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Hydrodynamic force acting on the structures, pipes and various forms of objects can generate destructive vibrations, and could cause acoustic and noise problems in industrial machineries. Such phenomenon is known as Flow-Induced Vibration (FIV), and it can obstruct smooth operation of engineering devices and could potentially cause serious consequences like system failures. The subject has become increasingly important problem in engineering industry in recent years for both single-phase and multi-phase flow cases, as well as for various flow orientations including external and internal flows. Present review paper summarizes the historical background of FIV research and how the phenomenon has been classified in both industrial and academic fields, particularly focusing on the progress of two-phase FIV research. Special attention was paid to the subject of internal two-phase FIV generated at industrial piping systems two-phase flow regimes. Based on the extensive and comprehensive literature survey, most up-to-date progress of the research in the area of two-phase flow induced vibration in piping system are thoroughly reviewed and presented in this article. [ABSTRACT FROM AUTHOR]

Thermal and hydraulic phenomena of the dryout and rewetting on a fuel rod were discussed, based on the transient process observed visually with the surface-temperature variations measured during the transition from dryout to rewetting with a high-speed CCD video movie under BWR-rated high pressure and temperature conditions of 7 MPa and ~559 K (~286°C) in a thermal-hydraulics experimental loop. Liquid film was observed falling under CCFL from the downstream spacer placed in upper side on the electrically heated rod after the dryout. The heater rod surface could be cooled with the liquid film falling in the early stage of the rewetting and then the surface temperatures recovered from unusually elevated levels with quench-rewetting. Graphical Abstract: [Figure not available: see fulltext.] [ABSTRACT FROM AUTHOR]

Abstract: Bent pipe wall thinning has been often found at the elbow of the drain line and the high-pressure secondary feed-water bent pipe in nuclear reactors. Liquid droplet impingement (LDI) erosion could be regarded as one of the major causes and is a significant issue of the thermal hydraulics and structural integrity in aging and life extension for nuclear power plant safety. In this paper a computational methodology is established for simulation of LDI erosion using computational fluid dynamics (CFD) simulation and theoretical calculation. Two-phase flow numerical simulations are conducted for standard elbow geometry, typically with the pipe diameter of 170mm. This computational fluid model is built up by incompressible Reynolds Averaged Navier–Stoke equations using standard k–ɛ turbulence model and the SIMPLE algorithm, and the numerical droplet model adopts the Lagrangian approach. The turbulence damping in vapor–droplets flow is theoretically analyzed by a damping function on the energy spectrum basis of single phase flow. Locally, a droplet impact angle function is employed to determine the overall erosion rate. Finally, the overall and local investigations are combined to purpose a general methodology of LDI erosion prediction procedure, which has been complemented into CFD code. Based on our more physical computational results, comparison with an available accident data was made to prove that our methodology could be an appropriate way to simulate and predict the bent pipe wall thinning phenomena. [Copyright &y& Elsevier]

Abstract: Calibration tests of UdFlow, the ultrasonic pulse-Doppler flowmeter manufactured by the Tokyo Electric Power Company, were conducted at the national standard loop in Mexico, CENAM (The Centro National de Metrologia), in order to evaluate the accuracy of the flowmeter. Four ultrasonic transducers were mounted circumferentially on the surface of 100 and 200 mm stainless steel pipes to measure four velocity profiles. Flow rates can be obtained by integrating each measuring line and averaging them. Air was injected upstream of the measuring point to provide bubbles as ultrasonic reflectors. Tests were conducted at five different flow rates with Reynolds numbers from 200,000 to 1,200,000. Tests were repeated six times at each flow rate to evaluate repeatability. In addition, a take-off and put-back test was carried out on the 100 mm pipe at a flow rate of 3000 L/min to evaluate reproducibility. The values of the CENAM loop are based on the average of weighing time while those of the ultrasonic-Doppler flow velocity-profile flowmeter are based on the time average of instantaneous values. The calibration tests found a deviation of less than 0.3% between the two devices in terms of the average of the values recorded in six rounds of measurement. Measurement at a different Reynolds number showed that the overall average deviation between the two devices was less than 0.3%. [Copyright &y& Elsevier]

Abstract: At hydraulic power stations, Pitot tubes have commonly been used to measure flow rates in steel penstocks for performance testing of hydraulic turbines. Due to the difficulties of Pitot tube installation, transit-time ultrasonic flow meters are becoming a popular replacement, but their accuracy is sensitive to velocity profiles that depend on Reynolds numbers and pipe surface roughness. Ultrasonic pulse Doppler flow meters have recently gained favor as suitable tools to measure flow rates in steel penstocks because they can measure instantaneous velocity profiles directly. Field tests were conducted at an actual hydraulic power plant using an ultrasonic pulse Doppler flow meter, and it was found capable of measuring velocity profiles in a large steel penstock with a diameter of over one meter and Reynolds number of more than five million. Furthermore, two ultrasonic transducers were placed on the pipe surface to validate the multi-line measurement of asymmetric flow. Each transducer recorded the velocity profile simultaneously from the pipe centerline to its far wall during plant operation. Velocity profiles were obtained from three-minute measurements to improve the accuracy of flow rate measurements. [Copyright &y& Elsevier]

