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This study focusses on investigating the effect of flow orientation on global and local two-phase flow parameters under two-phase flow conditions. Flow visualization, pressure measurement, and conductivity probe measurements are performed in a test facility made of 50.8 mm inner diameter acrylic pipes with flow visualization, pressure measurement, and conductivity probe measurement. Characteristics of flow regimes and interfacial structures, and frictional pressure drop prediction in the vertical upward and vertical downward two-phase flows are compared and analyzed. The results show that unique coring phenomenon and slug bubbles with off-centered noses appear in the vertical downward flow. The flow regime transition boundaries shift to the lower gas superficial velocities in the vertical downward flow compared to that in the vertical upward flow. Furthermore, the distribution and one-dimensional transport of the void fraction, interfacial area concentration, bubble velocity, bubble Sauter-mean diameter, and bubble frequency are acquired and compared to study the effect of flow orientation on interfacial structures. The interfacial structure and its development are found to be mainly affected by the lift force and the turbulence related bubble interactions. The frictional pressure drop is acquired and modeled by the Lockhart–Martinelli correlation. The frictional pressure drop is found to be larger in the vertical upward flow than that in the vertical downward flow. The recommended C value for both vertical upward and vertical downward flows are proposed.

Ocean conditions will cause ships to be affected by rolling, heaving and other motions, which introduce additional inertial forces to the marine nuclear power system, introduce additional pressure drop to the fluids in each loop, bring errors to the measurement of thermal hydraulic parameters such as pressure difference and flow rate. Therefore, the measurement of pressure difference and flow rate under ocean conditions is studied and corrected in this paper. We built a rolling platform and a heaving platform and measured the differential pressure and flow rate under various working conditions. Through analysis, the relationship between differential pressure, flow fluctuation and swaying, undulating and other movements under moving conditions was obtained. Besides, the calculation method of attached pressure drop was analyzed, and the following conclusions were obtained: The influence of motion on differential pressure and flow rate is mainly composed of two parts: The additional pressure drop caused by the swinging motion of working medium in the pipeline and the pressure lead-in pipe, and the periodic fluctuation of friction pressure drop caused by the periodic fluctuation of flow velocity caused by swinging motion. The attached pressure drop caused by motion can be calculated through theoretical analysis. Through the analysis of experimental data and the comparison of theoretical calculation methods, it is found that the pressure drop measured by the experiment is in good agreement with that fitted by the theoretical calculation, and the modification scheme of pressure difference under ocean conditions is feasible to some extent.

The rod bundle fuel is characterized by compact structure and narrow flow passage. The fragments and corrosion products, flowing with the coolant, can cause local blockage accident, threaten the integrity of the fuel cladding. Therefore, it is necessary to use the Particle Image Velocimetry (PIV) to visualize and measure the flow fields downstream of the blockages. The results show that partial blockages will cause flow reversal. In the backflow zone, vortices are generated downstream of the blockage, causing increase in the resistance. The length of backflow zone increases with the increase of the Reynolds number. The wake area formed downstream of the blockage presented periodic changes with the time and the period is about 0.8s. For the blockage of the interior subchannels, in the backflow zone, two rows of asymmetrically distributed vortices, and the vortices interfere with each other and cause squeeze deformation. For the blockage of the side and corner sub-channels, the formed vortices have irregular shape and nonuniformed distribution, and the flow field is more complex and changeable. This is believed to be caused by the high intensity turbulence and the influence of the wall.

The floating reactor system will be rolling, heaving and other movements affected by waves, wind, etc. These motions will introduce additional inertial force field into the rod bundle, thus affecting the flow characteristics in the rod bundle channel. In order to study the influence of rolling motion on the flow characteristics of rod bundle, a visualization research of the flow field of rod bundle channel with a pitch-to-diameter ratio of 1.326 was carried out under rolling motion. The results show that under a small driving force, the rolling motion has a significant effect on the flow field in the rod bundle, affecting the velocity distribution in different sub-channels, and there is transverse mixing between adjacent sub-channels. With the increase of driving force, the influence of rolling motion is gradually weakened. The flow field distribution under rolling motion is significantly different from that under pulsating flow. The experimental results show that the influence of rolling motion on the middle sub-channel of the rod bundle channel is small, and the influence on the edge sub-channel is large. The velocity field of the subchannels on both sides of the edge fluctuates periodically, and the wave phase is inverse. This study shows that the flow field variation caused by rolling motion is different from that caused by pulsating flow, and the flow field of fuel assembly in floating reactor system under ocean condition is further studied.

