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Abstract: In this paper, the two-phase flow instability between multi-channels (FIBM) operating under ocean conditions is studied theoretically. The physical and mathematical model of the multi-channel system which is based on Lee and Pan [Lee, J.D., Pan, C., 1999. Dynamics of multiple parallel boiling channel systems with forced flows. Nuclear Engineering and Design 192, 31–44], is extended to analyze the influence of ocean conditions. The influence of ocean conditions on the FIBM is analyzed, particularly with respect to the periodic total mass flow rate and rolling motion. Furthermore, the instability oscillation trajectories of the multi-channel system are obtained on the phase plane of the inlet velocity and boiling boundary. Some of the trajectories show chaotic characteristic. The instability zone of a nine-channel system under rolling motion is then obtained. [Copyright &y& Elsevier]

Thermal hydraulic characteristics of sodium boiling are significant in the safety analysis for the sodium cooled fast reactor, one of the Generation IV reactors. However, the knowledge on the mechanism of sodium boiling is quite limited. In this study, boiling experiments on sodium flowing through an annulus are performed. The annulus is of 1000 mm length, 8 mm inner diameter and 12 mm outer diameter. The heat flux varies from 80 to 500 kW/m 2 , with inlet subcooling from 63 to 285 °C, inlet flow velocity from 0.02 to 0.5 m/s and system pressure from 3.67 to 103 kPa. The boiling phenomena, two phase friction pressure drop and boiling heat transfer coefficient are investigated. Correlations for the two-phase friction multiplier factor ϕ l 2 and boiling heat transfer coefficient are proposed, respectively. Predicted results show good agreement with the experimental data. [ABSTRACT FROM AUTHOR]

Highlights • Single bubble motion characteristics in inclined tube bundles were analyzed. • A new hypothesis of velocity decomposition and resistance superposition was proposed. • Theoretical models were developed to predict single bubble deviation angles in inclined tube bundles. Abstract A theoretical investigation on single bubble motion in inclined tube bundles was carried out. Due to the shear flow caused by the nonhomogeneous velocity distribution in an inclined tube bundle, single bubble motion was affected by the lift force in the horizontal direction. As a result, the flow direction of single bubble may deviate from the buoyancy direction. Then, both force analyses and velocity decompositions of two-phase in inclined tube bundles were conducted. Based on several hypotheses, a theoretical single-region model for D b > D ins and a two-region model for D b < D ins were developed to predict deviation angles of single bubble motion in inclined tube bundles. In general, more than 75 percent of Todreas's data could be evaluated well by the new developed models with a devaition of ±50% when D b ≥ D c. However, the developed models can not present excellent predictions for several deviation angle data when D b < D c , especially when S L - d < D b < D c. This unsatisfied agreements may be related to two factors. The one is the unsatisfied repearability of Todreas's data. The other one is that several flow mechanisms of small bubbles in inclined tube bundles are not clear yet. It's suggested that further experimental and theoretical investigations on mechanisms of bubble rotation, bubble entrainment and bubble bouncing at the curve surfaces of tubes could be conducted. [ABSTRACT FROM AUTHOR]

An experimental research has been carried out for vertically upward two-phase flow of air-water across inverted half U-tube bundle with or without anti-vibration bars. Pressure drop characteristics in inverted half U-tube bundle and anti-vibration bars were obtained and analyzed. New correlations for total pressure drop of two-phase flow in inverted half U-tube bundle and local pressure drop of two-phase flow in anti-vibration bars were developed. The measured two-phase pressure drops in inverted half U-tube bundle of three walls in the test sections were unequal. The pressure drop in the side wall dominated the total pressure drop of two-phase flow in inverted half U-tube bundle for 57 ≤  G  < 114 kg·m −2 ·s −1 , while the pressure drop in the front wall dominated for 114 <  G  ≤ 143 kg·m −2 ·s −1 . An obvious mass velocity effect was observed for the total pressure drop of two-phase flow in inverted half U-tube bundle. For a given Martinelli parameter, the total pressure drop of two-phase flow in inverted half U-tube bundle decreased with an increasing mass velocity. The two-phase local pressure drop data of anti-vibration bars could be fitted well in terms of the Martinelli parameter and mass velocity. [ABSTRACT FROM AUTHOR]