Two advanced boiling water reactors (ABWRs) whose electric output power is 1356 MW have been commercially operated since 1996 and 1997 by the Tokyo Electric Power Company (TEPCO) in Japan. Features of an ABWR are reactor internal pumps (RIPs) placed in the lower plenum and downcomer, peripherally bottom-mounted on the reactor pressure vessel--which should require different modeling from the jet pumps and two recirculation pumps in the primary outer-loop recirculations of BWR-5. Efforts focused on modeling and simulating the ABWR with transient analyses by the point-kinetics model with the local reactivity modified by local importance weighting of the squared nodal power during start-up tests using the RETRAN-3D code, version MODO03 without three-dimensional kinetics. The core and reactor pressure vessel including ten ABWR RIPs and the steam lines were modeled, and simulations were carried out for the cases of the one-pump trip test, the changing-setpoint tests, the main-steam-isolation-valve-closure test, and the generator load rejection test with bypass. The analytical simulation with RETRAN-3D/ MOD003 well reproduced the measured data of the ABWR in operation for the RIP trip and the transient tests and could demonstrate its validation for applying to the ABWR with modeling of RIPs. [ABSTRACT FROM AUTHOR]

A BWR core noise monitoring system is developed for addressing core anomaly problems in future advanced core operation. In order to monitor in-core status from a limited number of signals, various up-to-date signal processing algorithms are introduced to compensate for a lack of information. These algorithms, such as independent component analysis, factor analysis and model based parameter estimation, are demonstrated to be effective through real plant data analysis to evaluate core and regional stability index, reactivity coefficients and core flow rate. Through these practices, we demonstrate that the core noise monitoring system is an effective general platform for providing a variety of monitoring tools to meet the requirements in future advanced core operation. [Copyright &y& Elsevier]

Abstract: Liquid droplet impingement (LDI) erosion could be regarded to be one of the major causes of unexpected troubles occasionally occurred in the inner bent pipe surface. Evaluating the LDI erosion is an important topic of the thermal hydraulics and structural integrity in aging and life extension for nuclear power plants. One of the causes of LDI erosion is the impact pressure by the impingement of droplets in the involved steam. We investigated a simple droplet impingement to a rigid wall using volume of fluid (VOF) model, which is a two-phase Eulerian–Eulerian approach. The impact of a single water droplet with a high velocity towards a solid surface is examined numerically. The high Reynolds number value implies inertia dominated the phenomena and supports an inviscid approach to the problem. The high Weber number is justifying that an assumption to neglect the surface tension effect is adopted. We show that the compressibility of the liquid medium plays a dominant role in the evolution of the phenomenon. Both generation and propagation of shock waves are directly computed by solving the fluid dynamics continuity and momentum equations. In the simulation we employed a front tracking solution procedure, which is particularly suitable for two-phase free surface computation. The numerical results show that critical maximum pressure is not highest at the center of droplet contact on the surface at the first instantaneous moment but highest behind the contact angle later before jet eruption. It agrees generally well (within 20%) with the mathematical analysis. Finally, a droplet impact angle function is proposed for the global LDI erosion prediction. [Copyright &y& Elsevier]

Abstract: Prediction of amount of entrained droplets or entrainment fraction in annular two-phase flow is essential for the estimation of dryout condition and analysis of post dryout heat transfer in light water nuclear reactors and steam boilers. In this study, air–water and organic fluid (Freon-113) annular flow entrainment experiments have been carried out in 9.4 and 10.2mm diameter test sections, respectively. Both the experiments covered three distinct pressure conditions and wide range of liquid and gas flow conditions. The organic fluid experiments simulated high pressure steam–water annular flow conditions. In each experiment, measurements of entrainment fraction, droplet entrainment rate and droplet deposition rate have been performed by using the liquid film extraction method. A simple, explicit and non-dimensional correlation developed by Sawant [Sawant, P.H., Ishii, M., Mori, M., 2008. Droplet entrainment correlation in vertical upward co-current annular two-phase flow. Nucl. Eng. Des. 238 (6), 1342–1352] for the prediction of entrainment fraction is further improved in this study in order to account for the existence of critical gas and liquid flow rates below which no entrainment is possible. Additionally, a new correlation is proposed for the estimation of minimum liquid film flow rate at the maximum entrainment fraction condition. The improved correlation successfully predicted the newly collected air–water and Freon-113 entrainment fraction data. Furthermore, the correlations satisfactorily compared with the air–water, helium–water and air–genklene experimental data measured by Willetts [Willetts, I.P., 1987. Non-aqueous annular two-phase flow. D.Phil. Thesis, University of Oxford]. However, comparison of the correlations with the steam–water data available in literature showed significant discrepancies. It is proposed that these discrepancies might have been caused due to the inadequacy of the liquid film extraction method used to measure the entrainment fraction or due to the change in mechanism of entrainment under high liquid flow conditions. [Copyright &y& Elsevier]