The present study proposes a new feature extraction method based on non-stationary conductivity probe signals. Two types of discriminative network models, i.e., the probabilistic neural network (PNN) and nonlinear support vector machine (SVM), are established for flow regime identification using small sample sets. The eigenvectors are composed of 16 feature quantities obtained by wavelet packet decomposition (WPD) and 8 feature quantities in the time-domain derived from the reconstructed low-frequency signals. The 8 features include maximum, minimum, standard deviation, arithmetic mean, kurtosis, peak factor, impulse factor and margin factor. The signals are normalized based on features rather than samples before flow regime identification. In the current study, WPD results show that the conductivity probe signals in two-phase flow are mostly in low frequency. The identification accuracy of the nonlinear SVM is 90.47%, which is better than 83.33% by the PNN method. This study verifies the superiority of nonlinear SVM in solving small samples and nonlinear flow regime classification problems. However, the accuracy of flow regime identification near flow regime transitional boundaries still remains questionable and needs further improvement.

The load distribution and coupling feedback between parallel loops is one of the issues in the multi-loop nuclear reactor system. In the present study, a dual-loop reactor natural circulation experimental system with typical structural features was designed and built based on the characteristics of a floating nuclear power plant. The system is used to study the system response characteristics under the conditions of heat load disturbance, resistance disturbance and system response time. The results show that the flow of the primary loop will be redistributed when the load disturbance was introduced into the natural circulation loop. The primary flow rate changes quickly when the heat load disturbance reaches the critical point. When the heat load disturbance is lost on one side, the total natural circulation flow rate of the loop decreases to 60% of the value in the steady-state flow. The response of the system after introduction of resistance disturbance is similar to the introduction of the load disturbance. The response of the system can be divided into two stages, i.e., the rapid response period and the slow adjustment period after the introduction of the disturbance. With the increase of disturbance, the negative slope of stage 1 increases, and the slope of stage 2 gradually changes from positive to negative. With the experimental, a model is developed to predict the response of the system. A characteristic number S is defined to characterize the symmetry of the system. Based on the experimental data, it is found that when S is between 0.5–2, the external disturbance will not cause asymmetric response of the system.

The spacer grid is widely used with the fuel rod buddle to stabilize the fuel assembly structure and reduce the vibration effect in the pressurized water reactor (PWR). To study the flow resistance characteristics of the working fluid under periodic fluctuation of the floating nuclear reactor (FNR), an experimental study of the pulsating flow resistance characteristics of a 5 × 5 fuel rod bundle (P/D = 1.326 and W/D = 1.268) was carried out. The ranges of the cycle-averaged Reynolds number Reta = 0.8 × 103–9 × 103, pulsating amplitude Au = 0.2–0.8 and dimensionless frequency ω ′ = 1.26–2.81 were covered in this experimental study. The experimental results indicated that the flow rate lags behind the pressure drop in the fuel rod bundle channel, and the phase lags increase with the increase of ω ′ but change slightly with Au and Reta. The instantaneous friction factor and local resistance coefficient are greater than the steady-state value at the end of deceleration and the whole acceleration stage in the pulsating flow cycle, and smaller than the steady-state value at the beginning of the deceleration stage. The present study compared the flow resistance of the accelerating and decelerating half-cycle with the steady-state value and found that the speed of the accelerating half-cycle increases gradually with the increase of Au, but the speed of the decelerating half-cycle flow resistance decreases gradually with the increase of Au. In addition, the obtained results have also indicated that the local resistance coefficient of the spacer grid is closer to the steady-state value than the friction factor in the accelerating and decelerating half-cycle. During the pulsating flow, the cycle-average friction factor and local resistance coefficient in the whole cycle are larger than the steady-state value under the corresponding Reynolds number. The normalization parameters of Cf = λ ta/λst and Csg = Kta/Kst are proposed to evaluate the effect of pulsating flow on the cycle-averaged flow resistance. Cf and Csg increase with the increase of Au and ω ′ , and decrease with the increase of Reta. In addition, it is found that Csg is smaller than Cf under the same pulsation parameters, which it means that the influence of fluid fluctuation on the resistance of the spacer grid is relatively small. Finally, the new correlations for predicting λta and Kta in the pulsating flow are established through dimensional analysis, and the prediction result is reasonably good.

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 4.5Dh) downstream of the spacer grid, indicated that the mixing effect of the spacer grid becomes weakened after 4.5Dh. 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.

