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Sample records for convective flow boiling

  1. Boiling Suppression in Convective Flow

    International Nuclear Information System (INIS)

    Aounallah, Y.

    2004-01-01

    The development of convective boiling heat transfer correlations and analytical models has almost exclusively been based on measurements of the total heat flux, and therefore on the overall two-phase heat transfer coefficient, when the well-known heat transfer correlations have often assumed additive mechanisms, one for each mode of heat transfer, convection and boiling. While the global performance of such correlations can readily be assessed, the predictive capability of the individual components of the correlation has usually remained elusive. This becomes important when, for example, developing mechanistic models for subcooled void formation based on the partitioning of the wall heat flux into a boiling and a convective component, or when extending a correlation beyond its original range of applications where the preponderance of the heat transfer mechanisms involved can be significantly different. A new examination of existing experimental heat transfer data obtained under fixed hydrodynamic conditions, whereby the local flow conditions are decoupled from the local heat flux, has allowed the unequivocal isolation of the boiling contribution over a broad range of thermodynamic qualities (0 to 0.8) for water at 7 MPa. Boiling suppression, as the quality increases, has consequently been quantified, thus providing valuable new insights on the functionality and contribution of boiling in convective flows. (author)

  2. An experimental study of forced convective flow boiling CHF in nanofluid

    International Nuclear Information System (INIS)

    Ahn, Hoseon; Kim, Seontae; Jo, Hangjin; Kim, Dongeok; Kang, Soonho; Kim, Moohwan

    2008-01-01

    Recently the enhancement of CHF (critical heat flux) in nanofluids under the pool boiling condition is known as a result of nanoparticle deposition on the heating surface. The deposition phenomenon of nanoparticles on the heating surface is induced dominantly by the vigorous boiling on the heating surface. Considering the importance of flow boiling conditions in various practical heat transfer applications, an experimental study was performed to verify whether or not the enhancement of CHF in nanofluids exists in a forced convective flow boiling condition. The nanofluid used in this research was Al 2 O 3 -water dispersed by the ultra-sonic vibration method in very low concentration (0.01% Vol). A heater specimen was made of a copper block easily detachable to look into the surface condition after the experiment. The heating method was a thermal-heating made with a conductive material. The flow channel took a rectangular type (10mm x 10mm) and had a length of 1.2 m to assure a hydrodynamically fully-developed region. In result, CHF in the nanofluid under the forced convective flow boiling condition has been enhanced distinctively along with the effect of flow rates. To reason the CHF increase in the nanofluids, the boiling surface was investigated thoroughly with the SEM image. (author)

  3. Forced convection flow boiling and two-phase flow phenomena in a microchannel

    Science.gov (United States)

    Na, Yun Whan

    2008-07-01

    The present study was performed to numerically analyze the evaporation phenomena through the liquid-vapor interface and to investigate bubble dynamics and heat transfer behavior during forced convective flow boiling in a microchannel. Flow instabilities of two-phase flow boiling in a microchannel were studied as well. The main objective of this research is to investigate the fundamental mechanisms of two-phase flow boiling in a microchannel and provide predictive tools to design thermal management systems, for example, microchannel heat sinks. The numerical results obtained from this study were qualitatively and quantitatively compared with experimental results in the open literature. Physical and mathematical models, accounting for evaporating phenomena through the liquid-vapor interface in a microchannel at constant heat flux and constant wall temperature, have been developed, respectively. The heat transfer mechanism is affected by the dominant heat conduction through the thin liquid film and vaporization at the liquid-vapor interface. The thickness of the liquid film and the pressure of the liquid and vapor phases were simultaneously solved by the governing differential equations. The developed semi-analytical evaporation model that takes into account of the interfacial phenomena and surface tension effects was used to obtain solutions numerically using the fourth-order Runge-Kutta method. The effects of heat flux 19 and wall temperature on the liquid film were evaluated. The obtained pressure drops in a microchannel were qualitatively consistent with the experimental results of Qu and Mudawar (2004). Forced convective flow boiling in a single microchannel with different channel heights was studied through a numerical simulation to investigate bubble dynamics, flow patterns, and heat transfer. The momentum and energy equations were solved using the finite volume method while the liquid-vapor interface of a bubble is captured using the VOF (Volume of Fluid

  4. Analysis of forced convective transient boiling by homogeneous model of two-phase flow

    International Nuclear Information System (INIS)

    Kataoka, Isao

    1985-01-01

    Transient forced convective boiling is of practical importance in relation to the accident analysis of nuclear reactor etc. For large length-to-diameter ratio, the transient boiling characteristics are predicted by transient two-phase flow calculations. Based on homogeneous model of two-phase flow, the transient forced convective boiling for power and flow transients are analysed. Analytical expressions of various parameters of transient two-phase flow have been obtained for several simple cases of power and flow transients. Based on these results, heat flux, velocity and time at transient CHF condition are predicted analytically for step and exponential power increases, and step, exponential and linear velocity decreases. The effects of various parameters on heat flux, velocity and time at transient CHF condition have been clarified. Numerical approach combined with analytical method is proposed for more complicated cases. Solution method for pressure transient are also described. (author)

  5. Suppression of saturated nucleate boiling by forced convective flow

    International Nuclear Information System (INIS)

    Bennett, D.L.; Davis, M.W.; Hertzler, B.L.

    1980-01-01

    Tube-side forced convective boiling nitrogen and oxygen and thin film shell-side forced convective boiling R-11 data demonstrate a reduction in the heat transfer coefficient associated with nucleate boiling as the two-phase friction pressure drop increases. Techniques proposed in the literature to account for nucleate boiling during forced convective boiling are discussed. The observed suppression of nucleate boiling for the tube-side data is compared against the Chen correlation. Although general agreement is exhibited, supporting the interactive heat transfer mechanism theory, better agreement is obtained by defining a bubble growth region within the thermal boundary layer. The data suggests that the size of the bubble growth region is independent of the friction drop, but is only a function of the physical properties of the boiling liquid. 15 refs

  6. Visualization of bubble behaviors in forced convective subcooled flow boiling

    International Nuclear Information System (INIS)

    Inaba, Noriaki; Matsuzaki, Mitsuo; Kikura, Hiroshige; Aritomi, Masanori; Komeno, Toshihiro

    2007-01-01

    Condensation characteristics of vapor bubble after the departure from a heated section in forced convective subcooled flow boiling were studied visually by using a high speed camera. The purpose of the present study was to measure two-phase flow parameters in subcooled flow boiling. These two-phase flow parameters are void fraction, interfacial area concentration and Sauter mean diameter, which express bubble interface behaviors. The experimental set-up was designed to measure the two-phase flow parameters necessary for developing composite equations for the two fluid models in subcooled flow boiling. In the present experiments, the mass flux, liquid subcooling and the heater were varied within 100-1000kg/m 2 s, 2-10K and 100-300kW/m 2 respectively. Under these experimental conditions, the bubble images were obtained by a high-speed camera, and analyzed paying attention to the condensation of vapor bubbles. These two-phase parameters were obtained by the experimental data, such as the bubble parameter, the bubble volume and the bubble surface. In the calculation process of the two phase flow parameters, it was confirmed that these parameters are related to the void fraction. (author)

  7. Forced convection and subcooled flow boiling heat transfer in asymmetrically heated ducts of T-section

    International Nuclear Information System (INIS)

    Abou-Ziyan, Hosny Z.

    2004-01-01

    This paper presents the results of an experimental investigation of heat transfer from the heated bottom side of tee cross-section ducts to an internally flowing fluid. The idea of this work is derived from the cooling of critical areas in the cylinder heads of internal combustion engines. Fully developed single phase forced convection and subcooled flow boiling heat transfer data are reported. Six T-ducts of different width and height aspect ratios are tested with distilled water at velocities of 1, 2 and 3 m/s for bulk temperatures of 60 and 80 deg. C, while the heat flux was varied from about 80 to 700 kW/m 2 . The achieved data cover Reynolds numbers in the range of 5.22 x 10 4 to 2.36 x 10 5 , Prandtl numbers in the range from 2.2 to 3.0, duct width aspect ratio between 2.19 and 3.13 and duct height aspect ratio from 0.69 to 2.0. The results revealed that the increase in either the width or height aspect ratio of the T-ducts enhances the convection heat transfer coefficients and the boiling heat fluxes considerably. The following comparisons are provided for coolant velocity of 2 m/s, bulk temperature of 60 deg. C, wall superheat of 20 K and wall to bulk temperature difference of 20 K. As the width aspect ratio increases by 43%, the convection heat transfer coefficient and the boiling heat flux increase by 27% and 39%, respectively. An increase in the height aspect ratio by 290% enhances the convection heat transfer coefficient and the boiling heat fluxes by 82% and 103%, respectively. When the coolant velocity changes from 1 to 2 m/s, the heat transfer coefficient increases by 60% and the boiling heat flux rises by 62-98% for the various tested ducts. The convection heat transfer coefficient increases by 12% and the boiling heat flux decreases by 31% as the bulk fluid temperature rises from 60 to 80 deg. C. A correlation was developed for Nusselt number as a function of Reynolds number, Prandtl number, viscosity ratio and some aspect ratios of the T-duct

  8. A study of forced convective subcooled flow boiling

    International Nuclear Information System (INIS)

    Serizawa, Akimi; Kenning, D.B.R.

    1979-01-01

    Based on a simple nucleation model, parameter survey technique is used to derive a predictive correlation for boiling initiation under forced convection. Results are expressed by a semi-empirical equation which considers effects of the flow turbulence on interfacial heat transfer coefficient for evaporation and condensation of vapour bubbles during their growth. This correlation agrees within +-25% with a variety of experimental water data presently available. The bubble departure diameter and the subcooling-dependence of active nucleation sites were examined, using experimental data available. Results are expressed by empirical equations. Finally, an analytical model is presented to predict conditions for the point of net vapour generation. The model is based on the formation and growth of a bubble boundary layer adjacent to the heated wall. It is shown that the point of net vapour generation is determined by the liquid subcooling at the boiling initiation and the subcooling-dependences of bubble departure diameter and bubble flux. The result implies that the bubble ejection from bubble layer is a possible mechanism for the significant void increase even at high velocities. (author)

  9. Modeling a forced to natural convection boiling test with the program LOOP-W

    International Nuclear Information System (INIS)

    Carbajo, J.J.

    1984-01-01

    Extensive testing has been conducted in the Simulant Boiling Flow Visualization (SBFV) loop in which water is boiled in a vertical transparent tube by circulating hot glycerine in an annulus surrounding the tube. Tests ranged from nonboiling forced convection to oscillatory boiling natural convection. The program LOOP-W has been developed to analyze these tests. This program is a multi-leg, one-dimensional, two-phase equilibrium model with slip between the phases. In this study, a specific test, performed at low power where non-boiling forced convection was changed to boiling natural convection and then to non-boiling again, has been modeled with the program LOOP-W

  10. A highly stable microchannel heat sink for convective boiling

    International Nuclear Information System (INIS)

    Lu, Chun Ting; Pan Chin

    2009-01-01

    To develop a highly stable two-phase microchannel heat sink, we experimented with convective boiling in diverging, parallel microchannels with different distributions of laser-etched artificial nucleation sites. Each microchannel had a mean hydraulic diameter of 120 µm. The two-phase flow visualization and the magnitudes of pressure drop and inlet temperature oscillations under boiling conditions demonstrated clearly the merits of using artificial nucleation sites to further stabilize the flow boiling in diverging, parallel microchannels. The stability map showed the plane of subcooling number versus phase change number. It illustrated that diverging, parallel microchannels with artificial nucleation cavities have a much wider stable region than parallel microchannels with uniform cross-sections or diverging, parallel microchannels without artificial nucleation cavities. In addition, the results revealed that the design with cavities distributed uniformly along the downstream half of the channel presented the best stability performance among the three distributions of nucleation sites. This particular design can be regarded as a highly stable microchannel heat sink for convective boiling

  11. Critical heat flux for free convection boiling in thin rectangular channels

    International Nuclear Information System (INIS)

    Cheng, Lap Y.; Tichler, P.R.

    1991-01-01

    A review of the experimental data on free convection boiling critical heat flux (CHF) in vertical rectangular channels reveals three mechanisms of burnout. They are the pool boiling limit, the circulation limit, and the flooding limit associated with a transition in flow regime from churn to annular flow. The dominance of a particular mechanism depends on the dimensions of the channel. Analytical models were developed for each free convection boiling limit. Limited agreement with data is observed. A CHF correlation, which is valid for a wide range of gap sizes, was constructed from the CHFs calculated according to the three mechanisms of burnout. 17 refs., 7 figs

  12. Basic Study for Active Nucleation Site Density Evaluation in Subcooled Flow Boiling

    International Nuclear Information System (INIS)

    Chu, In Cheol; Song, Chul Hwa

    2008-01-01

    Numerous studies have been performed on a active nucleation site density (ANSD) due to its governing influence on a heat transfer. However, most of the studies were focused on pool boiling conditions. Kocamustafaogullari and Ishii developed an ANSD correlation from a parametric study of the existing pool boiling data. Also, they extended the correlation to a convective flow boiling condition by adopting the nucleation suppression factor of Chen's heat transfer correlation. However, the appropriateness of applying the Chen's suppression factor to an ANSD correlation was not fully validated because there was not enough experimental data on ANSD in the forced convective flow boiling. Basu et al. performed forced convective boiling experiments and proposed a correlation of ANSD which is the only correlation based on experimental data for a forced convective boiling. They concluded that the ANSD is only dependent on the static contact angle and the wall superheat, and is independent of the flow rate and the subcooling, which contradict the general acceptance of the nucleation suppression in the forced convective boiling. It seems that no reliable ANSD correlation or model is available for a forced convective boiling. In the present study, the effect of the flow velocity on the suppression of the nucleation site was examined, and the effectiveness of a Brewster reflection technique for the identification of the nucleation site was also examined

  13. Experimental and theoretical study on forced convection film boiling heat transfer

    International Nuclear Information System (INIS)

    Liu, Qiusheng

    2001-01-01

    Theoretical solutions of forced convection film boiling heat transfer from horizontal cylinders in saturated liquids were obtained based on a two-phase laminar boundary layer film boiling model. It was clarified that author's experimental data for the cylinders with the nondimensional diameters, D, of around 1.3 in water and in Freon-113 agreed with the values of theoretical numerical solutions based on the two-phase laminar boundary layer model with the smooth vapor-liquid interface except those for low flow velocities. A forced convection film boiling heat transfer correlation including the radiation contribution from the cylinders with various diameters in saturated and subcooled liquids was developed based on the two-phase laminar boundary layer film boiling model and the experimental data for water and Freon-113 at wide ranges of flow velocities, surface superheats, system pressures and cylinder diameters. (author)

  14. Experimental study on forced convection boiling heat transfer on molten alloy

    International Nuclear Information System (INIS)

    Nishimura, Satoshi; Ueda, Nobuyuki; Nishi, Yoshihisa; Furuya, Masahiro; Kinoshita, Izumi

    1999-01-01

    In order to clarify the characteristics of forced convection boiling heat transfer on molten metal, basic experiments have been carried out with subcooled water flowing on molten Wood's alloy pool surface. In these experiments, water flows horizontally in a rectangular duct. A cavity filled with Wood's alloy is present in a portion of the bottom of the duct. Wood's alloy is heated by a copper conductor at the bottom of the cavity. The experiments have been carried out with various velocities and subcoolings of water, and temperature of Wood's alloy. Boiling curves on the molten alloy surface were obtained and compared with that on a solid heat transfer surface. It is observed that the boiling curve on molten alloy is in a lower superheat region than the boiling curve on a solid surface. This indicates that the heat transfer performance of forced convection boiling on molten alloy is enhanced by increase of the heat transfer area, due to oscillation of the surface and fragmentation of molten alloy

  15. Flow boiling heat transfer at low liquid Reynolds number

    International Nuclear Information System (INIS)

    Weizhong Zhang; Takashi Hibiki; Kaichiro Mishima

    2005-01-01

    Full text of publication follows: In view of the significance of a heat transfer correlation of flow boiling at conditions of low liquid Reynolds number or liquid laminar flow, and very few existing correlations in principle suitable for such flow conditions, this study is aiming at developing a heat transfer correlation of flow boiling at low liquid Reynolds number conditions. The obtained results are as follows: 1. A new heat transfer correlation has been developed for saturated flow boiling at low liquid Reynolds number conditions based on superimposition of two boiling mechanisms, namely convective boiling and nucleate boiling. In the new correlation, two terms corresponding to the mechanisms of nucleate boiling and convective boiling are obtained from the pool boiling correlation by Forster and Zuber and the analytical annular flow model by Hewitt and Hall-Taylor, respectively. 2. An extensive database was collected for saturated flow boiling heat transfer at low liquid Reynolds number conditions, including data for different channels geometries (circular and rectangular), flow orientations (vertical and horizontal), and working fluids (water, R11, R12, R113). 3. An extensive comparison of the new correlation with the collected database shows that the new correlation works satisfactorily with the mean deviation of 16.6% for saturated flow boiling at low liquid Reynolds number conditions. 4. The detailed discussion reveals the similarity of the newly developed correlation for flow boiling at low liquid Reynolds number to the Chen correlation for flow boiling at high liquid Reynolds number. The Reynolds number factor F can be analytically deduced in this study. (authors)

  16. A separate-effect-based new appraisal of convective boiling and its suppression

    International Nuclear Information System (INIS)

    Aounallah, Yacine

    2008-01-01

    The development of convective boiling heat transfer correlations and analytical models has been based almost exclusively on the knowledge of global heat transfer coefficients, while the predictive capabilities of the correlation constituting components (typically additive convection and boiling) have remained usually elusive. This becomes important when, for example, developing a mechanistic subcooled void model based on wall heat flux partitioning, or when applying a correlation beyond its developmental range. In the latter case, the preponderance of the individual heat transfer mechanisms, through the phenomenon of boiling suppression, can become significantly different, thus leading to uncharted uncertainty extrapolations. An examination of existing experimental data, obtained under fixed hydrodynamic conditions, has allowed the isolation of the boiling heat transfer contribution over a broad range of thermodynamic qualities (0 to 0.8) and mass fluxes (1,100 to 3,900 kg/(m 2 ·s)) for water at 7.2 MPa. Boiling suppression has been quantified, thus providing valuable new insights on the basic functional relationships of boiling in convective flows. This work has allowed a new interpretation and representation of the standard flow 'boiling map' (Collier's) to be developed. The convection enhancement and boiling suppression components (F and S) of the well-known Chen's correlation - an important constitutive relationship implemented in several best-estimate (realistic) thermal-hydraulics codes - have been individually determined, showing the pitfall of splitting the correlation for mechanistic boiling heat transfer modelling, and the important role of compensating errors in uncertainty extrapolation. An initial attempt to formulate a new correlation, based for the first time on segregated heat transfer components, is also included. (author)

  17. Mechanistic modeling of CHF in forced-convection subcooled boiling

    International Nuclear Information System (INIS)

    Podowski, M.Z.; Alajbegovic, A.; Kurul, N.; Drew, D.A.; Lahey, R.T. Jr.

