimental data measured by a multi-sensor conductivity probe are compared to both the IATE and flow regime based predictions. In total, 50 air–water vertical co-current upward and downward bubbly flow conditions in pipes with diameters ranging from 2.54 to 20.32cm are evaluated. It is found that TRACE, using the conventional flow regime relation, always underestimates a i by predicting a larger bubble size than observed in the experimental data. Additionally, the axial trend of the a i prediction is always linear because a i in the conventional code is predominantly determined by the pressure. However, TRACE with the one-group IATE significantly improves the prediction results, yielding a ±13.0% difference with the data. It is found that the non-linear developments observed in the experimental data, which reflect bubble interactions, are predicted well using the IATE. Moreover, in several conditions dominated by bubble interactions, the a i trend displayed is opposite to the effect of pressure. In these cases, the conventional TRACE relation predicts an incorrect trend in a i , while the IATE predicts the experimental data well. [Copyright &y& Elsevier]

Highlights: [•] Separate effects experiments on gravity-driven air-ingress in helium filled vessel. [•] Effect of pipe-breaks’ geometric parameters on air-ingress is investigated. [•] Gravity-driven air-ingress demonstrates three characteristic stages. [•] A unique non-dimensional density ratio characterizes the transition between stages. [•] Predictive models are developed for gravity-driven exchange between air and helium. [Copyright &y& Elsevier]

Abstract: The present study implements the one-dimensional interfacial area transport equation into the TRACE code, being developed by the U.S. Nuclear Regulatory Commission. The interfacial area transport equation replaces the conventional flow regime dependent correlations and the regime transition criteria for furnishing the interfacial area concentration in the two-fluid model. This approach allows dynamic tracking of the interfacial area concentration by mechanistically modeling bubble coalescence and disintegration mechanisms. Thus, it eliminates potential artificial bifurcations or numerical oscillations stemming from the use of conventional static correlations. To implement the interfacial area transport equation, a three-field version of TRACE is utilized, which is capable of tracking both the continuous liquid and gas fields as well as a dispersed gas field. To demonstrate the feasibility of the present approach, the steady-state one-group interfacial area transport equation applicable to adiabatic air–water bubbly two-phase flow is first tested in the present study. Data obtained in 18 different flow conditions from two vertical co-current upward air–water bubbly two-phase flow experiments performed in round pipes (25.4mm and 48.3mm) are used to help evaluate the implementation. Results obtained from TRACE with the interfacial area transport equation (TRACE-T) and those from TRACE without the transport equation (TRACE-NT) are compared to demonstrate the enhancement in prediction accuracy. The predictions made by TRACE-T agree well with the data with an average percent difference of approximately ±8%. It is also evident from the results that while TRACE-T accounts for dynamic interaction of bubbles along the flow field, the predictions made by TRACE-NT are attributed primarily to the pressure change. [Copyright &y& Elsevier]

• An comprehensive code evaluation and validation is carried out using COBRA-TF. • The predictions are validated by reflood data from the NRC/PSU RBHT test facility. • COBRA-TF agrees generally well with the RBHT data. • Prediction errors exist for specific thermal–hydraulic quantities, requiring further model improvement. Accurate prediction of the reflood transients during LOCA has long been a challenging task. One reason for this is the substantial cost incurred to perform large-scale reflood tests, which are formidable in experimental expenditure and intensive in technology as well as program management. Another reason preventing a comprehensive understanding of reflood transients is the high-level complexity involved in the two-phase flow mass and heat transfer processes, which makes the measurement extremely difficult and inefficient. The situation has been significantly improved with the design and operation of the Nuclear Regulatory Commission (NRC)/Pennsylvania State University (PSU) Rod Bundle Heat Transfer (RBHT) test facility. A variety of advanced instrumentations were developed and used at this facility and very high-resolution data has been obtained, especially for the liquid droplet field. In the current study taking advantage of the unique NRC/PSU RBHT data, an extensive and comprehensive code evaluation and validation is carried out using the thermal–hydraulic sub-channel analysis code COBRA-TF. The system parametric effects investigated include: the system pressure, inlet liquid subcooling temperature, inlet flooding rate and rod bundle power. A variety of thermal–hydraulic quantities predicted by the numerical code are evaluated including: the quench front propagation, cladding, spacer grid and vapor temperature variations, two-phase pressure drop, liquid droplet velocity and coolant void fraction. In addition, the prediction errors are presented for each of the quantities in great detail in order to have a comprehensive understanding of the code performance. In general, COBRA-TF agrees relatively well with the experimental data in terms of the overall quench front propagation, only predicting slightly earlier quench. However, under low inlet subcooling and high flooding rate conditions, significant discrepancies are observed. The comparison with rod bundle thermal–hydraulic parameters indicates that COBRA-TF is able to predict the cladding temperature before quench well within a 15% range. The spacer grid and vapor temperature predictions are within 20% error. While the overall droplet velocity prediction is found to be within 30% error, COBRA-TF is able to capture the decreasing trend for droplet velocity during reflood. In addition, the prediction of the dispersed flow film boiling void fraction shows that the code always under-predicts the void fraction. While the most prediction discrepancy is found to be the two-phase flow pressure drop after quench, which involves more than 50% under-prediction for the bulk liquid boiling regime. The present study clarifies the effect of different system parameters on the various two-phase flow quantities during reflood. The results obtained provide answers for many existing code modeling issues and thus can be instructive and useful for future model development and code upgrading. [ABSTRACT FROM AUTHOR]