Passive Residual Heat Removal Heat Exchanger (PRHR HX) immerged in the In-containment Refueling Water Storage Tank (IRWST) plays an important role in removing the core decay heat under non-Loss of Coolant Accident (LOCA) accident. In the initial stage of the Passive Residual Heat Removal System (PRHRS) activating, the natural convection is the main way to exchange heat between the secondary side of the PRHR HX with IRWST. However, it is almost impossible to acquire the thermohydraulic performance in IRWST since the influence of the natural convection and thermal stratification in IRWST happens simultaneously. In this paper, Hierarchical Two-Tiered Scaling (H2TS) method is employed to get scaling criteria based on control equations. Anda visualized water tank with five C-shape electrical heating tubes was scaled-down and applied to investigate the thermohydraulic performance for IRWST of CAP1400 Nuclear Power Plants (NPPs). Particle Image Velocimetry (PIV) measurement was used to investigate the evolution of the flow field of total twenty-five surfaces in three axial directions under three steady- and two variable-heating conditions. Meanwhile, three thermocouple bundles were adopted to acquire temperature data in three regions. The experimental results show that the temperature deviation in the same height of the IRWST is flattened by plenty of vortexes. As a result, local circulation and the temperature difference is less than 1 °C. While the thermal stratification along the vertical direction is obvious, the distribution of flow field and temperature demonstrate that a “dead zone” exists in bottom of the IRWST. Besides, a “thermal interface” region forms in the lower middle region as hot and cold fluid mixing. An obvious upwelling flow is observed near the PRHR HX heating rod bundle from flow fields of three axial directions. At the same time, the maximum height of the upwelling flow decreases with time increasing. On the one hand, the natural convection is induced by the thermal stratification. What is more, the thermal stratification will impair the scale and strength of natural convection, which will intensify the thermal stratification conversely.

To calculate the thermal characteristics of the rectangle liquid droplet radiator more accurately, a novel method, which estimated the characteristics of each droplet, was introduced. The shape factors of two and three droplets were theoretically deduced and verified by the Monte Carlo method. It was found that the shape factor of two droplets depends only on the ratio of the distance between the two droplets to the radii of droplets. According to this characteristic, the shape factor of two droplets was calculated and fitted to be a formula, which makes it convenient to calculate the shape factor in the following analysis. Based on the superposition rule of the shape factor, a shape factor matrix was introduced to calculate the effect of surrounding droplets when analyzing the heat transfer of the droplet sheet. The temperature distribution of the droplet sheet was calculated based on the shape factor matrix, which obtained using the combination of theoretical method and Monte Carlo method. The results showed that the temperature difference among the droplets on a same width-thickness plane was negligible. Therefore, a simplified model was developed based on this characteristic. The discrepancies of the temperature distributions obtained by the original model and the simplified model were found to be less than 0.35%, which demonstrates the feasibility of the simplified model to analyze the thermal characteristics of the rectangle droplet radiator. By further analyzing the heat transfer characteristics of the droplet sheet, proposals were made for design of the rectangle liquid droplet radiators.

In possible accident scenarios of the reactor coolant system, the convection and mixing characteristics of borated water have a great influence on reactivity variations in the reactor, and then impact the reactor safety. The mixing and diffusion characteristics of the boric acid solution in the annular downcomer is studied by laser-induced fluorescence (LIF). An aqueous solution of Rhodamine B (RhB) fluorescence dye, which is used to simulate boron solution, was injected in the coolant inlet to visualize the fluid mixing process in the annular volume. Laser sheet was formed by optical system to emit the fluorescence, a high-speed camera was used to capture the fluorescence information for reconstructing the concentration distribution of dye. In order to guarantee the experimental accuracy, LIF image processing and a calibration technology are applied. The concentration profiles were implemented to illuminate the fluid mixing between dye solution and diluted water in the downcomer. The coefficient of variation (COV) was conducted to quantify the spatial and temporal transport of dye in the annular downcomer. The results indicate that a higher flow improves the transport ability of dye in the transverse direction, and shortens the distance and time to uniform. The mass force can accelerate the dye transport in the vertical direction. The whole-field measurement technique implemented in this study is expected to provide an insight in the research of flow mixing in the complex geometry. The experimental results can provide a better understanding of mixing phenomena in annular downcomer, and it is also useful for the annular downcomer CFD validation.

