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The critical heat flux (CHF) is a vital parameter influencing the safety and efficiency of reactor cores. In this study, the Eulerian two-fluid model coupled with the extended wall boiling model in STAR-CCM+ was employed to simulate the departure from nucleate boiling (DNB) phenomenon in a 5 × 5 pressurized water reactor (PWR) fuel rod bundle channel with spacer grids under non-uniform heating conditions. The transition in boiling curves was used as the criterion of DNB occurrence, while the temperature distribution of rod surfaces was utilized for CHF location predictions. The predicted CHF value and CHF location exhibited good agreement with the experimental data. The deviation between calculated and experimental CHF values was within 15% and the deviation between predicted and experimental CHF locations was within one grid-to-grid span length. The results of this study suggested good prospects for the application of two-phase CFD model in predicting CHF in fuel assemblies with spacer grids. [ABSTRACT FROM AUTHOR]

The PRIUS program was established to generate an experimental database for the 6 × 4 and 12 × 6 rod bundle geometry. The database will be used to address the subchannel and CFD code analysis required for modeling and validation. This is necessary because Small Break Loss of Coolant Accident (SBLOCA) and Intermediate Break Loss of Coolant Accident (IBLOCA) present three-dimensional phenomena in the core due to the radial power profile, crossflow, and diffusion-dispersion. Therefore, specific experimental programs are required, especially during core reflooding, to investigate the large-scale three-dimensional effects. However, validating each sensitive model of the code separately in the presence of 3D effects is not possible due to the inability to implement instrumentation at high pressure and temperature steam-water flow conditions. The PRIUS test program uses a single-phase flow test to simulate a non-homogeneous velocity distribution and provide information on crossflow with radial mixing effects between subchannels. The CUPID code, which uses a macroscopic turbulence model, has been validated using the PRIUS-II experimental database. Existing macroscopic turbulence models were also validated for their prediction capabilities with different inlet flow conditions. However, the validation revealed significant errors in the shear region between subchannels. An improved macroscopic turbulence model showed promising results in predicting turbulence kinetic energy in porous media analysis.

The evaluation of available Look-Up Tables for prediction of heat transfer coefficient distribution in rod bundles cooled by supercritical water with the aim of their further use in computational analyses of various Fuel Assembles of Supercritical Water-Cooled Reactors is made. The comparison between the calculations based on Look-Up Tables with the values from empirical correlations and experimental data for smooth and wire-wrapped rod bundles was presented. The obtained results showed that Look-Up Table of the University of Ottawa, which was created to describe improved and deteriorated heat transfer regimes in round tubes, allows describing available data points with 30 % of the mean square deviation. It is noted that the presence of wire intensifies heat transfer exchange near pseudocritical temperature region but existing versions of Look-Up Tables cannot take into account this effect. Nevertheless, there is potential for further improvement in predicting the heat transfer coefficient using Look-Up Table by introducing additional correction factors.

AbstractAs part of an ongoing integrated research project (IRP), detailed investigations of the flow characteristics of an 84-pin hexagonal rod bundle representing a potential modular gas-cooled fast reactor design have been performed. The rod bundle features a pitch-to-diameter ratio of 1.5 and a large central guide tube, along with simple spacer grids and an outer enclosure. The primary focus of the current work is modeling and simulation of this geometry using multiple computational techniques. These include high-fidelity large eddy simulation (LES) and advanced Reynolds-averaged Navier-Stokes (RANS) approaches.The flow predictions are validated at a Reynolds number of 12 000 using matched index of refraction particle image velocimetry experimental data that were also generated by Texas A&M University as part of the IRP. The comparison data include velocity magnitude and turbulent kinetic energy (TKE) at different lateral locations and at multiple distances downstream of the spacer grid. Many characteristics of the experimental flow are replicated in the LES and RANS runs, and a general agreement between simulation and experiment is shown.Except for the immediate vicinity of the grid, LES is able to capture the velocity magnitude within a 10% error and the TKE to within the experimental uncertainty. The LES shows notably better agreement with the experimental data than RANS, particularly regarding the TKE decay behavior downstream of the grid. Both methods show significant departures from the experiment in their predictions close to the central guide tube. The experimental and high-fidelity data will be used to inform closures for novel multiscale methodologies. The long-term goal of the work is to use the high-fidelity data to yield improved multiscale thermal analysis techniques for solving fuel performance problems of direct relevance to industry. [ABSTRACT FROM AUTHOR]

The critical heat flux (CHF) is a vital parameter influencing the safety and efficiency of reactor cores. In this study, the Eulerian two-fluid model coupled with the extended wall boiling model in STAR-CCM+ was employed to simulate the departure from nucleate boiling (DNB) phenomenon in a 5 × 5 pressurized water reactor (PWR) fuel rod bundle channel with spacer grids under non-uniform heating conditions. The transition in boiling curves was used as the criterion of DNB occurrence, while the temperature distribution of rod surfaces was utilized for CHF location predictions. The predicted CHF value and CHF location exhibited good agreement with the experimental data. The deviation between calculated and experimental CHF values was within 15% and the deviation between predicted and experimental CHF locations was within one grid-to-grid span length. The results of this study suggested good prospects for the application of two-phase CFD model in predicting CHF in fuel assemblies with spacer grids.

Affected by sea wind and waves, the floating nuclear power plant (FNPP) would be in rolling motion during operation. The rolling motion will introduce fluctuated inertial force periodically, resulting in periodic thermal and hydraulic parameters variation, and may threaten the safety of the reactor. To investigate the influence of rolling motion on the heat transfer characteristics in the rod bundle channel under natural circulation condition, an experimental system with a 5×5 rod bundle channel was design and built on a rolling platform. An experimental investigation was conducted to study the influence of rolling motion on heat transfer in 5×5 rod bundle channels under rolling motion. The transient heat transfer characteristics of natural circulation in rod bundle were obtained. Experimental results demonstrate that the rolling motion induces periodic oscillations in the flow rate of the system. Furthermore, it is observed that with an increase in heating power, there is a reduction in the relative amplitude of these flow rate fluctuations. Consequently, this leads to periodic variations in Nu, accompanied by a corresponding decrease in relative fluctuation amplitude as well. The fluctuant flow disturbances disrupt the boundary layer and yield an enhancement of approximately 5% in the heat transfer coefficient. Regarding the local heat transfer characteristics, the incorporation of rolling motion introduces secondary flow which can significantly alter them. The presence of gravity and centrifugal force field fluctuations with a period half that of the rolling period leads to a local Nu curve with two peaks. With the center of the rod bundle channel as the axis, reverse phase fluctuations occur on both sides of the channel. The heat transfer performance is enhanced in subchannels near the rolling axis due to the rolling motion, but it diminishes away from that axis. The secondary flow induced by rolling motion amplifies velocity fluctuation amplitude in subchannels close to the rolling axis. Local Nu increases with an increase in power, whereas relative fluctuation amplitude decreases as power rises. The troughs of Nu are almost the same under different rolling periods, and the peak Nu fluctuates observed in subchannels become more pronounced as the intensity of rolling motion intensifies. In conclusion, this paper’s experimental results significantly contribute to advancing our understanding of heat transfer in rod bundle channels under oceanic conditions. The experimental results of this paper can provide experimental data for the study of wall temperature distribution of the rod bundle channel under rolling motion conditions.

