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• 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]

• 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]

• 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]

• A method to predict velocity profile in rod bundle has been proposed. • Gas velocity distribution can be simplified to turbulent velocity distribution. • Liquid velocity distribution can be derived by interfacial shear stress correlation. • The predictions show reasonable agreement with data. Annular flow in rod bundle occurs frequently in boiling water reactor (BWR) under normal operation and in pressurized water reactor (PWR) under accident conditions. When annular flow occurs, rod film thickness is one of the most important parameters for safety concern. In order to calculate rod film thickness distribution inside rod bundle, both gas and liquid velocity distribution inside rod bundle is necessary. Also, flow distribution inside rod bundle is of importance for calculation of interfacial shear stress and other parameters in each sub-channel. However, entrained droplets in gas core can sharpen the gas velocity profile while existing rods can flatten the gas velocity profile, making it different from that of single phase flow. The liquid velocity distribution inside rod bundle is hard to obtain from experiment because of the complex structure of its geometry. Due to these reasons, this paper deducts gas and liquid velocity profiles from existing adiabatic rod film thickness data. Even the gas velocity profiles for different flow conditions are different, there exists a gas velocity profile that can be applied to all current flow conditions with minimum uncertainties. Based on that gas velocity profile, liquid velocity profile can be calculated. From these flow velocity profiles, film thickness on each rod and interfacial shear stress in each sub-channel can be calculated. [ABSTRACT FROM AUTHOR]

• Boiling heat transfer to water in a three-rod bundle at near-critical pressure is investigated. • Circumferential wall temperature distributions are obtained. • Parametric effects on heat transfer characteristics in subcooled and saturated boiling regions are analyzed respectively. • Heat transfer correlations for the film boiling are assessed against the experimental data. Heat transfer experiments of flow boiling heat transfer in a three-rod bundle with spacer grids are performed at Shanghai Jiao Tong University. The test section consists of three heated rods and three unheated rod-like elements with an outer diameter of 15 mm and a pitch to diameter ratio of 1.2. The effective heated length is 1960 mm. The test conditions are as follows: pressure ranging from 20 to 21.5 MPa, mass flux varying from 800 to 1600 kg/(m2·s), heat flux between 400 and 700 kW/m2 and inlet quality ranging from −0.49 to 0.2. Circumferentially non-uniformed wall temperature distributions are observed and analyzed respectively in both subcooled and saturated boiling regions. Parametric effects of mass flux, heat flux and system pressure on heat transfer are discussed in detail. The heat transfer coefficient increases with the rise of mass flux and system pressure. However, due to the occurrence of DNB, the heat transfer coefficient decreases with the increase of heat flux, which is opposite to the nucleate boiling heat transfer at low pressure. Several correlations developed for the film boiling are selected for assessment against the experimental data. In the subcooled boiling, the Yagov and Dedov correlation gives the best prediction accuracy. In the saturated boiling, the prediction of Collier correlation agrees favorably with the test data. [ABSTRACT FROM AUTHOR]

• Pulsatile flow through rod bundles was investigated. • The axial and lateral velocities in a cycle were analyzed. • The accelerating and decelerating half-cycles are different. • Lateral velocity in pulsatile flow is lower than in steady flow. • Turbulence intensifies with increase in the mean velocity. The possibility of unsteady flow through rod bundles is currently gaining attention from researchers. Velocity components of pulsatile flow (time-averaged Reynolds number = 0.7–4.0 × 103) through 5 × 5 square array bare rod bundles (P/D = 1.326 and W/D = 1.268) has been investigated experimentally. The responses of axial and lateral velocities to the sinusoidal imposed pulsation were different. The axial velocity responded directly to the imposed pulsation while the lateral velocity diminished. Also, the response in the subchannels and gaps was different. The axial velocity in the gaps followed the imposed trend, while at the subchannel, the acceleration was steep but the deceleration steady. Lateral velocity in the gaps and subchannels differs by an average of 40%. Turbulence was higher on the rod surfaces compare to other points. The dissimilarity in the acceleration and deceleration half-cycles suggests that there is a difference in the instantaneous parameters during acceleration and deceleration. Further in situ measurements of the pressure field is a potential follow-up study. [ABSTRACT FROM AUTHOR]

• Flow resistance characteristics are obtained and fitted under vertical conditions. • Inclination increases the frictional resistance and correlations are given. • Natural circulation friction factors fluctuation differs from forced circulation. • A correlation is proposed to predict transient resistance for rolling conditions. Flow resistance characteristics in a 3 × 3 rod bundle channel are investigated under static vertical, inclined and rolling motion conditions. The circulation mode hardly influences the frictional and local resistance and the transition Reynolds number under forced and natural circulation conditions are both around 800. Correlations are proposed to predict the frictional and local resistance coefficients. Under inclined conditions, the friction factors are slightly larger than vertical values due to the radial convection and increase with the inclination angles for the same Reynolds number. Since the flow resistance increases clearly in the laminar regime, a correlation allowing for the inclination angles is proposed to predict the friction factors in the laminar regime under inclined conditions. The spacer grid local resistance is barely influenced by the inclination instead. The natural circulation flowrate and pressure drop fluctuations are quite different from the forced circulation results in previous studies. The amplitude is much larger and no obvious phase difference is observed under natural circulation conditions. The friction factor variation rules are different from the forced circulation results as well. A new correlation is proposed considering the rolling parameters through dimensionless numbers and rolling Reynolds number. It is tested with a mass of experimental data and applicable to friction factors prediction under given conditions. [ABSTRACT FROM AUTHOR]

• TR-PIV provides high-resolution data in time and space of rod bundle channel. • Mean flow properties and turbulent parameters nearby the spacer grid are discussed. • Swirl strength and PSD reveal the quasi-periodicity vortex downstream of the grid. • Turbulence scale is dealt with this paper through cross-correlation analysis. The performance of the CFD prediction of the flow details in the fuel assembly using Reynolds-averaged Naiver-Stokes (RANS) model is still poor. Evaluation and understanding of the complicated turbulent flow structure in the rod bundle channel with mixing devices are essential to optimize turbulence models. Therefore, the current study used the time-resolved particle image velocimetry (TR-PIV) as the measurement tool and combined with the matching index refractive (MIR) technology to obtain high-fidelity experimental data in pressurized water reactor (PWR) based 5 × 5 rod bundle channel. Mean flow properties, turbulent parameters, and spectral analysis are presented in this paper. The results show that a large velocity gradient appears downstream of the spacer grid. The turbulence intensity has been decayed sharply about 45 mm (about 4.5D h) downstream of the spacer grid, indicated that the mixing effect of the spacer grid becomes weakened after 4.5D h. Besides, the fluctuating velocity and instantaneous swirl strength are utilized to present the transport characteristics of the eddies downstream of the spacer grid. Combined with the power spectral density spectrum analysis of the transverse fluctuating velocity, the vortex shedding frequency of the mixing layer downstream of the spacer grid is obtained. These results indicate the existence of the quasi-periodic oscillations in the rod bundle channel. Finally, the integral scale on the rod bundle channel is calculated using the cross-correlation function of two reference points. This result reveals that the turbulence in the rod bundle channel is anisotropic. [ABSTRACT FROM AUTHOR]