    1997-05-01

    Because of the complexity of phenomena governing boiling heat transfer, the approach to solve practical problems has traditionally been based on experimental correlations rather than mechanistic models. The recent progress in computational fluid dynamics (CFD), combined with improved experimental techniques in two-phase flow and heat transfer, makes the use of rigorous physically-based models a realistic alternative to the current simplistic phenomenological approach. The objective of this paper is to present a new CFD model for critical heat flux (CHF) in low quality (in particular, in subcooled boiling) forced-convection flows in heated channels

  18. Evaluation of forced-convection nucleate boiling detection by acoustic emission

    International Nuclear Information System (INIS)

    Wells, R.P.; Paterson, J.A.

    1981-10-01

    Acoustic Emission techniques are being investigated for use as protection systems in neutral beam accelerators and water cooled beam dumps. For this purpose, the characteristics of the boiling curve for forced-convection surface boiling have been compared to the Acoustic Emission (AE) produced. Results indicate that AE, in the form of count-rate, is a sensitive indicator of nucleate boiling incipience and is relatively insensitive to flow velocity in the 0 to 12 m/s range

  19. Thermal-hydraulic performance of convective boiling jet array impingement

    International Nuclear Information System (INIS)

    Jenkins, R; De Brún, C; Kempers, R; Lupoi, R; Robinson, A J

    2016-01-01

    Jet impingement boiling is investigated with regard to heat transfer and pressure drop performance using a novel laser sintered 3D printed jet impingement manifold design. Water was the working fluid at atmospheric pressure with inlet subcooling of 7 o C. The convective boiling performance of the impinging jet system was investigated for a flat copper target surface for 2700≤Re≤5400. The results indicate that the heat transfer performance of the impinging jet is independent of Reynolds number for fully developed boiling. Also, the investigation of nozzle to plate spacing shows that low spacing delays the onset of nucleate boiling causing a superheat overshoot that is not observed with larger gaps. However, no sensitivity to the gap spacing was measured once boiling was fully developed. The assessment of the pressure drop performance showed that the design effectively transfers heat with low pumping power requirements. In particular, owing to the insensitivity of the heat transfer to flow rate during fully developed boiling, the coefficient of performance of jet impingement boiling in the fully developed boiling regime deteriorates with increased flow rate due to the increase in pumping power flux. (paper)

  20. Forced-convection boiling tests performed in parallel simulated LMR fuel assemblies

    International Nuclear Information System (INIS)

    Rose, S.D.; Carbajo, J.J.; Levin, A.E.; Lloyd, D.B.; Montgomery, B.H.; Wantland, J.L.

    1985-01-01

    Forced-convection tests have been carried out using parallel simulated Liquid Metal Reactor fuel assemblies in an engineering-scale sodium loop, the Thermal-Hydraulic Out-of-Reactor Safety facility. The tests, performed under single- and two-phase conditions, have shown that for low forced-convection flow there is significant flow augmentation by thermal convection, an important phenomenon under degraded shutdown heat removal conditions in an LMR. The power and flows required for boiling and dryout to occur are much higher than decay heat levels. The experimental evidence supports analytical results that heat removal from an LMR is possible with a degraded shutdown heat removal system

  1. A dry-spot model of critical heat flux and transition boiling in pool and subcooled forced convection boiling

    International Nuclear Information System (INIS)

    Ha, Sang Jun

    1998-02-01

    A new dry-spot model for critical heat flux (CHF) is proposed. The new concept for dry area formation based on Poisson distribution of active nucleation sites and the critical active site number is introduced. The model is based on the boiling phenomena observed in nucleate boiling such as Poisson distribution of active nucleation sites and formation of dry spots on the heating surface. It is hypothesized that when the number of bubbles surrounding one bubble exceeds a critical number, the surrounding bubbles restrict the feed of liquid to the microlayer under the bubble. Then a dry spot of vapor will form on the heated surface. As the surface temperature is raised, more and more bubbles will have a population of surrounding active sites over the critical number. Consequently, the number of the spots will increase and the size of dry areas will increase due to merger of several dry spots. If this trend continues, the number of effective sites for heat transport through the wall will diminish, and CHF and transition boiling occur. The model is applicable to pool and subcooled forced convection boiling conditions, based on the common mechanism that CHF and transition boiling are caused by the accumulation and coalescences of dry spots. It is shown that CHF and heat flux in transition boiling can be determined without any empirical parameter based on information on the boiling parameters such as active site density and bubble diameter, etc., in nucleate boiling. It is also shown that the present model well represents actual phenomena on CHF and transition boiling and explains the mechanism on how parameters such as flow modes (pool or flow) and surface wettability influence CHF and transition boiling. Validation of the present model for CHF and transition boiling is achieved without any tuning parameter always present in earlier models. It is achieved by comparing the predictions of CHF and heat flux in transition boiling using measured boiling parameters in nucleate

  2. Experimental and theoretical studies on subcooled flow boiling of pure liquids and multicomponent mixtures

    Energy Technology Data Exchange (ETDEWEB)

    Jamialahmadi, M.; Abdollahi, H.; Shariati, A. [The University of Petroleum Industry, Ahwaz (Iran); Mueller-Steinhagen, H. [Institute of Technical Thermodynamics, German Aerospace Center (Germany); Institute of Thermodynamics and Thermal Engineering, University of Stuttgart (Germany)

    2008-05-15

    To improve the design of modern industrial reboilers, accurate knowledge of boiling heat transfer coefficients is essential. In this study flow boiling heat transfer coefficients for binary and ternary mixtures of acetone, isopropanol and water were measured over a wide range of heat flux, subcooling, flow velocity and composition. The measurements cover the regimes of convective heat transfer, transitional boiling and fully developed subcooled flow boiling. Two models are presented for the prediction of flow boiling heat transfer coefficients. The first model is the combination of the Chen model with the Gorenflo correlation and the Schluender model for single and multicomponent boiling, respectively. This model predicts flow boiling heat transfer coefficients with acceptable accuracy, but fails to predict the nucleate boiling fraction NBF reasonably well. The second model is based on the asymptotic addition of forced convective and nucleate boiling heat transfer coefficients. The benefit of this model is a further improvement in the accuracy of flow boiling heat transfer coefficient over the Chen type model, simplicity and the more realistic prediction of the nucleate boiling fraction NBF. (author)

  3. A Correlation for Forced Convective Boiling Heat Transfer of Refrigerants in a Microfin Tube

    Science.gov (United States)

    Momoki, Satoru; Yu, Jian; Koyama, Shigeru; Fujii, Tetsu; Honda, Hiroshi

    The experimental study is reported on the forced convective boiling of pure refrigerants HCFC22, HFC134a and HCFC123 flowing in a horizontal microfin tube. The local heat transfer coefficient defined based on the actual inside surface area is measured in the ranges of mass velocity of 200 to 400 kg/m2s, heat flux of 5 to 64 kW/m2 and reduced pressure of 0.07 to 0.24. Using the Chen-type model, a new correlation for microfin tubes is proposed considering the enhancement effect of microfins on both the convective heat transfer and the nucleate boiling components. In the convective heat transfer component, the correlation to predict the heat transfer coefficient of liquid-only flow is determined from preliminary experiments on single-phase flow in microfin tubes, and the two-phase flow enhancement factor is determined from the present experimental data. For the nucleate boiling component, the correlation of Takamatsu et al. for smooth tube is modified. The prediction of the present correlation agrees well with present experimental data, and is available for several microfin tubes which were tested by other researchers.

  4. Transition from boiling to two-phase forced convection

    International Nuclear Information System (INIS)

    Maroti, L.

    1985-01-01

    The paper presents a method for the prediction of the boundary points of the transition region between fully developed boiling and two-phase forced convection. It is shown that the concept for the determination of the onset of fully developed boiling can also be applied for the calculation of the point where the heat transfer is effected again by the forced convection. Similarly, the criterion for the onset of nucleate boiling can be used for the definition of the point where boiling is completely suppressed and pure two-phase forced convection starts. To calculate the heat transfer coefficient for the transition region, an equation is proposed that applies the boundary points and a relaxation function ensuring the smooth transition of the heat transfer coefficient at the boundaries

  5. Direct numerical simulation of bubble dynamics in subcooled and near-saturated convective nucleate boiling

    International Nuclear Information System (INIS)

    Lal, Sreeyuth; Sato, Yohei; Niceno, Bojan

    2015-01-01

    Highlights: • We simulate convective nucleate pool boiling with a novel phase-change model. • We simulate four cases at different sub-cooling and wall superheat levels. • We investigate the flow structures around the growing bubble and analyze the accompanying physics. • We accurately simulate bubble shape elongation and enhanced wall cooling due to the sliding and slanting motions of bubbles. • Bubble cycle durations show good agreement with experimental observations. - Abstract: With the long-term objective of Critical Heat Flux (CHF) prediction, bubble dynamics in convective nucleate boiling flows has been studied using a Direct Numerical Simulation (DNS). A sharp-interface phase change model which was originally developed for pool boiling flows is extended to convective boiling flows. For physical scales smaller than the smallest flow scales (smaller than the grid size), a micro-scale model was used. After a grid dependency study and a parametric study for the contact angle, four cases of simulation were carried out with different wall superheat and degree of subcooling. The flow structures around the growing bubble were investigated together with the accompanying physics. The relation between the heat flux evolution and the bubble growth was studied, along with investigations of bubble diameter and bubble base diameter evolutions across the four cases. As a validation, the evolutions of bubble diameter and bubble base diameter were compared to experimental observations. The bubble departure period and the bubble shapes show good agreement between the experiment and the simulation, although the Reynolds number of the simulation cases is relatively low

  6. The mechanisms of transitions from natural convection and nucleate boiling to nucleate boiling or film boiling caused by rapid depressurization in highly subcooled water

    International Nuclear Information System (INIS)

    Sakurai, Akira; Shiotsu, Masahiro; Hata, Koichi; Fukuda, Katsuya

    1999-01-01

    The mechanisms of transient boiling process including the transitions to nucleate boiling or film boiling from initial heat fluxes, q in , in natural convection and nucleate boiling regimes caused by exponentially decreasing system pressure with various decreasing periods, τ p on a horizontal cylinder in a pool of highly subcooled water were clarified. The transient boiling processes with different characteristics were divided into three groups for low and intermediate q in in natural convection regime, and for high q in in nucleate boiling regime. The transitions at maximum heat fluxes from low q in in natural convection regime to stable nucleate boiling regime occurred independently of the τ p values. The transitions from intermediate and high q in values in natural convection and nucleate boiling to stable film boiling occurred for short τ p values, although those to stable nucleate boiling occurred for tong τ p values. The CHF and corresponding surface superheat values at which the transition to film boiling occurred were considerably lower and higher than the steady-state values at the corresponding pressure during the depressurization respectively. It was suggested that the transitions to stable film boiling at transient critical heat fluxes from intermediate q in in natural convection and from high q in in nucleate boiling for short τ p occur due to explosive-like heterogeneous spontaneous nucleation (HSN). The photographs of typical vapor behavior due to the HSN during depressurization from natural convection regime for short τ p were shown. (author)

  7. Forced convective and subcooled flow boiling heat transfer to pure water and n-heptane in an annular heat exchanger

    International Nuclear Information System (INIS)

    Peyghambarzadeh, S.M.; Sarafraz, M.M.; Vaeli, N.; Ameri, E.; Vatani, A.; Jamialahmadi, M.

    2013-01-01

    Highlights: ► The cooling performance of water and n-heptane is compared during subcooled flow boiling. ► Although n-heptane leaves the heat exchanger warmer it has a lower heat transfer coefficient. ► Flow rate, heat flux and degree of subcooling have direct effect on heat transfer coefficient. ► The predictions of some correlations are evaluated against experimental data. - Abstract: In this research, subcooled flow boiling heat transfer coefficients of pure n-heptane and distilled water at different operating conditions have been experimentally measured and compared. The heat exchanger consisted of vertical annulus which is heated from the inner cylindrical heater with variable heat flux (less than 140 kW/m 2 ). Heat flux is varied so that two different flow regimes from single phase forced convection to nucleate boiling condition are created. Meanwhile, liquid flow rate is changed in the range of 2.5 × 10 −5 –5.8 × 10 −5 m 3 /s to create laminar up to transition flow regimes. Three subcooling levels including 10, 20 and 30 °C are also considered. Experimental results demonstrated that subcooled flow boiling heat transfer coefficient increases when higher heat flux, higher liquid flow rate and greater subcooling level are applied. Furthermore, influence of the operating conditions on the bubbles generation on the heat transfer surface is also discussed. It is also shown that water is better cooling fluid in comparison with n-heptane

  8. Heat transfer coefficient correlation for convective boiling inside plain and micro fin tubes using genetic algorithms

    International Nuclear Information System (INIS)

    Picanco, Marco Antonio Silva; Bandarra Filho, Enio Pedone; Passos, Julio Cesar

    2006-01-01

    Two-phase flow heat transfer has been exhaustively studied over recent years. However, in this field several questions remain unanswered. Heat transfer coefficient prediction related to nucleate and convective boiling have been studied using different approaches, numerical, analytical and experimental. In this work, an experimental analysis, data representation and heat transfer coefficient prediction on two-phase heat transfer on nucleate and convective boiling are presented. An empirical correlation is obtained based on genetic algorithms search engine over a dimensional analysis of the two-phase flow heat transfer problem. (author)

  9. Modelling of subcooled boiling and DNB-type boiling crisis in forced convection

    International Nuclear Information System (INIS)

    Bricard, Patrick

    1995-01-01

    This research thesis aims at being a contribution to the modelling of two phenomena occurring during a forced convection: the axial evolution of the vacuum rate, and the boiling crisis. Thus, the first part of this thesis addresses the prediction of the vacuum rate, and reports the development of a modelling of under-saturated convection in forced convection. The author reports the development and assessment of two-fluid one-dimensional model, the development of a finer analysis based on an averaging of local equations of right cross-sections in different areas. The second part of this thesis addresses the prediction of initiation of a boiling crisis. The author presents generalities and motivations for this study, reports a bibliographical study and a detailed analysis of mechanistic models present in this literature. A mechanism of boiling crisis is retained, and then further developed in a numerical modelling which is used to assess some underlying hypotheses [fr

  10. Micro-channel convective boiling heat transfer with flow instabilities

    International Nuclear Information System (INIS)

    Consolini, L.; Thome, J.R.

    2009-01-01

    Flow boiling heat transfer in micro-channels has attracted much interest in the past decade, and is currently a strong candidate for high performance compact heat sinks, such as those required in electronics systems, automobile air conditioning units, micro-reactors, fuel cells, etc. Currently the literature presents numerous experimental studies on two-phase heat transfer in micro-channels, providing an extensive database that covers many different fluids and operating conditions. Among the noteworthy elements that have been reported in previous studies, is the sensitivity of micro-channel evaporators to oscillatory two-phase instabilities. These periodic fluctuations in flow and pressure drop either result from the presence of upstream compressibility, or are simply due to the interaction among parallel channels in multi-port systems. An oscillating flow presents singular characteristics that are expected to produce an effect on the local heat transfer mechanisms, and thus on the estimation of the two-phase heat transfer coefficients. The present investigation illustrates results for flow boiling of refrigerants R-134a, R-236fa, and R-245fa in a 510 μm circular micro-channel, exposed to various degrees of oscillatory compressible volume instabilities. The data describe the main features of the fluctuations in the temperatures of the heated wall and fluid, and draw attention to the differences in the measured unstable time-averaged heat transfer coefficients with respect to those for stable flow boiling. (author)

  11. Micro-channel convective boiling heat transfer with flow instabilities

    Energy Technology Data Exchange (ETDEWEB)

    Consolini, L.; Thome, J.R. [Ecole Polytechnique Federale de Lausanne (Switzerland). Lab. de Transfert de Chaleur et de Masse], e-mail: lorenzo.consolini@epfl.ch, e-mail: john.thome@epfl.ch

    2009-07-01

    Flow boiling heat transfer in micro-channels has attracted much interest in the past decade, and is currently a strong candidate for high performance compact heat sinks, such as those required in electronics systems, automobile air conditioning units, micro-reactors, fuel cells, etc. Currently the literature presents numerous experimental studies on two-phase heat transfer in micro-channels, providing an extensive database that covers many different fluids and operating conditions. Among the noteworthy elements that have been reported in previous studies, is the sensitivity of micro-channel evaporators to oscillatory two-phase instabilities. These periodic fluctuations in flow and pressure drop either result from the presence of upstream compressibility, or are simply due to the interaction among parallel channels in multi-port systems. An oscillating flow presents singular characteristics that are expected to produce an effect on the local heat transfer mechanisms, and thus on the estimation of the two-phase heat transfer coefficients. The present investigation illustrates results for flow boiling of refrigerants R-134a, R-236fa, and R-245fa in a 510 {mu}m circular micro-channel, exposed to various degrees of oscillatory compressible volume instabilities. The data describe the main features of the fluctuations in the temperatures of the heated wall and fluid, and draw attention to the differences in the measured unstable time-averaged heat transfer coefficients with respect to those for stable flow boiling. (author)

  12. Comparative analysis of heat transfer correlations for forced convection boiling

    International Nuclear Information System (INIS)

    Guglielmini, G.; Nannei, E.; Pisoni, C.

    1978-01-01

    A critical survey was conducted of the most relevant correlations of boiling heat transfer in forced convection flow. Most of the investigations carried out on partial nucleate boiling and fully developed nucleate boiling have led to the formulation of correlations that are not able to cover a wide range of operating conditions, due to the empirical approach of the problem. A comparative analysis is therefore required in order to delineate the relative accuracy of the proposed correlations, on the basis of the experimental data presently available. The survey performed allows the evaluation of the accuracy of the different calculating procedure; the results obtained, moreover, indicate the most reliable heat transfer correlations for the different operating conditions investigated. This survey was developed for five pressure range (up to 180bar) and for both saturation and subcooled boiling condition

  13. Mechanisms of convective and boiling heat transfer enhancement via ultrasonic vibration

    International Nuclear Information System (INIS)

    Kim, Yi Gu; Kim, Ho Young; Kang, Seoung Min; Kang, Byung Ha; Lee, Jin Ho

    2003-01-01

    This work experimentally studies the fundamental mechanisms by which the ultrasonic vibration enhances convection and pool boiling heat transfer. A thin platinum wire is used as both a heat source and a temperature sensor. A high speed video imaging system is employed to observe the behavior of cavitation and thermal bubbles. It is found that when the liquid temperature is below its boiling point, cavitation takes place due to ultrasonic vibration while cavitation disappears when the liquid reaches the boiling point. Moreover, when the gas dissolved in liquid is removed by pre-degassing, the cavitation arises only locally. Depending on the liquid temperature, heat transfer rates in convection, subcooled boiling and saturated boiling regimes are examined. In convection heat transfer regime, fully agitated cavitation is the most efficient heat transfer enhancement mechanism. Subcooled boiling is most enhanced when the local cavitation is induced after degassing. In saturated boiling regime, acoustic pressure is shown to be a dominant heat transfer enhancement mechanism

  14. Forced convective transition boiling: review of literature and comparison of prediction methods

    International Nuclear Information System (INIS)

    Groeneveld, D.C.; Fung, K.K.