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

Fast and accurate evaluation of flow and heat transfer phenomena in rod bundles is a problem of long-standing interest in nuclear engineering. Computational fluid dynamics (CFD) can provide accurate but relatively time-intensive estimates, such that simulations of very long transients with high spatial detail are infeasible. On the other hand, subchannel codes require relatively low computational time and can provide pin-level estimates, but have substantial empiricism in evaluating aspects such as crossflows and turbulent mixing coefficients. A multiscale method (SC+) for bridging this accuracy/speed gap, based on the Subchannel CFD (SubChCFD) method of Liu et al. [Nucl. Eng. Design, Vol. 355, paper 110318 (2019)], is demonstrated and developed here with a focus on a 5 × 5 square rod bundle geometry. The method has been newly implemented into the commercial code STAR-CCM+ and benchmarked against Liu et al.'s data. The 5 × 5 geometry was chosen in part due to the availability of direct numerical simulation (DNS) data at a relevant Reynolds number of a similar configuration for comparison. The SC+ results are variously compared against results from DNS, large eddy simulation, wall-resolved Reynolds-averaged Navier-Stokes, and coarse-mesh CFD methods. As an expansion to the original SubChCFD approach, a simple Hi2Lo approach is demonstrated using the DNS data as a correction to the original friction factor correlations employed. This is verified to improve the predictions. Additional test cases with geometric perturbations are pursued, illustrating the flexibility of SC+. The potential of this method for modeling the narrow gap vortex instability, which would represent an advancement over standard subchannel approaches, is also assessed. The method is expanded to include transverse flow losses, which was found to improve the results for modeling the gap instability. Initial extensions of SC+ for hexagonal rod bundles are also presented; some inaccuracies for the coarsest meshes prompted a detailed investigation of the mesh convergence behavior of the method. Geometric correction factors were devised that provided substantial improvement on these very coarse meshes, improving the prospects of SC+ for wider usage. Future work plans are to expand the methodology to wire-wrapped rod bundles and to implement the method into Pronghorn, with a unified pipeline via the Cardinal wrapper between the codes NekRS, Pronghorn, and BISON, to solve fuel performance problems of direct interest to industry. [ABSTRACT FROM AUTHOR]

Flow circulation and heat removal through shield and reflector assemblies can have major impacts on safety in long transients for sodium fast reactors (SFRs). These transients are typically categorized by reduced flow rates and large-scale organized flow patterns, including potential intra-assembly circulation. Such low-flow cases can provide challenges for experiments because of complications in measuring the flow rates and temperatures with high accuracy in different areas. This consequently also raises the uncertainty of many modeling approaches for these phenomena. In an effort to address some of these issues, high-fidelity large eddy simulations are performed using the highly parallel solver NekRS. A 19-pin configuration of a tight-lattice wire-wrapped hexagonal bundle (pitch-to-diameter ratio = 1.07), representing a prototypical internal configuration of a shield assembly, was investigated. The sodium flow was set at a bundle Reynolds number of 2000, with simulations being performed for modified Richardson numbers of 0.0 (i.e., no buoyancy), 0.01, and 0.04, where mixed-convection effects are anticipated. The flow and temperature fields for these cases are discussed in detail. The high-fidelity data should prove useful as reference data for expanding and improving on various reduced-resolution approaches. A basic framework for combining subchannel and computational fluid dynamics methodologies in SFRs is also presented, with preliminary results from simulations of light water reactor bundles and a discussion of changes that need to be made for potential application to SFRs. [ABSTRACT FROM AUTHOR]

This study reports the effects of the grid spacer with mixing vane (MV) on gas velocity for air single-phase flow and liquid film thickness for air-water two-phase annular flow in a 3 × 3 rod bundle channel. To investigate the effects of the shape of the spacers, three kinds of spacers were installed into the channel: spacer without MV (without MV), spacer with four MVs (4-MV30), and spacer with two MVs (2-MV30). The 4-MV30 and 2-MV30 had vanes that were inclined 30 deg from the vertical axis. Gas velocity was measured with a hot-wire anemometer and liquid film thickness was measured with the constant electric current method. From the gas velocity measurement, it was found that the irregularities in the velocity distribution became smaller toward downstream of the spacer. For the 2-MV30, the flow distribution was asymmetric and the flow was biased. From the liquid film thickness measurement, it was found that the liquid film thickness became thicker downstream of the spacer. [ABSTRACT FROM AUTHOR]

High-resolution two-phase flow data in the rod bundle are important in the development and validation of high-fidelity models for computational fluid dynamics and subchannel codes, in particular, those pertaining to light water reactor cooling systems. The Michigan Adiabatic Rod Bundle Flow Experiment (MARBLE) has been constructed as a modular assembly of an 8 × 8 lattice rod bundle to simulate scaled pressurized water reactor and boiling water reactor subchannel assemblies. To establish a high-spatial resolution database of the void fraction in the reactor fuel assembly geometries, tomographic measurements were performed with the High-Resolution Gamma-ray Tomography System, which was designed and built in house; the detector system has a spatial resolution of less than 1.0 mm using 240 LYSO (Lu1.8Y0.2SiO5) scintillators with a fan-beam array. In the present study, the local void fraction was measured with the MARBLE facility under various air-water flow conditions (jg = 0.04 to 0.85 m/s and jl = 0.12 to 0.77 m/s) covering from bubbly to cap-turbulent flows. The local void fraction was also successfully measured under nonuniform and asymmetric air bubble distribution conditions with an investigation of the effect of spacer grids and mixing vanes on void drift across subchannels. [ABSTRACT FROM AUTHOR]

Copyright of Atomic Energy Science & Technology is the property of Editorial Board of Atomic Energy Science & Technology and its content may not be copied or emailed to multiple sites or posted to a listserv without the copyright holder's express written permission. However, users may print, download, or email articles for individual use. This abstract may be abridged. No warranty is given about the accuracy of the copy. Users should refer to the original published version of the material for the full abstract. (Copyright applies to all Abstracts.)