Highlights • This study presents experimental studies of the single-phase resistance. • Study of resistance characteristics in rod bundles under low Reynolds number. • An improved correlation of the frictional resistance coefficient is obtained. • Introduce kinematic viscosity correction factor for correction. • This newly developed correlation predicts data with a relative error within ±5%. Abstract A considerable amount of research has gone into understanding the thermal-hydraulics characteristics of rod bundle channels which is of great importance for the nuclear industry. This paper will show that the development of the correlation of the frictional resistance coefficient in rod bundle channels under natural circulation conditions is well understood. This paper presents experimental studies of the single-phase resistance and obtains an improved correlation of the frictional resistance coefficient under natural circulation conditions based upon thermal-hydraulics characteristics which are well understood. It is also shown that this newly developed correlation closely predicts data. The experimental results show that the frictional resistance coefficient of a rod bundle channel under natural circulation is related to the physical properties of fluid and the size of the rod bundle channel. According to different flow regions, the relationship between the frictional resistance coefficient and Reynolds number is discussed separately, and the expressions of frictional resistance coefficient with Reynolds number in laminar flow region and transition region are obtained. After considering the influence of fluid properties on the frictional resistance coefficient, this paper introduces the kinematic viscosity correction factor to correct the frictional resistance coefficient. Finally, the complete expression of the frictional resistance coefficient was obtained. This newly developed predictive expression is in good agreement with experimental data in both the laminar and transition flow regions, with a relative error within ±5%. Thus we now have a predictor of the frictional resistance coefficient under natural circulation conditions based upon more complete thermal-hydraulic parameters in a static state. [ABSTRACT FROM AUTHOR]

• PIV used to study flow fields & turbulence in 5x5 rod bundle across 22 cases (Reynolds: 310-12296). • Higher turbulence intensity at lower Reynolds due to significant relative velocity gradient in rod bundle. • Transition Reynolds ≈ 900 in rod bundle, less distinct than in pipes. • Fluctuation velocity decreases with Reynolds, abrupt rise at transition Reynolds. • Non-uniform transition Reynolds leads to varying turbulence intensity in rod bundle. Utilizing the particle image velocimetry (PIV) technique, this study investigates the fully developed flow field and turbulence statistics in a 5×5 rod bundle across 22 cases, with Reynolds numbers ranging from 310 to 12296.Time-averaged, statistical, and spectral analyses were performed on the full field and instantaneous velocity vectors obtained through PIV measurements to examine the velocity distribution, turbulence characteristics, and spectral features in different subchannels under various Reynolds numbers. In conjunction with a differential pressure transmitter, the flow properties and fluid transitions were also explored.The experimental results reveal that the relative velocity gradient in the rod bundle channel is more significant at lower Reynolds numbers. As the Reynolds number increases, the relative velocity distribution becomes more uniform. A comparison of experimental results from different planes indicates that the fluctuation velocity in the gap subchannel is greater than that in the inner subchannel, suggesting that tightly arranged rods can generate higher turbulence intensity. Furthermore, the transition observed through the friction factor in the rod bundle channel is less distinct than those in pipes, with the transition Reynolds number being approximately 900. A low Reynolds number effect is present in the rod bundle channel, where the dimensionless fluctuation velocity decreases as the Reynolds number increases. Upon reaching the transition Reynolds number, the area-averaged dimensionless fluctuation velocity abruptly rises. By examining the turbulence intensity variations at different positions within the rod bundle, it becomes evident that the transition Reynolds numbers vary across different subchannel positions. Ultimately, the experimental data can be employed to validate the applicability of turbulence models for different Reynolds numbers. [ABSTRACT FROM AUTHOR]

Highlights • Assessment of RANS models is performed for a loosely spaced rod bundle with P/D = 1.32. • The flow configuration has a Reynolds number of 54600 and a Prandtl number of 0.016. • Five different linear, non-linear and differential transport turbulence models are assessed. • All the models over-predict the wall shear stress with respect to LES result. • None of the models is able to predict the gap vortex street. Abstract In the present study, a wide range of steady-state and unsteady Reynolds Averaged Navier Stokes (RANS) simulations have been performed to reproduce the turbulent flow and heat transfer in a fuel assembly of the ALFRED reactor. The considered fuel assembly consists of a loosely spaced bare rod bundle with pitch to rod diameter ratio of P/D = 1.32. It is cooled by liquid lead, hence the Prandtl number of the working fluid is very low, i.e. Pr = 0.016. The obtained results are extensively compared with the available Large Eddy Simulation (LES) results. This article is divided in two main parts. In the first part, a thorough assessment of eddy viscosity (i.e. linear and non-linear) and Reynolds stress turbulence models have been performed against the reference LES results. Flow and thermal fields have been qualitatively and quantitatively analysed. In general, RANS results are in a good agreement with the LES, however, most of the models severely overpredict wall shear stress and the corresponding pressure drop in the rod bundle. In the second part, URANS simulations have been performed to reproduce a gap vortex street in the considered loosely spaced bare rod bundle. For this purpose, two turbulence models, that gave the best steady-state RANS predictions, i.e. the k-ε cubic and the elliptic blending Reynolds-stress model, are tested as well. However, the gap vortex street in this particular loosely spaced bare rod bundle turned out to be too weak to be predicted by the URANS approach. [ABSTRACT FROM AUTHOR]

Prediction of rod film thickness and its distribution in annular flow in rod bundle is of essential importance to nuclear power industry. This is attributed to the possible film dryout on fuel rods under certain conditions, which will result in a drastically increasing cladding temperature and possible fuel damage. Though there exist a lot of models or correlations to predict film thickness for annular flow in pipes, there are few models to predict rod film thickness inside rod bundle. The one-dimensional models or correlations for pipe flow cannot be used for rod bundle directly due to different geometry and flow dynamics. In relation to this, a new approach to predict the rod film thickness inside rod bundle has been developed. This approach first divides the rod bundle cross-section into three regions according to their characteristic length scales. Then, the gas and liquid flow distribution inside rod bundle has been given with reasonable assumptions. Based on the flow distribution profile, the sub-channel area-averaged gas and liquid volumetric fluxes have been given. Finally, the correlation developed for pipe flow is adopted for each sub-channel based on the local two-phase parameters in the rod bundle. Compared to experimental data obtained in-house and found in the literature, the newly proposed approach can predict rod film thickness with satisfactory accuracy. [ABSTRACT FROM AUTHOR]