    1976-06-01

    This report reviews the published information on transition boiling heat transfer under forced convective conditions. It was found that transition boiling data have been obtained only within a limited range of conditions and many data are considered unreliable. The data do not permit the derivation of a correlation; however the parametric trends can be isolated from the data. Several authors have proposed correlations valid in the transition boiling region. Most of the correlations are valid only within a narrow range of conditions. A comparison with the data shows that in general agreement is poor. Hsu's correlation is tentatively recommended for low flows and pressures. (author)

  15. Prediction model for initial point of net vapor generation for low-flow boiling

    International Nuclear Information System (INIS)

    Sun Qi; Zhao Hua; Yang Ruichang

    2003-01-01

    The prediction of the initial point of net vapor generation is significant for the calculation of phase distribution in sub-cooled boiling. However, most of the investigations were developed in high-flow boiling, and there is no common model that could be successfully applied for the low-flow boiling. A predictive model for the initial point of net vapor generation for low-flow forced convection and natural circulation is established here, by the analysis of evaporation and condensation heat transfer. The comparison between experimental data and calculated results shows that this model can predict the net vapor generation point successfully in low-flow sub-cooled boiling

  16. Flow with boiling in four-cusp channels simulating damaged core in PWR type reactors

    International Nuclear Information System (INIS)

    Esteves, M.M.

    1985-01-01

    The study of subcooled nucleate flow boiling in non-circular channels is of great importance to engineering applications in particular to Nuclear Engineering. In the present work, an experimental apparatus, consisting basically of a refrigeration system, running on refrigerant-12, has been developed. Preliminary tests were made with a circular tube. The main objective has been to analyse subcooled flow boiling in four-cusp channels simulating the flow conditions in a PWR core degraded by accident. Correlations were developed for the forced convection film coefficient for both single-phase and subcooled flow boiling. The incipience of boiling in such geometry has also been studied. (author) [pt

  17. Reynolds analogy for subcooled surface boiling under forced convection

    International Nuclear Information System (INIS)

    Avdeev, A.A.

    1982-01-01

    For the case of subcooled surface boiling under forced convection the analytic expression of analogy between the heat transfer and carry pulse (Reynolds analogy) is derived. It is concluded that the obtained dependence creates the basis for solution of a series of problems of surface boiling physics. On the basis of the performed analysis the method of coordinate calculation of the origin of intensive vapour generation is developed and the formula for calculation of the broken-off-bubble radius under forced convection is derived [ru

  18. Forced convective boiling heat transfer of water in vertical rectangular narrow channel

    International Nuclear Information System (INIS)

    Chen, Chong; Gao, Pu-zhen; Tan, Si-chao; Chen, Han-ying; Chen, Xian-bing

    2015-01-01

    Highlights: • Chen correlation cannot well predict the coefficient of rectangular channel. • Kim and Mudawar correlation is the best one among the Chen type correlations. • Lazarek and Black correlation predicted 7.0% of data within the ±30% error band. • The new correlation can well predict the coefficient with a small MAE of 14.4%. - Abstract: In order to research the characteristics of boiling flows in a vertical rectangular narrow channel, a series of convective boiling heat transfer experiments are performed. The test section is made of stainless steel with an inner diameter of 2 × 40 mm and heated length of 1100 mm. The 3194 experimental data points are obtained for a heat flux range of 10–700 kW/m 2 , a mass flux range of 200–2400 kg/m 2 s, a system pressure range of 0.1–2.5 MPa, and a quality range of 0–0.8. Eighteen prediction models are used to predict the flow boiling heat transfer coefficient of the rectangular narrow channel and the predicted value is compared against the database including 3194 data points, the results show that Chen type correlations and Lazarek and Black type correlations are not suitable for the rectangular channel very much. The Kim and Mudawar correlation is the best one among the 18 models. A new correlation is developed based on the superposition concept of nucleate boiling and convective boiling. the new correlation is shown to provide a good prediction against the database, evidenced by an overall MAE of 14.4%, with 95.2% and 98.6% of the data falling within ±30% and ±35% error bands, respectively

  19. Modeling of subcooled boiling in the vertical flow

    International Nuclear Information System (INIS)

    Koncar, B.; Mavko, B.

    1999-01-01

    A two-dimensional model of subcooled boiling in a vertical channel was developed. Its basic idea is that the vapor phase generation has a similar effect on the flow field as a hypothetical liquid phase generation. The bubble volume, generated due to evaporation process, was filled with liquid and included as a source term in the continuity equation for the liquid phase. Thus, the single-phase from of transport equations was preserved and bubbles were retained in the boundary layer near the heated surface. Time development of subcooled boiling was simulated and effects of governing physical mechanisms (evaporation, condensation, vapor-phase convection, vapor-phase diffusion) on the flow field and pressure drop were analyzed. The Results of the proposed two-dimensional model were compared with experimental data and RELAP5/MOD3.2 calculations. The presented model represents a contribution to the two-dimensional simulation of the subcooled boiling phenomenon.(author)

  20. Heat transfer in pool boiling liquid neon, deuterium and hydrogen, and critical heat flux in forced convection of liquid neon

    International Nuclear Information System (INIS)

    Astruc, J.M.

    1967-12-01

    In the first part, free-convection and nucleate pool boiling heat transfer (up to burn-out heat flux) between a platinum wire of 0.15 mm in diameter in neon, deuterium and hydrogen has been studied at atmospheric pressure. These measurements were continued in liquid neon up to 23 bars (Pc ≅ 26.8 b). Film boiling heat transfer coefficients have been measured in pool boiling liquid neon at atmospheric pressure with three heating wires (diameters 0.2, 0.5, 2 mm). All the results have been compared with existing correlations. The second part is devoted to measurements of the critical heat flux limiting heat transfer with small temperature differences between the wall and the liquid neon flowing inside a tube (diameters 3 x 3.5 mm) heated by joule effect on 30 cm of length. Influences of flow stability, nature of electrical current, pressure, mass flow rate and subcooling are shown. In conclusion, the similarity of the heat transfer characteristics in pool boiling as well as in forced convection of liquid neon and hydrogen is emphasized. (author) [fr

  1. Flow patterns and heat transfer coefficients in flow-boiling and convective condensation of R22 inside a micro fin of new design

    International Nuclear Information System (INIS)

    Muzzio, A.; Niro, A.; Garaviglia, M.

    1998-01-01

    Saturated flow boiling and convective condensation experiments for oil-free refrigerant R22 been carried out with a micro fin tube of new design and with a smooth tube. Both tube have the same outer diameter of 9.52 mm and are horizontally operated. Two-phase flow pattern data have been obtained in addition of heat transfer coefficient and pressure drops; more-over, adiabatic tests have been also performed in order for flow pattern map to cover even adiabatic flows. Data are for mass fluxes ranging from about 90 to 400 Kg/s m 2 . In boiling tests, the nominal saturation temperature is 5 degree C, with inlet quality varying from 0.2 to 0.6 and the quality change ranging from 0.1 to 0.5. In condensation, results are for saturation temperature equal to 35 degree C, with inlet quality between 0.8 and 0.4, and quality change within 0.6 and 0.2. The comparison shows a large heat transfer augmentation with a moderate increment of pressure drops, especially in evaporation were the enhancement factor comes up to 4 while the penalty factor is about 1.4 at the most. Heat transfer coefficients both in evaporation and condensation are compared to the predictions of some recent correlations specifically proposed or modified for micro fin tube

  2. A phenomenological model of the thermal hydraulics of convective boiling during the quenching of hot rod bundles

    International Nuclear Information System (INIS)

    Nelson, R.A.; Unal, C.

    1991-01-01

    In this paper, a phenomenological model of the thermal hydraulics of convective boiling in the post-critical-heat-flux (post-CHF) regime is developed and discussed. The model was implemented in the TRAC-PF1/MOD2 computer code (an advanced best-estimate computer program written for the analysis of pressurized water reactor systems). The model was built around the determination of flow regimes downstream of the quench front. The regimes were determined from the flow-regime map suggested by Ishii and his coworkers. Heat transfer in the transition boiling region was formulated as a position-dependent model. The propagation of the CHF point was strongly dependent on the length of the transition boiling region. Wall-to-fluid film boiling heat transfer was considered to consist of two components: first, a wall-to-vapor convective heat-transfer portion and, second, a wall-to-liquid heat transfer representing near-wall effects. Each contribution was considered separately in each of the inverted annular flow (IAF) regimes. The interfacial heat transfer was also formulated as flow-regime dependent. The interfacial drag coefficient model upstream of the CHF point was considered to be similar to flow through a roughened pipe. A free-stream contribution was calculated using Ishii's bubbly flow model for either fully developed subcooled or saturated nucleate boiling. For the drag in the smooth IAF region, a simple smooth-tube correlation for the interfacial friction factor was used. The drag coefficient for the rough-wavy IAF was formulated in the same way as for the smooth IAF model except that the roughness parameter was assumed to be proportional to liquid droplet diameter entrained from the wavy interface. The drag coefficient in the highly dispersed flow regime considered the combined effects of the liquid droplets within the channel and a liquid film on wet unheated walls. 431 refs., 6 figs., 4 tabs

  3. Free convection film flows and heat transfer

    CERN Document Server

    Shang, Deyi

    2010-01-01

    Presents development of systematic studies for hydrodynamics and heat and mass transfer in laminar free convection, accelerating film boiling and condensation of Newtonian fluids, and accelerating film flow of non-Newtonian power-law fluids. This book provides a system of analysis models with a developed velocity component method.

  4. Heat-transfer correlations for natural convection boiling

    International Nuclear Information System (INIS)

    Stephan, K.; Abdelsalam, M.

    1980-01-01

    To-date there exists no comprehensive theory allowing the prediction of heat-transfer coefficients in natural convection boiling, in spite of the many efforts made in this field. In order to establish correlations with wide application, the methods of regression analysis were applied to the nearly 500 existing experimental data points for natural convection boiling heat transfer. As demonstrated by the analysis, these data can best be represented by subdividing the substances into four groups (water, hydrocarbons, cryogenic fluids and refrigerants) and employing a different set of dimensionless numbers for each group of substances, because certain dimensionless numbers important for one group of substances are unimportant to another. One equation valid for all substances could be built up, but its accuracy would be less than that obtained for the individual correlations without adding undesirable complexity. (author)

  5. Visualization study for forced convection heat transfer of supercritical carbon dioxide near pseudo-boiling point

    International Nuclear Information System (INIS)

    Sakurai, K.; Ko, H.S.; Okamoto, K.; Madarame, H.

    2001-01-01

    For development of new reactor, supercritical water is expected to be used as coolant to improve thermal efficiency. However, the thermal characteristics of supercritical fluid is not revealed completely because its difficulty for experiment. Specific phenomena tend to occur near the pseudo-boiling point which is characterised by temperature corresponding to the saturation point in ordinary fluid. Around this point, the physic properties such as density, specific heat and thermal conductivity are drastically varying. Although there is no difference between gas and liquid phases in supercritical fluids, phenomena similar to boiling (with heat transfer deterioration) can be observed round the pseudo-boiling point. Experiments of heat transfer have been done for supercritical fluid in forced convective condition. However, these experiments were mainly realised inside stainless steel cylinder pipes, for which flow visualisation is difficult. Consequently, this work has been devoted to the development of method allowing the visualisation of supercritical flows. The experiment setup is composed of main loop and test section for the visualisation. Carbon dioxide is used as test fluid. Supercritical carbon dioxide flows upward in rectangular channel and heated by one-side wall to generate forced convection heat transfer. Through window at mid-height of the test section, shadowgraphy was applied to visualize density gradient distribution. The behavior of the density wave in the channel is visualized and examined through the variation of the heat transfer coefficient. (author)

  6. SAS3A analysis of natural convection boiling behavior in the Sodium Boiling Test Facility

    International Nuclear Information System (INIS)

    Klein, G.A.

    1979-01-01

    An analysis of natural convection boiling behavior in the Sodium Boiling Test (SBT) Facility has been performed using the SAS3A computer code. The predictions from this analysis indicate that stable boiling can be achieved for extensive periods of time for channel powers less than 1.4 kW and indicate intermittent dryout at higher powers up to at least 1.7 kW. The results of this anaysis are in reasonable agreement with the SBT Facility test results

  7. New flow boiling heat transfer model for hydrocarbons evaporating inside horizontal tubes

    International Nuclear Information System (INIS)

    Chen, G. F.; Gong, M. Q.; Wu, J. F.; Zou, X.; Wang, S.

    2014-01-01

    Hydrocarbons have high thermodynamic performances, belong to the group of natural refrigerants, and they are the main components in mixture Joule-Thomson low temperature refrigerators (MJTR). New evaluations of nucleate boiling contribution and nucleate boiling suppression factor in flow boiling heat transfer have been proposed for hydrocarbons. A forced convection heat transfer enhancement factor correlation incorporating liquid velocity has also been proposed. In addition, the comparisons of the new model and other classic models were made to evaluate its accuracy in heat transfer prediction

  8. Convection in multiphase fluid flows using lattice Boltzmann methods

    NARCIS (Netherlands)

    Biferale, L.; Perlekar, P.; Sbragaglia, M.; Toschi, F.

    2012-01-01

    We present high-resolution numerical simulations of convection in multiphase flows (boiling) using a novel algorithm based on a lattice Boltzmann method. We first study the thermodynamical and kinematic properties of the algorithm. Then, we perform a series of 3D numerical simulations changing the

  9. Forced convective boiling of water inside helically coiled tube. Characteristics of oscillation of dryout point

    International Nuclear Information System (INIS)

    Nagai, Niro; Sugiyama, Kenta; Takeuchi, Masanori; Yoshikawa, Shinji; Yamamoto, Fujio

    2006-01-01

    The helically coiled tube of heat exchanger is used for the evaporator of prototype fast breeder reactor 'Monju'. This paper aims at the grasp of two-phase flow phenomena of forced convective boiling of water inside helical coiled tube, especially focusing on oscillation phenomena of dryout point. A glass-made helically coiled tube was used to observe the inside water boiling behavior flowing upward, which was heated by high temperature oil outside the tube. This oil was also circulated through a glass made tank to provide the heat source for water evaporation. The criterion for oscillation of dryout point was found to be a function of inlet liquid velocity and hot oil temperature. The observation results suggest the mechanism of dryout point oscillation mainly consists of intensive nucleate boiling near the dryout point and evaporation of thin liquid film flowing along the helical tube. In addition, the oscillation characteristics were experimentally confirmed. As inlet liquid velocity increases, oscillation amplitude also increases but oscillation cycle does not change so much. As hot oil temperature increases, oscillation amplitude and cycle gradually decreases. (author)

  10. Flow boiling heat transfer on nanowire-coated surfaces with highly wetting liquid

    International Nuclear Information System (INIS)

    Shin, Sangwoo; Choi, Geehong; Kim, Beom Seok; Cho, Hyung Hee

    2014-01-01

    Owing to the recent advances in nanotechnology, one significant progress in energy technology is increased cooling ability. It has recently been shown that nanowires can improve pool boiling heat transfer due to the unique features such as enhanced wetting and enlarged nucleation sites. Applying such nanowires on a flow boiling, which is another major class of boiling phenomenon that is associated with forced convection, is yet immature and scarce despite its importance in various applications such as liquid cooling of energy, electronics and refrigeration systems. Here, we investigate flow boiling heat transfer on surfaces that are coated with SiNWs (silicon nanowires). Also, we use highly-wetting dielectric liquid, FC-72, as a working fluid. An interesting wetting behavior is observed where the presence of SiNWs reduces wetting and wicking that in turn leads to significant decrease of CHF (critical heat flux) compared to the plain surface, which opposes the current consensus. Also, the effects of nanowire length and Reynolds number on the boiling heat transfer are shown to be highly nonmonotonic. We attempt to explain such an unusual behavior on the basis of wetting, nucleation and forced convection, and we show that such factors are highly coupled in a way that lead to unusual behavior. - Highlights: • Observation of suppressed wettability in the presence of surface roughness (nanowires). • Significant reduction of critical heat flux in the presence of nanowires. • Nonmonotonic behavior of heat transfer coefficient vs. nanowire length and Reynolds number

  11. Dual-zone boiling process

    International Nuclear Information System (INIS)

    Bennett, D.L.; Schwarz, A.; Thorogood, R.M.

    1987-01-01

    This patent describes a process for boiling flowing liquids in a heat exchanger wherein the flowing liquids is heated in a single heat exchanger to vaporize the liquid. The improvement described here comprises: (a) passing the boiling flowing liquid through a first heat transfer zone of the heat exchanger comprising a surface with a high-convective-heat-transfer characteristic and a higher pressure drop characteristic; and then (b) passing the boiling flowing liquid through a second heat transfer zone of the heat exchanger comprising an essentially open channel with only minor obstructions by secondary surfaces, with an enhanced nucleate boiling heat transfer surface and a lower pressure drop characteristic

  12. PSI-BOIL, a building block towards the multi-scale modeling of flow boiling phenomena

    International Nuclear Information System (INIS)

    Niceno, Bojan; Andreani, Michele; Prasser, Horst-Michael

    2008-01-01

    Full text of publication follows: In these work we report the current status of the Swiss project Multi-scale Modeling Analysis (MSMA), jointly financed by PSI and Swissnuclear. The project aims at addressing the multi-scale (down to nano-scale) modelling of convective boiling phenomena, and the development of physically-based closure laws for the physical scales appropriate to the problem considered, to be used within Computational Fluid Dynamics (CFD) codes. The final goal is to construct a new computational tool, called Parallel Simulator of Boiling phenomena (PSI-BOIL) for the direct simulation of processes all the way down to the small-scales of interest and an improved CFD code for the mechanistic prediction of two-phase flow and heat transfer in the fuel rod bundle of a nuclear reactor. An improved understanding of the physics of boiling will be gained from the theoretical work as well as from novel small- and medium scale experiments targeted to assist the development of closure laws. PSI-BOIL is a computer program designed for efficient simulation of turbulent fluid flow and heat transfer phenomena in simple geometries. Turbulence is simulated directly (DNS) and its efficiency plays a vital role in a successful simulation. Having high performance as one of the main prerequisites, PSIBOIL is tailored in such a way to be as efficient a tool as possible, relying on well-established numerical techniques and sacrificing all the features which are not essential for the success of this project and which might slow down the solution procedure. The governing equations are discretized in space with orthogonal staggered finite volume method. Time discretization is performed with projection method, the most obvious a the most widely used choice for DNS. Systems of linearized equation, stemming from the discretization of governing equations, are solved with the Additive Correction Multigrid (ACM). methods. Two distinguished features of PSI-BOIL are the possibility to

  13. An investigation of transition boiling mechanisms of subcooled water under forced convective conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kwang-Won, Lee; Sang-Yong, Lee

    1995-09-01

    A mechanistic model for forced convective transition boiling has been developed to investigate transition boiling mechanisms and to predict transition boiling heat flux realistically. This model is based on a postulated multi-stage boiling process occurring during the passage time of the elongated vapor blanket specified at a critical heat flux (CHF) condition. Between the departure from nucleate boiling (DNB) and the departure from film boiling (DFB) points, the boiling heat transfer is established through three boiling stages, namely, the macrolayer evaporation and dryout governed by nucleate boiling in a thin liquid film and the unstable film boiling characterized by the frequent touches of the interface and the heated wall. The total heat transfer rates after the DNB is weighted by the time fractions of each stage, which are defined as the ratio of each stage duration to the vapor blanket passage time. The model predictions are compared with some available experimental transition boiling data. The parametric effects of pressure, mass flux, inlet subcooling on the transition boiling heat transfer are also investigated. From these comparisons, it can be seen that this model can identify the crucial mechanisms of forced convective transition boiling, and that the transition boiling heat fluxes including the maximum heat flux and the minimum film boiling heat flux are well predicted at low qualities/high pressures near 10 bar. In future, this model will be improved in the unstable film boiling stage and generalized for high quality and low pressure situations.