Shell-side two-phase pressure drops under cross-flow condition play an important role in the operation of heat exchangers. However, no well-verified general correlation for predicting two-phase friction multipliers in horizontal tube bundles is available in the open literatures. In the present paper, based on a combination of Kanizawa and Ribastki flow regime criteria with the modified Mao and Hibiki flow regime criteria, the flow patterns of two-phase vertically upward flow across horizontal tube bundles in the open literatures were identified. The results show that flow patterns have a significant effect on two-phase friction multipliers. For the bubbly flow, finely dispersed bubbly flow, and annular flow, the mass velocity effect on two-phase friction multipliers can be neglected. However, the two-phase friction multipliers of cap bubbly flow and churn flow decrease with an increasing mass velocity. Then, a general dimensionless correlation in terms of the Martinelli parameter and the two-phase Froude number were developed, which is able to accurately predicting gas–liquid two-phase flow resistances of churn flow in staggered horizontal tube bundle under cross-flow condition. New correlations for predicting two-phase friction multipliers of bubbly flow, finely dispersed bubbly flow and annular flow both in staggered and in-line horizontal tube bundles were also developed and can give excellent representations for the existing data in the open literatures. [ABSTRACT FROM AUTHOR]

Single-phase convection and steam-water two-phase flow boiling heat transfer experiments in an electrically heated 4 × 16 staggered horizontal rod bundle under cross-flow condition have been carried out for the following range: 45 kg m −2  s −1  ≤  G  ≤ 383 kg m −2  s −1 , 20.0 kW m −2  ≤  q  ≤ 55.0 kW m −2 , 0.02 ≤  x out  ≤ 0.57 and 112 kPa ≤  P in  ≤ 190 kPa. The major effects on local flow boiling heat transfer coefficients in horizontal rod bundles were analyzed. Firstly, the row number has a significant effect on local flow boiling heat transfer coefficients since the vapor bubbles generated from the lower rods impinge on the surface of the upper rods and enhance the turbulence there. Secondly, an increasing heat flux contributes to an increase of the average bundle flow boiling heat transfer coefficient at all mass velocities. Thirdly, the local quality can also slightly enhance the flow boiling heat transfer. Fourthly, the flow boiling heat transfer can be enhanced with an increasing mass velocity under lower heat fluxes. In a general, a Chen-type correlation was developed to predict local flow boiling heat transfer coefficients in horizontal rod bundles with a maximum deviation of ±20%. [ABSTRACT FROM AUTHOR]

This experimental study was aimed at turbulence researches in ECC T-junction structure and it was conducted on ECC condensation heat transfer in a horizontal T-junction pipe. By using sub-cooled water in different flow rates, various flow patterns had a comparison with saturate steam in the cold leg. In the researches, the first part is dedicated to obtain the effects of hydraulic jump and back flow on T-junction thermal mixing. This research, which was based on momentum method, tries to find the influence of different flow patterns on the temperature distribution and capture the flow patterns' picture as reference. The equations, which can be employed to analyze temperature change, have been deduced after the research. And the second part of the researches focused on two-phase flow status after SI coolant mixed. The conclusion was deduced according to the pure steam SI injected condition. With some reasonable assumptions, the analysis included both steam Reynolds number and liquid Reynolds number correlations. The length of downstream is longer than that of the scaled-down distance from SI line to downcomer in PWR. And this could contribute to PWR safety analysis and find the relation between the flow status and mixed length. In addition, the test also figured out the relative velocity changes in T-junction thermal mixing following the Reynolds number research on mixed two-phase flow. This correlation could contribute to structure design improvement in PWR and prevent excessive steam from rushing to downcomer. All equations and correlations show good agreement within experiment data. [ABSTRACT FROM AUTHOR]

Single-phase convective heat transfer and steam-water two-phase flow boiling heat transfer experiments have been carried out with a vertical seven-rod bundle at low flow conditions near atmospheric pressure. With respect to single-phase experimental results, new correlations were developed to predict convective heat transfer coefficients of laminar flow and Reynolds numbers at the transition from the laminar region to the turbulent region. Two-phase experimental results show that the rod bundle geometry enhanced heat transfer compared with circular tubes. An obvious mass velocity effect on the flow boiling heat transfer coefficient was observed. This conclusion indicates that forced convection, instead of nucleate boiling heat transfer, dominated two-phase flow boiling heat transfer under the experimental conditions. In addition, two-phase flow boiling heat transfer coefficients in the vertical rod bundle increased first and then decreased with the quality increases for a given mass velocity. By taking the enhancement factor of forced convection and the suppression factor of nucleate boiling into account, a new correlation in terms of Boiling number, Reynolds number and Martinelli number was developed to be successful in predicting predict two-phase flow boiling heat transfer coefficients. [ABSTRACT FROM AUTHOR]