• A noise-removal method for X-ray CT data collected in rod bundles was developed. • Characteristics of void fraction covariance in rod bundles were analyzed. • Characteristics of relative velocity covariance in rod bundles were analyzed. • Covariance correlations for rod bundle channels were developed. • Correlation applicability was examined for rod bundles with a large water rod. Accurately modeling gas-liquid two-phase flows in rod bundles is critical for process engineering, equipment optimization, and safety assessments involving rod bundle channels. Void fraction covariance and relative velocity covariance are key parameters that indicate the effect of non-uniform void fraction distribution on the value of area-averaged relative velocity. Reliable correlations for these covariances are essential for accurately modeling two-phase flows in system analysis codes. Developing constitutive relations for void fraction and relative velocity covariance requires accurate experimental data. Under light water-cooled reactor conditions, X-ray computed tomography (CT) is suitable for measuring gas-liquid two-phase flows. However, measurement noise is often present in the void fraction distribution data collected by X-ray CT. To address this, a post-processing algorithm was developed to remove noise from the X-ray CT data for rod bundle channels under BWR prototypical conditions. Based on the optimized data, constitutive correlations for void fraction covariance and relative velocity covariance in rod bundles were established for both bulk and subcooled boiling flows. The mean absolute relative deviations of these correlations for void fraction covariance and relative velocity covariance were 1.03 % and 0.956 %, respectively. Furthermore, the developed correlation was validated for a rod bundle containing a large water rod, with mean absolute relative deviations of 0.851 % for void fraction covariance and 1.31 % for relative velocity covariance. [ABSTRACT FROM AUTHOR]

• A two-phase flow parameter database downstream of MVSG was set up. • Two-phase flow rules and interface transport features after MVSG were revealed. • Additional IAC source/sink term with MVSG effect was separated using IATE. This study investigated the axial two-phase flow evolution and interfacial area transport characteristics downstream of the mixing vane spacer grid (MVSG) in a tight lattice rod bundle channel with a subchannel hydraulic diameter of D h = 17.27 mm. The experiment was conducted under the air-water two-phase flow at room temperature and atmospheric pressure. The phase distributions of 6 positions downstream of MVSG (Z / D h = 9.15, 14.94, 20.73, 26.52, 32.31, and 61.26) were measured utilizing a self-developed double-layer wire-mesh sensor (WMS). 42 cases were obtained, involving the bubbly flow, cap-bubbly flow, and slug flow. Void fraction, bubble velocity, interfacial area concentration (IAC), and bubble size distribution (BSD) databases were built. Under the group-1 (G-1) flow, due to the swirling flow generated by MVSG, the G-1 bubbles were drawn from the bundle surface and the gap region into the subchannel center to form a spindle core-peak distribution. BSD results demonstrate that the swirling flow promotes bubbles' coalescence. The one-dimensional (1-D) void fraction and IAC downstream of the MVSG decrease first and then recover, which contrasts with the non-mixing vane spacer grid (N-MVSG) effect. It was attributed to the increment of the bubbles' velocity after MVSG due to the bubbles' migration to the channel center and coalescence. Under the group-2 (G-2) flow, MVSG intensifies the core peak distribution of void fraction, and the void fraction profile is also spindle-shaped. The obvious bubbles' break-up occurs at Z / D h = 14.94 attributed to the swirling decay. The 1-D void fraction and IAC transport indicate, that with the liquid-velocity increment, the MVSG effect is different from the N-MVSG effect, which is mainly achieved by affecting the G-1 bubbles' dynamic behaviors. The model evaluation utilizing the interfacial area transport equation (IATE) closing bubble interaction source/sink terms indicates that the advection effect dominates the IAC evolution and the contribution of the pressure effect is weak. The remarkable bubbles' coalescence occurs under the high liquid velocity. In future studies, the additional bubble break-up and coalescence source/sink term model considering the MVSG effect should be developed based on these cases to improve the IATE prediction ability. [Display omitted] [ABSTRACT FROM AUTHOR]

Tight lattice rod bundles can increase equipment compactness and improve heat exchange efficiency, and they are widely adopted in heat exchange engineering. Understanding the flow characteristics of gas-liquid two-phase flow in tight lattice subchannels is essential for developing heat exchangers and fuel assemblies with these tight bundles. The drift-flux model (DFM) is a crucial two-phase flow model extensively used in subchannel analysis codes. Investigating the DFM in tight lattice subchannels benefits the advancement of these codes. For two-phase flow interfacial structures, significant two-group characteristics exist, with bubbles divided into small (group-one) and large (group-two) bubbles. The substantial differences in flow characteristics between these two groups provide a solid foundation for developing a two-group DFM. This study examined the two-group characteristics of two-phase flow in a tight lattice interior subchannel and developed a corresponding two-group DFM. The two-group correlations for the distribution parameters and drift velocities at the tight lattice interior subchannel level were proposed and verified using experimental data. The accuracy of the developed two-group DFM in predicting group-wise void fraction and gas velocity was also verified. The standard relative deviation between the model predictions and experimental data was 8.11 % and 13.1 % for group-one and group-two void fractions and 8.33 % and 11.7 % for group-one and group-two gas velocities, respectively. This indicates satisfactory accuracy for the newly developed two-group DFM in a tight lattice interior subchannel. • The distinct two-group flow structure in a tight lattice subchannel was discussed. • The two-group distribution parameters were modeled. • The two-group drift velocity models were experimentally validated. • The two-group drift-flux model for a tight lattice subchannel was established. • Results by this two group drift-flux model agreed with experimental data. [ABSTRACT FROM AUTHOR]