Abstract: Upward annular two-phase flow in a vertical tube is characterized by the presence of liquid film on the tube wall and entrained droplet laden gas phase flowing through the tube core. Entrainment fraction in annular flow is defined as a fraction of the total liquid flow flowing in the form of droplets through the central gas core. Its prediction is important for the estimation of pressure drop and dryout in annular flow. In the following study, measurements of entrainment fraction have been obtained in vertical upward co-current air–water annular flow covering wide ranges of pressure and flow conditions. Comparison of the experimental data with the existing entrainment fraction prediction correlations revealed their inadequacies in simulating the trends observed under high flow and high pressure conditions. Furthermore, several correlations available in the literature are implicit and require iterative calculations. Analysis of the experimental data showed that the non-dimensional numbers, Weber number (We = ρ g〈j g〉2 D/σ(Δρ/ρ g)1/4) and liquid phase Reynolds number (Re f = ρ f〈j f〉D/μ f), successfully collapse the data. In view of this, simple, explicit correlation was developed based on these non-dimensional numbers for the prediction of entrainment fraction. The new correlation successfully predicted the trends under the high flow and high pressure conditions observed in the current experimental data and the data available in open literature. However, in order to use the proposed correlation it is necessary to predict the maximum possible entrainment fraction (or limiting entrainment fraction). In the current analysis, an experimental data based correlation was used for this purpose. However, a better model or correlation is necessary for the maximum possible entrainment fraction. A theoretical discussion on the mechanism and modeling of the maximum possible entrainment fraction condition is presented. [Copyright &y& Elsevier]

Several researchers have reported the detachment of vapor bubbles from a vertical heated wall in upward flow boiling of water. In these experiments, the bubble shape was frequently flattened along the wall at inception but became more rounded before the onset of detachment. In this study, the motion of two-dimensional bubbles is investigated numerically to elucidate the mechanism of bubble detachment. The numerical results show that the change in bubble shape due to surface tension force induces the liquid flow from the periphery to the base of bubble and consequently the liquid inertia lifts the bubble from the wall. [ABSTRACT FROM AUTHOR]

New constitutive models for the interfacial forces acting on bubbles were developed for accurately predicting the lateral phase distribution in turbulent bubbly two-phase flow in vertical channels. Several experimental measurements have revealed that the lateral void profile in bubbly two-phase flow varies from the void peaking near the wall to the almost flat distributions as the liquid velocity increases. However, within the authors'' knowledge, the effect of liquid velocity on the void profile has not been successfully predicted by the existing models; this would indicate the strong limitation of the existing multidimensional two-phase flow models. In view of these, the validity of the present constitutive models was tested in varied conditions of the liquid velocity as well as the bubble size. Since several assumptions were required in the models mainly due to the insufficient knowledge of the bubble motion, further improvements should still be needed. Nevertheless, the predicted lateral phase distributions were found to be in reasonably good agreement with available experimental data. It is hence expected that the present constitutive models can effectively be used in the practical applications and also be the base of the more sophisticated ones. [Copyright &y& Elsevier]

• Evaluation of decay heat removal systems of FHR during thermal-hydraulic transient was performed. • Heat removal performances of FHR systems are investigated. • FHR's decay heat removal system shows sufficient heat removal capability to avoid catastrophic accidents. The Fluoride-salt-cooled High-temperature Reactor (FHR) is one of the Generation IV nuclear reactor concepts being investigated mainly in the U.S. and China. Liquid salt with high boiling point of over 1400 °C is used as a coolant and its baseline fuel is the graphite-matrix coated-particle fuel developed for high-temperature gas-cooled reactors. To establish advanced safety systems in FHR, passive decay heat removal system which utilizes fluid's natural circulation is considered. Three systems are proposed: Direct Reactor Air Cooling System (DRACS), Reactor Vessel Air Cooling System (RVACS), and Silo Cooling System (SCS). Although these three safety systems were previously developed for other types of reactors, their heat removal performances need to be evaluated for the FHR plant. The current study provides evaluation of decay heat removal systems of FHR during thermal-hydraulic transient. In addition, FHR's grace period to start the reactor cooling was determined under the condition where no operable heat removal systems are available. The current study shows that FHR's decay heat removal systems possess sufficient heat removal capability to avoid catastrophic accidents. The construction of Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is currently being planned and its development is a key step to realize the commercialization in near future. This study provides performance evaluation of decay heat systems for FHTR under thermal-hydraulic transient, and also suggests the adequate operation procedure suitable for FHTR to avoid a severe accident. [ABSTRACT FROM AUTHOR]

Utilization of one-dimensional system analysis code such as TRACE, RELAP5, TRAC-BF1 to evaluate gas-liquid two-phase flow behaviors in nuclear power plants is crucial for the plant-level safety assessment. In the two-fluid model, interfacial momentum transfer between two phases is expressed under interfacial drag term in the momentum equation. For the rigorous and accurate expression of interfacial drag term and drift flux parameter, covariance due to the area averaging of void fraction distribution must be considered. In the present paper, an effect of the covariance on void fraction prediction in pipe flow was numerically assessed by implementing Hibiki and Ozaki's model into the interfacial drag term in the one-dimensional two-fluid model. For the low flow rate with high void fraction conditions, it was found that the inclusion of covariance model slightly underestimated void fraction value than that calculated by the drift-flux model. This underestimation comes from the momentum source term in the two-fluid model, which was derived under the assumption of uniform void fraction distribution. Therefore, in this paper, momentum source term was rederived with consideration of void fraction covariance and a complete set of momentum equation and constitutive formulations for the one-dimensional two-fluid model is presented. [ABSTRACT FROM AUTHOR]