The present study investigates the air-water two-phase flow across a 90-degree vertical-upward elbow with the computational fluid dynamics (CFD) simulation. The Eulerian-Eulerian two-fluid model and the Multi Size Group (MUSIG) model are used to predict the development of the detailed interfacial structures between the two phases. The axial development of the void fraction and the interfacial area concentration are investigated and benchmarked with the experimental data measured using the four-sensor conductivity probe. It is concluded that CFD simulation can predict the characteristics distributions of void fraction and interfacial area concentration and their development downstream of the elbow. The double-peaked void fraction distribution is found to be caused by the secondary flow induced by the elbow. The liquid phase on the outer curvature moves to the inner curvature and forms a double counter rotating vortex, entraining the bubbles to form a double-peaked distribution. The elbow effects become dissipated between 33 and 63 hydraulic diameters. The simulation results of liquid-phase and gas-phase parameters can be used to develop the theoretical two-phase flow models for the elbow region.

The Pressurizer (PRZ) is an important device used to control the system pressure in nuclear power plants (NPP). The weight and volume of the PRZ are important factors affecting the weight and layout of nuclear power plants. Reducing the weight and volume of the PRZ while satisfying the demand has important theoretical and practical significance for the design and operation of NPP. In this work, an improved non-dominated sorting genetic algorithm, INSGAII, is proposed adopting several improvement strategies. Comparing to the NSGAII, the improved algorithm INSGAII has shown great performance. The INSGAII is devoted to the optimal design of the Pressurizer aiming at reducing its weight and volume. Mathematical modeling of the PRZ of the Qinshan nuclear power plant phase II was established. The weight and volume of the PRZ is optimized with the proposed algorithm INSGAII. The results show that the volume and weight of the optimized PRZ are reduced at least 6.11% and 15.06%, respectively, as compared to the prototype PRZ.

The present study investigates the two-phase flow pressure drop across 90° vertical elbows. The elbows, positioned in the vertical-upward to horizontal and horizontal to vertical-downward orientations, have an inner diameter of 50.8 mm and a radius of curvature of 152.4 mm. Experiments were performed in a wide range of flow conditions at different axial locations and flow orientations to measure the local gauge pressure and void fraction. It is observed that increasing the liquid-phase or gas-phase velocity increases the local gauge pressure. The experimental data are used to study the elbow orientation effect on pressure drop and develop predictive method for two-phase flow pressure drop across 90° vertical elbows. The pressure components due to friction and geometry effects (form loss), gravity and acceleration or deceleration of the two phases are estimated accurately with the detailed experimental data. Then, predictive methods for each pressure components are investigated. For estimation of the pressure component due to friction and geometry effects, six models are evaluated, and the model coefficients are determined from the experimental data. The evaluation results show that the modified Kim et al. correlation and Bowden and Yang correlation can predict pressure drops across both vertical elbows through adjusting the parameter C and the newly introduced minor loss factor. Both models demonstrate negligible dependence on flow regimes. However, the Kim et al. correlation gives closer prediction to the experimental data than the Bowden and Yang correlation, and thus is recommended for prediction of the pressure component due to friction and geometry effects. The other models need further modification to predict the two-phase flow pressure drop across elbows.

The current work seeks to develop the interfacial area transport models for horizontal bubbly flow applicable to various pipe sizes. A steady-state one-dimensional, one-group interfacial area transport equation (IATE) for horizontal adiabatic air-water bubbly flow is presented. The models discussed include: (a) frictional pressure drop prediction based on Lockhart–Martinelli approach; (b) drift-flux based closure relations for void-weighted bubble velocity and void fraction; (c) bubble interaction mechanisms; and (d) local void fraction reconstruction to calculate covariance parameters associated with asymmetric bubble distribution in horizontal bubbly flow. To develop these models, experiments are performed to establish local databases in bubbly flow using the four-sensor conductivity probe in horizontal test facilities with three different inner diameters (38.1 mm, 50.8 mm, and 101.6 mm). In total, 23 test conditions with superficial liquid and gas velocities in the ranges of 3.00–6.00 m/s and 0.08–1.00 m/s, respectively, are employed in the current work. The range of void fraction for the established database is 0.017–0.193. It is demonstrated that, the void weighted bubbly velocity ( >), void fraction ( ), and frictional pressure loss can be predicted very well with an averaged percent difference of ±4.8%, ±4.7%, and ±2.1%, respectively. The covariance parameters can be predicted generally within ±30% using the current approach for void fraction reconstruction. As a result, the IATE can predict the measured interfacial area concentration with an averaged percent difference of ±5.9%, except for the test conditions with large bubbles generated as flow develops. Meanwhile, it is found that decreasing liquid velocity decreases the contributions to the total change of interfacial area concentration by random collision induced bubble coalescence, and turbulent impact induced bubble breakup.