Two-phase flows are common phenomena in numerous industrial equipment. The complex and changeable phase interface determines the characteristics such as flow pattern evolution and energy-mass transport between the two phases. As a key component of rod bundle fuel assembly in nuclear reactors, the spacer grid plays important roles, including fixing the rod position, promoting mixing, and intensifying heat transfer. This study investigated the axial two-phase flow evolution and interfacial area transport characteristics downstream of the non-mixing vane spacer grid (N-MVSG) inside a tight lattice rod bundle fuel assembly under air-water two-phase flows. A self-developed double-layer wire-mesh sensor (WMS) was adopted to survey 6 positions downstream of the N-MVSG (Z / D h = 9.15, 14.94, 20.73, 26.52, 32.31, and 61.26), and 42 cases were acquired from bubbly to cap-bubbly and to slug flow. Eventually, void fraction, bubble velocity, interfacial area concentration (IAC), and bubble size distribution (BSD) databases were established. Under the group-1 (G-1) flow, some small bubbles migrate toward the subchannel center, and the void fraction gap-region peak distribution gets weakened. The 1-D void fraction and IAC after the N-MVSG increase first and then gradually recover to its upstream level. Under the two-group flow, the N-MVSG clearly cut group-2 (G-2) bubbles, and the time-averaged void fraction profile downstream of N-MVSG undergoes the evolution of the square distribution, and ring distribution, and then reverts to the upstream core peak distribution. The analysis of BSD results indicates that the G-2 bubbles' break-up is mainly caused by the increment of the flow-field turbulence, rather than the cutting of the N-MVSG. The 1-D void fraction and IAC transport show the total void fraction and IAC increase first and then recover downstream of the N-MVSG. Meanwhile, the G-2 void fraction and IAC reach the maximum before the G-1's due to the coalescence of the G-1 and G-2 bubbles at the N-MVSG's exit. The model evaluation using the IATE without bubble interaction source/sink terms shows that, for most cases, the advection effect and pressure effect dominate the void fraction and IAC evolution. The remarkable bubbles' break-up occurs under the transition cases and the two-group cases with high liquid velocity. In future studies, the additional source/sink term models considering the N-MVSG effect should be developed based on these cases to improve the IATE prediction performance. [ABSTRACT FROM AUTHOR]

• Flow regime-independent two-group void fraction model proposed for rod bundles. • Two-group void fraction model validated for rod bundles. • Appropriate rod bundle characteristic length for group two IAC modeling proposed. • Two-group Santer mean diameter model developed for rod bundles. • Two-group IAC model developed and validated for rod bundles. Interfacial area concentration (IAC) is an essential geometrical parameter in immiscible gas-liquid two-phase flow since the IAC influences the interfacial mass and heat transfer rates and the interfacial drag force between the two phases. The IAC characteristics heavily rely on bubble shape and size, which differ with flow regimes and flow channel geometries. Establishing the two-group IAC model to treat bubbles in two groups, based on their sizes, is the prevalent approach for straightforwardly modeling the IAC in dispersed flows in various channel geometries. Spherical and distorted bubbles are classified as group one. Cap, slug, and churn-turbulent bubbles are classified as group two. The present study targeted developing the two-group IAC model for dispersed flows in rod bundles. The fidelity of the predicted two-group IAC depends on the robustness of the two-group void fraction and Sauter mean diameter models. A robust and flow regime-independent methodology was proposed to calculate the two-group void fraction based on the validated two-group drift-flux model. Moreover, the present study established a new two-group Sauter mean diameter model by introducing a new characteristic length for estimating group two bubble size in rod bundles. The new characteristic length accounts for the large group two bubble interface distortion caused by rod arrays. The two-group IAC was calculated based on two-group void fraction and Sauter mean diameter models, and it was validated against IAC datasets in dispersed flows in an 8 × 8 rod bundle. The model provides high accuracy and strong reliability in predicting the two-group void fraction, Sauter mean diameter, and IAC in rod bundles. [ABSTRACT FROM AUTHOR]

• The research scope has been expanded to the gap region primarily dominated by heat conduction. • The thinning effect of the spacer grid on the boundary layer thickness is analyzed quantitatively. • The effect of Reynolds number on characteristics of boundary layer has been studied. • The velocity and temperature variations upstream and downstream of the spacer grid has been studied. The flow and heat transfer performance of the coolant have a significant impact on the safe operation of nuclear reactor. In this study, the numerical simulation of the thermal–hydraulic characteristics in a 3 × 3 rod bundle has been conducted. The accuracy of different turbulence models in predicting the results is examined, and the optimal turbulence model suitable for the study is obtained. The calculated and experimental temperatures are compared and are in good agreement. The study explores the velocity and temperature variations in cross sections, as well as the velocity, temperature, and vortex structures in the subchannels. The velocity and temperature distributions in gap region are analyzed quantitatively, and the thinning effect of the spacer grid on boundary layer thickness is studied. The research scope has been expanded to the gap region, and the relevant research results can further enrich the understanding of thermal–hydraulic characteristics in the rod bundle channels. [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]

The knowledge of the phenomenology of loss of coolant/cooling scenarios in spent fuel pools is of prime interest since a huge amount of fission product can be released in accident conditions. In this context, the ASPIC facility is dedicated to the study of the heat transfers that occur at the scale of one rod bundle stored in a spent fuel pool cell. Investigations have been carried out on a full high assembly composed of 17 × 17 electrically heated rods. The water level in the cell is controlled and set, in this test series, so that about one meter of the assembly is uncovered. Therefore, the upper part of the assembly gets overheated unless some counter measure is undertaken. In this contribution, the ability of a spray system to limit the temperature rise of the uncovered part of the rod bundle is assessed. It has been observed that the water spray reduces the temperature level on the whole cross section of the assembly and lowers significantly the risk of degradation of the rod bundle. Besides, for a bundle heat power of 20 kW, a spray mass flow rate of 10 gs−1 or higher leads to rod temperatures lower than 480 °C. • Experimental investigations on spent fuel pool loss of cooling/coolant accidents. • Spray system efficiency analysis to cool down an overheated 17x17 rod bundle. • Thermal evolution analysis of the heated assembly and fluid flows [ABSTRACT FROM AUTHOR]

The supercritical water cooled reactor (SCWR) is stands out among the Generation IV reactors with high safety and economy. Understanding the fluid flow and heat transfer characteristics within rod bundles under near-critical conditions is crucial for ensuring the safety of the SCWR core. In this study, computational models utilizing the SST k-ω turbulence model and boundary-resolved grids were validated against heat transfer data for near-critical water within a rod bundle. Subsequently, the validated model was utilized to examine the impacts of system pressure, fluid temperature at inlet, mass flux, and heat flux on the heat transfer characteristics of rod bundle with spacer grids under subcritical and supercritical conditions. The study revealed that the heat-transfer coefficient under supercritical conditions is significantly larger than that under subcritical conditions. Additionally, the heat transfer coefficient increases with mass flux, while independent from pressure and heat flux. [ABSTRACT FROM AUTHOR]