  14. A critical review of predictive models for the onset of significant void in forced-convection subcooled boiling

    International Nuclear Information System (INIS)

    Dorra, H.; Lee, S.C.; Bankoff, S.G.

    1993-06-01

    This predictive models for the onset of significant void (OSV) in forced-convection subcooled boiling are reviewed and compared with extensive data. Three analytical models and seven empirical correlations are considered in this review. These models and correlations are put onto a common basis and are compared, again on a common basis, with a variety of data. The evaluation of their range of validity and applicability under various operating conditions are discussed. The results show that the correlations of Saha-Zuber seems to be the best model to predict OSV in vertical subcooled boiling flow

  15. Enhanced Natural Convection in a Metal Layer Cooled by Boiling Water

    International Nuclear Information System (INIS)

    Cho, Jae-Seon; Suh, Kune Y.; Chung, Chang-Hyun; Park, Rae-Joon; Kim, Sang-Baik

    2004-01-01

    An experimental study is performed to investigate the natural convection heat transfer characteristics and the solidification of the molten metal pool concurrently with forced convective boiling of the overlying coolant to simulate a severe accident in a nuclear power plant. The relationship between the Nusselt number (Nu) and the Rayleigh number (Ra) in the molten metal pool region is determined and compared with the correlations in the literature and experimental data with subcooled water. Given the same Ra condition, the present experimental results for Nu of the liquid metal pool with coolant boiling are found to be higher than those predicted by the existing correlations or measured from the experiment with subcooled boiling. To quantify the observed effect of the external cooling on the natural convection heat transfer rate from the molten pool, it is proposed to include an additional dimensionless group characterizing the temperature gradients in the molten pool and in the external coolant region. Starting from the Globe and Dropkin correlation, engineering correlations are developed for the enhancement of heat transfer in the molten metal pool when cooled by an overlying coolant. The new correlations for predicting natural convection heat transfer are applicable to low-Prandtl-number (Pr) materials that are heated from below and solidified by the external coolant above. Results from this study may be used to modify the current model in severe accident analysis codes

  16. Local heat transfer estimation in microchannels during convective boiling under microgravity conditions: 3D inverse heat conduction problem using BEM techniques

    Science.gov (United States)

    Luciani, S.; LeNiliot, C.

    2008-11-01

    Two-phase and boiling flow instabilities are complex, due to phase change and the existence of several interfaces. To fully understand the high heat transfer potential of boiling flows in microscale's geometry, it is vital to quantify these transfers. To perform this task, an experimental device has been designed to observe flow patterns. Analysis is made up by using an inverse method which allows us to estimate the local heat transfers while boiling occurs inside a microchannel. In our configuration, the direct measurement would impair the accuracy of the searched heat transfer coefficient because thermocouples implanted on the surface minichannels would disturb the established flow. In this communication, we are solving a 3D IHCP which consists in estimating using experimental data measurements the surface temperature and the surface heat flux in a minichannel during convective boiling under several gravity levels (g, 1g, 1.8g). The considered IHCP is formulated as a mathematical optimization problem and solved using the boundary element method (BEM).

  17. Multi-scale modeling and analysis of convective boiling: towards the prediction of CHF in rod bundles

    International Nuclear Information System (INIS)

    Niceno, B.; Sato, Y.; Badillo, A.; Andreani, M.

    2010-01-01

    In this paper we describe current activities on the project Multi-Scale Modeling and Analysis of convective boiling (MSMA), conducted jointly by the Paul Scherrer Institute (PSI) and the Swiss Nuclear Utilities (Swissnuclear). The long-term aim of the MSMA project is to formulate improved closure laws for Computational Fluid Dynamics (CFD) simulations for prediction of convective boiling and eventually of the Critical Heat Flux (CHF). As boiling is controlled by the competition of numerous phenomena at various length and time scales, a multi-scale approach is employed to tackle the problem at different scales. In the MSMA project, the scales on which we focus range from the CFD scale (macro-scale), bubble size scale (meso-scale), liquid micro-layer and triple interline scale (micro-scale), and molecular scale (nano-scale). The current focus of the project is on micro- and meso- scales modeling. The numerical framework comprises a highly efficient, parallel DNS solver, the PSI-BOIL code. The code has incorporated an Immersed Boundary Method (IBM) to tackle complex geometries. For simulation of meso-scales (bubbles), we use the Constrained Interpolation Profile method: Conservative Semi-Lagrangian 2nd order (CIP-CSL2). The phase change is described either by applying conventional jump conditions at the interface, or by using the Phase Field (PF) approach. In this work, we present selected results for flows in complex geometry using the IBM, selected bubbly flow simulations using the CIP-CSL2 method and results for phase change using the PF approach. In the subsequent stage of the project, the importance of effects of nano-scale processes on the global boiling heat transfer will be evaluated. To validate the models, more experimental information will be needed in the future, so it is expected that the MSMA project will become the seed for a long-term, combined theoretical and experimental program

  18. Heat transfer with water in forced convection without boiling in small diameter tubes

    International Nuclear Information System (INIS)

    Ricque, Roger; Siboul, Roger

    1969-01-01

    This note presents the measurements performed for the establishment of an empirical heat transfer law for water in forced convection without boiling in small diameter tubes (2 and 4 mm), with high flow velocity and strong heat flux, and for relatively low fluid temperatures. A correlation of experimental points is obtained with a very small maximum dispersion: Nu fl = 0,0092 Re fl 0,88 Pr 0,5 (μ fl /μ p ) 0,14 . A correlation for the fiction coefficient is also presented [fr

  19. Numerical simulation of bubble growth and departure during flow boiling period by lattice Boltzmann method

    International Nuclear Information System (INIS)

    Sun, Tao; Li, Weizhong; Yang, Shuai

    2013-01-01

    Highlights: • The bubble departure diameter is proportional to g −0.425 in quiescent fluid. • The bubble release frequency is proportional to g 0.678 in quiescent fluid. • The simulation result supports the transient micro-convection model. • The bubble departure diameter has exponential relation with inlet velocity. • The bubble release frequency has linear relation with inlet velocity. -- Abstract: Nucleate boiling flows on a horizontal plate are studied in this paper by a hybrid lattice Boltzmann method, where both quiescent and slowly flowing ambient are concerned. The process of a single bubble growth on and departure from the superheated wall is simulated. The simulation result supports the transient micro-convection model. The bubble departure diameter and the release frequency are investigated from the simulation result. It is found that the bubble departure diameter and the release frequency are proportional to g −0.425 and g 0.678 in quiescent fluid, respectively, where g is the gravitational acceleration. Nucleate boiling in slowly flowing ambient is also calculated in consideration of forced convection. It is presented that the bubble departure diameter and the release frequency have exponential relationship and linear relationship with inlet velocity in slowly flowing fluid, respectively

  20. Void Fraction Measurement in Subcooled-Boiling Flow Using High-Frame-Rate Neutron Radiography

    International Nuclear Information System (INIS)

    Kureta, Masatoshi; Akimoto, Hajime; Hibiki, Takashi; Mishima, Kaichiro

    2001-01-01

    A high-frame-rate neutron radiography (NR) technique was applied to measure the void fraction distribution in forced-convective subcooled-boiling flow. The focus was experimental technique and error estimation of the high-frame-rate NR. The results of void fraction measurement in the boiling flow were described. Measurement errors on instantaneous and time-averaged void fractions were evaluated experimentally and analytically. Measurement errors were within 18 and 2% for instantaneous void fraction (measurement time is 0.89 ms), and time-averaged void fraction, respectively. The void fraction distribution of subcooled boiling was measured using atmospheric-pressure water in rectangular channels with channel width 30 mm, heated length 100 mm, channel gap 3 and 5 mm, inlet water subcooling from 10 to 30 K, and mass velocity ranging from 240 to 2000 kg/(m 2 .s). One side of the channel was heated homogeneously. Instantaneous void fraction and time-averaged void fraction distribution were measured parametrically. The effects of flow parameters on void fraction were investigated

  1. An Experimental Study on the Convective Heat Transfer in Narrow Rectangular Channels for Downward Flow to Predict Onset of Nucleate Boiling

    International Nuclear Information System (INIS)

    Song, Junghyun; Jeong, Yong Hoon; Lee, Juhyung; Chang, Soon Heung

    2014-01-01

    Research reactor is the nuclear reactor serves neutron source for many research fields such as neutron scattering, non-destructive testing, radioisotope treatment and so on. Due to that characteristic of research reactor, as many people work around the research reactor, research reactor should be designed to have much more conservative margin for normal operation. Boiling heat transfer is the one of the most efficient type in heat transfer modes, however, research reactor needs to avoid onset of nucleate boiling (ONB) in normal operation as IAEA recommend for research reactors to have enough ONB margin to maintain the normal operation state in 'IAEA-TECDOC-233' (1980) for the same reason explained above. Jordan Research and Training Reactor (JRTR) operates under downward flow in narrow rectangular channel in fuel assembly. There isn't sufficient heat transfer data under downward flow condition and only few ONB prediction correlation as well. In the present work, not only a new ONB prediction model would be developed, but also comparison between heat transfer data with several heat transfer correlations could be shown. In addition, as Sudo and Omar S. proposed differently about the Nusselt number behaviors in upward and downward convective heat transfer, the study of convective heat transfer should be conducted continuously to determine it exactly. In this paper, single-phase heat transfer data is analyzed by several heat transfer correlations before developing ONB prediction correlation. In this study, an experiment on the single-phase heat transfer was conducted. As shown in Fig. 5, comparison between experimental data and existing correlations shows quite huge difference as about 40%. Additional experiments on single-phase heat transfer at low heat flux are necessary to clarify the tendency of Nusselt number among heat flux and to develop new correlation for single-phase heat transfer

  2. Nucleate pool boiling, film boiling and single-phase free convection at pressures up to the critical state. Part I: Integral heat transfer for horizontal copper cylinders

    Energy Technology Data Exchange (ETDEWEB)

    Gorenflo, Dieter; Baumhoegger, Elmar; Windmann, Thorsten; Herres, Gerhard [Institut fuer Energie- und Verfahrenstechnik, Universitaet Paderborn, Warburger Str. 100, D-33098 Paderborn (Germany)

    2010-11-15

    Transcritical working cycles for refrigerants have led to increased interest in heat transfer near the Critical State. In general, experimental results for this region differ significantly from those far from it because some fluid properties vary much more there than at a greater distance. In this paper, measurements for two-phase and single-phase free convective heat transfer from an electrically heated copper tube with 25 mm O.D. to refrigerant R125 are discussed for fluid states very close to the Critical Point and far from it. It is shown that heat transfer for film boiling slightly below and for free convection slightly above the critical pressure is very similar. The new - and also previous - experimental data for nucleate boiling, film boiling, and single-phase free convection are compared with calculated results between atmospheric and critical pressure. It can be concluded that the Principle of Corresponding States in its simplest form is very well suited to transfer the results to other refrigerants. In Part II, particular attention will be given to a minimum superheat for nucleate boiling and a maximum superheat for film boiling and single-phase free convection within the circumferential variation of the isobaric wall superheat on the lower parts of the tube. (author)

  3. Instability in flow boiling in microchannels

    CERN Document Server

    Saha, Sujoy Kumar

    2016-01-01

    This Brief addresses the phenomena of instability in flow boiling in microchannels occurring in high heat flux electronic cooling. A companion edition in the SpringerBrief Subseries on Thermal Engineering and Applied Science to “Critical Heat Flux in Flow Boiling in Microchannels,” and "Heat Transfer and Pressure Drop in Flow Boiling in Microchannels,"by the same author team, this volume is idea for professionals, researchers, and graduate students concerned with electronic cooling.

  4. Numerical simulation of flow boiling for organic fluid with high saturation temperature in vertical porous coated tube

    Energy Technology Data Exchange (ETDEWEB)

    Yang Dong, E-mail: dyang@mail.xjtu.edu.cn [State Key Laboratory of Multiphase Flow in Power Engineering, Xi' an Jiaotong University, Xi' an, Shaanxi Province 710049 (China); Pan Jie; Wu Yanhua; Chen Tingkuan [State Key Laboratory of Multiphase Flow in Power Engineering, Xi' an Jiaotong University, Xi' an, Shaanxi Province 710049 (China); Zhou, Chenn Q. [Department of Mechanical Engineering, Purdue University Calumet, Hammond, IN 46323 (United States)

    2011-08-15

    Highlights: > A model is developed for the prediction of flow boiling in vertical porous tubes. > The model assumes that the nucleate boiling plays an important role. > The present model can predict most of the experimental values within {+-}20%. > The results indicate the nucleate boiling contribution decreases from 50% to 15%. - Abstract: A semi-analytical model is developed for the prediction of flow boiling heat transfer inside vertical porous coated tubes. The model assumes that the forced convection and nucleate boiling coexist together in the annular flow regime. Conservations of mass, momentum, and energy are used to solve for the liquid film thickness and temperature. The heat flux due to nucleate boiling consists of those inside and outside micro-tunnels. To close the equations, a detailed analysis of various forces acting on the bubble is presented to predict its mean departure diameter. The active nucleation site density of porous layer is determined from the pool boiling correlation by introducing suppression factor. The flow boiling heat transfer coefficients of organic fluid (cumene) with high saturation temperature in a vertical flame-spraying porous coated tube are studied numerically. It is shown that the present model can predict most of the experimental values within {+-}20%. The numerical results also indicate that the nucleate boiling contribution to the overall heat transfer coefficient decreases from 50% to 15% with vapor quality increasing from 0.1 to 0.5.

  5. Correlation development of natural convection heat transfer in consideration of aspect ratio change and coolant boiling

    International Nuclear Information System (INIS)

    Park, L. J.; Cho, Y. L.; Kang, K. H.; Kim, S. B.; Kim, H. D.; Cho, J. S.; Jung, C. H.

    1999-01-01

    A new correlation on natural convection heat transfer with crust formation in the molten metal pool has been developed in consideration of coolant boiling effect and of aspect ratio change by an increase in crust thickness. Two test results of the convection cooling case, natural and forced convection cooling cases, and of the boiling case were used in the present study. The experimental results have shown that the Nusselt number of the case with boiling condition in the molten metal pool is greater than that of the case with non-boiling condition at the same Rayleigh number. Even though the Rayleigh number rapidly decreases due to an increase of the crust thickness, the Nusselt number does not rapidly decrease because of the aspect ratio effect. From the experimental results, the new correlation between the Nusselt number and Rayleigh number in the molten metal pool with the crust formation has been developed as Nu 0.051(Ra) 1/3 (AR) . 0 .2441 (Φ) 0.025 using Globe and Dropkin correlation

  6. Boiling in porous media

    International Nuclear Information System (INIS)

    1998-01-01

    This conference day of the French society of thermal engineers was devoted to the analysis of heat transfers and fluid flows during boiling phenomena in porous media. This book of proceedings comprises 8 communications entitled: 'boiling in porous medium: effect of natural convection in the liquid zone'; 'numerical modeling of boiling in porous media using a 'dual-fluid' approach: asymmetrical characteristic of the phenomenon'; 'boiling during fluid flow in an induction heated porous column'; 'cooling of corium fragment beds during a severe accident. State of the art and the SILFIDE experimental project'; 'state of knowledge about the cooling of a particulates bed during a reactor accident'; 'mass transfer analysis inside a concrete slab during fire resistance tests'; 'heat transfers and boiling in porous media. Experimental analysis and modeling'; 'concrete in accidental situation - influence of boundary conditions (thermal, hydric) - case studies'. (J.S.)

  7. Some specific features of subcooled boiling heat transfer and crisis at extremely high heat flux densities

    International Nuclear Information System (INIS)

    Gotovsky, M.A.

    2001-01-01

    Forced convection boiling is the process used widely in a lot of industry branches including NPP. Heat transfer intensity under forced convection boiling is considered in different way in dependence on conditions. One of main problems for the process considered is an influence of interaction between forced flow and boiling on heat transfer character. For saturated water case a transition from ''pure'' forced convection to nucleate boiling can be realized in smooth form. (author)

  8. Prediction of flow instability during natural convection

    International Nuclear Information System (INIS)

    Farhadi, Kazem

    2005-01-01

    The occurrence of flow excursion instability during passive heat removal for Tehran Research Reactor (TRR) has been analyzed at low-pressure and low-mass rate of flow conditions without boiling taking place. Pressure drop-flow rate characteristics in the general case are determined upon a developed code for this purpose. The code takes into account variations of different pressure drop components caused by different powers as well as different core inlet temperatures. The analysis revealed the fact that the instability can actually occur in the natural convection mode for a range of powers per fuel plates at a predetermined inlet temperature with fixed geometry of the core. Low mass rate of flow and high sub-cooling are the two important conditions for the occurrence of static instability in the TRR. The calculated results are compared with the existing data in the literature. (author)

  9. Interface tracking computations of bubble dynamics in nucleate flow boiling

    International Nuclear Information System (INIS)

    Giustini, G.

    2015-01-01

    The boiling process is of utter importance for the design and operation of water-cooled nuclear reactors. Despite continuous effort over the past decades, a fully mechanistic model of boiling in the presence of a solid surface has not yet been achieved. Uncertainties exist at fundamental level, since the microscopic phenomena governing nucleate boiling are still not understood, and as regards 'component scale' modelling, which relies heavily on empirical representations of wall boiling. Accurate models of these phenomena at sub-milli-metric scale are capable of elucidating the various processes and to produce quantitative data needed for up-scaling. Within this context, Direct Numerical Simulation (DNS) represents a powerful tool for CFD analysis of boiling flows. In this contribution, DNS coupled with an Interface Tracking method (Y. Sato, B. Niceno, Journal of Computational Physics, Volume 249, 15 September 2013, Pages 127-161) are used to analyse the hydrodynamics and heat transfer associated with heat diffusion controlled bubble growth at a solid substrate during nucleate flow boiling. The growth of successive bubbles from a single nucleation site is simulated with a computational model that includes heat conduction in the solid substrate and evaporation from the liquid film (micro-layer) present beneath the bubble. Bubble evolution is investigated and the additional (with respect to single phase convection) heat transfer mechanisms due to the ebullition cycle are quantified. The simulations show that latent heat exchange due to evaporation in the micro-layer and sensible heat exchange during the waiting time after bubble departure are the main heat transfer mechanisms. It is found that the presence of an imposed flow normal to the bubble rising path determines a complex velocity and temperature distribution near the nucleation site. This conditions can result in bubble sliding, and influence bubble shape, departure diameter and departure frequency

  10. A dry-spot model for the prediction of critical heat flux in water boiling in bubbly flow regime

    International Nuclear Information System (INIS)

    Ha, Sang Jun; No, Hee Cheon

    1997-01-01

    This paper presents a prediction of critical heat flux (CHF) in bubbly flow regime using dry-spot model proposed recently by authors for pool and flow boiling CHF and existing correlations for forced convective heat transfer coefficient, active site density and bubble departure diameter in nucleate boiling region. Without any empirical constants always present in earlier models, comparisons of the model predictions with experimental data for upward flow of water in vertical, uniformly-heated round tubes are performed and show a good agreement. The parametric trends of CHF have been explored with respect to variation in pressure, tube diameter and length, mass flux and inlet subcooling

  11. Flow boiling in microgap channels experiment, visualization and analysis

    CERN Document Server

    Alam, Tamanna; Jin, Li-Wen

    2013-01-01

    Flow Boiling in Microgap Channels: Experiment, Visualization and Analysis presents an up-to-date summary of the details of the confined to unconfined flow boiling transition criteria, flow boiling heat transfer and pressure drop characteristics, instability characteristics, two phase flow pattern and flow regime map and the parametric study of microgap dimension. Advantages of flow boiling in microgaps over microchannels are also highlighted. The objective of this Brief is to obtain a better fundamental understanding of the flow boiling processes, compare the performance between microgap and c

  12. Heat transfer coefficient for flow boiling in an annular mini gap

    Directory of Open Access Journals (Sweden)

    Hożejowska Sylwia

    2016-01-01

    Full Text Available The aim of this paper was to present the concept of mathematical models of heat transfer in flow boiling in an annular mini gap between the metal pipe with enhanced exterior surface and the external glass pipe. The one- and two-dimensional mathematical models were proposed to describe stationary heat transfer in the gap. A set of experimental data governed both the form of energy equations in cylindrical coordinates and the boundary conditions. The models were formulated to minimize the number of experimentally determined constants. Known temperature distributions in the enhanced surface and in the fluid helped to determine, from the Robin condition, the local heat transfer coefficients at the enhanced surface – fluid contact. The Trefftz method was used to find two-dimensional temperature distributions for the thermal conductive filler layer, enhanced surface and flowing fluid. The method of temperature calculation depended on whether the area of single-phase convection ended with boiling incipience in the gap or the two-phase flow region prevailed, with either fully developed bubbly flow or bubbly-slug flow. In the two–phase flow, the fluid temperature was calculated by Trefftz method. Trefftz functions for the Laplace equation and for the energy equation were used in the calculations.