In CFD simulations of boiling flow and CHF phenomenon, the solid domains are usually neglected to reduce the complexity and computational time. However, based on the simulation of three test sections, the azimuthal heat conduction in the solid domains is found to have a significant influence on the numerical results, especially on the wall temperature distribution. This technical note tries to give the researchers some suggestions about the azimuthal heat conduction effects on boiling flow by CFD simulation of steady state flows in vertical pipes and rod bundle geometry. The CFD modeling methodology was developed using FLUENT. Eulerian two-phase flow is used to model the flow and heat transfer phenomena of boiling flow. Three boundary conditions are tested to investigate the effect of azimuthal heat conduction. The results show that the three boundary conditions are nearly equivalent to each other for subcooled boiling flow in centrosymmetric and non-centrosymmetric geometry, like vertical pipe and rod bundle. But for the CHF phenomenon in rod bundle, the simplified BC overestimates the azimuthal inhomogeneity and the peak temperature of the rod surface. During this process, azimuthal heat conduction plays a key role in this result. The conclusions require researchers should model the entire domain in the CFD simulation and set the heating boundary according to the experiment as accurately as possible to better predict the flow and heat transfer characteristics in the boiling flow. [ABSTRACT FROM AUTHOR]

Spacer grids induce local flow disturbance in fuel assembly including lateral cross flow and elevated turbulence fluctuation. In order to investigate the cross flow downstream of spacer grids and to provide the high-fidelity data for the CFD validation of rod bundle flow, a flow measurement method has been proposed to measure the flow in the full cross section of the 5 × 5 rod bundle. A two-dimensional (2D) particle image velocimetry (PIV) system is utilized for the flow measurement. A telecentric lens is employed to minimize the perspective angle when the CCD camera is located distant from the measured cross section. The matched refractive index technique is also introduced to eliminate the non-uniform laser illumination behind the rods which is resulted from the laser refraction on the convex rod surface. Such configuration is examined for the cross flow measurement of a 5 × 5 rod bundle in a small scale test facility. It has been shown that the cross flow pattern can be well depicted in the entire cross section when the measuring plane is within 10 D h downstream of a spacer grid. The obtained distribution of sampled flow velocity appears in good agreement with the normal distribution. [ABSTRACT FROM AUTHOR]

Interest in unsteady flow in floating systems has increased in recent times. This paper presents the results of the experimental investigation on the resistance characteristics of pulsating flow through square array bare rod bundles with P/D  = 1.326 and W/D  = 1.268 conducted in the range of time-averaged Reynold’s number Re ta  = 1–7 × 10 3 and dimensionless acceleration ( α ) of 0.001–0.013. Pulsatile flow affects the time-averaged friction factor because it changes the velocity field close to the wall. The dimensionless frequency parameter ( ω ′ ) relates the frequency to the viscous effects while the dimensionless acceleration ( α ) is the product of the velocity amplitude ( A u ), diameter ( D ) of the channel and the frequency ( ω ) normalized by the mean velocity ( U m ). Ratios ( C λ ) of the time-averaged friction factor (λ p ) in pulsatile flow to the friction factor (λ s ) in a steady flow and ( C p ) of their average pressure drops were used to compare the two flows. C λ and C p increase with an increase in A u , ω ′ and α ; but decreases with increase in Re ta . The dynamics of the extra turbulence generated as a result of pulsation was also analyzed in the subchannels. [ABSTRACT FROM AUTHOR]

In reactor fuel assembly in PWR, spacer grids with mixing vanes are used in rod bundles to enhance flow mixing and heat transfer. Thus, the spacer grids have significant influence on the economy and safety of reactors. This paper focuses on visualization research on the single phase flow in rod bundle with spacer grids by Particle Image Velocimetry (PIV), and a measurement of multi-subchannels in the both transverse and longitudinal directions was conducted at the Reynolds number of 10,400. Matching Index of Refraction (MIR) technique was implemented to accurately measure the internal flow field in the rod bundle. The flow statistics of rod bundle were obtained, such as mean velocity, turbulence intensity and Reynolds stress. The velocity distribution and turbulence statistics in the longitudinal flow field are presented downstream the spacer grid. The streamwise development of symmetrical vortex array in the transverse flow field is also evaluated. A comparison of Reynolds stresses in the longitudinal and the transverse flow field is performed to reveal the momentum transfer process in three-dimensional flow field. Experimental results benefit the development and evaluation of spacer grid, and also provide the whole-field validation for numerical simulations. [ABSTRACT FROM AUTHOR]

The IAEA is organizing a Coordinated Research Project (CRP) on the application of Computational Fluid Dynamics (CFD) for nuclear reactor design. Within this CRP, an open benchmark is organized on fluid mixing in rod bundles. KAERI has offered precise experimental data of the velocity field downstream of a mixed twist/split type mixing grid. Additionally, temperature measurements in various axial and azimuthal locations of a partly heated rod were made available by KAERI. The experimental set up consists of a 2000 mm long bundle of 4 × 4 rods with pitch to diameter ratio of 1.35, placed in a square housing of 142 × 142 mm. Each rod has a diameter of 25.4 mm. The mixing grid is located at 2/3 height of the channel with 5 spacer grids located upstream and 3 downstream of the mixing grid. These experiments are analysed with the TrioCFD code. Turbulence is treated with the RANS approach by using a non-linear eddy viscosity model. The comparison of measured and calculated mean velocity fields is presented. Measured profiles of velocity mean values were calculated correctly for the central sub-channel and the gap of rod-to-rod both close and far from the mixing grid. The measured temperature is underestimated by the calculations, on one hand due to the localization of the thermocouples inside the wall of the rod and on the other hand due to the used thermal boundary condition called “imposed heat flux”. A first order correction procedure helps to understand the detected discrepancy between experiment and calculation. [ABSTRACT FROM AUTHOR]

Fuel subassemblies of a fast reactor consist of fuel pin bundle with helically wound spacer wires, arranged in a triangular pitch within a hexagonal wrapper. The fuel pins are located within the subassembly. Further the subassembly comprises of a diffuser where the cross section changes from cylindrical to hexagonal, mixing plenum before the exit of pin bundle and a specially designed blockage adapter. Accurate assessment of the pressure drop in the fuel subassembly is essential to ensure adequate core cooling and design of sodium pump. Experimental determination of pressure drop characteristics in the subassembly by simulating the hydraulic condition in the subassemblies of the reactor core is considered essential as a better choice as correlations reported in the literature cannot be directly used for all the complex regions present in the subassembly. This is due to the fact that flows in the interconnecting sections are highly under developed. Further, the flow regime in a fuel subassembly varies from laminar (during shutdown heat removal under natural convection) to completely turbulent under full power condition. To understand the hydraulic characteristics of the 500 MWe Proto type Fast Breeder Reactor (PFBR) fuel subassembly, an experimental facility has been commissioned. Experiments on full scale subassembly with dummy fuel pins have been performed using water as simulant. Experiments have been conducted covering a wide range of Reynolds number encompassing laminar, transition and turbulent regimes. In the rod bundle, no abrupt changes in friction factor were observed when the flow changes from laminar to turbulent. The experimental results have been transposed using Euler number similarity, to determine the pressure drop for sodium flow in the reactor. Possible uncertainties in the experimental data have been quantified. Useful pressure drop correlations have been proposed for various section of the subassembly of a fast breeder reactor. Finally the results for fuel pin bundle region are compared with the data reported in the literature and a satisfactory agreement has been observed. [ABSTRACT FROM AUTHOR]