A tight lattice rod (or tube) bundle is a promising design concept to enhance heat transfer efficiency in heat exchangers using a phase change. Understanding gas-liquid two-phase flow structures in subchannels is crucial for the development of advanced subchannel analysis codes. The flow characteristics in tight lattice subchannels are expected to differ from those of conventional rod bundle subchannels. This study comprehensively analyzed the flow parameters of gas-liquid two-phase flows using bubbly and transitional bubbly flow data collected in a test section simulating a tight lattice subchannel. The considered flow parameters included the distribution parameter, drift velocity, void fraction covariance, relative velocity covariance, and interfacial area concentration. These key flow parameters were obtained based on the high spatial resolution flow distribution data measured by a dual wire-mesh sensor. The effect of the characteristic channel length on a newly developed drift-flux model was discussed. Compared with the hydraulic diameter, using the channel casing size provided better prediction results for the void fraction in the tight lattice subchannel. The accuracy of the Hibiki-Ishii correlation for IAC prediction in the tight lattice subchannel was evaluated, and the deviation was within 30 %, consistent with the reported accuracy. • Distinct flow structure in a tight lattice subchannel was discussed. • Newly developed drift-flux model for the subchannel agreed well with the data. • Characteristic length for the subchannel was identified. • Interfacial area model agreed well with the data. • Covariance models were developed for the tight lattice subchannel. [ABSTRACT FROM AUTHOR]

Flow pattern is very key parameter in the study of flow characteristics and heat transfer characteristics of two-phase flow. In this Paper, two-phase boiling experiment in a 3×3 rod bundle was conducted within the following parameter range: 0≤G≤600kg/(m2⋅s),0.1≤p≤0.12MPa,10≤ΔTsub≤30℃. We first observed the four typical flow patterns, including bubble flow, bubble-churn flow, churn flow, annular flow, and the physical characteristics were studied and the corresponding differential pressure fluctuation signals measured for the four typical flow patterns. The dynamic characteristic of typical flow pattern was investigated by the of hierarchical entropy algorithm. Then the hierarchical entropy value of different flow patterns was calculated. The results indicate that the hierarchical entropy has a better ability to resist noise, and can reveal dynamics complexity of different flow pattern in rod bundle.

The crossflow is the key phenomenon in turbulent flow inside rod bundles. In order to establish confidence on application of computational fluid dynamics (CFD) to simulate the crossflow in rod bundles, three Reynolds-Averaged Navier Stokes (RANS) models i.e. the realizable k-ε model, the k-ω SST model and the Reynolds stress model (RSM), and the Large Eddy simulations (LES) with the Wall-Adapting Local Eddy-viscosity (WALE) model are validated based on the Particle Image Velocimetry (PIV) flow measurement experiment in a 5 × 5 rod bundle. In order to investigate effects of periodic boundary condition in the gap, the numerical results obtained with four inner subchannels are compared with that obtained with the whole 5 × 5 rod bundle. The results show that periodic boundaries in the gaps produce strong errors far downstream of the spacer grid, and therefore the full 5 × 5 rod bundle should be simulated. Furthermore, it can be concluded, that the realizable k-ε model can only provide reasonable results very close to the spacer grid, while the other investigated models are in good agreement with the experimental data in the whole downstream flow in the rod bundle. The LES approach shows superiority to the RANS models.

Unlike ordinary nuclear power reactors, the floating nuclear power plants are constantly affected by sea waves and are often under inclined condition. The flow and heat transfer characteristics in the rod bundle channel for inclined condition is crucial for the design and operation of floating nuclear power plants. This paper experimentally investigates the flow resistance and heat transfer in an inclined 3 × 3 rod bundle channel at 10∼15 MPa in a natural circulation system and the inclination angle is 10°–30°. The mass flux becomes smaller as the experimental loop is inclined due to the decrease of the natural circulation driving force. As the system inclination angle becomes larger, the mass flux decreases more. Re and the outlet mass quality also affect the mass flux decreasing amplitude for inclined condition. For single phase flow, the friction coefficient correlation in a rod bundle channel is developed and the MAE (mean absolute error) is 2.82 %. The friction coefficients when the channel is inclined are larger than the friction coefficients when the channel is vertical. Correlations for Nu in the static vertical rod bundle channel are also developed and the MAE is less than 2 %. The heat transfer coefficient is enhanced for inclined condition if the effect of mass flux change is eliminated. For flow boiling, inclination has little effect on the number and size of the bubbles for highly subcooled boiling, while inclination has large influence on the flow pattern for saturated boiling. When the inclination angle becomes larger, the two phase multiplier increases. As the outlet mass quality increases, the increasing amplitude of two phase multiplier will become even larger when the inclination angle is larger. The heat transfer coefficient of flow boiling increases when the channel is inclined, and the enhancement is larger when the mass flux is smaller. When the inclination angle is 30°, the boiling heat transfer coefficient could increase by 25.7 % at a low mass flux. The paper reveals the quantitive influence of inclination on both single phase flow and flow boiling in the rod bundle channel. The working pressure and temperature have reached the operating conditions of the floating nuclear power plants, and the results can be applied to the design of nuclear power floating plants and pressurized water reactor using rod bundle fuel assembly. [ABSTRACT FROM AUTHOR]

• A conductivity probe design allows detailed measurements in the near-wall zone. • The electrodiffusion method was adapted to measure shear stress in a rod assembly. • The spacer grid changes the void fraction distribution from saddle to dome-shaped. • Such redistribution leads to non-monotonic change in wall shear stress along flow. This study focuses on the experimental investigation of local characteristics of upward bubble flow in a 3 × 3 square rod bundle assembly. The experiments were conducted at Reynolds numbers ranging from 4000 to 11000 and gas void fractions up to 10%. The distribution of the gas phase around the central rod of the assembly was measured using a frontal conductivity probe, while the wall shear stress on the central rod was obtained through the electrodiffusion method. It is shown that the gas phase distribution does not exhibit the symmetry characteristic of the assembly cross-section. Furthermore, as the distance from the spacer grid increases, a reconfiguration of void fraction profiles is observed. This explains the non-monotonic dependence of wall shear stress and wall shear stress fluctuations on the distance to the spacer grid. [ABSTRACT FROM AUTHOR]