  13. Free convection film flows and heat transfer laminar free convection of phase flows and models for heat-transfer analysis

    CERN Document Server

    Shang, De-Yi

    2012-01-01

    This book presents recent developments in our systematic studies of hydrodynamics and heat and mass transfer in laminar free convection, accelerating film boiling and condensation of Newtonian fluids, as well as accelerating film flow of non-Newtonian power-law fluids (FFNF). These new developments provided in this book are (i) novel system of analysis models based on the developed New Similarity Analysis Method; (ii) a system of advanced methods for treatment of gas temperature- dependent physical properties, and liquid temperature- dependent physical properties; (iii) the organically combined models of the governing mathematical models with those on treatment model of variable physical properties; (iv) rigorous approach of overcoming a challenge on accurate solution of three-point boundary value problem related to two-phase film boiling and condensation; and (v) A pseudo-similarity method of dealing with thermal boundary layer of FFNF for greatly simplifies the heat-transfer analysis and numerical calculati...

  14. Contribution to the multidimensional modelling of convective high pressure boiling flows for pressurised water reactors

    International Nuclear Information System (INIS)

    Gueguen, J.

    2013-01-01

    This study is a contribution to the modelling of multidimensional high pressure boiling flows relative to PWR. Numerical simulation of such two-phase flows is considered to be an interesting way for the DNB understanding. The first part of this study exposes a two-dimensional steady state two-phase flows model able to predict velocity and temperature profiles in tube. The mixture balanced equations are used with the eddy diffusivity concept to close the turbulent transport terms. The second part is devoted to the development of the model in the general two dimensional case. Contrary to the steady state model, this model is independent of experimental data and implies the use of an original local homogeneous relaxation model (HRM). The results obtained from the comparison with the data bank DEBORA reveals that in a mixture approach two sub models are sufficient to obtain a physical good description of turbulent boiling flows. Some limitations appear at conditions close to DNB conditions. The turbulent closures and the relaxation time in the HRM model have been clearly identified as the most important and sensitive parameters in the model. (author) [fr

  15. A dry-spot model for the prediction of critical heat flux in water boiling in bubbly flow regime

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Sang Jun; No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    This paper presents a prediction of critical heat flux (CHF) in bubbly flow regime using dry-spot model proposed recently by authors for pool and flow boiling CHF and existing correlations for forced convective heat transfer coefficient, active site density and bubble departure diameter in nucleate boiling region. Without any empirical constants always present in earlier models, comparisons of the model predictions with experimental data for upward flow of water in vertical, uniformly-heated round tubes are performed and show a good agreement. The parametric trends of CHF have been explored with respect to variations in pressure, tube diameter and length, mass flux and inlet subcooling. 16 refs., 6 figs., 1 tab. (Author)

  16. A dry-spot model for the prediction of critical heat flux in water boiling in bubbly flow regime

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Sang Jun; No, Hee Cheon [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents a prediction of critical heat flux (CHF) in bubbly flow regime using dry-spot model proposed recently by authors for pool and flow boiling CHF and existing correlations for forced convective heat transfer coefficient, active site density and bubble departure diameter in nucleate boiling region. Without any empirical constants always present in earlier models, comparisons of the model predictions with experimental data for upward flow of water in vertical, uniformly-heated round tubes are performed and show a good agreement. The parametric trends of CHF have been explored with respect to variations in pressure, tube diameter and length, mass flux and inlet subcooling. 16 refs., 6 figs., 1 tab. (Author)

  17. Subcooled flow boiling heat transfer of ethanol aqueous solutions in vertical annulus space

    Directory of Open Access Journals (Sweden)

    Sarafraz M.M.

    2012-01-01

    Full Text Available The subcooled flow boiling heat-transfer characteristics of water and ethanol solutions in a vertical annulus have been investigated up to heat flux 132kW/m2. The variations in the effects of heat flux and fluid velocity, and concentration of ethanol on the observed heat-transfer coefficients over a range of ethanol concentrations implied an enhanced contribution of nucleate boiling heat transfer in flow boiling, where both forced convection and nucleate boiling heat transfer occurred. Increasing the ethanol concentration led to a significant deterioration in the observed heat-transfer coefficient because of a mixture effect, that resulted in a local rise in the saturation temperature of ethanol/water solution at the vapor-liquid interface. The reduction in the heat-transfer coefficient with increasing ethanol concentration is also attributed to changes in the fluid properties (for example, viscosity and heat capacity of tested solutions with different ethanol content. The experimental data were compared with some well-established existing correlations. Results of comparisons indicate existing correlations are unable to obtain the acceptable values. Therefore a modified correlation based on Gnielinski correlation has been proposed that predicts the heat transfer coefficient for ethanol/water solution with uncertainty about 8% that is the least in comparison to other well-known existing correlations.

  18. Natural convection and boiling heat transfer of a liquid metal in a magnetic field

    International Nuclear Information System (INIS)

    Seki, Masahiro; Kawamura, Hiroshi

    1983-02-01

    A liquid metal is often assumed as a coolant and a breeding material of a Tokamak fusion reactor. However, many problems on the thermo-hydraulics of a liquid metal in a magnetic field are still remained to be studied. In the present report, natural convection and boiling of a liquid metal in a strong magnetic field are studied to examine a fundamental feasibility of a fusion reactor cooled by a liquid metal. In the experimental study of the natural convection, the circulation of a liquid metal was found to be surpressed even by a magnetic field parallel to the gravity. A numerical study has confirmed the conclusion drawn by the experiment. In the study of boiling heat transfer, stable boiling of a liquid metal has been found also in a strong magnetic field. The burnout heat flux hardly affected by the magnetic field. These indicate a fundamental feasibility of the liquid-metal cooling for a Tokamak fusion reactor. (author)

  19. Flow boiling in expanding microchannels

    CERN Document Server

    Alam, Tamanna

    2017-01-01

    This Brief presents an up to date summary of details of the flow boiling heat transfer, pressure drop and instability characteristics; two phase flow patterns of expanding microchannels. Results obtained from the different expanding microscale geometries are presented for comparison and addition to that, comparison with literatures is also performed. Finally, parametric studies are performed and presented in the brief. The findings from this study could help in understanding the complex microscale flow boiling behavior and aid in the design and implementation of reliable compact heat sinks for practical applications.

  20. Development of Flow Boiling and Condensation Experiment on the International Space Station- Normal and Low Gravity Flow Boiling Experiment Development and Test Results

    Science.gov (United States)

    Nahra, Henry K.; Hall, Nancy R.; Hasan, Mohammad M.; Wagner, James D.; May, Rochelle L.; Mackey, Jeffrey R.; Kolacz, John S.; Butcher, Robert L.; Frankenfield, Bruce J.; Mudawar, Issam; hide

    2013-01-01

    Flow boiling and condensation have been identified as two key mechanisms for heat transport that are vital for achieving weight and volume reduction as well as performance enhancement in future space systems. Since inertia driven flows are demanding on power usage, lower flows are desirable. However, in microgravity, lower flows are dominated by forces other than inertia (like the capillary force). It is of paramount interest to investigate limits of low flows beyond which the flow is inertial enough to be gravity independent. One of the objectives of the Flow Boiling and Condensation Flight Experiment sets to investigate these limits for flow boiling and condensation. A two-phase flow loop consisting of a Flow Boiling Module and two Condensation Modules has been developed to experimentally study flow boiling condensation heat transfer in the reduced gravity environment provided by the reduced gravity platform. This effort supports the development of a flow boiling and condensation facility for the International Space Station (ISS). The closed loop test facility is designed to deliver the test fluid, FC-72 to the inlet of any one of the test modules at specified thermodynamic and flow conditions. The zero-g-aircraft tests will provide subcooled and saturated flow boiling critical heat flux and flow condensation heat transfer data over wide range of flow velocities. Additionally, these tests will verify the performance of all gravity sensitive components, such as evaporator, condenser and accumulator associated with the two-phase flow loop. We will present in this paper the breadboard development and testing results which consist of detailed performance evaluation of the heater and condenser combination in reduced and normal gravity. We will also present the design of the reduced gravity aircraft rack and the results of the ground flow boiling heat transfer testing performed with the Flow Boiling Module that is designed to investigate flow boiling heat transfer and

  1. Studies in boiling heat transfer in two phase flow through tube arrays: nucleate boiling heat transfer coefficient and maximum heat flux as a function of velocity and quality of Freon-113

    International Nuclear Information System (INIS)

    Rahmani, R.

    1983-01-01

    The nucleate boiling heat-transfer coefficient and the maximum heat flux were studied experimentally as functions of velocity, quality and heater diameter for single-phase flow, and two-phase flow of Freon-113 (trichlorotrifluorethane). Results show: (1) peak heat flux: over 300 measured peak heat flux data from two 0.875-in. and four 0.625-in.-diameter heaters indicated that: (a) for pool boiling, single-phase and two-phase forced convection boiling the only parameter (among hysteresis, rate of power increase, aging, presence and proximity of unheated rods) that has a statistically significant effect on the peak heat flux is the velocity. (b) In the velocity range (0 0 position or the point of impact of the incident fluid) and the top (180 0 position) of the test element, respectively

  2. Study on boiling heat transfer of high temperature liquid sodium

    International Nuclear Information System (INIS)

    Sakurai, Akira

    1978-01-01

    In the Intitute of Atomic Energy, Kyoto University, fundamental studies on steady state and non-steady state heat flow are underway in connection with reactor design and the safety in a critical accident in a sodium-cooled fast breeder reactor. First, the experimental apparatus for sodium heat transfer and the testing system are described in detail. The apparatus is composed of sodium-purifying section including the plugging meter for measuring purity and cold trap, the pool boiling test section for experimenting natural convection boiling heat transfer, the forced convection boiling test section for experimenting forced convection boiling heat transfer, and gas system. Next, the experimental results by the author and the data obtained so far are compared regarding heat transfer in sodium natural convection and stable nucleating boiling and critical heat flux. The effect of liquid head on a heater on boiling heat transfer coefficient and critical heat flux under the condition of low system pressure in most fundamental pool boiling was elucidated quantitatively, which has been overlooked in previous studies. It was clarified that this is the essentially important problem that can not be overlooked. From this point of view, expressions on heat transfer were also re-investigated. (Wakatsuki, Y.)

  3. Flow Boiling on a Downward-Facing Inclined Plane Wall of Core Catcher

    International Nuclear Information System (INIS)

    Kim, Hyoung Tak; Bang, Kwang Hyun; Suh, Jung Soo

    2013-01-01

    In order to investigate boiling behavior on downward-facing inclined heated wall prior to the CHF condition, an experiment was carried out with 1.2 m long rectangular channel, inclined by 10 .deg. from the horizontal plane. High speed video images showed that the bubbles were sliding along the heated wall, continuing to grow and combining with the bubbles growing at their nucleation sites in the downstream. These large bubbles continued to slide along the heated wall and formed elongated slug bubbles. Under this slug bubble thin liquid film layer on the heated wall was observed and this liquid film prevents the wall from dryout. The length, velocity and frequency of slug bubbles sliding on the heated wall were measured as a function of wall heat flux and these parameters were used to develop wall boiling model for inclined, downward-facing heated wall. One approach to achieve coolable state of molten core in a PWR-like reactor cavity during a severe accident is to retain the core melt on a so-called core catcher residing on the reactor cavity floor after its relocation from the reactor pressure vessel. The core melt retained in the core catcher is cooled by water coolant flowing in an inclined cooling channel underneath as well as the water pool overlaid on the melt layer. Two-phase flow boiling with downward-facing heated wall such as this core catcher cooling channel has drawn a special attention because this orientation of heated wall may reach boiling crisis at lower heat flux than that of a vertical or upward-facing heated wall. Nishikawa and Fujita, Howard and Mudawar, Qiu and Dhir have conducted experiments to study the effect of heater orientation on boiling heat transfer and CHF. SULTAN experiment was conducted to study inclined large-scale structure coolability by water in boiling natural convection. In this paper, high-speed visualization of boiling behavior on downward-facing heated wall inclined by 10 .deg. is presented and wall boiling model for the

  4. Two-phase flow boiling pressure drop in small channels

    International Nuclear Information System (INIS)

    Sardeshpande, Madhavi V.; Shastri, Parikshit; Ranade, Vivek V.

    2016-01-01

    Highlights: • Study of typical 19 mm steam generator tube has been undertaken in detail. • Study of two phase flow boiling pressure drop, flow instability and identification of flow regimes using pressure fluctuations is the main focus of present work. • Effect of heat and mass flux on pressure drop and void fraction was studied. • Flow regimes identified from pressure fluctuations data using FFT plots. • Homogeneous model predicted pressure drop well in agreement. - Abstract: Two-phase flow boiling in small channels finds a variety of applications in power and process industries. Heat transfer, boiling flow regimes, flow instabilities, pressure drop and dry out are some of the key issues related to two-phase flow boiling in channels. In this work, the focus is on pressure drop in two-phase flow boiling in tubes of 19 mm diameter. These tubes are typically used in steam generators. Relatively limited experimental database is available on 19 mm ID tube. Therefore, in the present work, the experimental set-up is designed for studying flow boiling in 19 mm ID tube in such a way that any of the different flow regimes occurring in a steam generator tube (from pre-heating of sub-cooled water to dry-out) can be investigated by varying inlet conditions. The reported results cover a reasonable range of heat and mass flux conditions such as 9–27 kW/m 2 and 2.9–5.9 kg/m 2 s respectively. In this paper, various existing correlations are assessed against experimental data for the pressure drop in a single, vertical channel during flow boiling of water at near-atmospheric pressure. A special feature of these experiments is that time-dependent pressures are measured at four locations along the channel. The steady-state pressure drop is estimated and the identification of boiling flow regimes is done with transient characteristics using time series analysis. Experimental data and corresponding results are compared with the reported correlations. The results will be

  5. Development of natural convection heat transfer correlation for liquid metal with overlying boiling coolant

    International Nuclear Information System (INIS)

    Cho, Jae Seon; Suh, Kune Y.; Chung, Chang Hyun; Park, Rae Joon; Kim, Sang Baik

    1999-01-01

    Experimental study was performed to investigate the natural convection heat transfer characteristics and the crust formation of the molten metal pool concurrent with forced convective boiling of the overlying coolant. Tests were performed under the condition of the bottom surface heating in the test section and the forced convection of the coolant being injected onto the molten metal pool. The constant temperature and constant heater input power conditions were adopted for the bottom heating. Test results showed that the temperature distribution and crust layer thickness in the metal layer are appreciably affected by the heated bottom surface temperature of the test section, but not much by the coolant injection rate. The relationship between the Nu number and Ra number in the molten metal pool region is determined and compared with the correlations in the literature, and the experiment without coolant boiling. A new correlation on the relationship between the Nu number and Ra number in the molten metal pool with crust formation is developed from the experimental data

  6. Prediction of flow boiling curves based on artificial neural network

    International Nuclear Information System (INIS)

    Wu Junmei; Xi'an Jiaotong Univ., Xi'an; Su Guanghui

    2007-01-01

    The effects of the main system parameters on flow boiling curves were analyzed by using an artificial neural network (ANN) based on the database selected from the 1960s. The input parameters of the ANN are system pressure, mass flow rate, inlet subcooling, wall superheat and steady/transition boiling, and the output parameter is heat flux. The results obtained by the ANN show that the heat flux increases with increasing inlet sub cooling for all heat transfer modes. Mass flow rate has no significant effects on nucleate boiling curves. The transition boiling and film boiling heat fluxes will increase with an increase of mass flow rate. The pressure plays a predominant role and improves heat transfer in whole boiling regions except film boiling. There are slight differences between the steady and the transient boiling curves in all boiling regions except the nucleate one. (authors)

  7. Flow dynamics of volume-heated boiling pools

    International Nuclear Information System (INIS)

    Ginsberg, T.; Jones, O.C.; Chen, J.C.

    1979-01-01

    Safety analyses of fast breeder reactors require understanding of the two-phase fluid dynamic and heat transfer characteristics of volume-heated boiling pool systems. Design of direct contact three-phase boilers, of practical interest in the chemical industries also requires understanding of the fundamental two-phase flow and heat transfer behavior of volume boiling systems. Several experiments have been recently reported relevant to the boundary heat-loss mechanisms of boiling pool systems. Considerably less is known about the two-phase fluid dynamic behavior of such systems. This paper describes an experimental investigation of the steady-state flow dynamics of volume-heated boiling pool systems

  8. Augmentation of forced flow boiling heat transfer by introducing air flow into subcooled water flow

    International Nuclear Information System (INIS)

    Koizumi, Y.; Ohtake, H.; Yuasa, T.; Matsushita, N.