Steam injector (SI) operates as passive jet pump and heat exchanger, which is activated only by the high-pressure steam and water. SI possess simple geometry and does not require any external power supplies nor rotating machinery for the operation. It is capable of discharging subcooled water at higher pressure than the inlet, and operable as a passive jet pump. In addition, direct contact condensation heat transfer between steam and water-jet provides superior heat exchanging characteristics, more than 1,000 times that of shell and tube heat exchangers. Due to these advantageous features, SI has a great potential to be applied as a passive safety system of nuclear power plants including existing light water reactors and next generation reactors. In the present study, SI experiments targeting for the operating range and the pumping performance were carried out at 0.02–0.81 MPa inlet steam pressure and 0.21–0.80 kg/s inlet water flow rate. The SI body was manufactured by stainless steel equipped with overflow port. The water injection nozzle was designed with shaft-driving mechanism to adjust the axial location of the water jet and steam inlet area. During the experiment, SI's pump performance and operation mode were investigated by changing the inlet conditions. Maximum attainable discharge pressure and operation state of the SI were recorded at each inlet condition. In this study, liquid jet break-up length was considered to assess the SI's operation mode. Clear boundaries to explain the SI's operational range were obtained using the jet stability analysis. The research results presented here provides significant aspect to properly design operable SI as the passive safety system. [ABSTRACT FROM AUTHOR]

In this paper, experimental investigation of the molten metal spreading behavior that was carried out at Hokkaido University using high frequency inductive heater is presented to address the fundamental behavior of the molten metal spreading and deposition behaviors on dry flat plate. Molten copper was utilized as a test sample, and dataset was obtained for the falling molten metal on dry stainless-steel plate at various elevations, nozzle sizes and initial temperatures. During the spreading transient, high-speed thermo-camera was utilized to measure the molten metal’s surface temperature. Immediately after the solidification, solidified molten metal’s spread area and deposition thickness were measured. Based on the database obtained, dimensional analysis was conducted to identify the key parameters responsible for the molten metal spreading. From the obtained database, new experimental correlation was developed which is capable of predicting the spreading area at reasonable accuracy. Present analysis provides characteristic information of molten metal spreading and deposition behaviors which will be useful for the corium relocation problem in severe accident analysis. [ABSTRACT FROM AUTHOR]

In this paper, experimental investigation of the molten metal jet's colliding and spreading behaviors on the flat steel surface covered with water layer was carried out. High-frequency induction heating system was utilized to produce the molten metal sample and it was released to the wet surface from a fixed elevation. As the molten metal collides against the surface, it rapidly goes through solidification while spreading on the wet surface. High-speed thermo-camera was utilized to measure the molten metal's surface temperature during the spreading transient. Once the molten metal completely solidifies, molten metal's spread area and thickness were measured. From the obtained database, a dimensional analysis was conducted to investigate the key parameters responsible for the molten metal spreading on the wet surface. Based on the key non-dimensional parameters identified in the current analysis, the new empirical correlation was proposed. Its predictive capability was found to be 18.9% in mean absolute relative deviation. [ABSTRACT FROM AUTHOR]

Flow boiling heat transfer characteristics of surfactant solution was studied and compared to that of pure water. The tested fluid was an aqueous solution of cetyltrimethyl ammonium chloride (CTAC) with addition of sodium salicylate (NaSal) at the same mass concentration. Tests were performed with 6.0 × 3.5 mm cross-section channel over flow rate range of 192–403 ml/min, at inlet temperature of 80 °C and an outlet pressure of 101.3 kPa. Liquid column method and resistance method of temperature determination were adopted to obtain accurate experimental data. The present experimental results show that, the flow boiling in CTAC/NaSal solutions at an appropriate concentration (100 ppm, ppm refers to part per million) behaved a good flow boiling performance. The heat transfer coefficient of 100 ppm CTAC/NaSal solution at the critical heat flux (CHF) was increased to 1.36 times compared to that of pure water. Besides, the CHF of 0.08 mm thick Nickel foil was increased to about 126% by surfactant addition when the concentration was beyond 100 ppm. The addition of CATC/NaSal increased the nucleation site density and inhibits bubble coalescence which were likely to be causes of enhancing the boiling heat transfer coefficient and CHF of subcooled flow boiling of CATC/NaSal aqueous solutions. [ABSTRACT FROM AUTHOR]