• The boiling two-phase pressure drop experiments of liquid sodium are conducted in a 7-rod bundle. • Based on the two-phase pressure drop data, the existing correlations for predicting the two-phase frictional multiplier of liquid sodium are evaluated. • A new friction multiplier for calculating the boiling two-phase flow in rod bundle channels is proposed. • The predicted value of the developed correlation is within 30% of the experimental data. The two-phase pressure drop characteristics of liquid sodium are very significant for the safety and serious incident program development analysis of sodium-cooled fast reactor (SFR). In this paper, the boiling two-phase pressure drop experiments of liquid sodium were conducted in a 7-rod bundle with the mass flux range of 80 ∼ 390 kg/(m2·s), heat flux up to 120 kW/m2 and the absolute pressure range of 7.5 ∼ 100 kPa. The experimental results show that the frictional multiplier of boiling two-phase liquid sodium in the rod bundle channel decreases with the increase of X LM . Some existing two-phase pressure drop correlations in the literatures were assessed and compared with the experimental data obtained in the present study. Results showed that these correlations could not predict the current experiments well because of the different geometries and measurements of high temperature environment. The new correlation for the two-phase friction multiplier of liquid sodium was developed in the 7 rod bundle. The predicted value of the new correlation was within 30 % of the experimental data in the rod bundle, which sufficiently demonstrates that the new correlation developed in this paper can well predict the boiling two-phase frictional pressure drop of liquid sodium in the rod bundle. [ABSTRACT FROM AUTHOR]

• The boiling two-phase heat transfer experiments of liquid sodium are carried out. • The variation of parameters in typical boiling process is obtained. • The influence law of thermal parameters on heat transfer coefficient is revealed. • The new correlation of the heat transfer coefficient of boiling two-phase liquid sodium is developed. Sodium boiling will occur in some hypothetical accident cases (such as additional reactivity introduction and heat discharge capacity deterioration, and Local fault propagation). Thus sodium boiling experiments are necessary for the safety analysis and the development of serious accident programs in sodium-cooled fast reactors. In the present study, the boiling two-phase heat transfer experiments of liquid sodium were carried out on the boiling liquid sodium test loop. The experimental results showed that the entire boiling process can be divided into four stages. The heat transfer coefficient of the boiling two phases in the rod bundle channel increases gradually with the increase of heat flux and grows very slowly with the increase of system pressure. Based on the 54 groups of rod channels obtained, the new correlation of transfer heat coefficient of boiling two-phase liquid sodium was developed in a 7-rod bundle. The new correlation can well predict the experimental data of boiling two-phase liquid sodium obtained by other scholars, and the prediction error is within ± 50 \% . Finally, mean relative deviation (MRD), absolute mean relative deviation (MARD), and root mean square deviation (RMSRD) were introduced to evaluate the accuracy of the correlation quantitatively. [ABSTRACT FROM AUTHOR]

• Flow resistance under rolling motion is investigated by numerical simulation. • The influence mechanism of rolling motion on friction resistance is analyzed. • Additional pressure drops mainly affect the fluctuation amplitude. Flow resistance characteristics in a 3 × 3 rod bundle channel under rolling motion are investigated by numerical simulation to overcome the limitation of experimental conditions and measurement methods. Severe rolling conditions are selected to make up for the lack of test range. The local parameters, including flow field, temperature field and wall shear stress are given to analyze the influence mechanism of the rolling motion on friction resistance of the rod bundle channel. Under rolling conditions, the absolute value of flow rate does not influence the fluctuation amplitude of frictional pressure drop when inlet flow rate does not fluctuate. Additional pressure drop is the main factor affecting the fluctuation amplitude of frictional pressure drop under rolling conditions. When the rolling motion becomes severe, the velocity deviation increases, while it has no obvious influence on velocity peak value within cross section. Increasing the flow rate weakens the effect of rolling motion on axial velocity distribution. The variation of flow rate does not affect the magnitude and distribution of radial velocity, and the radial secondary flow intensity is mainly determined by the intensity of rolling motion. Different from velocity distribution, the wall shear stress distribution is obviously affected by rolling motion and is unconnected with flow rate, while the flow rate affects the relative value of wall shear stress. Under the same rolling parameters, the variation range of boundary layer thickness increases with decreasing flow rate, which leads to an increase in the relative variation range of wall shear stress with decreasing flow rate. [ABSTRACT FROM AUTHOR]

An experimental study on heat transfer to supercritical water flowing in a 2 × 2 rod bundle test section with two channels is carried out at Shanghai Jiao Tong University. The bundle consists of four heated rods with an OD of 10 mm and a pitch-to-diameter ratio of 1.18. The 2 × 2 rod bundle is inserted into a square assembly box with rounded corners by which the test section is separated into two channels. Water flows downward in the first channel between the pressure tube and the assembly box and then turns upward in the second channel inside the assembly box to cool the four heated rods, which are directly heated by DC power. The experimental conditions are as follows: pressure ranging from 23 to 26 MPa, mass flux ranging from 400 to 1200 kg/m 2 s, heat flux between 300 and 1000 kW/m 2 , and bulk fluid temperatures ranging from 200 to 480 °C. The heat transfer through the assembly box is strongly affected by the temperature difference between the two channels. The minimum fluid enthalpy increase in the first channel appears near the pseudo-critical temperature region. Non-uniformity of circumferential wall-temperature distribution around the heated rod is observed. The effects of various parameters on heat transfer behavior inside the 2 × 2 rod bundle are similar to effects observed in tubes. Six developed correlations are compared with the test data. The correlations of Jackson and Fewster (1975) and Bishop et al. (1964) provide the best predictions of the test data among the selected correlations. [ABSTRACT FROM AUTHOR]