• A new spacer grid pressure drop correlation is developed for the DFFB regime. • The model is validated by reflood data obtained from the NRC/PSU RBHT test facility. • The correlation improves the spacer grid pressure drop predication within a 25% error span. Flow blockages such as the spacer grids play an important role in a two-phase flow system as they modify the flow and heat transfer behavior. Their existence, on the one hand, increase the turbulent mixing and heat transfer capability of fluid directly downstream; while on the other hand, create additional pressure loss within the system thus requiring a detailed and comprehensive understanding. In the present work, a new two-phase flow pressure drop correlation induced by spacer grids in the dispersed flow film boiling (DFFB) regime is developed based on the unique reflood experimental data obtained from the Nuclear Regulatory Commission/Pennsylvania State University rod bundle heat transfer test facility. The test conditions cover wide ranges of system parameters for two-phase flow system which can be used for model development. It is found that the most dominating system parameter is the mixture mass flow rate across the spacer grid location. As a result, the spacer grid loss coefficient is correlated with the mixture Reynolds number and a blockage geometry correction factor. The current correlation is found to significantly improve the prediction of the spacer grid pressure drop in the DFFB regime generally within a 25% error span, while the other exiting correlations failed to capture the two-phase flow hydraulic behavior involved during the reflood transients. The current correlation developed can be readily implemented into existing thermal-hydraulic analysis codes to further improve their prediction capabilities in the DFFB regime with the presence of flow blockages. [ABSTRACT FROM AUTHOR]

• A new spacer grid induced droplet breakup correlation was developed. • Conservation of mass, energy, and important flow quantities are considered. • The NRC/PSU RBHT reflood data is used for model development. • New correlation predicts the droplet size after breakup within a 17% error span. • New correlation predicts the validation data set with RMSE of 0.1060 and standard deviation of 0.0990. The liquid droplet breakup process at the leading edge of a dry spacer grid during reflood transients is studied theoretically based on the experimental observations in the present work. Various rod bundle quenching experiments were carried out at the Rod Bundle Heat Transfer (RBHT) test facility to simulate the reflood transients of a Loss of Coolant Accident (LOCA) for Light Water Reactors. In the experiment, variation of the droplet size distributions across a spacer grid are measured and analyzed. The size of droplet generated downstream of a spacer grid is found to be affected by the condition of the spacer grid as well as the quench front propagation. In addition, the substantially reduced droplet size at downstream of the spacer grid indicates that droplet breakup takes place. Based on available experimental data obtained in the Rod Bundle Heat Transfer tests and in literature, a theoretical model incorporating two droplet groups (large and small droplet groups) is developed for the dry spacer grid droplet breakup by considering the conservation of liquid mass, energy transport, and including various important flow and heat transfer parameters including the Nusselt number, Reynolds number, Jakob number, Weber number, dimensionless radiation number, distance from the initial breakup point, incoming droplet size, and spacer grid blockage ratio. Moreover, a liquid droplet mass loss coefficient was proposed for small droplet group to specially account for the liquid mass evaporation due to interfacial heat transfer and radiation. The model so developed is found to be able to predict the RBHT data within 17% error with RMSE of 0.1060 and standard deviation of 0.0990 for the validation data set, showing significantly increased accuracy. Comparison made with other experimental results indicates that the current model developed correctly captures the mass and energy transport process during breakup. In addition, sensitivity analyses are performed to investigate the overall performance and relative importance of various parameters that are related to the breakup process in the current study. The current model developed can be incorporated into nuclear reactor thermal-hydraulic and safety analysis codes to enhance their capabilities of predicting the droplet behaviors and the fuel rod peak cladding temperature in the dispersed flow film boiling regime during reflood transients of postulated accident scenarios such as LOCA for light water reactors. [ABSTRACT FROM AUTHOR]