An experimental investigation of horizontal forced vibration effect on void fraction variation of subcooled boiling flow was carried out in this study. In order to simulate the fuel assembly subchannel of a boiling water reactor (BWR), an annular test section with inner and our diameters of 19.1 and 38.1 mm was utilized for subcooled boiling tests under atmospheric pressure. The annular test section was attached to an eccentric-cam vibrator, which was driven by a low-speed motor and can produce horizontal forced vibrations with frequency up to 20 Hz and maximum displacement of 22.2 mm. The inlet liquid velocity and subcooling were set as v f , in = 0.25–1.00 m/s and Δ T Sub = 5–20 °C. Different heat fluxes of q ″ = 0.058–0.193 MW/m 2 were loaded through the center heater rod, and the void fraction and fluid temperature were measured during the tests under stationary (no vibration) and vibration conditions. Test results show that in the subcooled boiling region, the void fraction and fluid temperature can vary under horizontal forced vibrations, and the variation trends were presented in N Zu - N Sub and v f , in -〈 α 〉 plots. These variations can be explained by the potential changes of thermal boundary layers (TBL) and the heat transfer enhancement under vibrations. In addition, no significant change of void fraction and fluid temperature was found in or near the saturated boiling conditions under vibrations. [ABSTRACT FROM AUTHOR]

On March 11, 2011, huge earthquake and tsunami attacked Fukushima Daiichi Nuclear Power Plant. After the accident, research on plant decommissioning has become actively worldwide. Several research institutes have performed experiments to investigate methods of identifying the location and spreading/deposition behaviors of molten core debris in the bottom of Primary Containment Vessel (PCV) using Severe Accident (SA) analysis codes. Nevertheless, knowledge of spreading and deposition behaviors of corium is not sufficient, especially phenomena involving collision against the floor surface. In this study, experimental investigations on molten metal spreading and depositing behaviors on the steel plate were carried out. Zinc and copper were utilized for the molten metal samples and spreading behaviors were carefully observed using high speed video camera. Immediately after the collision between falling molten metal and steel surface, initial pause on spreading was observed. New scaling relation based on Dinh et al. (2000) was developed by focusing on the initial spreading pause of the molten metal droplet. Proposed correlation is capable to predicting the spread and deposition of falling molten metal at average error of 18.1%. [ABSTRACT FROM AUTHOR]

In order to investigate the potential seismic vibrations effect on two-phase flow in an annular channel, experimental tests with air-water two-phase flow under horizontal vibrations were carried out. A low-speed eccentric-cam vibration module capable of operating at motor speed of 45–1200 rpm ( f  = 0.75–20 Hz) was attached to an annular channel, which was scaled down from a prototypic BWR fuel sub-channel with inner and outer diameters of 19.1 mm and 38.1 mm, respectively. The two-phase flow was operated in the ranges of 〈 j f 〉 = 0.25–1.00 m/s and 〈 j g 〉 = 0.03–1.46 m/s with 27 flow conditions, and the vibration amplitudes controlled by cam eccentricity ( E ) were designed for the range of 0.8–22.2 mm. Ring-type impedance void meters were utilized to detect the area-averaged time-averaged void fraction under stationary and vibration conditions. A systematic experimental database was built and analyzed with effective maps in terms of flow conditions (〈 j g 〉-〈 j f 〉) and vibration conditions ( E - f and f - a ), and the potential effects were expressed by regions on the maps. In the 〈 j g 〉-〈 j f 〉 maps, the void fraction was found to potentially decrease under vibrations in bubbly flow regime and relatively lower liquid flow conditions, which may be explained by the increase of distribution parameter. Whereas and the void fraction may increase at the region closed to bubbly-to-slug transition boundary under vibrations, which may be explained by the changes of drift velocity due to flow regime change from bubbly to slug flows. No significant change in void fraction was found in slug flow regime under the present test conditions. [ABSTRACT FROM AUTHOR]

A filtered containment venting system (FCVS) prevents over pressurization of containment vessel and releasing of radioactive materials during the severe accidents in nuclear power plants. During the venting process, it has been reported that the two-phase mixture level in a wet FCVS tends to swell and fluctuate. The behaviors depend on inlet/boundary conditions and physical properties of the injected gas, which vary as the accident progresses. Proper controlling and monitoring of the FCVS pool water level is crucial because it affects filtration performance including scrubbing process and thermal-hydraulic stability. In order to investigate this phenomenon, the current study proposes a set of nitrogen and steam injection experiments using a vertical pipe with a diameter of 105 mm to evaluate the effects of flow conditions and physical properties of gases. Drift flux analysis was carried out to predict the two-phase mixture water level and its fluctuations. The experimental two-phase mixture level was consistent with the values predicted by the drift flux model for nitrogen and steam injection, and the model's capability was confirmed for the system pressure ranging from atmospheric to 0.20 MPa and initial water level ranging from 0.6 to 2.6 m for both small and large diameter pipe configurations. The fluctuation amplitudes in the current experiment were smaller than those observed in experiments conducted on small-diameter pipes. The mean two-phase mixture water level increased upon pressurization of the scrubbing pool. However, it was found that the effect of pressurization on the two-phase mixture level fluctuation amplitude was negligibly small. [ABSTRACT FROM AUTHOR]