The characteristics of the heat transfer and two-phase flow resistance in a rod bundle channel are closely related to flow patterns. In the present study, two-phase flow patterns for vapor–water flows in a 3 by 3 rod bundle channel were obtained at atmospheric pressure and relatively low mass flow condition. Under the current experimental conditions, slug flow found in circular tubes was not observed. Comparisons with some existing flow pattern maps and transition criteria were likewise conducted. Results show that the transition boundary of circular tube deviated from the transition boundary of the boiling two-phase flow of the rod bundle channel. Using the differential pressure signal of the vapor–liquid two-phase flow, multi-scale entropy algorithm is employed to reveal the dynamic characteristics of the different flow patterns in the rod bundle. The multi-scale entropy rate could effectively classify the flow patterns in the rod bundle channel. Results suggest that multi-scale entropy can be an effective method to reveal the dynamic details of macro and local vantages. [ABSTRACT FROM AUTHOR]

CFD analysis is carried out for supercritical water (SCW) heat transfer in a 2 × 2 rod bundle with three turbulence models, i.e. the SSG, the ω Reynolds stress model ( ω -RSM) and the baseline Reynolds stress model (BSL-RSM). The CFD calculation results show that the models predict similar cross-flow pattern between the adjacent subchannels. However, there is apparent discrepancy in the predicted turbulent mixing. Due to the strong cross-flow, the bulk parameters predicted by the models agree well. Comparison between the CFD results and the experiment data reveals that the SSG underpredicts the wall temperature, while the ω -RSM and BSL-RSM tend to overpredict the wall temperature remarkably in the region where the buoyancy effect is dominant near the wall. Such difference can be mainly attributed to the modeling approach near the wall. It may be concluded that the appropriate modeling of buoyancy effect is the key factor to improve the prediction when the wall region is resolved with very fine mesh. The bulk flow parameter obtained in the CFD calculation which cannot be easily measured in the experiment are used to evaluate the SCW heat transfer correlation. An uncertainty analysis shows that the uncertainty caused by the selection of turbulence models is negligible for the correlation evaluation. [ABSTRACT FROM AUTHOR]

• A series of numerical equations and calculational methods are investigated and concluded. • The equations are implemented into the calculation of the natural circulation characteristics. • The experiments are conducted with 3 × 3 rod bundles, and the experimental data is used to validate the code. • The phenomenon of the natural circulation characteristics under motion conditions are analyzed with both the experimental and calculational results. Natural circulation system is very important in the nuclear reactor system, especially for the safety in the operation of nuclear reactor. The natural circulation characteristics are different when the reactor is based in the ocean rather than on land. In this study, a set of numerical models is established on the natural circulation characteristics of fuel rods nuclear power system under motion conditions. The models are implemented into the calculation code compiled by Fortran 90, which includes the mass flowrate model of the natural circulation system, the heat exchanger model, the preheater model, the tube model and the Nitrogen pressurizer model. The code is validated with the experimental investigation and calculation error is within ±12%, which suggests that the code can simulate the natural circulation characteristics under stationary and motion conditions accurately. The analysis includes the natural circulation characteristics under stationary, rolling, heaving and inclining conditions. The influential factors include the rolling period, rolling angular acceleration amplitude, rolling angular amplitude, heaving period, heaving acceleration and inclining angles. The calculation results provide the reactor with an evaluation of reactor natural circulation system characteristics. [ABSTRACT FROM AUTHOR]

• The deviations between CHFs predicted by current model and the lookup table are less than 20%. • The introduce of the guide thimble can upgrade the CHF for the assembly by 3.45%. • The guide thimble can alter the transverse location of boiling crisis. • The inter-channel lateral mixing intensity between the innermost channels and the adjacent channels was suppressed by the guide thimble. Rod bundle with guide thimble is a typical structure in the PWR fuel assembly. The thermohydraulics in the channel adjacent to the guide thimble are affected by the non-heated tube, introducing different characteristics in critical heat flux (CHF). In the current work, the CHF of 5 × 5 rod bundle with guide thimble at the center was predicted using the Eulerian two-fluid framework coupled with the critical heat flux model for heat transfer deterioration at the heated wall. The local thermal-hydraulic parameters as well as CHF values were compared with these without guide thimble. It was found that the CHF in the fuel assembly with thimble was slightly larger than that in the assembly without thimble. The lateral locations of boiling crisis were altered by the existence of the guide thimble. The inter-channel cross flow was also suppressed by the guide thimble due to its effect on the flow and thermal boundaries of the subchannels. The results obtained in this work can be used to support the safety analysis in the fuel assembly with guide thimble. [ABSTRACT FROM AUTHOR]

As a typical ocean condition, rolling motion has a significant effect on the two-phase flow systems. The influences of rolling motion on the bubble rising velocity were studied theoretically and experimentally. Five following rolling conditions were investigated: ϑm5°T8s, ϑm10°T8s, ϑm15°T8s, ϑm15°T12s, and ϑm15°T16s, where ϑm is the maximum rolling angle and T represents the rolling period. The theoretical results show that the fluctuation amplitude of axial tangential acceleration is much larger than that of axial centrifugal acceleration. Although the amplitude of (aroll)a is much lower than the value of axial gravitational acceleration, it shows evident influence on the phase curve of axial resultant acceleration. The difference between adjacent troughs in the phase curve of aa increases as a result of a decreased rolling period or increased rolling amplitude and parameter y', where y' represents the horizontal coordinate that is attached to the frame and is in the plane in which the rolling side-to-side motion takes place. The experiments were conducted in a 3 × 3 rod bundle under stagnant condition. The experimental results show that the fluctuation amplitude of bubble rising velocity increases with the increase in rolling amplitude or the decrease in rolling period. In addition, the effect of rolling amplitude is much more significant than that of rolling period. The fluctuation of bubble rising velocity presents one peak in one rolling period under the conditions of ϑm5°T8s, ϑm10°T8s and ϑm15°T8s. However, it presents two peaks under the conditions of ϑm15°T12s and ϑm15°T16s. The axial resultant acceleration gives rise to the fluctuation of bubble rising velocity. [ABSTRACT FROM AUTHOR]

Highlights: [•] CFD is used to model triangular-pitch rod bundle. [•] The Reynolds stress model with different pressure–strain models is used in this investigation. [•] Predicting the secondary flow characteristics and made comparisons with measured data. [•] Predicting the position of vortex center and the distribution of normal stresses. [•] The distribution trends of normal stresses have good agreements. [ABSTRACT FROM AUTHOR]

Abstract: In this paper, a computational fluid dynamics (CFD) methodology is proposed to investigate the thermal–hydraulic characteristics in a rod bundle with split-vane pair grids. The SST k–ω turbulence model is primarily adopted in the present CFD model. Thickness of grids is considered in order to precisely simulate their effects in reduction of the effective flow area and increase of the flow velocity and the heat transfer capability. In addition, the uniform heat generation rate is also used to model the heat generation rate within the fuel rod, rid of constant heat flux assumption as the previous simulation works. Existing data of Nusselt (Nu) number distributions in the axial and azimuthal directions obtained by Holloway et al. (2008) have been employed to validate the CFD model. The Nu decreasing trend downstream of a split-vane pair grid and the Nu variation around the circumferential surface of rod can be explained by the predicted flow characteristics, including flow swirling, detachment, separation, etc. [Copyright &y& Elsevier]