    2001-01-01

    The effect of air injection into a subcooled water flow on boiling heat transfer and a critical heat flux (CHF) was examined experimentally. Experiments were conducted in the range of subcooling of 50 K, a superficial velocity of water and air Ul = 0.17 ∼ 3.4 and Ug = 0 ∼ 15 m/s, respectively. A test heat transfer surface was a 5 mm wide, 40 mm long and 0.5 mm thick stainless steel sheet embedded on the bottom wall of a 10 mm high and 20 mm wide rectangular flow channel. Nine times enhancement of the heat transfer coefficient in the non-boiling region was attained at the most by introducing an air flow into a water single-phase flow. The heat transfer improvement was prominent when the water flow rate was low and the air introduction was large. The present results of the non-boiling heat transfer were well correlated with the Lockhart-Martinelli parameter X tt ; h TP /h L0 = 5.0(1/ X tt ) 0.5 . The air introduction has some effect on the augmentation of heat transfer in the boiling region, however, the two-phase flow effect was little and the boiling was dominant in the fully developed boiling region. The CHF was improved a little by the air introduction in the high water flow region. However, that was rather greatly reduced in the low flow region. Even so, the general trend by the air introduction was that qCHF increased as the air introduction was increased. The heat transfer augmentation in the non-boiling region was attained by less power increase than that in the case that only the water flow rate was increased. From the aspect of the power consumption and the heat transfer enhancement, the small air introduction in the low water flow rate region seemed more profitable, although the air introduction in the high water flow rate region and also the large air introduction were still effective in the augmentation of the heat transfer in the non-boiling region. (author)

  9. Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun [Seoul National University, Seoul (Korea, Republic of); Paark, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    This paper presents results of experimental studies on the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. Ad a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 deg C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleigh number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer. 10 refs., 4 figs., 1 tab. (Author)

  10. Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun [Seoul National University, Seoul (Korea, Republic of); Paark, Rae Joon; Kim, Sang Baik [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper presents results of experimental studies on the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. Ad a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 deg C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleigh number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer. 10 refs., 4 figs., 1 tab. (Author)

  11. Natural convection heat transfer characteristics of the molten metal pool with solidification by boiling coolant

    International Nuclear Information System (INIS)

    Cho, Jae Seon; Suh, Kune Yull; Chung, Chang Hyun; Park, Rae Joon; Kim, Sang Baik

    1997-01-01

    This paper presents results of experimental studies on the heat transfer and solidifcation of the molten metal pool with overlying coolant with boiling. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. As a result, the crust, which is a solidified layer, may form at the top of the molten metal pool. Heat transfer is accomplished by a conjugate mechanism, which consists of the natural convection of the molten metal pool, the conduction in the crust layer and the convective boiling heat transfer in the coolant. This work examines the crust formation and the heat transfer rate on the molten metal pool with boiling coolant. The simulant molten pool material is tin (Sn) with the melting temperature of 232 .deg. C. Demineralized water is used as the working coolant. The crust layer thickness was ostensibly varied by the heated bottom surface temperature of the test section, but not much affected by the coolant injection rate. The correlation between the Nusselt number and the Rayleight number in the molten metal pool region of this study is compared against the crust formation experiment without coolant boiling and the literature correlations. The present experimental results are higher than those from the experiment without coolant boiling, but show general agreement with the Eckert correlation, with some deviations in the high and low ends of the Rayleigh number. This discrepancy is currently attributed to concurrent rapid boiling of the coolant on top of the metal layer

  12. Boiling curve in high quality flow boiling

    International Nuclear Information System (INIS)

    Shiralkar, B.S.; Hein, R.A.; Yadigaroglu, G.

    1980-01-01

    The post dry-out heat transfer regime of the flow boiling curve was investigated experimentally for high pressure water at high qualities. The test section was a short round tube located downstream of a hot patch created by a temperature controlled segment of tubing. Results from the experiment showed that the distance from the dryout point has a significant effect on the downstream temperatures and there was no unique boiling curve. The heat transfer coefficients measured sufficiently downstream of the dryout point could be correlated using the Heineman correlation for superheated steam, indicating that the droplet deposition effects could be neglected in this region

  13. Experimental study and modelling of transient boiling

    International Nuclear Information System (INIS)

    Baudin, Nicolas

    2015-01-01

    A failure in the control system of the power of a nuclear reactor can lead to a Reactivity Initiated Accident in a nuclear power plant. Then, a power peak occurs in some fuel rods, high enough to lead to the coolant film boiling. It leads to an important increase of the temperature of the rod. The possible risk of the clad failure is a matter of interest for the Institut de Radioprotection et de Securite Nucleaire. The transient boiling heat transfer is not yet understood and modelled. An experimental set-up has been built at the Institut de Mecanique des Fluides de Toulouse (IMFT). Subcooled HFE-7000 flows vertically upward in a semi annulus test section. The inner half cylinder simulates the clad and is made of a stainless steel foil, heated by Joule effect. Its temperature is measured by an infrared camera, coupled with a high speed camera for the visualization of the flow topology. The whole boiling curve is studied in steady state and transient regimes: convection, onset of boiling, nucleate boiling, critical heat flux, film boiling and rewetting. The steady state heat transfers are well modelled by literature correlations. Models are suggested for the transient heat flux: the convection and nucleate boiling evolutions are self-similar during a power step. This observation allows to model more complex evolutions, as temperature ramps. The transient Hsu model well represents the onset of nucleate boiling. When the intensity of the power step increases, the film boiling begins at the same temperature but with an increasing heat flux. For power ramps, the critical heat flux decreases while the corresponding temperature increases with the heating rate. When the wall is heated, the film boiling heat transfer is higher than in steady state but it is not understood. A two-fluid model well simulates the cooling film boiling and the rewetting. (author)

  14. Analysis of the fragmentation of hot drops with film boiling in a water flow

    International Nuclear Information System (INIS)

    Malmazet, Erik de

    2009-01-01

    The goal of this work is to study different aspects of the fragmentation of very hot drops placed in a uniform flow, a phenomenon related to vapor explosion studies. First, a theoretical study of the isothermal hydrodynamic fragmentation of drops by the Boundary Layer Stripping (BLS) mechanism is done by developing two models. The first model, contrary to past studies which dismissed the BLS, includes deformation and acceleration effects and this is shown to greatly enhance the mass loss by BLS, which enables this mechanism to become a much more effective mechanism when the external flow is gaseous. But it is still ineffective in the liquid case. The second model describes transient aspects of the BLS, and by coupling it with a stripping criteria for the internal boundary layer, it is possible to predict the time of the initiation of fragmentation. Then, a model for film boiling over horizontal cylinders and axisymmetric bodies which is able to properly describe the inertial and convection terms in the vapor flow is presented. This has never been done before, although these terms cannot be neglected in physical conditions close to vapor explosions. The model is able to predict all the experimental results of TREPAM, the only existing forced convection film boiling experiment in conditions close to a vapor explosion, and which results could not be predicted by other models. In the last part, an experimental study of the fragmentation of hot tin drops in a water flow which uses digital fast camera and flash X ray imagery is presented. This study has allowed the observation of several new features of the drop fragmentation mechanism. (author) [fr

  15. Boiling induced mixed convection in cooling loops

    International Nuclear Information System (INIS)

    Knebel, J.U.; Janssens-Maenhout, G.; Mueller, U.

    2000-01-01

    This article describes the SUCO program performed at the Forschungszentrum Karlsruhe. The SUCO program is a three-step series of scaled model experiments investigating the possibility of a sump cooling concept for future light water reactors. In case of a core melt accident, the sump cooling concept realises a decay heat removal system that is based on passive safety features within the containment. The article gives, first, results of the experiments in the 1:20 linearly scaled SUCOS-2D test facility. The experimental results are scaled-up to the conditions in the prototype, allowing a statement with regard to the feasibility of the sump cooling concept. Second, the real height SUCOT test facility with a volume and power scale of 1:356 that is aimed at investigating the mixed single-phase and two-phase natural circulation flow in the reactor sump, together with first measurement results, are discussed. Finally, a numerical approach to model the subcooled nucleate boiling phenomena in the test facility SUCOT is presented. Physical models describing interfacial mass, momentum and-heat transfer are developed and implemented in the commercial software package CFX4.1. The models are validated for an isothermal air-water bubbly flow experiment and a subcooled boiling experiment in vertical annular water flow. (author)

  16. Single-phase flow and flow boiling of water in horizontal rectangular microchannels

    OpenAIRE

    Mirmanto

    2013-01-01

    This thesis was submitted for the degree of Doctor of Philosophy and awarded by Brunel University The current study is part of a long term experimental project devoted to investigating single-phase flow pressure drop and heat transfer, flow boiling pressure drop and heat transfer, flow boiling instability and flow visualization of de-ionized water flow in microchannels. The experimental facility was first designed and constructed by S. Gedupudi (2009) and in the present study; ...

  17. Study on boiling heat transfer of subcooled flow under oscillatory flow condition

    International Nuclear Information System (INIS)

    Ohtake, Hiroyasu; Yamazaki, Satoshi; Koizumi, Yasuo

    2004-01-01

    The Onset of Nucleate Boiling, the point of Net Vapor Generation and Critical Heat Flux on subcooled flow boiling under oscillatory flow, focusing on liquid velocity, amplitude and frequency of oscillatory flow were investigated experimentally and analytically. Experiments were conducted using a copper thin-film and subcooled water in a range of the liquid velocity from 0.27 to 4.07 m/s at 0.10MPa. The liquid subcooling was 20K. Frequency of oscillatory flow was 2 and 4 Hz, respectively; amplitude of oscillatory flow was 25 and 50% in a ratio of main flow rate, respectively. Temperatures at Onset of Nuclear Boiling and Critical Heat Flux obtained in the experiments decreased with the oscillatory flow. The decrease of liquid velocity by oscillatory flow caused the ONB and the CHF to decrease. On the other hand, heat flux at Net Vapor Generation decreased with oscillatory flow; the increase of liquid velocity by oscillatory flow caused the NVG to decrease. (author)

  18. Validation of system codes RELAP5 and SPECTRA for natural convection boiling in narrow channels

    Energy Technology Data Exchange (ETDEWEB)

    Stempniewicz, M.M., E-mail: stempniewicz@nrg.eu; Slootman, M.L.F.; Wiersema, H.T.

    2016-10-15

    Highlights: • Computer codes RELAP5/Mod3.3 and SPECTRA 3.61 validated for boiling in narrow channels. • Validated codes can be used for LOCA analyses in research reactors. • Code validation based on natural convection boiling in narrow channels experiments. - Abstract: Safety analyses of LOCA scenarios in nuclear power plants are performed with so called thermal–hydraulic system codes, such as RELAP5. Such codes are validated for typical fuel geometries applied in nuclear power plants. The question considered by this article is if the codes can be applied for LOCA analyses in research reactors, in particular exceeding CHF in very narrow channels. In order to answer this question, validation calculations were performed with two thermal–hydraulic system codes: RELAP and SPECTRA. The validation was based on natural convection boiling in narrow channels experiments, performed by Prof. Monde et al. in the years 1990–2000. In total 42 vertical tube and annulus experiments were simulated with both codes. A good agreement of the calculated values with the measured data was observed. The main conclusions are: • The computer codes RELAP5/Mod 3.3 (US NRC version) and SPECTRA 3.61 have been validated for natural convection boiling in narrow channels using experiments of Monde. The dimensions applied in the experiments were performed for a range that covers the values observed in typical research reactors. Therefore it is concluded that both codes are validated and can be used for LOCA analyses in research reactors, including natural convection boiling. The applicability range of the present validation is: hydraulic diameters of 1.1 ⩽ D{sub hyd} ⩽ 9.0 mm, heated lengths of 0.1 ⩽ L ⩽ 1.0 m, pressures of 0.10 ⩽ P ⩽ 0.99 MPa. In most calculations the burnout was predicted to occur at lower power than that observed in the experiments. In several cases the burnout was observed at higher power. The overprediction was not larger than 16% in RELAP and 15% in

  19. Flow-excursion-induced dryout at low-heat-flux

    International Nuclear Information System (INIS)

    Khatib-Rahbar, M.; Cazzoli, E.G.

    1983-01-01

    Flow-excursion-induced dryout at low-heat-flux natural-convection boiling, typical of liquid-metal fast-breeder reactors, is addressed. Steady-state calculations indicate that low-quality boiling is possible up to the point of Ledinegg instability leading to flow excursion and subsequent dryout in agreement with experimental data. A flow-regime-dependent dryout heat flux relationship based upon saturated boiling criterion is also presented. Transient analysis indicates that premature flow excursion can not be ruled out and sodium boiling is highly transient dependent. Analysis of a high-heat-flux forced convection, loss-of-flow transient shows a significantly faster flow excursion leading to dryout in excellent agreement with parallel calculations using the two-dimensional THORAX code. 17 figures

  20. Burnout in subcooled flow boiling of water. A visual experimental study

    Energy Technology Data Exchange (ETDEWEB)

    Celata, G.P.; Mariani, A.; Zummo, G. [ENEA, Engineering Div., National Institute of Thermal Fluid-Dynamics, Rome (Italy); Cumo, M. [University of Rome la Sapienza, Rome (Italy)

    2000-12-01

    The objective of the present work is to perform a photographic study of the burnout in highly subcooled flow boiling, in order to provide a qualitative description of the flow pattern under different conditions of boiling regime: ONB (onset of nucleate boiling), subcooled flow boiling and thermal crisis. In particular, the flow visualisation is focused on the phenomena occurring on the heated wall during the thermal crisis up to the physical burnout of the heater. Vapour bubble parameters are measured from flow images recorded, while the wall temperature is measured with an indirect method, by recording the heater elongation during all flow regimes studied. The combination of bubble parameters and wall temperature measurements as well as direct observations of the flow pattern, for all flow regimes, are collected in graphs which provide a useful global point of view of boiling phenomena, especially during boiling crisis. Under these conditions, a detailed analysis of the mechanisms leading to the critical heat flux is reported, and the so called events sequence, from thermal crisis occurrence up to heater burnout, is illustrated. (authors)

  1. Burnout in subcooled flow boiling of water. A visual experimental study

    International Nuclear Information System (INIS)

    Celata, G.P.; Mariani, A.; Zummo, G.; Cumo, M.

    2000-01-01

    The objective of the present work is to perform a photographic study of the burnout in highly subcooled flow boiling, in order to provide a qualitative description of the flow pattern under different conditions of boiling regime: ONB (onset of nucleate boiling), subcooled flow boiling and thermal crisis. In particular, the flow visualisation is focused on the phenomena occurring on the heated wall during the thermal crisis up to the physical burnout of the heater. Vapour bubble parameters are measured from flow images recorded, while the wall temperature is measured with an indirect method, by recording the heater elongation during all flow regimes studied. The combination of bubble parameters and wall temperature measurements as well as direct observations of the flow pattern, for all flow regimes, are collected in graphs which provide a useful global point of view of boiling phenomena, especially during boiling crisis. Under these conditions, a detailed analysis of the mechanisms leading to the critical heat flux is reported, and the so called events sequence, from thermal crisis occurrence up to heater burnout, is illustrated. (authors)

  2. A one-dimensional semi-empirical model considering transition boiling effect for dispersed flow film boiling

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Yu-Jou [Institute of Nuclear Engineering and Science, National Tsing Hua University, Hsinchu 30013, Taiwan, ROC (China); Pan, Chin, E-mail: cpan@ess.nthu.edu.tw [Institute of Nuclear Engineering and Science, National Tsing Hua University, Hsinchu 30013, Taiwan, ROC (China); Department of Engineering and System Science, National Tsing Hua University, Hsinchu 30013, Taiwan, ROC (China); Low Carbon Energy Research Center, National Tsing Hua University, Hsinchu 30013, Taiwan, ROC (China)

    2017-05-15

    Highlights: • Seven heat transfer mechanisms are studied numerically by the model. • A semi-empirical method is proposed to account for the transition boiling effect. • The parametric effects on the heat transfer mechanisms are investigated. • The thermal non-equilibrium phenomenon between vapor and droplets is investigated. - Abstract: The objective of this paper is to develop a one-dimensional semi-empirical model for the dispersed flow film boiling considering transition boiling effects. The proposed model consists of conservation equations, i.e., vapor mass, vapor energy, droplet mass and droplet momentum conservation, and a set of closure relations to address the interactions among wall, vapor and droplets. The results show that the transition boiling effect is of vital importance in the dispersed flow film boiling regime, since the flowing situation in the downstream would be influenced by the conditions in the upstream. In addition, the present paper, through evaluating the vapor temperature and the amount of heat transferred to droplets, investigates the thermal non-equilibrium phenomenon under different flowing conditions. Comparison of the wall temperature predictions with the 1394 experimental data in the literature, the present model ranging from system pressure of 30–140 bar, heat flux of 204–1837 kW/m{sup 2} and mass flux of 380–5180 kg/m{sup 2} s, shows very good agreement with RMS of 8.80% and standard deviation of 8.81%. Moreover, the model well depicts the thermal non-equilibrium phenomenon for the dispersed flow film boiling.

  3. Burnout in boiling heat transfer. Part III. High-quality forced-convection systems

    International Nuclear Information System (INIS)

    Bergles, A.E.

    1979-01-01

    This is the final part of a review of burnout during boiling heat transfer. The status of burnout in high-quality forced-convection systems is reviewed, and recent developments are summarized in detail. A general guide to the considerable literature is given. Parametric effects and correlations for water in circular and noncircular ducts are presented. Other topics discussed include transients, steam-generator applications, correlations for other fluids, fouling, and augmentation

  4. Prediction of incipient flow boiling from a uniformly heated surface

    International Nuclear Information System (INIS)

    Yin, S.T.; Abdelmessih, A.H.

    1977-01-01

    This study was undertaken to investigate the phenomenon of liquid superheat during incipient boiling in a uniformly heated forced convection channel. Experimental data were obtained using Freon 11 as the test medium. Based on existing theories, an analytical method was developed for predicting the point of termination of nucleate boiling, observed during a decreasing heat flux process with a nucleation activated surface. The method may also be used to predict the point of boiling incipience, observed during an increasing heat flux process with a non-activated surface; this point does not appear to have been treated analytically in previous work. It can be shown that some of the existing models are special cases of the present formulation

  5. Application of Sub-cooled Boiling Model to Thermal-hydraulic Analysis Inside a CANDU-6 Fuel Channel

    International Nuclear Information System (INIS)

    Kim, Man Woong; Lee, Sang Kyu; Kim, Hyun Koon; Yoo, Kun Joong; Kang, Hyoung Chul; Yoo, Seong Yeon

    2007-01-01

    Forced convection nucleate boiling is encountered in heat exchangers during normal and non-nominal modes of operation in pressurized water or boiling water reactors (PWRs or BWRs). If the wall temperature of the piping is higher than the saturation temperature of the nearby liquid, nucleate boiling occurs. In this regime, bubbles are formed at the wall. Their growth is promoted by the wall superheat (the difference between the wall and saturation temperatures), and they depart from the wall as a result of gravitational and liquid inertia forces. If the bulk liquid is subcooled, condensation at the bubble-liquid interface takes place and the bubble may collapse. This convection nucleate boiling is called as a sub-cooled nucleate boiling. As for the fuel channel of a CANDU 6 reactor, forced convection nucleate boiling models for flows along fuel elements enclosed inside typical CANDU-6 fuel channel has encountered difficulties due to the modeling of local effects along the horizontal channel. Therefore, the subcooled nucleate boiling has been modeled through temperature driven boiling heat and mass transfer, using a model developed at Rensselaer Polytechnic Institute. The objectives of this study are: (i) to investigate a proposed sub-cooled boiling model developed at Rensselaer Polytechnic Institute and (ii) to apply against a experiment and (iii) to predict local distributions of flow fields for the actual fuel channel geometries of CANDU-6 reactors. The numerical implementation is conducted using by the FLUENT 6.2 CFD computer code

  6. Applications of artificial neutral network for the prediction of flow boiling curves

    International Nuclear Information System (INIS)

    Su Guanghui; Jia Dounan; Fukuda, Kenji; Morita, Koji; Pidduck, Mark; Matsumoto, Tatsuya; Akasaka, Ryo

    2002-01-01

    An artificial neural network (ANN) was applied successfully to predict flow boiling curves. The databases used in the analysis are from the 1960's, including 1,305 data points which cover these parameter ranges: pressure P=100-1,000 kPa, mass flow rate G=40-500 kg/m 2 ·s, inlet subcooling ΔT sub =0-35degC, wall superheat ΔT w =10-300degC and heat flux Q=20-8,000 kW/m 2 . The proposed methodology allows us to achieve accurate results, thus it is suitable for the processing of the boiling curve data. The effects of the main parameters on flow boiling curves were analyzed using the ANN. The heat flux increases with increasing inlet subcooling for all heat transfer modes. Mass flow rate has no significant effects on nucleate boiling curves. The transition boiling and film boiling heat fluxes will increase with an increase in the mass flow rate. Pressure plays a predominant role and improves heat transfer in all boiling regions except the film boiling region. There are slight differences between the steady and the transient boiling curves in all boiling regions except the nucleate region. The transient boiling curve lies below the corresponding steady boiling curve. (author)

  7. A study on boiling heat transfer with mixture boiling from vertical rod fin

    International Nuclear Information System (INIS)

    Kim, M.C.