Pressure drop experiments were conducted for a vertical two-phase air–water flow across a horizontal staggered rod bundle under conditions of air quality ranging from 0.001 to 0.3 and mass velocity ranging from 80 to 500 kg/(m 2 s). The experimental data were thoroughly analyzed and compared with predictions from existing correlations. New developed expressions provided the most accurate predictions. The two-phase friction multipliers for G ⩽ 200 kg/(m 2 s), which show a strong mass velocity effect, could be fit well in terms of the Martinelli parameter and the dimensionless mass velocity. The new correlation developed for the two-phase friction multiplier was successfully tested in predicting the experimental data for G ⩽ 200 kg/(m 2 s) in previous experiments of other researchers. Distributions of velocity and turbulent intensity were also obtained from the numerical simulation on single-phase flow of water using the CFD code ANSYS FLUENT 14.0. The SST k–ω model was verified to be the best turbulent model to gain flow characteristics in the horizontal rod bundle under experimental conditions. [ABSTRACT FROM AUTHOR]

Abstract: The transient thermal-hydraulic code THACOS is under development for analysis of China Demonstration Fast Reactor. Applying modular technology, the code contains the core module, the pump module, the sodium pool module and the heat exchanger module and each module could operate separately. It can provide one-dimensional thermal-hydraulic simulation for the primary sodium coolant loop. The point reactor kinetics equations with six-group delayed neutrons have been applied to calculate the core power considering reactivity feedbacks caused by the Doppler effect, coolant density, axial expansion of fuel rods and radial expansion of the core. Multiple-channel model is applied to depict the core. Compressible homogenous flow model is used for the two-phase flow of sodium. The calculated results show that sodium boiling will occur quickly under the ULOF accident without any shutdown rods insertion. While, with the insertion of three hydraulically suspended shutdown rods, the core could be shut down safely and boiling will not occur in a short period of time. Obviously, the passive hydraulically suspended shutdown rods could keep the core safe under ULOF accidents. [Copyright &y& Elsevier]

Abstract: In the present study, we preformed a two-dimensional numerical simulation of the motion and coalescence of bubble pairs rising in the stationary liquid pool, using the moving particle semi-implicit (MPS) method. Moving particles were used to describe the liquid phase and the vapor phase was evaluated using real vapor sate equation. The bubble–liquid interface was set to be a free surface boundary which could be captured according to the motion and location of interfacial particles. The behaviors of coalescence between two identical bubbles predicted by the MPS method were in good agreement with the experimental results reported in the literature. Numerical results indicated that the rising velocity of the trailing bubble was larger than that of the leading bubble. Both of the leading bubble and the trailing bubble rose faster than the isolated bubble. After coalescence, the coalesced bubble showed velocity and volume oscillations. The time of the volume oscillations increased with increasing initial bubble diameter. The wake flow and vortex would form behind the coalesced bubble. [Copyright &y& Elsevier]

Abstract: Two-phase flow instability of two-channel system has been theoretically studied in the present study. Based on the homogeneous flow model, the parallel channels model and system control equations are established by using the control volume integrating method. Gear method is used to solve the system control equations. The marginal stability boundary (MSB) at different system pressure conditions is obtained. The typical MSB shape is usually a classical inclination “L” at some operation condition (i.e. the system pressure is low and the inlet resistance coefficient is small). The three-dimensional instability spaces (or instability reefs) with different inlet resistance coefficients are obtained. The three coordinates consists of phase change number (N pch ), subcooling number (N sub ) and nondimensional pressure (P +). The lower part of the instability space is larger than the upper one. Increasing the system pressure or inlet resistance coefficient can strengthen the system stability. However, increasing the heating power destabilize the system stability. The influence of inlet subcooling on the system stability is multi-valued. [Copyright &y& Elsevier]