Abstract: An annulus can be seen as a rod bundle with a single rod, which has no thermal–hydraulic interaction with other rods. Thus the interaction in rod bundles can be obtained with the help of certain annuli. The heat transfer rates of both triangular and square lattices were studied based on the correlation of annuli, and modifications were done to take into account the influences of temperature difference, hydraulic diameter, interaction and shape difference, etc. The temperature difference correction was calculated with the section temperature distribution experimental data for annuli, while the other three were determined with geometric parameters. Correlations for both lattices were compared with available experimental data and the results showed good agreement. Comparison of these two correlations indicated that there were important differences between the two arrangement patterns if the pitch-to-diameter ratio was less than 1.25. Furthermore, the interaction correction was found to be generally equal to the thermal mixing length scale used in the sub-channel analysis, which may be used as a new correlation for the interaction correction. [Copyright &y& Elsevier]

Abstract: The effect of fuel bundle geometry has been evaluated using the subchannel code ASSERT to enhance the thermalhydraulic performance. Based on the configuration of a standard 37-element fuel bundle, the element diameter in each ring has been changed while those of other rings are kept as the original size. The dryout power of each element in a fuel bundle has been obtained for the modified fuel bundle and compared with that of a standard fuel bundle. The results obtained show that the modification of element diameter largely affects the thermal characteristics of the subchannel on the upper region of a modified element due to the buoyancy drift effect. The explicit effect of the element size on the dryout power of a fuel bundle is investigated by acquiring comprehensive numerical solutions. It is found that the dryout power of a fuel bundle has a tendency to increase as the element size decreases, except for the case where the outer ring is modified. The dependency of the transverse interchange model on the present results has been assessed by examining the dryout power of a bundle for different mixing coefficients and buoyancy drift models. [Copyright &y& Elsevier]

Abstract: An analytical calculation has been performed to predict the turbulent friction factor in a rod bundle. For each subchannel constituting a rod bundle, the geometry parameters are analytically derived by integrating the law of the wall over each subchannel with the consideration of a local shear stress distribution. The correlation equations for a local shear stress distribution are supplied from a numerical simulation for each subchannel. The explicit effect of a subchannel shape on the geometry parameter and the friction factor is reported. The friction factor of a corner subchannel converges to a constant value, while the friction factor of a central subchannel steadily increases with a rod distance ratio. The analysis for a rod bundle shows that the friction factor of a rod bundle is largely affected by the characteristics of each subchannel constituting a rod bundle. The present analytic calculations well predict the experimental results from the literature with rod bundles in circular, hexagonal, and square channels. [ABSTRACT FROM AUTHOR]

• LDV was verified to be used to measure the flow induced vibration of rod bundle. • Flow induced vibration of a 4x4 rod bundle with space grids was tested. • Both turbulence and vortex excited vibration of the rod and space grid were obtained. • Strong vortex induced resonance was found when the flow velocity was 7.35 m/s. In order to develop a feasible method for measuring FIV of both spacer grid and rod, two experiments are conducted in this study. First, the simultaneous verification tests between the LDV and piezoelectric accelerometer have been performed. For the case of vibration occurring in a channel with water, the measured signal by the LDV need to be modified while not affecting the measured vibration value. The results show that the LDV is a good candidate for measuring vibration under flow conditions. Then, the LDV is employed to measure the FIV of a specified 4 × 4 rod bundle with spacer grids. Both turbulence and vortex excite the rod bundle, and a strong vortex induced resonance is found when the flow velocity is 7.35 m/s. The vibration intensity is amplifying with increasing the flow velocity. Moreover, 'lock-in' phenomenon and two types of vortex shedding with St numbers of 0.15 and 0.26 are found. [ABSTRACT FROM AUTHOR]

• The flow feature of supercritical water in a rod bundle was studied experimentally. • The friction factor shows a valley at the vicinity of the pseudo-critical temperature. • The effects of pressure and heat flux on pressure drop and friction factor are weak. • An improved correlation was proposed based on the Filonenko correlation. An experiment has been performed to study the pressure drop and friction factor of supercritical water in a 2 × 2 rod bundle. A fuel-assembly simulator with four heated tubes was installed inside a square channel with round corners. The outer diameter of each heated tube is 8 mm with a pitch-to-diameter ratio of 1.18. The full length between the inlet and outlet pressure measuring holes is 550 mm. Experimental parameter covered the pressures of 23–28 MPa, heat fluxes of 200–1000 kW/m2 and mass flux of 1000 kg/m2s. The variations of total pressure drop, frictional pressure drop, gravitational pressure drop and acceleration pressure drop with bulk enthalpy were studied. It was found that the pressure drop varies notably in the pseudo-critical and high-enthalpy regions due to the strong variations of the thermophysical properties. The friction factor exhibits a local valley as the bulk temperature approaches the pseudo-critical temperature. The effect of heat flux on the frictional pressure drop and friction factor is weak, while the effect of pressure is mainly restricted to the high-enthalpy region. It was found that the friction factor of the 2 × 2 rod bundle is slightly higher than that of a smooth tube, but is significantly lower than those of sub-channel or internally-ribbed tube. Eight correlations were assessed against the current set of experimental friction factor. It was observed that the Filonenko correlation is superior to other correlations. An improved correlation for predicting the friction factor was proposed in support of the hydrodynamic analysis of rod bundles at supercritical pressures. [ABSTRACT FROM AUTHOR]

• Total, gravitational and frictional pressure drops were obtained by measurement methods. • The effects of pressure, mass flux and thermodynamic quality on pressure drop were investigated. • Two phase friction multiplier correlations were assessed against the present data. • A newly-developed correlation based on the Tran correlation was proposed. In this paper an experimental study on two-phase flow pressure drop characteristics has been performed in a triangular-array rod bundle geometry under various flow conditions (P = 5–9 MPa and G = 100–350 kg·m−2·s−1). Total, gravitational and frictional pressure drops are obtained by measurement methods. The effects of pressure, mass flux and thermodynamic quality on pressure drop were investigated in detail, and the proportions of gravitational and frictional pressure drop in total pressure drop are analyzed. Meanwhile, an assessment of eight existing correlations for two-phase friction multiplier calculation is carried out against the present data. The results show that Tran correlation has a relatively better performance when predicting the present data, while the accuracy is still not good enough. In views of this, a newly-developed correlation based on the Tran correlation has been proposed for the two-phase friction multiplier calculation in rod bundle geometries, which is proved to agree with the present data well. [ABSTRACT FROM AUTHOR]