Following Part (I) of the current study, which presents the experimental results of the elbow effects on two-phase flow parameters in bubbly flow, Part (II) develops models and correlations to predict the evolution of these parameters across and downstream of 90° vertical-upward and vertical-downward elbows. To quantify the length requires for the effects of the elbows to dissipate (or the dissipation length), the strength of elbows is defined as the variance of the local void fraction distribution. The axial development of the elbow-strength is modeled by an exponential function of the axial development length based on the experimental data. Then, the dissipation length of the elbow is determined by characterizing the evolution of the elbow-strength parameter. The elbow-strength parameter is also used to correlate the void-weighted bubble velocity and covariance terms in the interfacial area transport equation (IATE). The two-phase pressure drop across vertical elbows is modeled with a modified Lockhart-Martinelli correlation which considers the additional pressure drop induced by elbows. To evaluate the above developed models and correlations, they are implemented into the IATE applicable to the elbow-influenced region. The established IATE together with the available IATE of different flow orientations in straight channels are implemented to predict the interfacial area transport from the vertical-upward to horizontal to vertical-downward two-phase flow across elbows. It is found that the interfacial area concentration predictions are in good agreement with the experimental data with an average absolute percent difference of ±6% throughout the test section. The individual contributions to the interfacial area concentration transport due to each source and sink term in the IATE are discussed, demonstrating that the models and correlations developed for the two-phase flow parameters in elbow regions are reliable.

This study performs experimental investigation on two-phase bubbly flow through 90° vertical elbows going from vertical-upward to horizontal and horizontal to vertical-downward orientations. The elbows have an inner diameter of 50.8 mm and a radius of curvature of 152.4 mm. A detailed database for local and global two-phase flow parameters in different flow orientations and across vertical elbows is established with a state-of-the-art four-sensor conductivity probe, a pressure transducer, and a high-speed video camera. The evolution of void fraction, interfacial area concentration, bubble velocity, and two-phase pressure drop along the axial development length in different orientations and across elbows are presented. It is observed that elbows create secondary flow and cause redistribution of the bubbles, which significantly affect the development of two-phase flow parameters. The effects occur across the elbows and remain for a significant length downstream. The effects on the void fraction distribution downstream of the elbow are discussed in detail for both elbow orientations. Characteristic differences in the elbow effects due to the elbow orientation are observed. For the vertical-upward elbow, a similar trend in the evolution of the void fraction distribution is observed for all flow conditions investigated while two different trends are observed downstream of the vertical-downward elbow. It is found that the effects of the vertical-downward elbow dissipate faster than that of the vertical-upward elbow. The elbow effects on interfacial area concentration, bubble velocity, and two-phase pressure drop are also discussed. The evolution of the interfacial area concentration is similar to that of the void fraction evolution in the straight pipe sections. However, it increases significantly across both elbows due to the promotion of the bubble breakup induced by the secondary flow. The void-weighted area-averaged bubble velocity is subject to significant changes across and immediately downstream of the elbows due to the redistribution of the bubbles. Additionally, the two-phase pressure gradients across the vertical-upward and vertical-downward elbows are approximately 1.5 and 2 times of that in the straight pipe sections.

This study performs an experimental investigation of frictional pressure drop in air–water two-phase flows in straight pipes. A reliable experimental database for the two-phase pressure drop and void fraction is established with a differential pressure transducer and a four-sensor conductivity probe, respectively. The two-phase flow investigated focuses on gas-dispersed flow regimes in different pipe diameters of 38.1 mm, 50.8 mm, and 101.6 mm. Systematic study on the effects of flow orientation, flow regime and pipe size is performed. The most commonly used predictive models for the two-phase frictional pressure drop are evaluated with the newly established database and the existing databases found in the literature. It is demonstrated that both the conventional Lockhart-Martinelli approach and the ϕ f – correlation can generally predict the two-phase frictional pressure drop very well with different suggested values of coefficients C and n for different flow orientations, based on the established data. Meanwhile, the results show that the values of C and n are independent of the pipe size and the flow regime. The homogeneous flow model is evaluated with four β (ratio of volumetric flow rates) based mixture viscosity correlations. The predictions with the Beattie and Whalley mixture viscosity correlation are found to be the best regardless of the flow orientation. The Lockhart-Martinelli approach with the coefficient C calculated by correlation employed in the nuclear system analysis code RELAP5-3D and the Muller-Steinhagen and Heck correlation are also evaluated. It is found that these two modeling approaches as well as the homogeneous flow model tend to underestimate most of the experimental data. Improvements for pressure drop prediction in nuclear reactor safety analysis codes are observed.