    1981-01-01

    The purpose of the present study is concerned with the boiling characteristic of variations of the length-diameter ratio on the heat transfer rate where the nucleate boiling and natural convection occurred simultaneously. Circular fins were made with copper rod 32 mm in diameter, and those surfaces were mirror finished. The length-diameter ratio was varied 1 to 6. As a boiling liquid, the distilled water was used in this experiment. The results of this experiment were obtained as below. 1) From the observations, it was confirmed that nucleate boiling and natural convection occurred simultaneously. 2) As the length-diameter ratio increased, the boiling heat transfer rate also augmented. (author)

  8. Boiling and burnout phenomena under transient heat input, 1

    International Nuclear Information System (INIS)

    Aoki, Shigebumi; Kozawa, Yoshiyuki; Iwasaki, Hideaki.

    1976-01-01

    In order to simulate the thermo-hydrodynamic conditions at reactor power excursions, a test piece was placed in a forced convective channel and heated with exponential power inputs. The boiling heat transfer and the burnout heat flux under the transient heat input were measured, and pressure and water temperature changes in the test section were recorded at the same time. Following experimental results were obtained; (1) Transient boiling heat transfer characteristics at high heat flux stayed on the stationary nucleate boiling curve of each flow condition, or extrapolated line of the curves. (2) Transient burnout heat flux increased remarkably with decreasing heating-time-constant, when the flow rate was lower and the subcooling was higher. (3) Transient burnout phenomena were expressed with the relation of (q sub(max) - q sub(sBO)) tau = constant at several flow conditions. This relation was derived from the stationary burnout mechanism of pool boiling. (auth.)

  9. Flow film boiling heat transfer in water and Freon-113

    International Nuclear Information System (INIS)

    Liu, Qiusheng; Shiotsu, Masahiro; Sakurai, Akira

    2002-01-01

    Experimental apparatus and method for film boiling heat transfer measurement on a horizontal cylinder in forced flow of water and Freon-113 under pressurized and subcooled conditions were developed. The experiments of film boiling heat transfer from single horizontal cylinders with diameters ranging from 0.7 to 5 mm in saturated and subcooled water and Freon-113 flowing upward perpendicular to the cylinders were carried out for the flow velocities ranging from 0 to 1 m/s under system pressures ranging from 100 to 500 kPa. Liquid subcoolings ranged from 0 to 50 K, and the cylinder surface superheats were raised up to 800 K for water and 400 K for Freon-113. The film boiling heat transfer coefficients obtained were depended on surface superheats, flow velocities, liquid subcoolings, system pressures and cylinder diameters. The effects of these parameters were systematically investigated under wider ranges of experimental conditions. It was found that the heat transfer coefficients are higher for higher flow velocities, subcoolings, system pressures, and for smaller cylinder diameters. The observation results of film boiling phenomena were obtained by a high-speed video camera. A new correlation for subcooled flow film boiling heat transfer was derived by modifying authors' correlation for saturated flow film boiling heat transfer with authors' experimental data under wide subcooled conditions. (author)

  10. Mechanism of flow choking at shock boiling-up of a liquid

    International Nuclear Information System (INIS)

    Labuntsov, D.A.; Avdeev, A.A.

    1982-01-01

    The theory of the outflow of a saturated or non-heated liquid with thermodynamic parameters reaching the critical point from diaphragms and short nozzles has been developed basing on the concept of the boiling-up jump. Three characteristic flow conditions have been revealed: hydraulic, conditions when boiling-up jump is formed, and conditions of radial expansion of the flow. If the initial flow's parameters are low, the hydraulic conditions are realized. The expansion of the flow-passage cross-section of flow small jets by the final value takes place when the spinoidal overheating is reached near the exit cut-off at a small distance equal to the thickness of the boiling-up zone; and that causes the intensive jet dispersion in the radial direction. In case of overheatings close to the thermodynamic critical point, a boiling-up jump is formed inside the channel. The mechanism of flow choking has been analyzed; recommendations on calculation of the critical flow rate of a boiling-up liquid are given. The studied mechanism of flow choking at shock boiling-up of the flow permits to draw a rather detailed physical picture of the phenomenon and to give an explanation of the majority of experimentally-observed effects

  11. Critical heat flux in flow boiling in microchannels

    CERN Document Server

    Saha, Sujoy Kumar

    2015-01-01

    This Brief concerns the important problem of critical heat flux in flow boiling in microchannels. A companion edition in the SpringerBrief Subseries on Thermal Engineering and Applied Science to “Heat Transfer and Pressure Drop in Flow Boiling in Microchannels,” by the same author team, this volume is idea for professionals, researchers, and graduate students concerned with electronic cooling.

  12. Boiling in microchannels: a review of experiment and theory

    International Nuclear Information System (INIS)

    Thome, John R.

    2004-01-01

    A summary of recent research on boiling in microchannels is presented. The review addresses the topics of macroscale versus microscale heat transfer, two-phase flow regimes, flow boiling heat transfer results for microchannels, heat transfer mechanisms in microchannels and flow boiling models for microchannels. In microchannels, the most dominant flow regime appears to be the elongated bubble mode that can persist up to vapor qualities as high as 60-70% in microchannels, followed by annular flow. Flow boiling heat transfer coefficients have been shown experimentally to be dependent on heat flux and saturation pressure while only slightly dependent on mass velocity and vapor quality. Hence, these studies have concluded that nucleate boiling controls evaporation in microchannels. Instead, a recent analytical study has shown that transient evaporation of the thin liquid films surrounding elongated bubbles is the dominant heat transfer mechanism as opposed to nucleate boiling and is able to predict these trends in the experimental data. Newer experimental studies have further shown that there is in fact a significant effect of mass velocity and vapor quality on heat transfer when covering a broader range of conditions, including a sharp peak at low vapor qualities at high heat fluxes. Furthermore, it is concluded that macroscale models are not realistic for predicting flowing boiling coefficients in microchannels as the controlling mechanism is not nucleate boiling nor turbulent convection but is transient thin film evaporation (also, microchannel flows are typically laminar and not turbulent as assumed by macroscopic models). A more advanced three-zone flow boiling model for evaporation of elongated bubbles in microchannels is currently under development that so far qualitatively describes all these trends. Numerous fundamental aspects of two-phase flow and evaporation remain to be better understood and some of these aspects are also discussed

  13. Analysis of boiling

    International Nuclear Information System (INIS)

    Kolev, N.I.

    2011-01-01

    This paper summarizes the author's results in boiling analysis obtained in the last 17 years. It demonstrates that more information can be extracted from the analysis by incorporating even of gross turbulence characteristics consistently in the analysis and appropriate local volume and time averaging. The main findings are: Even in large scale analysis (no direct numerical simulation) the steady and transient averaged turbulence characteristics are necessary to increase the quality of predicting heat and mass transfer. It allows simulating the heat transfer change behind spacer grids analytically which is not the practice up to now. This allows also to simulate the change of the deposition behind the spacer grid and therefore this bring us closer to the mechanistic prediction of dry out. Accurate boiling heat transfer predictions require knowledge on the nucleation characteristics of each particular surface. The pulsation characteristics at the wall controlling the heat transfer are associated with the bubble departure frequencies but not identical with them. Considering the mutual interactions of the bubbles leads to the surprising analytical prediction of the departure from nucleate boiling just by using the mechanisms acting during flow boiling only. The performance of the author's analytical two-phase convection model combined with its analytical nuclide boiling model is proven to have the accuracy of the empirical Chen's model by having the advantage of predicting analytically the internal characteristics of the flow each of it validated by experiment. This is also important for the future use in multiphase CFD where details about the flow field generation have to be also predicted by constitutive relation as summarized in this paper. (author)

  14. Analysis of boiling

    International Nuclear Information System (INIS)

    Kolev, Nikolay Ivanov

    2011-01-01

    This paper summarizes the author's results in boiling analysis obtained in the last 17 years. It demonstrates that more information can be extracted from the analysis by incorporating even of gross turbulence characteristics consistently in the analysis and appropriate local volume and time averaging. The main findings are: Even in large scale analysis (no direct numerical simulation) the steady and transient averaged turbulence characteristics are necessary to increase the quality of predicting heat and mass transfer. It allows to simulate the heat transfer change behind spacer grids analytically which is not the practice up to now. This allows also to simulate the change of the deposition behind the spacer grid and therefore this bring us closer to the mechanistic prediction of dry out. Accurate boiling heat transfer predictions require knowledge on the nucleation characteristics of each particular surface. The pulsation characteristics at the wall controlling the heat transfer are associated with the bubble departure frequencies but not identical with them. Considering the mutual interactions of the bubbles leads to the surprising analytical prediction of the departure from nucleate boiling just by using the mechanisms acting during flow boiling only. The performance of the author's analytical two-phase convection model combined with its analytical nuclide boiling model is proven to have the accuracy of the empirical Chen's model by having the advantage of predicting analytically the internal characteristics of the flow each of it validated by experiment. This is also important for the future use in multiphase CFD where details about the flow field generation have to be also predicted by constitutive relation as summarized in this paper. (author)

  15. Critical heat flux for flow boiling of water in mini-channels

    International Nuclear Information System (INIS)

    Zhang, Weizhong; Mishima, Kaichiro; Hibiki, Takashi

    2007-01-01

    Critical heat flux (CHF) is a limiting factor when flow boiling is applied to dissipate high heat flux in mini-channels. In view of practical importance of critical heat flux correlations in engineering design and prediction, this study presents an evaluation of existing CHF correlations for flow boiling of water with available databases taken from small-diameter tubes, and then develops a new, simple CHF correlation for flow boiling in mini-channel. Three correlations by Bowring, Katto and Shah are evaluated with available CHF data in the literature for saturated flow boiling, and three correlations by Inasaka-Nariai, Celata et al. and Hall-Mudawar evaluated with the CHF data for subcooled flow boiling. The Hall-Mudawar correlation and the Shah correlation appear to be the most reliable tools for CHF prediction in the subcooled and saturated flow boiling regions, respectively. In order to avoid the defect of predictive discontinuities often encountered when applying previous correlations, a simple, nondimensional, inlet conditions dependent CHF correlation for saturated flow boiling has been formulated. Its functional form is determined by application of the artificial neural network and parametric trend analyses to the collected database. Superiority of this new correlation has been verified by the collected database. It has a mean deviation of 16.8% for this collected databank, smallest among all tested correlations. Compared to many inordinately complex correlations, this new correlation consists only of one single equation. (author)

  16. 1995 national heat transfer conference: Proceedings. Volume 12: Falling films; Fundamentals of subcooled flow boiling; Compact heat exchanger technology for the process industry; HTD-Volume 314

    International Nuclear Information System (INIS)

    Sernas, V.; Boyd, R.D.; Jensen, M.K.

    1995-01-01

    The papers in the first section cover falling films and heat transfer. Papers in the second section address issues associated with heat exchangers, such as: plate-and-frame heat exchanger technology; thermal design issues; condensation; and single-phase flows. The papers in the third section deal with studies related to: the turbulent velocity field in a vertical annulus; the effects of curvature and a dissolved noncondensable gas on nucleate boiling heat transfer; the effects of flow obstruction on the onset of a Ledinegg-type flow instability; pool boiling from a large-diameter tube; and two-dimensional wall temperature distributions and convection in a single-sided heated vertical tube. Separate abstracts were prepared for most papers in this volume

  17. Acceleration of two-phase flow by boiling, 1

    International Nuclear Information System (INIS)

    Hara, Toshitsugu; Uchida, Motokazu; Mitani, Akio; Mori, Yasuo; Hijikata, Kunio.

    1975-01-01

    This paper reports on the experimental results concerning the acceleration mechanism of the liquid used for liquid metal magnetohydrodynamic power generation. The experiment simulated two-component flow by injecting low boiling point liquid (R113) which is not soluble in main high temperature flow (hot water). From the boiling of this two component flow, the relations among the acceleration performance of the liquid, the number and frequency of bubbles generated from liquid drops, and the growth velocity of the bubbles have been investigated. All the injected liquid drops did not necessarily boil even if they were heated above the saturation temperature. The probability of boiling of the liquid drops becomes larger as the temperature difference between two liquids becomes larger. The bubble generation frequency distributed around the mean elapsed time of the liquid drops. The larger temperature difference between two liquids presents sharper distribution. The radius of bubbles increased proportionally to the two-thirds power of the elapsed time and also to two-thirds power of the temperature difference. The liquid acceleration performance by bubbles increased as the bubble generation frequency distribution becomes sharpe. (Tai, I.)

  18. CFD simulation of subcooled flow boiling at low pressure

    International Nuclear Information System (INIS)

    Koncar, B.; Mavko, B.

    2001-01-01

    An increased interest to numerically simulate the subcooled flow boiling at low pressures (1 to 10 bar) has been aroused in recent years, pursued by the need to perform safety analyses of research nuclear reactors and to investigate the sump cooling concept for future light water reactors. In this paper the subcooled flow boiling has been simulated with a multidimensional two-fluid model used in a CFX-4.3 computational fluid dynamics (CFD) code. The existing model was adequately modified for low pressure conditions. It was shown that interfacial forces, which are usually used for adiabatic flows, need to be modeled to simulate subcooled boiling at low pressure conditions. Simulation results are compared against published experimental data [1] and agree well with experiments.(author)

  19. Predictions of void fraction in convective subcooled boiling channels using a one-dimensional two-fluid model

    International Nuclear Information System (INIS)

    Hu, Lin-Wen; Pan, Chin

    1995-01-01

    Subcooled nucleate boiling under forced convective conditions is of considerable interest for many disciplines, such as nuclear reactor technology and other energy conversion systems, due to its high heat transfer capability. For such applications, the liquid entering the heating channel is usually in a subcooled state and nucleate boiling is initiated at some distance from the entrance. Further downstream from the boiling incipient point, the bubbles may depart from the heating wall. The point of first bubble departure is called the net vapor generation (NVG) point, because after this point, significant void is present in the subcooled liquid and the void fraction rises very rapidly even though the bulk liquid may still be in a highly subcooled state. The presence of vapor bubbles, which are at a temperature near the saturation temperature, in a subcooled liquid shows the existence of thermal nonequilibrium, which complicates the analysis of this boiling regime. 13 refs., 4 figs

  20. Bubble and boundary layer behaviour in subcooled flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Maurus, Reinhold; Sattelmayer, Thomas [Lehrstuhl fuer Thermodynamik, Technische Universitaet Muenchen, 85747 Garching (Germany)

    2006-03-15

    Subcooled flow boiling is a commonly applied technique for achieving efficient heat transfer. In the study, an experimental investigation in the nucleate boiling regime was performed for water circulating in a closed loop at atmospheric pressure. The horizontal orientated test-section consists of a rectangular channel with a one side heated copper strip and good optical access. Various optical observation techniques were applied to study the bubble behaviour and the characteristics of the fluid phase. The bubble behaviour was recorded by the high-speed cinematography and by a digital high resolution camera. Automated image processing and analysis algorithms developed by the authors were applied for a wide range of mass flow rates and heat fluxes in order to extract characteristic length and time scales of the bubbly layer during the boiling process. Using this methodology, the bubbles were automatically analysed and the bubble size, bubble lifetime, waiting time between two cycles were evaluated. Due to the huge number of observed bubbles a statistical analysis was performed and distribution functions were derived. Using a two-dimensional cross-correlation algorithm, the averaged axial phase boundary velocity profile could be extracted. In addition, the fluid phase velocity profile was characterised by means of the particle image velocimetry (PIV) for the single phase flow as well as under subcooled flow boiling conditions. The results indicate that the bubbles increase the flow resistance. The impact on the flow exceeds by far the bubbly region and it depends on the magnitude of the boiling activity. Finally, the ratio of the averaged phase boundary velocity and of the averaged fluid velocity was evaluated for the bubbly region. (authors)

  1. Flow Boiling and Condensation Experiment (FBCE) for the International Space Station

    Science.gov (United States)

    Mudawar, Issam; O'Neill, Lucas; Hasan, Mohammad; Nahra, Henry; Hall, Nancy; Balasubramaniam, R.; Mackey, Jeffrey

    2016-01-01

    An effective means to reducing the size and weight of future space vehicles is to replace present mostly single-phase thermal management systems with two-phase counterparts. By capitalizing upon both latent and sensible heat of the coolant rather than sensible heat alone, two-phase thermal management systems can yield orders of magnitude enhancement in flow boiling and condensation heat transfer coefficients. Because the understanding of the influence of microgravity on two-phase flow and heat transfer is quite limited, there is an urgent need for a new experimental microgravity facility to enable investigators to perform long-duration flow boiling and condensation experiments in pursuit of reliable databases, correlations and models. This presentation will discuss recent progress in the development of the Flow Boiling and Condensation Experiment (FBCE) for the International Space Station (ISS) in collaboration between Purdue University and NASA Glenn Research Center. Emphasis will be placed on the design of the flow boiling module and on new flow boiling data that were measured in parabolic flight, along with extensive flow visualization of interfacial features at heat fluxes up to critical heat flux (CHF). Also discussed a theoretical model that will be shown to predict CHF with high accuracy.

  2. Changes of enthalpy slope in subcooled flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Collado, Francisco J.; Monne, Carlos [Universidad de Zaragoza-CPS, Departamento de Ingenieria Mecanica-Motores Termicos, Zaragoza (Spain); Pascau, Antonio [Universidad de Zaragoza-CPS, Departamento de Ciencia de los Materiales y Fluidos-Mecanica de Fluidos, Zaragoza (Spain)

    2006-03-01

    Void fraction data in subcooled flow boiling of water at low pressure measured by General Electric in the 1960s are analyzed following the classical model of Griffith et al. (in Proceedings of ASME-AIChE heat transfer conference, 58-HT-19, 1958). In addition, a new proposal for analyzing one-dimensional steady flow boiling is used. This is based on the physical fact that if the two phases have different velocities, they cannot cover the same distance - the control volume length - in the same time. So a slight modification of the heat balance is suggested, i.e., the explicit inclusion of the vapor-liquid velocity ratio or slip ratio as scaling time factor between the phases, which is successfully checked against the data. Finally, the prediction of void fraction using correlations of the net rate of change of vapor enthalpy in the fully developed regime of subcooled flow boiling is explored. (orig.)

  3. Burnout in a high heat-flux boiling system with an impinging jet

    International Nuclear Information System (INIS)

    Monde, M.; Katto, Y.