• A visualized CHF experiment was carried out under the similar range and geometry with PWRs. • The CHF database, experimental trends and comparisons to existing CHF prediction methods were presented. • Three boiling phenomena are observed at CHF, and the possible CHF mechanism was explained. • Quantitative analysis of bubble layer thickness was carried with/without mixing-vanes spacer. In this study, a critical heat flux (CHF) experiment using R-134a as working fluid was carried out in a vertical single rod geometry. A unique feature of this experiment are high speed visualization studies obtained at a viewing port located at the CHF location. The 9.5 mm O.D. heated rod was placed in a 19 x19mm square channel to simulate the Pressurized Water Reactors (PWRs) fuel lattice. Test conditions were selected and scaled to cover a wide range of water conditions relevant to PWRs. The CHF database, experimental trends and comparisons to existing CHF prediction methods are presented. The visualization results of the flow boiling at the CHF location and the CHF mechanism are discussed with a particular emphasis on the impact of grid spacers on the boiling phenomena. Quantitative analysis of bubble layer thickness is carried out, including the relationship between bubble layer thickness and CHF. [ABSTRACT FROM AUTHOR]

• The influence of buoyancy force on heat transfer in a 5 × 5 rod bundle at low flow rate is studied. • It is reported that there is stronger non-uniformity of circumferential heat transfer in heat transfer impairment region. • The heat transfer impairment degree of rod bundles in mixed convection is discussed. • A new buoyancy parameter and a mixed convection heat transfer correlation for rod bundles are proposed. A correlation is developed to predict mixed convection heat transfer of water flow in rod bundles. The rod bundles are 5 × 5 and the Re number ranges from 1 × 103 to 3 × 104 in the experiment. In the forced convection region, the Dittus-Boelter correlation predicts well with the experimental data. However, in the mixed convection region, the correlations reported in the literature are all deviated from the measured data. Moreover, based on the classical form of buoyancy parameter, Gr*/Re3.425/Pr0.8, the experimental data of circular channels extracted from literature are compared with the measured data. The comparison shows that the buoyancy parameters of the onset point of heat transfer impairment (HTI), turning point of HTI and onset point of heat transfer enhancement (HTE) in rod bundles are very close to that in circular channels, but the maximum degree of HTI in bundles is significantly weaker than that in circular channels. The experimental data indicate that the circumferential heat transfer is significantly uneven in HTI region, which result in the differences of buoyancy parameter of turning point in different circumferential directions and eventually lead to the weaker overall degree of HTI in bundles. Considering the derivation method of the classical form of buoyancy parameter proposed by Rouai, a new buoyancy parameter form for rod bundle channels is derived based on the Dittus-Boelter correlation for forced convection heat transfer and the Cheng and Todreas's correlation for friction factors in rod bundles. Based on the new buoyancy parameter form, a new correlation to predict mixed convection heat transfer in rod bundles is proposed and provides good prediction accuracy. [ABSTRACT FROM AUTHOR]

• Rolling natural circulation loop with rod bundle channel is studied numerically. • The coupled simulation predict the flow characteristics well. • The coupled simulation can be performed independently of experiments. • This study makes a foundation for future coupling researches. • The coupling method is described in detail. The numerical simulation of the one-dimensional user code and the three-dimensional software STAR-CCM + coupling is performed on a natural circulation loop of a 3 × 3 rod bundle channel under rolling conditions. The coupling simulation method in this paper utilizes our one-dimensional user code, which is compiled and linked to generate a dynamic link library (DLL). The DLL is then imported into STAR-CCM+, so as to realize the coupling simulation of the one-dimensional user code and the three-dimensional Computerized Fluid Dynamics (CFD) code. In the one-dimensional user code, the mathematical model under rolling conditions is established to simulate the natural circulation loop except for the rod bundle channel. Adding a rotational motion to the STAR-CCM + enables a rolling simulation of the rod bundle channel. In order to compare the rationality of the coupled simulation, this paper also reports on the results carried out in an uncoupled simulation to study the rod bundle channel under the rolling conditions. The coupled simulation is a simulation that can be performed independently of an experiment, but the uncoupled simulation is not. That fact further demonstrates the wide applicability of the coupled simulation. The flow velocity and pressure drop characteristics in the 3 × 3 rod bundle channel are investigated. The results show that the coupled simulation method can predict the flow characteristics of the rolling rod bundle channel in the natural circulation loop well. The paper describes the coupling method in detail, which can make a foundation for future coupling research. [ABSTRACT FROM AUTHOR]

• Longitudinal flow through rod bundle of ETRR2 Cobalt/Iridium irradiation device was studied. • Study of optimal performance was carried out. • Contours of fluid temperature , fluid velocity, and temperature of Ir rods were illustrated. • Determination of the conditions at which minimum entropy generation was defined. • The present results can be correlated well with best fit correlations. In the present paper, three dimension CFD analysis was presented for longitudinal flow through rod bundle of ETRR2 Cobalt/Iridium irradiation device. Study of optimal performance for both heat transfer and flow characteristics was performed by the determination of the conditions at which minimum entropy generation exist. The study was performed for Reynolds number (Re) in the range 2 × 104 ≤ Re ≤ 5.5 × 105, and Prandtl number (Pr) in the range 3.2 ≤ Pr ≤ 5.2 under both the conditions of uniform and non uniform (cosine shape) heat flux. Contours of fluid temperature and fluid velocity at irradiation box outlet as well as temperature contours of Ir rods at different values of inlet velocity were illustrated. Two correlations one relating the Nusselt number (Nu) with Re and Pr and the other relating skin friction factor with Re was presented in the studied range of Re and Pr. Determination of the applicability of the existing correlations for turbulent forced convection heat transfer and friction coefficient are also presented. Comparison between the flow and heat transfer characteristics through Ir rods of different locations in the irradiation box is presented. [ABSTRACT FROM AUTHOR]

This study offers state-of-the-art rod bundle facility development and testing with modified coolant, with the prospects and challenges for the small modular reactor (SMR). The thermal-hydraulics characteristics of a 3 × 3 square array rod bundle tested under a fully-developed single-phase turbulent upflow condition using pure water and water-mixed alumina nano-size particles. This rod bundle with specially-designed grid spacers and flow distributors developed to simulate a unit block of SMR core. Test data collected for a wide range of inlet flow conditions (Reynolds numbers from 21,000 to 100,000) with and without heating and nanoparticle mixing in water. Then, the heat transfer performance in the form of Nusselt number, Nu, and pressure drop was calculated and compared with the estimation of well-established models and correlations. The pressure drop test data of water appropriately matched with estimation data. In the heat transfer study, the value of Nu increased with the percentile inclusion of nanofluid. But the addition of nanoparticle is recommended to a limit as it increased the pressure drop value, which is reasonably high for some of the SMRs. [ABSTRACT FROM AUTHOR]