Particle Image Velocimetry (PIV) technology is used in current paper to investigated turbulent flow of central blocked accidents in a heated rectangular channel. Proper Orthogonal Decomposition (POD) method and Power Spectral Density (PSD) method are also used to expose the turbulent characteristics in the wake. The velocity field at different Reynolds numbers and blockage ratios is analyzed. During the blockage experiments, the movement of eddies are captured in detail in the instantaneous flow field. The vortex gradually decreases with the increase of Reynolds number, but gradually increases with the increase of blockage ratio. The dominant vortex structures in the first two POD modes gradually become sharper at higher blockage ratio. Subsequently, the peak frequencies of the instantaneous flow field are the same as those of POD modes at various blockage ratios.

In the dispersed flow film boiling (DFFB) regime during the reflood stage of a postulated loss of coolant accident (LOCA), liquid droplets are entrained by a continuous vapor phase moving upward within the rod bundle. The liquid droplets may interact strongly with the vapor flow and the resulting thermal–hydraulic behavior of the two-phase mixture could significantly affect the evolution of peak cladding temperature (PCT). Therefore, a clear and comprehensive understanding of liquid droplet behavior is crucial for reactor safety analysis. As an attempt to visualize and quantify the droplet behavior during reflood, an advanced image processing technique is used in the present study to capture the distributions of droplet size and velocity at the RBHT test facility. In order to ensure that the experimental droplet data obtained is reliable as well as to better utilize the data for subsequent thermal–hydraulic analyses, the image processing based measurement uncertainties have to be investigated and analyzed. In this paper, various statistical methods are explored to determine the minimum droplet sample size required as a function of standard deviation at a given confidence level. Accordingly, an appropriate droplet count lower limit is selected. Based on this lower limit, the 95% confidence intervals are determined for two typical reflood tests. In addition, the repeatability in droplet measurement of Rod Bundle Heat Transfer (RBHT) reflood tests is carefully examined by comparing data sets from two identical tests. Results show that a lower required level of precision and a smaller scattering in the measured droplet data generally require a smaller sample size. For typical RBHT reflood tests, it is found that the droplet size measured downstream of a spacer grid near the peak power location always has better accuracy compared with that at the upstream location. For the tests investigated, the maximum relative error in the liquid droplet Sauter mean diameter is found to be generally smaller than 0.15 (15%) when the droplet count lower limit is 30. Comparison of the results from two identical tests indicates that the RBHT reflood test results are highly repeatable.

In this research, the numerical simulation and analysis of flow behavior in 5 × 5 fuel rod bundle with consideration of spacer grid has been conducted. ANSYS FLUENT 19.0 with the consideration of the transition SST (Shear-Stress Transport) with SAS (Scale-Adaptive Simulation) turbulence model is used for the numerical simulation. The obtained simulation results are reasonable compared with the experimental data based on the comparison in the aspect of velocity distribution. The effect of the spacer grid on the flow behavior in the fuel rod bundle has also been investigated. The mixing vanes in the spacer grid can cause various strong flow branches; these flow branches develop and cause the complex flow behavior in the fuel rod bundle. Additionally, the secondary flow, the swirling factor and the mixing factor were investigated for the quantitative analysis of the spacer grid effect on the flow mixing behavior in fuel rod bundle.

Thermal stratification is a common phenomenon in the surge line of Pressurized Water Reactors (PWRs). A study using Computational Fluid Dynamics (CFD) has been performed to analyze the thermal stratification in the pressurized surge line with different surge line and main pipe velocity conditions. This study is typical of considering both the flow subjected to the buoyancy and the conjugate heat transfer between the fluid and the pipe wall. What makes the research complex is exactly the impact of buoyancy on thermal stratification. To validate the accuracy of CFD calculations, necessary comparison is performed with experiment results. Models with different computational conditions resulting from various similarity criterion parameters are used. More meaningful information than that from the experimental results can be obtained. At the same time, many actual operating conditions which are difficult for experiment can be also accomplished by applying CFD mode. Temperature distributions inside and on the outer surface of the surge line are provided by the CFD mode with different turbulent models. Compared to the standard k-ɛ turbulence model and Reynolds stress model (RSM), the SST k-ω turbulent model is more adapted for this study. Based on the conclusion, the numerical results with the SST k-ω turbulence model for the prototype on the actual nuclear power plant unit were mainly described.