The Great East Japan Earthquake occurred on March 11, 2011 fatally damaged the Fukushima Daiichi Nuclear Power Plant (NPP), caused prolonged station blackout (SBO). Following the SBO, the reactor water level gradually dropped due to the increase in steam discharge from safety relief valves and eventually led to nuclear fuel melt down. Almost four years have passed since the accident, and official reports by the Japanese regulatory have given the general description of causes and progressions of the fatal accident. Even after the Fukushima accident, more than 430 nuclear power plants are currently operating and over 80 units are under construction worldwide. From this fact alone, it is extremely important to learn from the Fukushima accident and enhance the safety culture of the reactor operation to completely eliminate the possibilities of catastrophic accidents seen in 2011. In this study, the best estimate transient thermal-hydraulics code, RETRAN-03/MOD04 was utilized to focus on the effectiveness of the Isolation Condenser (IC) installed on Advanced Boiling Water Reactors (ABWR). The ICs turned out to be one of the very few operable safety systems during the Fukushima accident, and this simple yet reliable safety system should be utilized to secure ABWR from possible reactor core damages. In the present paper, several case studies conducted utilizing the ICs are presented and methods of countermeasures to improve light water reactor safety level in design and operation features are proposed. [ABSTRACT FROM AUTHOR]

Steam injector (SI) is a simply designed passive jet pump which does not require external power source or internal mechanical parts. The SI utilizes direct contact condensation between steam and water as an operational mechanism and is capable of producing higher pressure water than the inlet fluid pressures. The accident in Fukushima Daiichi Nuclear Power Plant caused setback to the credibility and reliability of nuclear power. One way to regain its trust from the global community, it is suggested to develop and install passive coolant injection systems that are operable even during the station black out. In this review paper, thorough and complete review of the SI system was completed and applicability of the SI system as the passive core cooling system is discussed in details. Due to its high heat removal capability, the system can possibly be applied as a high efficiency heat exchanger as well. Its design and operational mechanisms, and fundamental thermal-hydraulic theory utilized in the analysis and experimental work are reviewed. In addition, its possible application towards existing nuclear power plant systems is reviewed. [ABSTRACT FROM AUTHOR]

Abstract: The bent pipe wall thinning has been often found at the elbow of the drain line and the high-pressure secondary feed-water bent pipe in the nuclear reactors. The liquid droplet impingement (LDI) erosion could be regarded to be one of the major causes and is a significant issue of the thermal hydraulics and structural integrity in aging and life extension for nuclear power plants safety. In this paper two-phase numerical simulations are conducted for standard elbow geometry, typically the pipe diameter is 170mm. The turbulence attenuation in vapor-droplets flow is analysed by a damping function on the energy spectrum basis of single phase flow. Considering the vapor turbulent kinetic energy attenuation due to the involved droplets, a computational fluid dynamic (CFD) tool has been adopted by using two-way vapor-droplet coupled system. This computational fluid model is built up by incompressible Reynolds Averaged Navier–Stoke equations using standard k–ε model and the SIMPLE algorithm, and the numerical droplet model adopts the Lagrangian approach, a general LDI erosion prediction procedure for bent pipe geometry has been performed to supplement the CFD code. The liquid droplets diameter, velocity, volume concentration are evaluated for the effects of carrier turbulence attenuation. The result shows that carrier turbulence kinetic energy attenuation is proved to be an important effect for LDI erosion rate when investigating the bent pipe wall thinning phenomena. [Copyright &y& Elsevier]

Abstract: Axial developments of the local void fraction, interfacial area concentration and bubble Sauter mean diameter were measured in subcooled boiling flow of water in a vertical internally heated annulus using the double-sensor conductivity probe technique. Measurements were performed under varying conditions of heat flux, inlet liquid velocity and inlet liquid temperature. A total of 10 data sets were acquired. Based on these measurements with the previous data obtained in the present test loop, the influence of flow condition on the profiles of local two-phase flow parameters was discussed. The measured average void fraction and interfacial area concentration were compared with the predictions by existing correlations for drift-flux parameters and interfacial area concentration. Also, the recently proposed bubble layer thickness model in subcooled boiling was evaluated for the measurement data. [Copyright &y& Elsevier]

Abstract: Disturbance waves play an important role in interfacial transfer of mass, momentum and energy in annular two-phase flow. In spite of their importance, majority of the experimental data available in literature on disturbance wave properties such as velocity, frequency, wavelength and amplitude are limited to near atmospheric conditions (Azzopardi, B.J., 1997. Drops in annular two-phase flow. International Journal of Multiphase Flow, 23, 1–53). In view of this, air–water annular flow experiments have been conducted at three pressure conditions (1.2, 4.0 and 5.8bar) in a tubular test section having an inside diameter 9.4mm. At each pressure condition liquid and gas phase flow rates are varied over a large range so that the effects of density ratio, liquid flow rate and gas flow rate on disturbance wave properties can be studied systematically. A liquid film thickness is measured by two flush mounted ring shaped conductance probes located 38.1mm apart. Disturbance wave velocity, frequency, amplitude and wavelength are estimated from the liquid film thickness measurements by following the statistical analysis methods. Parametric trends in variations of disturbance wave properties are analyzed using the non-dimensional numbers; liquid phase Reynolds number (Re f), gas phase Reynolds number (Re g), Weber number (We) and Strouhal number (Sr). Finally, the existing correlations available for the prediction of disturbance wave velocity and frequency are analyzed and a new, improved correlation is proposed for the prediction of disturbance wave frequency. The new correlation satisfactorily predicted the current data and the data available in literature. [Copyright &y& Elsevier]