The lead cooled fast reactor is considered as one of the most potential Generation IV nuclear systems due to the low working pressure, the appreciable neutron economy, and the considerable passive characteristics. Meanwhile, the lead bismuth eutectic (LBE) has the related advantages of lead with the weaker corrosiveness, which has been paid much attention by recent decades. Moreover, the sub-channel code is a necessary analysis tool for the reactor thermal-hydraulic design and safety analysis, which has been developed combined with the accumulation of LBE experimental data and the understanding of physical phenomena. In this study, a sub-channel code available for LBE was developed, and the corresponding geometric characterization method of typical sub-channels was described in detail, including the fuel assembly with wire spacer. As for this sub-channel code, the transversal thermal conduction through gap was taken into account. In addition, the physical properties, the heat transfer model, the flow resistance model and the turbulent mixing model were analyzed. Finally, the thermal-hydraulic experiments of LBE conducted on THEADES (THErmal-hydraulics and Ads DESign) were selected as the evaluation data of this sub-channel code, including 19 rods with grid spacer and wire spacer, and the calculated results were in good agreement with the experimental results. [ABSTRACT FROM AUTHOR]

• The turbulent flow in a rod bundle was studied experimentally and numerically. • First- and second-order turbulence statistics were provided. • Large-scale quasi-periodic flow pulsation was detected by both PIV and CFD. • EB-RSM model showed acceptable performance in predicting the flow in rod bundles. The turbulent flow in a fuel bundle affects the heat transfer from the fuel rods to the coolant, which is a concerned issue in the design of fuel assemblies. In this paper, an experiment was performed using particle image velocimetry to measure the velocity distribution of water in a 3 × 3 rod bundle at various Reynolds numbers. The technique of matching index of refraction was utilized to fully visualize the flow field. Emphases were put on the flow inside a bare bundle that is free from the effects of grid spacer. It was found that the axial mean velocity in the subchannel is 30% higher than that in the gap center due to the cross-sectional nonuniform flow area. However, the axial root-mean-square velocity showed an inverse trend. The Reynolds shear stress exhibited a wavy variation in the lateral direction. A quasi-periodic large-scale flow pulsation was observed across the gap, and its wave length increases with Reynolds number. It was also found that the turbulent mixing coefficient was higher than that predicted by classical correlations. A numerical simulation was conducted signifying the acceptable performance of the EB-RSM model in predicting the flow behavior of water in the 3 × 3 rod bundle. Transient simulation with this turbulence model is able to capture the main flow phenomenon in the rod bundle. [ABSTRACT FROM AUTHOR]

• The transient friction characteristics of pulsatile flow in rod bundles are analyzed. • The transient flow fields of pulsatile flow in rod bundle are visualized using PIV. • The evolution characteristics of pulsatile flow pattern in rod bundles are analyzed. • The velocity gradient and turbulent shear stress in the whole period are analyzed. Rod bundles are widely used in various devices and energy engineering including heat exchangers, nuclear fuel assembly and so on. The transient fluctuation of flow rate in the thermal devices is a common phenomenon under accident scenarios or ocean condition, the study of pulsatile flows has significant practical relevance for the design and safety of thermal devices. In order to determine the flow characteristics of pulsatile flows in rod bundles, the experimental investigations for transient friction loss and evolution of flow field of pulsatile flows of a 5 × 5 rod bundle (P/D = 1.326) are performed. The experiments were conducted under the ranges of time-averaged Reynolds number Re ta = 2000–7000, pulsatile amplitude Au = 0.2–0.6 and pulsatile period T = 10 s–100 s. A comparison of friction factors under steady flows and pulsatile flows is performed to determine the influence of pulsatile flows on frictional characteristics. The results show that the transient friction factors versus flow rate under pulsatile flow presents an "oval" shape, and the "oval" shape has a variation with pulsatile parameters changing. To further reveal the details of pulsatile flow fields, the evolution of flow pattern in the pulsatile flow of Re ta = 5200, Au = 0.6, T = 10 s is visualized in rod bundles by Particle Image Velocimetry (PIV). The velocity distribution in different stages of pulsatile flows are displayed and discussed. The influence of velocity gradient and turbulent shear stress in the whole pulsatile period for frictional loss is analyzed, and the variation of transient friction factors under different pulsatile parameters is revealed. The results describe transient behaviors to contribute a better understanding for transient flows in fuel assemblies. [ABSTRACT FROM AUTHOR]

• Chun model would severely underestimate grid local loss coefficient for small Re. • A two-phase friction multiplier correlation was first proposed for rod bundle. • A new way to model the two-phase multiplier of spacer grid was proposed. This paper presents the experimental results and analyses of the friction and local pressure loss coefficient of a 5 × 5 rod bundle with spacer grid. Based on the experimental data, friction and local pressure loss models were assessed and the role of spacer grid was discussed. The results showed that Cheng and Todreas model could predict bare rod friction loss well, and Chun model applied for high Re but would severely underestimate grid local loss for small Re. For two-phase condition, a universal correlation was proposed for bare rod two-phase friction multiplier, and it was found that all existing models cannot predict grid local loss multiplier very well. High prediction accuracy could be obtained by combining homogenous flow model, the liquid-mixture density ratio and a modification factor as a function of the grid parameter ε (axial projected area of grid strip to flow area) and void fraction. [ABSTRACT FROM AUTHOR]

• Thom and new Shah correlation can be applied in rod bundle with error of 30%. • Empirical correlations better predict subcooled boiling in rod bundle. • Semi-empirical correlations are not suitable for rod bundle. To meet the requirement of passive safety, natural circulation has been applied in cooling the nuclear reactor core. In comparison with the great successes in predicting subcooled boiling in simple geometries such as pipes, annuli and rectangular channels, few researches have been performed for rod bundle. We performed subcooled boiling heat transfer experiments in a vertical 3 × 3 rod bundle under upward natural circulation flow condition. The experiments were performed under pressure from 0.2 MPa to 1.0 MPa, mass flux from 162 to 411 kg/m2/s, heat flux from 31 to 271 kW/m2, local subcooling from 0 to 25 °C and thermodynamic vapor quality from −0.05 to 0. The experimental results were compared with the predictions of 14 subcooled boiling correlations. The comparisons showed that the new Shah correlation and Thom correlation can accurately predict the heat transfer coefficient of natural circulation flow in the rod bundle within a deviation of 30%. The average deviations of new Shah correlation and Thom correlation are 3.98% and 2.54% respectively. According to the discussion of three kinds of correlation, we found that empirical correlations can better predict the subcooled boiling than semi-empirical correlations. [ABSTRACT FROM AUTHOR]