This paper focuses on investigating the geometric effects of inlets on global and local two-phase flow parameters in vertical-downward air–water two-phase flow. Flow visualization, frictional pressure loss analysis, and local experiments are performed in a test facility constructed from 50.8 mm inner diameter acrylic pipes. Three types of inlets of interest are studied: (1) two-phase flow injector without a flow straightener (Type A), (2) two-phase flow injector with a flow straightener (Type B), and (3) injection through a horizontal-to-vertical-downward 90° vertical elbow (Type C). A detailed flow visualization study is performed to characterize flow regimes including bubbly, slug, churn-turbulent, and annular flow. Flow regime maps for each inlet are developed and compared to identify the effects of each inlet. Frictional pressure loss analysis shows that the Lockhart–Martinelli method is capable of correlating the frictional loss data acquired for Type B and Type C inlets with a coefficient value of C = 25, but additional data may be needed to model the Type A inlet. Local two-phase flow parameters measured by a four-sensor conductivity probe in four bubbly and near bubbly flow conditions are analyzed. It is observed that vertical-downward two-phase flow has a characteristic center-peaked void profile as opposed to a wall-peaked profile as seen in vertical-upward flow. Furthermore, it is shown that the Type A inlet results in the most pronounced center-peaked void fraction profile, due to the coring phenomenon. Type B and Type C inlets provide a more uniform distribution of the void fraction profile with a reduced coring effect.

This study investigates elbow effects on single-phase and two-phase flows. ANSYS CFX is used to simulate the evolution of the single-phase velocity distribution and two-phase interfacial structure downstream of vertical elbows. Simulations are benchmarked with existing LDA data and detailed four-sensor conductivity probe data. It is found that elbows create secondary flow, containing two counter-rotating vortices in both single-phase and two-phase flows. The vortices shift to pipe center and develop into parabolic velocity profile downstream. The elbow effect dissipates exponentially and dissipates 90% of its initial maximum value at approximately 11D downstream, no matter the elbow direction. In two-phase flow, bubbles are entrained by the secondary flow to form double-peaked void fraction and interfacial area concentration distributions. The double-peaked distribution moves upward and becomes single-peaked distribution downstream. Due to the additional buoyant force, the elbow effect takes much longer length to dissipate in two-phase flow than in single-phase flow.

This study develops a one-group interfacial area transport equation (IATE) for vertical-upward-to-horizontal air–water bubbly two-phase flows through a 90° elbow with a non-dimensional centerline radius of curvature of three. In order to develop the model, an extensive database is established by acquiring local two-phase flow parameters using a four-sensor conductivity probe upstream and downstream of the elbow. The data show there exist three characteristic regions in void distribution, including a bimodal-to-bimodal region, a bimodal-to-single-peaked region, and a developed horizontal flow region with void accumulated at the top of the pipe cross-section. Using the database, the preliminary dissipation length model developed by Yadav et al. (2014b) is improved by including the transition region near the exit of the elbow in addition to the dissipation region. To close the IATE model, the bubble velocity advection term and bubble interaction terms in the IATE are correlated with the parameter characterizing the “elbow-strength”. The two-phase pressure drop across the elbow is modeled using the modified Lockhart–Martinelli correlation which takes into account the minor loss effect. The closed IATE model is implemented to predict interfacial area transport in vertical-upward-to-horizontal two-phase flow. It is found that the developed model is capable of predicting interfacial area concentration with an average percent difference of less than ±6%.

In this paper, the time-resolved particle image velocimetry (TR-PIV) and match index refractive (MIR) techniques were used to study the flow field in a large range (0 – 22 Dh) downstream a spacer grid (SG) in a 5 × 5 rod bundle channel at different Reynolds number. The sodium chloride solution (1%) is used as the working fluid to reduce the refractive index error of fluorinated ethylene propylene (FEP) and water. The proper orthogonal decomposition (POD) background removal technique was used to minimize the FEP reflection. These methods greatly reduced the interference of background noise and improved the accuracy of cross-correlation calculation. For TR-PIV velocity fields downstream of the mixing vanes, time-averaged, statistical, spectral, and cross-correlation analysis were performed for the instantaneous full-field experimental data. The transport characteristics of coherent structures in different subchannels of rod bundles are calculated and discussed. The results show that the SG caused a relatively large transverse velocity and reduces the axial velocity. With the increase of the Reynolds number, the SG promotes the generation of transverse flow and has a great resistance to the axial flow. There is relatively large turbulence intensity downstream of the SG due to the mixing effect. The attenuation of transverse turbulence intensity component is slower than the axial component. Moreover, spectrum analysis shows that cross-arranged mixing vanes will generate periodic vortices but single mixing vane will not. These periodic vortices gradually propagate downstream along the inner subchannel and dissipate in the gap subchannel due to the effect of viscosity. The cross-correlation analysis shows that the mixing effect of the SG will reduce the scale of the coherent structure, and increase the convection velocity. The results of current research are helpful for understanding the strong anisotropic turbulence in the rod bundle channel with SG. Finally, the experimental results can be utilized to benchmark the applicability of turbulence models under different Reynolds number and the performance of partially averaged Naiver–Stokes or multiple RANS algorithms downstream of the SG.