Understanding the spreading behavior of a molten core is important for predicting the progression of severe accidents and ensuring the smooth decommissioning of severe accident-experienced nuclear power plants. In the current study, molten-metal drop experiments were performed using tin and copper. These experiments were conducted to expand the fundamental knowledge on spreading and deposition behaviors to verify the analysis model. In the case of a nozzle diameter of 4 mm, the deposited metal was uniformly relocated; however, the center was extremely thin and Sn demonstrated a circular shape for a nozzle diameter of 10 mm, while Cu spread non-uniformly. The unevenness ratios for Cu were much higher than that of the Sn independent of the dropped height and the nozzle diameters. The effects of hydraulic and thermal dimensionless numbers on the deposition and spreading behaviors were analyzed. A correlation between the spreading and deposition behaviors and the dimensionless numbers of Re , We , Pe , Oh , and Bo was observed. In particular, the Oh and Bo numbers had significantly high correlation coefficients with the dimensionless spreading area, where effects of inertial force on the spreading process were relatively low. The experimental correlation of the dimensionless spreading area was proposed using the Re and Bo numbers. The prediction capability was improved, in which the predicted values were within 28% for the current experiments. [ABSTRACT FROM AUTHOR]

Abstract: For the purpose of accurate flow measurement, an automatic three-dimensional (3D) sound field measurement system has been developed, and an experimental study has been conducted on ultrasonic properties. By using this system, ultrasonic sound pressure distributions and radiation angles in water have been measured. According to Snell’s law, the ultrasonic transmission properties can be obtained on the basis of incidence angle, acoustic impedance, basic frequency of ultrasound, and material and thickness of the metallic plate. However, this law cannot be applied to certain cases where an ultrasonic incident wave passes through a metallic plate and turns into a longitudinal wave, a shear wave and a Lamb wave. Consequently, the ultrasonic propagation paths have been investigated experimentally at various angles of incidence. From the experiments, it was confirmed that the ultrasonic beam paths change with incidence angles. Hence, the most suitable incidence angle has been determined from the result of measurements. Velocity measurements using an ultrasonic velocity profiler were made at various incidence angles. The accuracy of measuring flow rates changed with the incidence angles. The optimal incidence angle determined from 3D field measurements was found to yield the most accurate flow rates. [Copyright &y& Elsevier]

Abstract: Forced convective subcooled boiling flow experiments were conducted in a BWR-scaled vertical upward annular channel. Water was used as the testing fluid, and the tests were performed at atmospheric pressure. A high-speed digital video camera was applied to capture the dynamics of the bubble nucleation process. Bubble lift-off diameters were obtained from the images for a total of 91 test conditions. A force balance analysis of a growing bubble was performed to predict the bubble lift-off size. The dimensionless form of the bubble lift-off diameter was formulated to be a function of Jacob number and Prandtl number. The proposed model agreed well with the experimental data within the averaged relative deviation of ±35.2%. [Copyright &y& Elsevier]

Abstract: A photographic study was carried out for the subcooled flow boiling of water to elucidate the rise characteristics of single vapor bubbles after the departure from nucleation sites. The test section was a transparent glass tube of 20mm in inside diameter and the flow direction was vertical upward; liquid subcooling was parametrically changed within 0–16K keeping system pressure and liquid velocity at 120kPa and 1m/s, respectively. The bubble rise paths were analyzed from the video images that were obtained at the heat flux slightly higher than the minimum heat flux for the onset of nucleate boiling. In the present experiments, all the bubbles departed from their nucleation sites immediately after the inception. In low subcooling experiments, bubbles slid upward and consequently were not detached from the vertical heated wall; the bubble size was increased monotonously with time in this case. In moderate and high subcooling experiments, bubbles were detached from the wall after sliding for several millimeters and migrated towards the subcooled bulk liquid. The bubbles then reversed the direction of lateral migration and were reattached to the wall at moderate subcooling while they collapsed due to the condensation at high subcooling. It was hence considered that the mechanisms of the heat transfer from heated wall and the axial growth of vapor volume were influenced by the difference in bubble rise path. It was observed after the inception that bubbles were varied from flattened to more rounded shape. This observation suggested that the bubble detachment is mainly caused by the change in bubble shape due to the surface tension force. [Copyright &y& Elsevier]