Subchannel code is one of the effective simulation tools for thermal-hydraulic analysis in nuclear reactor core. In order to reduce the computational cost and improve the calculation efficiency, empirical correlation of turbulent mixing coefficient is employed to calculate the lateral mixing velocity between adjacent subchannels. However, correlations utilized currently are often fitted from data achieved in central channel of fuel assembly, which would simply neglect the wall effects. In this paper, the CFD approach based on spectral element method is employed to predict turbulent mixing phenomena through gaps in 3 × 3 bare tight lattice rod bundle and investigate the flow pulsation through gaps in different positions. Re = 5000,10000,20500 and P/D = 1.03 and 1.06 have been covered in the simulation cases. With a well verified mesh, lateral velocities at gap center between corner channel and wall channel (W–Co), wall channel and wall channel (W–W), wall channel and center channel (W–C) as well as center channel and center channel (C–C) are collected and compared with each other. The obvious turbulent mixing distributions are presented in the different channels of rod bundle. The peak frequency values at W–Co channel could have about 40%–50% reduction comparing with the C–C channel value and the turbulent mixing coefficient β could decrease around 25%. corrections for β should be performed in subchannel code at wall channel and corner channel for a reasonable prediction result. A preliminary analysis on fluctuation at channel gap has also performed. Eddy cascade should be considered carefully in detailed analysis for fluctuating in rod bundle.

A subchannel void sensor (SCVS) acquires the two-phase flow in a rod bundle as the time-series data of cross-sectional distributions. Herein, the temperature and pressure ranges of an SCVS were extended to include the rated conditions of boiling water reactors. The improved SCVSs were installed in a 5 × 5 heated rod bundle at eight height levels. In a boiling experiment using the rod bundle, the three-dimensional distributions of the boiling two-phase flow were measured over a wide pressure range (up to 7.2 MPa). The new experimental data were compared with existing experimental data and the results of a subchannel analysis. Experimental results were consistent with those of a high-energy X-ray computed tomography study of a heated rod bundle with the same geometry and under the same heat and flow conditions as those used in our study. The subchannel analysis code reproduced the experimental results fairly well, and the obtained database is applicable for validating and improving thermal-hydraulic analysis codes. [ABSTRACT FROM AUTHOR]

An effective and accurate methodology is developed to create an inverse surrogate model for the mass reduction analysis of the rod in the rod bundle inserted in the crossflow. The performance of two surrogate modelling approaches has been evaluated. These models are the Response Surface Method and Legendre polynomial approximations. The relationship between dominant frequencies, support stiffness and rod mass derived from Computational Fluid Dynamics simulations is used as input data for approximations. The selection of sample points is implemented with a new type of orthogonal design. The results have shown that the proposed methodology can reliably replace the finite volume model and drastically reduce computational time. [ABSTRACT FROM AUTHOR]

Investigations of heat transfer characteristics in channels with supercritical fluids are crucial for power cycle applications, such as supercritical water-cooled reactors. To ensure the safety of the system, heat transfer deterioration (HTD) should be avoided. The criteria to identify the onset of HTD in heated rod bundles are still not determined. This study summarizes indicators for the onset of HTD in a rod bundle with supercritical water flowing upward based on observations from experimental data. Existing criteria for the tube flow are assessed by comparing predicted critical heat flux (CHF) values with experimentally identified values. Considering effects of different operating parameters, a new criterion for predicting HTD behavior in rod bundles is developed. The results indicate that the new criterion has a better performance than all existing criteria, with a mean average relative deviation of 0.26 %. Predicted CHF values are within ± 20 % of the respective experimentally identified CHF values. [Display omitted] • Sudden wall temperature gradient increase indicates heat transfer deterioration. • Existing criteria for predicting critical heat flux underestimate that in rod bundles. • Criteria considering only mass flux does not perform well. • Proposed new criterion performs better than existing criteria with errors within ± 20 %. [ABSTRACT FROM AUTHOR]

• Predicted the occurrence position and power of the rod bundles' CHF. • Compared the influence of cladding materials' properties on the rod bundles' CHF. • Provided a detailed analysis of the impact of grid spacers on the rod bundle' CHF. The critical heat flux (CHF) is one of the safety standards for core thermal design and is crucial for the normal operation of the reactor. The CHF problem of rod bundles with grid spacers is a hot topic in related research. The successful design of grid spacers can improve channel CHF by affecting the flow of downstream coolant, and this detection analysis can be achieved through numerical simulation methods. In addition, most simulation studies have not considered the solid region of the cladding, so there is little research on the effect of cladding materials on CHF. The role of grid spacers in rod bundle CHF is the focus of this paper, and the influence of cladding materials is another discussion direction. In this paper, following meticulous mesh generation and computational model validation, a study on the CHF of the 5 × 5 rod bundle with multiple grid spacers and different cladding materials is carried out by using CFD software. The verification results indicated that the physical model established in this study can reasonably predict the generation location and critical power value in the 5 × 5 rod bundle with grid spacers. Furthermore, the study identified that the grid spacer exerts a significant influence on coolant flow within the channel, while the weakening of the mixing effect of the grid spacer on the coolant flow is the main factor to lead the boiling crisis upstream of the grid spacer. By comparing the effects of thermal properties of different materials on CHF, it was found that when the thermal conductivity ratio of the two materials is 1.08, CHF is almost identical under the condition of a reference heat flux of 0.1 MW/m2 as a step size. In addition to thermal properties, further attention may be paid to the influence of surface characteristics of materials on CHF in future work. [ABSTRACT FROM AUTHOR]

• A newly subchannel analysis method is developed to model the bowed rod bundle geometry. • A newly developed CHF prediction method for rod bowing bundle is developed, which is independent of the empirical penalty factor and had a relatively strong versatility. • The effect mechanism of rod bowing on CHF is the variation of local parameters and the decrease of hydraulic-diameter. The penalty of fuel rod bowing on the departure from the nucleate boiling ratio limits must be considered in the thermal–hydraulic design. Different from the traditional penalty factor method, in this study, a relatively strong versatile method is proposed to predict the critical heat flux in rod bowing geometries by coupled subchannel analysis and mechanistic model. A subdivisional subchannel method is used to obtain more detailed local flow parameter and a distributed resistance model for rod bowing is developed to quantify the effect of rod bowing on local flow field. The bubble crowding model is coupled to the subchannel code to predict the CHF of the rod bowing bundles. The results showed that the present method can predict the CHF in rod bowing bundle well, even in the bowed-to-contact geometries. The effect mechanism of rod bowing on CHF is the variation of local parameter and the decrease of hydraulic-diameter. [ABSTRACT FROM AUTHOR]