The flow rate of the primary coolant in a nuclear reactor will fluctuate during accident conditions or in floating reactors that are affected by inertial forces in the ocean. These fluctuations may have a substantial impact on heat transfer by the primary coolant in the reactor. In the current research, the flow field distribution and turbulence structure at different planes of a 5 × 5 rod bundle under pulsating flow conditions were measured using phase-locked particle image velocimetry and matching index refractive. The local ensemble-averaged velocity and root-mean-square (RMS) component distributions of different pulsating flow phases were obtained from the measurement results. This experimental study covered a pulsation period T range of 3 – 7 s and an imposed pulsation amplitude Au range of 0.31 – 0.65. For quasi-steady, intermediate-frequency, and high-frequency pulsating flows, the ensemble-averaged velocity of different phases and the distribution of velocity amplitude modulation in different subchannels were studied. Moreover, the phase shifts of the velocity modulation amplitude and RMS components of the rod bundle channel relative to the imposed pulsating flow with different pulsation parameters (Au, T) were also analyzed. Results for the rod bundle channel under the pulsating flow condition and for a conventional channel were compared; the discrepancies and agreements are discussed in detail. These research results can help to understand pulsating flow in a rod bundle channel better, and the experimental results can provide reference and validation data for numerical simulations.

Thermal stratified flow in a pressurizer surge line was investigated on a 1/3 scaled model based on the Richardson number similarity. A database which could be used to verify the thermal stratification simulation was established. Temperature distributions inside and on the outer surface of a surge line pipe with different surge line inlet flow rates were monitored and compared. Thermal stratification occupied a large part of the surge line. Temperature differences at all the measured test sections reached half of the maximum inlet temperature difference. As the surge line inlet velocity increased, the pattern of temperature distribution showed a similar trend and the cross-section maximum temperature difference changed little. This was consistent with the existence of thermal stratification at operating conditions in the corresponding pressurized water reactor.

Although much research has been conducted on investigating the flow boiling heat transfer of low saturation temperature refrigerants, there are few experimental data and theory about the flow boiling heat transfer of high saturation temperature organic mixture which exists widely in the petrochemical industry. To investigate the characteristics of flow boiling heat transfer of high saturation temperature organic mixture, experiments of glycol-water solution flow boiling in a vertical porous surface tube and a vertical smooth tube are conducted. Test tubes are uniformly heated by electrical current with a heated length of 2,000 mm. The mass flux in the experiment ranges from 500 to 1,500t⋅h−1 and the heat flux on test tubes ranges from 10 to 40 kW⋅m−2. The flow boiling heat transfer coefficients and two-phase frictional pressure drops of the two types of tubes are obtained and compared. The results indicate that: the flow boiling heat transfer coefficient in the porous surface tube is 3.8∼5.7 times of that in smooth tube and the pressure drop of the porous surface tube is 0.99∼1.007 times of that in the smooth tube. The physical mechanisms of the enhanced heat transfer characteristics of flow boiling in the porous surface tube are analyzed. By the regression analysis of the experimental data, correlations predicting the flow boiling heat transfer coefficient and pressure drop of glycol-water solution within the error range of ±20% are established. The experimental results can be used to guide the design of heat exchange equipment using the porous surface tube as heat transfer elements under these test conditions.

Properties of supercritical water will undergo drastic changes near the "pseudo critical" point which gives rise to a serious concern about flow instabilities under supercritical conditions. To investigate the fluid flow behaviors, experiments with different system pressures (near-critical and supercritical pressure) were conducted on the Xi'an Jiao tong University's (XJTU) supercritical water test facility. Large amplitudes of mass flow flux as well as inlet pressure were found in various operation conditions. Mechanism of this unusual flow instability behavior is discussed. The oscillation behavior occurring in the present work has also been compared with that under the subcritical pressures and parameters effects are also studied.