Abstract: Rise characteristics of vapor bubbles after the departure from a nucleation site in forced convective subcooled flow boiling were studied visually using two synchronized high speed video cameras. The test section was a transparent glass tube of 20mm in inside diameter, filtrated and deionized tap water was used as a working fluid, and the flow direction adopted was vertical upward. The outer surface of test section tube was electrically heated to generate vapor bubbles inside of the tube. In the present experiments, the mass flux and liquid subcooling were varied within 94–1435kg/m2 s and 2.2–10K, respectively. Since the observations were performed at low heat fluxes to avoid the significant increase in the number of active nucleation sites, the obtained bubble images were clear enough to carry out the detailed image analysis for the rise characteristics of individual bubbles. The following three different bubble rise paths were observed after the departure from nucleation sites: some bubbles slid upward the vertical wall for long distance, while other bubbles were detached from the wall after sliding for several millimeters and then migrated toward the bulk liquid; after the migration, some of the detached bubbles were collapsed in subcooled liquid but others remained close to the wall and were reattached to the wall. The results of detailed image analyses suggested that the variation in bubble shape from flattened to more rounded was of primary importance for the occurrence of bubble detachment from the wall. [Copyright &y& Elsevier]

Abstract: Visual study was conducted to elucidate the rise characteristics of vapor bubbles after the departure from a nucleation site in forced convective boiling. The flow direction was vertical upward and nearly saturated water was used as a working fluid. The outer surface of a transparent glass tube of 20 mm in inner diameter was electrically heated to generate vapor bubbles inside of the tube. The number density of bubbles in the test section was kept low to observe the behavior of individual bubbles using high speed cameras. The cross-sectional area-averaged velocity of bulk flow was set at 0.5 and 1.0 m/s. The approximate maximum diameter of each bubble was 1.6–2.9 mm. The following observations were made on the behavior of typical bubbles generated in the present experimental conditions: (i) bubbles grew at a nucleation site for a short period of time less than 1 ms with the shape flattened along the vertical heating surface; (ii) bubbles then slid upward the vertical wall for a few millimeters with the gradual increase in their size; (iii) after that, bubbles were detached from the heating surface and migrate towards the bulk flow but they remained close to the wall; (iv) at the instant of detachment, bubble rise velocity was already comparable with the local velocity of surrounding liquid and the rapid growth seen immediately after the nucleation was already completed; (v) after the detachment, bubble size was decreased in slightly subcooled bulk fluid and some of them were collapsed; (vi) bubbles that did not experience the collapse in subcooled liquid reversed the direction of radial migration and was eventually reattached to the wall; (vii) after the reattachment, bubbles generally slid upward the vertical surface for a long distance with the gradual growth due to the heat from the heating wall, though the bubbles that experienced larger gap with the wall after the first detachment showed rebounding motion with the reduced amplitude. The detachment from the wall and the reattachment to the wall of bubbles were clearly observed as the typical bubble behavior after the departure from a nucleation site in the present experiments. It was expected that the inertia force and shear induced lift force have the important roles in these phenomena. [Copyright &y& Elsevier]

Forced convection subcooled water boiling experiments were conducted in a vertical annular channel. A high-speed digital video camera was applied to record the dynamics of the subcooled boiling process. The flow visualization results show that the bubble departure frequency generally increases as the heat flux increases. For some cases, the departure frequency may reach a limit around 1000 bubbles/s. In addition, bubble lift-off diameter, bubble growth rate and bubble velocity after bubble lift-off were determined by analyzing the images. The experimental data obtained from this study can be used in modeling the bubble departure frequency, bubble lift-off diameter, and bubble dynamics in forced convection subcooled boiling. [Copyright &y& Elsevier]

Experimental study was made on the flow structure around a bubble in air–water bubbly flow. In order to measure velocity profiles around a bubble, an Ultrasonic Velocity Profile monitor was employed, which can obtain an instantaneous velocity profile along its measuring line across a channel. The experiments were carried out in a 100×10 mm2 rectangular channel for the air–water counter-current bubbly flow whose void fraction smaller than 7%. The bubble Reynolds number was ranged between 700 and 1000. Most bubbles had ellipsoidal shapes and rose up with wobbling motions. Our experimental results plotted in the form of non-dimensional velocity profiles show that the velocity field around a bubble has a structure similar to the turbulent boundary layer on a solid wall. On the other hand, an earlier analytical study by Moore [J. Fluid Mech. 16 (1963) 161] used an assumption of a spherical bubble rising in liquid irrotationally, and the solution was derived that the flow around a bubble being composed of a thin boundary layer and its outer main stream in potential flow. In this paper, the relation between these two types of boundary layer structures is discussed. [Copyright &y& Elsevier]

As the investigation on the decommissioning of Fukushima Daichi Power Plant (F1) advances further, deepening the knowledge on the reactor building behavior under the different debris removal method is a vital assessment to be made. As the submersion-method has been well investigated in the past, the main scope of the current paper will be on the dry-method where it is assumed that the debris cooling takes place under contact with a small quantity of water or with the air. In this paper, following the discussion of the effectiveness of the dry retrieval method for F1, the feasibility of the method is assessed under the different debris location assumption and time-evolving decay heat from the debris. From the analysis result, potential improvement, as well as the applicability of dry-method, will be discussed. [ABSTRACT FROM AUTHOR]