This research explored novel experimental research on critical heat flux (CHF) in a hexagonal 7-rod bundle channel at low flow and low-pressure conditions. The heater rod in the experiment was composed of Inconel-625 alloy and the heated length was 390 mm, and the diameter was 37.2 mm. The experimental parameters of inlet temperature, outlet pressure as well as mass flux ranged from 39.7–71.7 °C, 130–197 kPa and 32.3–109.4 kg m2 s−1, respectively. More significantly, the influence of temperature, outlet pressure and mass flow on exit gas quality and CHF were investigated. It was found that CHF simultaneously increased with the rise of the mass flux and outlet pressure, but decreased with the drop of inlet temperature. Additionally, the exit quality increased along with the rise of inlet temperature and outlet pressure, but decreased with the increase of mass flux. Moreover, the CHF data measured in this experiment were also compared with the predicted results of several empirical correlations. The results indicated that the general correlations underestimated the experimental values, and there is a significant relative error between the look up table and the experimental values. Finally, an empirical correlation was developed to predict the CHF of the 7-rod bundle channel and showed that the proposed correlation could predict experimental results well, with a relative error of within ±10%. [ABSTRACT FROM AUTHOR]

Laser Doppler anemometry (LDA) measurement of axial flow velocity and pressure drop measurement has been carried out in a 5 × 5 rod bundle installed with two split-mixing-vane grids. The measured results are utilized to validate the CFD simulation. The realizable and non-linear k-ε turbulence model is utilized in the CFD computation, while the two-layer wall treatment is employed with both models. Mesh sensitivity investigation shows that the pressure drop is weakly affected by local mesh refinement in the spacer grid, while its effect on the velocity in the near wake of spacer grid is apparent. The validation shows that prediction on the mean axial velocity is relatively poor in the near wake of the spacer grid where the measured result shows relatively smooth distribution of axial velocity. Refinement of mesh in the spacer grid eliminates some peaks for the non-linear model. Comparing with the effect of mesh refinement, the difference caused by the turbulence models is relatively weak in the near wake region. In the far-wake region it is still difficult to judge which model shows definite superiority to the other. For pressure drop, prediction of non-linear k-ε is closer to the experiment.

Gas–liquid two-phase flow in a stagnant pool is an important phenomenon in designing and operating industrial facilities. When gas is mixed or boiling occurs in stagnant water, the actual water level appears higher than the original water level. The actual water level is called a two-phase mixture level and largely depends on the flow channel geometries, dimensions, and flow conditions. This study focuses on the influence of channel geometries, circular pipes and rod bundles, on the two-phase mixture level and its fluctuation behavior. An air–water experiment using circular pipes with inner diameters of 50 and 224 mm and 5 × 5 and 10 × 10 rod bundles was conducted, and the two-phase mixture level swell was visually observed. As the inlet gas flow rate increased, the two-phase mixture level basically increased regardless of the channel geometry. The fluctuation amplitude was remarkably increased by formulating the slug bubbles covering the entire diameter in the small pipe with a diameter of up to 50 mm. In the rod bundles and large pipe with a diameter of 224 mm, no slug bubble was sustained, and the two-phase water level and its fluctuation amplitude were relatively small compared with those of the small pipe.

Thermohydraulic behavior in spent fuel pool is quite important in evaluating safety of a nuclear reactor under accidental conditions. When the liquid level of spent fuel pool decreases, fuel rods are exposed, heated up and finally failed. In order to avoid overheating of fuel rods, spray cooling is planned. Spay water is expected to enter rod bundle and cool down fuel rod. Therefore, quantity of spray water flowing into rod bundle has to be evaluated. A series of experiments were carried out for this phenomenon. The results showed that the water flowing into rod bundles is objected to the flooding and correlated by flooding correlations. However, the parameters of flooding correlations depended upon spray water flow rate and showed higher values compared with previous correlations. [ABSTRACT FROM AUTHOR]

Subchannel code analysis is one of the key thermal-hydraulic approaches for nuclear reactor design and safety analysis. At present, subchannel codes are employed to compute local thermal-hydraulic conditions on the rod bundle fuel assemblies of nuclear reactor cores and to predict the performance of nuclear cores during normal and hypothetical accident conditions. Currently, the subchannel code is still the main tool for thermal-hydraulic analysis in the process of nuclear fuel licensing. For inter-subchannel transfer, the widely accepted key mechanisms are (1) single- and two-phase cross flow, (2) single- and two-phase turbulent mixing, and (3) two-phase void drift. Turbulent mixing has been recognized as a vortex train moving along the gap between rods. As one of the key phenomena, the turbulent mixing model has been embedded in the subchannel code for decades. Originally, the turbulent mixing model was developed based on various adiabatic and diabatic subchannel turbulent mixing tests. Numerous correlations or coefficients have been developed for different codes. For commercial applications, the large-scale rod bundle tests of thermal mixing and critical heat flux (CHF) are the main approaches to obtain a specific model for a particular fuel/spacer design. The turbulent mixing coefficient and other parameters are determined in this process for the specific mixing vane grid design. In this process, various approaches to obtain the turbulent mixing coefficient have been proposed. Conventionally, in the subchannel codes the combined bare rod mixing and spacer grid–enhanced turbulent mixing effects on coolant have been represented by the turbulent mixing coefficient. The lack of a grid-dependent directional cross-flow model has always led to the prediction bias of local condition, especially for the hot channel where CHF generally occurs. However, in recent years, modified grid models with directional diversion cross flow have been developed to improve the prediction of spacer grid performance. In recent years, owing to the very fast improvement and rapid growth of computational resources, computational fluid dynamics (CFD) has gained popularity and advancement in the model development of subchannel codes. To substitute the costly and time consuming tests, instead of a simple turbulent mixing coefficient in the lumped parameter approach, various CFD approaches for turbulent mixing model development in subchannel codes have been proposed. CFD takes great advantage of lower cost, high resolution, and versatility. Though verification and validation are still required, CFD will be a very important tool for developing turbulent mixing models for subchannel codes. In this critical review, the development and application of turbulent mixing models in various subchannel codes for liquid metal-cooled reactor analysis are reviewed and summarized. The codes, models, tests, simulations, and future modifications are reviewed in detail. [ABSTRACT FROM AUTHOR]