• A 2-D multiphase fluid dynamics code ACENA has been preliminarily developed. • This code is intended to simulate multiphase flow during LFR postulated accidents. • The models adopted in this code has been validated against several experiments. • The prediction results of this code agree well with experimental results. To simulate the heavy liquid metal-gas (HLM-gas) two phase flow phenomena that may occur during postulated accidents of Lead-based fast reactors (LFRs), a 2D multiphase fluid dynamics code, ACENA, is preliminarily developed and validated in this study. A brief description of theoretical models, including governing equations, auxiliary models and material properties are listed in this study. The capability of simulating flow and heat transfer characteristics between gas and liquid phases has been examined by typical air-water bubble column experiment, LBE-N 2 two phase flow experiment and Lead-Ar heat transfer experiment, and the comparison results show that this code could attain acceptable accuracy in simulating flow and heat transfer characteristics of HLM-gas two phase flow with appropriate interfacial drag coefficient and heat transfer coefficient models. [ABSTRACT FROM AUTHOR]

Experimental investigations on atmospheric steam-water two-phase flow regimes and transitions in two transparent narrow rectangular channels with cross-section areas of 1 mm × 20 mm and 2 mm × 20 mm were carried out under conditions of 0.007 ≤ J f ≤ 0.27 m s−1, 0.05 ≤ J g ≤ 65.2 m s−1, 0 < q ≤ 18 kW m−2. Four flow patterns were observed in narrow rectangular horizontal channels under heated condition, i.e., bubbly flow, slug flow, churn flow and annular flow. New dryout phenomenon of annular flow was observed under high wall heat flux conditions. The liquid phase occupies half of the middle of narrow channel above the bottom rectangular plate. Below the top plate, liquid phase is concentrated to a slender streamline in the middle of upper half narrow channel by high speed steam flow. For the flow regime transitions, the critical values of void fraction for transitions from bubbly flow to slug flow, slug flow to churn flow, and churn flow to annular flow in narrow rectangular horizontal channels under heated condition are 0.7, 0.8 and 0.9, respectively. The heat flux decays flow regime transitions because the generated bubbles from heated wall surfaces promote the separation of bubbles in the upstream and downstream. Flow regime transitions under larger surface tension condition need greater vapor phase velocities to disturb the mainflow. In a general, new flow regime transition criteria were developed and can present excellent predictions for the present data of flow regimes and transitions in narrow rectangular horizontal channels under heated condition. Image 1 • Flow pattern data of steam-water two-phase in narrow rectangular channels are obtained. • The effect of heat flux on flow pattern transition is analyzed. • New flow regime transition criterias in narrow rectangular horizontal channels are developed. [ABSTRACT FROM AUTHOR]

• CHF data in a vertical tube with an inner diameter of 6.0 mm under oscillatory flow conditions were obtained. • The effect of flow oscillation on periodic and continuous dryout was analyzed. • CHF data were compared with predictions calculated by previous correlations. Experimental investigations on single-phase convection and critical heat flux in a vertical tube with an inner diameter of 6.0 mm under inlet flow oscillation conditions were carried out for the following range: 255 ≤ G av ≤ 455 kg·m−2·s−1, 0.5 ≤ P ≤ 3.0 MPa, 0 < Δ G / G av ≤ 3.0 and 0 < τ (period of flow oscillation) ≤ 10.6 s. Single-phase convective results show that the fluctuation amplitude of single-phase pressure drop decreases with an increasing τ. Besides, there may be a critical value τ c. The average value of wall temperature under steady state condition is smaller than that under oscillatory flow condition when τ > τ c while the former is larger than the latter when τ < τ c. With respect to two-phase results, the effect of flow oscillation on fluctuation amplitude of wall temperatures under flow boiling conditions is quite small. In addition, flow oscillation deteriorates flow boiling heat transfer in vertical tubes to some extent when 1.04 ≤ τ ≤ 10.6 s. The normalized periodic CHF (F p) decreases with an increasing average mass velocity, especially for short periods of flow oscillation. In addition, the declining rate of F p as a function of Δ G / G av was decayed with an increasing period of flow oscillation. In a general, 2006 CHF lookup-table and Kim et al. correlation are recommended to calculate CHF under steady state condition and periodic dryout heat flux under inlet flow sinusoidal oscillation condition, respectively. [ABSTRACT FROM AUTHOR]