Rod bundle flow is an important research field related to reactor cooling in nuclear power plants. Owing to the rapid development of computerized performance assessments, interest in coolant flow analysis using computational fluid dynamics has garnered research interest. Rod bundle flow research data compared with experimental results under various conditions are thus needed. To address this, a boiling model verification study was conducted with reference to experiments. This study adopts the Reynolds-averaged Navier–Stokes equation, a practical analysis method compared to direct numerical simulation and large eddy simulation, including turbulence modeling, to predict the flow of coolant inside a rod bundle. This study also investigates void behavior in low-pressure subcooled flow boiling using a Eulerian approach (two-fluid model). The rod bundle has a length of 0.59 m and a hydraulic diameter of approximately 14.01 mm. At the cross-section at a height of 0.58 m, near the exit, numerical results were compared with the experimental values of the volume fraction of vapor and interfacial area concentration. The simulation results showed good agreement with the experimental data for six different initial conditions with constant density. [ABSTRACT FROM AUTHOR]

In view of the importance of one-dimensional thermal-hydraulic analysis for rod bundle geometry, extensive efforts to improve constitutive equations have been made in recent years. In the present review article, state-of-the-art of the rod-bundle constitutive equations for flow regime map, void fraction, covariance, and interfacial area concentration models is reviewed. Among them, the constitutive relations for covariance and interfacial area concentration models may have the potential to improve the robustness of the conventional analysis method. Some sensitivity analysis results using TRAC-BF1 code with the new constitutive equations are summarized and reviewed. [ABSTRACT FROM AUTHOR]

During operation of a nuclear power plant, the bending deformation of fuel rods usually occurs and has a substantial effect on the heat transfer and flow behavior of coolant. To assess the influence of rod bowing on thermal-hydraulic phenomena in a nuclear reactor, it is crucial to consider the void fraction characteristics in rod bundles. In this study, experiments were conducted using an irregular sub-channel impedance void meter and high-speed camera in 5 × 5 rod bundle channels with three different bow deformation ratios, as well as a normal straight rod bundle channel. The results show that when the bowing deformation of rods occurs in a rod bundle channel, bubble coalescence enhances. Bubbles which are far away from the bowing rod bundle subchannel tend to move laterally towards the region near the bowing rod bundle subchannel, leading to an increase in void fraction compared to a normal straight subchannel. Besides, as the liquid flow rate increases, the influence of the rod bowing on void fraction within the subchannels increases. Moreover, the axial development characteristics of the void fraction are increasingly affected by the bowing rod with the increase of bowing deformation ratio. [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 review of experimental and CFD methods used to analyze the thermal-hydraulic characteristics of sub-channels in rod bundle geometries is presented. • Both adiabatic and boiling two-phase flow studies are examined. • The effects of the design and flow parameters on the two-phase flow hydrodynamics and heat transfer performance are summarized. • Based on previous numerical studies, a basic CFD model for two-phase flows in rod bundle geometries is proposed. Thermal-hydraulic analysis of two-phase flow plays a crucial role in the optimal design of nuclear fuel assemblies and in nuclear safety. This paper presents the state of the art in two-phase flow hydrodynamics and heat transfer in nuclear fuel assemblies. A comprehensive review of experimental and CFD methods used in the last 30 years to analyze the thermal-hydraulic characteristics of sub-channels in rod bundle geometries under different flow conditions (adiabatic and boiling two-phase flow) is presented. The effects of design parameters, e.g. rod-pitch length, mixing vane angle, and flow parameters, e.g. pressure, mass flux and heat flux, on flow hydrodynamics and heat transfer performance are summarized. In addition, the capabilities of CFD modelling of two-phase in rod bundles are discussed. Moreover, based on the existing studies, some recommendations for future studies are given. [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]

Through the use of CFD, a thermal-hydraulic analysis was carried out on a set of 4 rod assemblies. The common wire-wrapped spacers on the fuel assemblies are wound in the same direction for all rods. Because of this design, it is anticipated that there would be an adverse flow in all of the sub-channels. In the course of this analysis, this flow was demonstrated to be real. In order to achieve desirable flow characteristics for coolant mixing, a novel configuration for wire-wrapped spacers is proposed here. This pattern designates a group of 4 rods, the rods have the wrapping system of clockwise and anticlockwise pattern. It was determined that this pattern had an exceptional mixing effect as well as a uniform flow. This pattern has a maximum temperature area that is approximately 8.5 °C lower than that of the conventional wire-wrapped fuel bundle arrangement. When compared to the regular pattern, the pressure drop caused by the new arrangement was approximately 1.9% less severe. The redesigned layout of the rod assembly improves heat transfer while also reducing the amount of pressure loss that occurs. • A CFD simulation was performed for a 2 × 2 rod bundle with a modified wire spacer pattern. • A user-defined function was created for all the thermophysical properties of supercritical water. • The new wire spacer pattern reduces the peak temperature of the measured section by 8.5 °C. [ABSTRACT FROM AUTHOR]

• The turbulent flow in a rod bundle was studied using large eddy simulation. • The first- and second-order turbulence statistics were provided. • The cross-sectional mixing and vortexes were captured and analyzed. • Turbulent energy budget was studied by decomposing the Reynolds stress. The turbulent flow and interchannel mixing in rod bundles is a concerned issue in the thermal-hydraulic design of fuel assemblies. In this paper, a large eddy simulation was performed to study the turbulent flow of water in a 2 × 2 rod bundle. The diameter of each rod and the pitch-to-diameter ratio were set to 20 mm and 1.25, respectively. The grid resolutions were carefully studied by examining the two-point correlation of velocity components along the streamwise direction, and by comparing the power spectral density with Kolmogorov's −5/3 law. The numerical results showed that the root-mean-square (RMS) velocity is highly anisotropic in the gap area. The secondary flow on the cross-section of the rod bundle was nonuniform, and the intensity is strong close to the wall where small eddies were generated. The distributions of Reynolds normal and shear stresses and the budget of primary normal stress were presented. [ABSTRACT FROM AUTHOR]