    1978-01-01

    An experimental study has been made on the fully-developed nucleate boiling at atmospheric pressure in a simple forced-convection boiling system, which consists of a heated flat surface and a small, high-speed jet of water or of freon-113 impinging on the heated surface. A generalized correlation for burnout heat flux data, that is applied to either water or freon-113 is successfully evolved, and it is shown that surface tension has an important role for the onset of burnout phenomenon, not only in the ordinary pool boiling, but also in the present boiling system with a forced flow. (author)

  4. Advanced Wall Boiling Model with Wide Range Applicability for the Subcooled Boiling Flow and its Application into the CFD Code

    International Nuclear Information System (INIS)

    Yun, B. J.; Song, C. H.; Splawski, A.; Lo, S.

    2010-01-01

    Subcooled boiling is one of the crucial phenomena for the design, operation and safety analysis of a nuclear power plant. It occurs due to the thermally nonequilibrium state in the two-phase heat transfer system. Many complicated phenomena such as a bubble generation, a bubble departure, a bubble growth, and a bubble condensation are created by this thermally nonequilibrium condition in the subcooled boiling flow. However, it has been revealed that most of the existing best estimate safety analysis codes have a weakness in the prediction of the subcooled boiling phenomena in which multi-dimensional flow behavior is dominant. In recent years, many investigators are trying to apply CFD (Computational Fluid Dynamics) codes for an accurate prediction of the subcooled boiling flow. In the CFD codes, evaporation heat flux from heated wall is one of the key parameters to be modeled for an accurate prediction of the subcooled boiling flow. The evaporate heat flux for the CFD codes is expressed typically as follows, q' e = πD 3 d /6 ρ g h fg fN' where, D d , f ,N' are bubble departure size, bubble departure frequency and active nucleation site density, respectively. In the most of the commercial CFD codes, Tolubinsky bubble departure size model, Kurul and Podowski active nucleation site density model and Ceumem-Lindenstjerna bubble departure frequency model are adopted as a basic wall boiling model. However, these models do not consider their dependency on the flow, pressure and fluid type. In this paper, an advanced wall boiling model was proposed in order to improve subcooled boiling model for the CFD codes

  5. Bubble behaviour and mean diameter in subcooled flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Zeitoun, O.; Shoukri, M. [McMaster Univ., Hamilton, Ontario (Canada)

    1995-09-01

    Bubble behaviour and mean bubble diameter in subcooled upward flow boiling in a vertical annular channel were investigated under low pressure and mass flux conditions. A high speed video system was used to visualize the subcooled flow boiling phenomenon. The high speed photographic results indicated that, contrary to the common understanding, bubbles tend to detach from the heating surface upstream of the net vapour generation point. Digital image processing technique was used to measure the mean bubble diameter along the subcooled flow boiling region. Data on the axial area-averaged void fraction distributions were also obtained using a single beam gamma densitometer. Effects of the liquid subcooling, applied heat flux and mass flux on the mean bubble size were investigated. A correlation for the mean bubble diameter as a function of the local subcooling, heat flux and mass flux was obtained.

  6. Void fraction and incipient point of boiling during the subcooled nucleate flow boiling of water

    International Nuclear Information System (INIS)

    Unal, H.C.

    1977-01-01

    Void fraction has been determined with high-speed photography for subcooled nucleate flow boiling of water. The data obtained and the data of various investigators for adiabatic flow of stream-water mixtures and saturated bulk boiling of water have yielded a correlation which covers the following conditions: geometry: vertically orientated circular tubes, rectangular channels and annuli; pressure: 2 to 15.9 MN/m 2 ; mass velocity: 388 to 3500 kg/m 2 s; void fraction: 0 to 99%; hydraulic diameter: 0.0047 to 0.0343 m; heat flux: adiabatic and 0.01 to 2.0 MW/m 2 . The accuracy of the correlation is estimated to be 12.5%. The value of the so-called distribution (or flow) parameter has been experimentally determined and found to be equal to 1 for a vertical small-diameter circular tube. The incipient point of boiling for subcooled nucleate flow boiling of water has been determined with high-speed photography. The data obtained and the data available in the literature have yielded a correlation which covers the following conditions: geometry: plate, circular tube and inner tube-heated, outer tube-heated and inner - and outer tube heated annulus; pressure: 0.15 to 15.9 MN/m 2 ; mass velocity: 470 to 17355 kg/m 2 s; hydraulic diameter: 0.00239 to 0.032 m; heat flux: 0.13 to 9.8 MW/m 2 ; subcooling: 2.6 to 108 K; material of heating surface: stainless steel and nickel. The accuracy of the correlation is estimated to be 27.5%. Maximum bubble diameters have been measured at the incipient point of boiling. These data and the data from literature have been correlated for the pressure range of 0.1 to 15.9 MN/m 2 . (author)

  7. Visualization of Two Phase Natural Convection Flow in a Vertical Pipe using the Sulfuric Acid - Copper Sulfate Electroplating System

    Energy Technology Data Exchange (ETDEWEB)

    Ohk, Seung-Min; Chae, Myeong-Seon; Chung, Bum-Jin [Kyung Hee Univ., Yongin (Korea, Republic of)

    2014-10-15

    The passive containment cooling system (PCCS) driven by natural forces convection gain draws research interests after Fukushima NPP accident. The PCCS was classified into three categories: Containment pressure suppression, Containment passive heat removal/pressure suppression systems and Passive containment spray. Among the types of containment passive heat removal/pressure suppression systems, the system composed of an internal heat exchanger and an external coolant tank is considered. In a severe accident condition, the heat from the containment atmosphere is transferred to the outer surface of the heat exchanger by the convection and condensation of the mixture of steam and gases. On the other hand, the heat is transferred to external pool by single phase or two phase natural convection inside of heat exchanger pipes. The study aimed at investigating the influence of the diameter (D) and height (H) of the heat exchanger pipes on the single phase and two phase natural convection heat transfer. As the initial stage of the study, the two phase natural convection flow inside a vertical pipe is visualized. In order to achieve the aim with ample test rig, a sulfuric acid - cooper sulfate electroplating system was employed based on the analogy between heat and mass transfer. The reduction of hydrogen ion at the cathode surface at high potential was used to simulate the boiling phenomena. This study tried to visualize the boiling heat transfer inside a vertical pipe using a cupric acid-copper sulfate (H{sub 2}SO{sub 4}-CuSO{sub 4}) electroplating system. This seems to be successful so far. However further study has to be done to compare the result with real two phase flow situation. The surface tension and surface characteristics are to be tuned to simulate the real situation.

  8. Heat transfer in vapour-liquid flow of carbon dioxide

    International Nuclear Information System (INIS)

    Yagov, V.V.

    2009-01-01

    During the last decade a number of studies of boiling heat transfer in carbon dioxide notably increase. As a field of CO 2 practical using corresponds to high reduced pressures, and a majority of available experimental data on CO 2 flow boiling even in submillimetric channels relate to turbulent liquid flow regimes, a possibility arises to develop sufficiently general method for HTC predicting. Under the above conditions nucleate boiling occurs up to rather high flow quality, even in annular flow regime due to extremely small size of an equilibrium vapour bubble. This conclusion is in agreement with the available experimental data. The predicting equation for nucleate boiling heat transfer developed by the present author in 1988 is valid for any nonmetallic liquid. A contribution of forced convection in heat transfer is calculated according to the Petukhov et al. equation with correction factor, which accounted for an effect of velocity increase due to evaporation. This effect can be essential at relatively small heat fluxes and rather high mass flow rates. The Reynolds analogy and homogeneous model are used in order to account for the convective heat transfer augmentation in two-phase flow. Due to low ratio of liquid and vapour densities at high reduced pressures the homogeneous approximation of two-phase flow seems to be warranted. A total heat transfer coefficient is calculated as an interpolated value of boiling and convective HTCs. The experimental data on CO 2 flow boiling related to regimes before heated wall dryout incipience are in rather good agreement with the calculations. (author)

  9. Subcooled flow boiling heat transfer from microporous surfaces in a small channel

    International Nuclear Information System (INIS)

    Yan, Sun; Li, Zhang; Hong, Xu; Xiaocheng, Zhong

    2011-01-01

    The continuously increasing requirement for high heat transfer rate in a compact space can be met by combining the small channel/microchannel and heat transfer enhancement methods during fluid subcooled flow boiling. In this paper, the sintered microporous coating, as an efficient means of enhancing nucleate boiling, was applied to a horizontal, rectangular small channel. Water flow boiling heat transfer characteristics from the small channel with/without the microporous coating were experimentally investigated. The small channel, even without the coating, presented flow boiling heat transfer enhancement at low vapor quality due to size effects of the channel. This enhancement was also verified by under-predictions from macro-scale correlations. In addition to the enhancement from the channel size, all six microporous coatings with various structural parameters were found to further enhance nucleate boiling significantly. Effects of the coating structural parameters, fluid mass flux and inlet subcooling were also investigated to identify the optimum condition for heat transfer enhancement. Under the optimum condition, the microporous coating could produce the heat transfer coefficients 2.7 times the smooth surface value in subcooled flow boiling and 3 times in saturated flow boiling. The combination of the microporous coating and small channel led to excellent heat transfer performance, and therefore was deemed to have promising application prospects in many areas such as air conditioning, chip cooling, refrigeration systems, and many others involving compact heat exchangers. (authors)

  10. A microgravity boiling and convective condensation experiment

    Science.gov (United States)

    Kachnik, Leo; Lee, Doojeong; Best, Frederick; Faget, Nanette

    1987-12-01

    A boiling and condensing test article consisting of two straight tube boilers, one quartz and one stainless steel, and two 1.5 m long glass-in-glass heat exchangers, on 6 mm ID and one 10 mm ID, was flown on the NASA KC-135 0-G aircraft. Using water as the working fluid, the 5 kw boiler produces two phase mixtures of varying quality for mass flow rates between 0.005 and 0.1 kg/sec. The test section is instrumented at eight locations with absolute and differential pressure transducers and thermocouples. A gamma densitometer is used to measure void fraction, and high speed photography records the flow regimes. A three axis accelerometer provides aircraft acceleration data (+ or - 0.01G). Data are collected via an analog-to-digital conversion and data acquisition system. Bubbly, annular, and slug flow regimes were observed in the test section under microgravity conditions. Flow oscillations were observed for some operating conditions and the effect of the 2-G pullout prior to the 0-G period was observed by continuously recording data throughout the parabolas. A total fo 300 parabolas was flown.

  11. Critical heat flux in subcooled and low quality boiling

    International Nuclear Information System (INIS)

    Maroti, L.

    1976-06-01

    A semi-empirical relationship for critical heat flux prediction in a light water cooled power reactor core is developed. The method of developing this relationship is the extension of the analysis of pool boiling crisis for forced convective boiling. In the calculations the energy conservation equation is used together with additional condition for the crisis. Assuming that in the vicinity of the crisis the heat is transported only by the latent heat of the vapour this condition for the crisis can be characterized by the maximum departure velocity of the vapour. Because only flow boiling crisis associating with bubbling at the heating surface is considered the model could be applied only for low quality boiling crisis. The calculated results are compared to the available experimental ones. (Sz.N.Z.)

  12. Acceleration of a two-phase flow by boiling, (3)

    International Nuclear Information System (INIS)

    Mori, Yasuo; Hijikata, Kunio; Iwata, Shoichiro

    1976-01-01

    Acceleration of two-component, two-phase flow has been studied, and a method using the volume expansion by boiling for accelerating fluid has been investigated. In this study, the phenomena of atomizing and boiling were separated, and the liquid with low boiling point was injected into water at lower than the saturation temperature, and was atomized. Then, this was mixed with high temperature liquid and was boiled. The uniform buffle flow was produced, and the phenomena were observed with a high speed camera. The process of acceleration and the acceleration performance were compared with the results of theoretical analysis described in the second report. The experiment was carried out with liquid R113, and at first, the mechanism of atomizing was studied. The atomizing was caused when the relative velocity between R113 and water was more than 4 m/s irrespective of water velocity. The distribution of the diameter of fine liquid drops was almost normal distribution. When the fine drops of R113 were mixed with the high temperature water, bubbles were produced, and the production rate showed definite dependence on the degree of overheating. The flow of bubbles was uniform. However, some of R113 did not become bubbles. The efficiency of acceleration was 1.0 which was independent of the degree of overheating. A further problem is to reduce the quantity of the liquid which does not boil. (Kato, T.)

  13. Void fraction in horizontal bulk flow boiling at high flow qualities

    International Nuclear Information System (INIS)

    Collado, Fancisco J.; Monne, Carlos; Pascau, Antonio

    2008-01-01

    In this work, a new thermodynamic prediction of the vapor void fraction in bulk flow boiling, which is the core process of many energy conversion systems, is analyzed. The current heat balance is based on the flow quality, which is closely related to the measured void fraction, although some correlation for the vapor-liquid velocity ratio is needed. So here, it is suggested to work with the 'static' or thermodynamic quality, which is directly connected to the void fraction through the densities of the phases. Thus, the relation between heat and the mixture enthalpy (here based on the thermodynamic quality instead of the flow one) should be analyzed in depth. The careful void fraction data taken by Thom during the 'Cambridge project' for horizontal saturated flow boiling with high flow qualities (≤0.8) have been used for this analysis. As main results, first, we have found that the applied heat and the increment of the proposed thermodynamic enthalpy mixture throughout the heated duct do not agree, and for closure, a parameter is needed. Second, it has been checked that this parameter is practically equal to the classic velocity ratio or 'slip' ratio, suggesting that it should be included in a true thermodynamic heat balance. Furthermore, it has been clearly possible to improve the 'Cambridge project' correlations for the 'slip' ratio, here based on inlet pressure and water velocity, and heat flux. The calculated void fractions compare quite well with the measured ones. Finally, the equivalence of the suggested new heat balance with the current one through the 'slip' ratio is addressed. Highlighted is the same new energetic relation for saturated flow boiling that has been recently confirmed by the authors for Knights data, also taken during the 'Cambridge project', which include not only horizontal but also vertical upwards flows with moderate outlet flow quality (≤0.2)

  14. Convection flow study within a horizontal fluid layer under the action of gas flow

    Directory of Open Access Journals (Sweden)

    Kreta Aleksei

    2016-01-01

    Full Text Available Experimental investigation of convective processes within horizontal evaporating liquid layer under shear–stress of gas flow is presented. It is found the structures of the convection, which move in opposite direction relative to each other. First convective structure moves in reverse direction with the flow of gas, and the second convective structure moves towards the gas flow. Convection flow within the liquid layer is registered with help of PIV technique. Average evaporation flow rate of Ethanol liquid layer under Air gas flow is measured. Influence of the gas velocity, at a constant temperature of 20 °C, on the evaporation flow rate has been studied.

  15. An improved mechanistic critical heat flux model for subcooled flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Young Min [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    Based on the bubble coalescence adjacent to the heated wall as a flow structure for CHF condition, Chang and Lee developed a mechanistic critical heat flux (CHF) model for subcooled flow boiling. In this paper, improvements of Chang-Lee model are implemented with more solid theoretical bases for subcooled and low-quality flow boiling in tubes. Nedderman-Shearer`s equations for the skin friction factor and universal velocity profile models are employed. Slip effect of movable bubbly layer is implemented to improve the predictability of low mass flow. Also, mechanistic subcooled flow boiling model is used to predict the flow quality and void fraction. The performance of the present model is verified using the KAIST CHF database of water in uniformly heated tubes. It is found that the present model can give a satisfactory agreement with experimental data within less than 9% RMS error. 9 refs., 5 figs. (Author)

  16. An improved mechanistic critical heat flux model for subcooled flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Young Min [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of); Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1997-12-31

    Based on the bubble coalescence adjacent to the heated wall as a flow structure for CHF condition, Chang and Lee developed a mechanistic critical heat flux (CHF) model for subcooled flow boiling. In this paper, improvements of Chang-Lee model are implemented with more solid theoretical bases for subcooled and low-quality flow boiling in tubes. Nedderman-Shearer`s equations for the skin friction factor and universal velocity profile models are employed. Slip effect of movable bubbly layer is implemented to improve the predictability of low mass flow. Also, mechanistic subcooled flow boiling model is used to predict the flow quality and void fraction. The performance of the present model is verified using the KAIST CHF database of water in uniformly heated tubes. It is found that the present model can give a satisfactory agreement with experimental data within less than 9% RMS error. 9 refs., 5 figs. (Author)

  17. Flow boiling heat transfer of carbon dioxide inside a small-sized microfin tube

    Energy Technology Data Exchange (ETDEWEB)

    Dang, Chaobin; Haraguchi, Nobori; Hihara, Eiji [Department of Human and Engineered Environmental Studies, Graduate School of Frontier Sciences, The University of Tokyo, Kashiwanoha, Kashiwa-shi, Chiba 277-8563 (Japan)

    2010-06-15

    This study investigated the flow boiling heat transfer of carbon dioxide inside a small-sized microfin tube (mean inner diameter: 2.0 mm; helix angle: 6.3 ) at a saturation temperature of 15 C, and heat and mass flux ranges of 4.5-18 kW m{sup -2} and 360-720 kg m{sup -2} s{sup -1}, respectively. Although, experimental results indicated that heat flux has a significant effect on the heat transfer coefficient, the coefficient does not always increase with mass flux, as in the case of conventional refrigerants such as HFCs or HCFCs. Under certain conditions, the heat transfer coefficient at a high mass flux was lower than that at a lower mass flux, indicating that convective heat transfer had a suppression effect on nucleate boiling. The heat transfer coefficients in the microfin tubes were 1.9{proportional_to}2.3 times the values in smooth tubes of the same diameter under the same experimental conditions, and the dryout quality was much higher, ranging from 0.9 to 0.95. The experimental results indicated that using microfin tubes may considerably increase the overall heat transfer performance. (author)

  18. Prediction of heat transfer coefficients for forced convective boiling of N2-hydrocarbon mixtures at cryogenic conditions using artificial neural networks

    Science.gov (United States)

    Barroso-Maldonado, J. M.; Belman-Flores, J. M.; Ledesma, S.; Aceves, S. M.

    2018-06-01

    A key problem faced in the design of heat exchangers, especially for cryogenic applications, is the determination of convective heat transfer coefficients in two-phase flow such as condensation and boiling of non-azeotropic refrigerant mixtures. This paper proposes and evaluates three models for estimating the convective coefficient during boiling. These models are developed using computational intelligence techniques. The performance of the proposed models is evaluated using the mean relative error (mre), and compared to two existing models: the modified Granryd's correlation and the Silver-Bell-Ghaly method. The three proposed models are distinguished by their architecture. The first is based on directly measured parameters (DMP-ANN), the second is based on equivalent Reynolds and Prandtl numbers (eq-ANN), and the third on effective Reynolds and Prandtl numbers (eff-ANN). The results demonstrate that the proposed artificial neural network (ANN)-based approaches greatly outperform available methodologies. While Granryd's correlation predicts experimental data within a mean relative error mre = 44% and the S-B-G method produces mre = 42%, DMP-ANN has mre = 7.4% and eff-ANN has mre = 3.9%. Considering that eff-ANN has the lowest mean relative error (one tenth of previously available methodologies) and the broadest range of applicability, it is recommended for future calculations. Implementation is straightforward within a variety of platforms and the matrices with the ANN weights are given in the appendix for efficient programming.

  19. Burnout experiment in subcooled forced-convection boiling of water for beam dumps of a high power neutral beam injector

    International Nuclear Information System (INIS)

    Horiike, Hiroshi; Kuriyama, Masaaki; Morita, Hiroaki

    1982-01-01

    Experimental studies were made on burnout heat flux in highly subcooled forced-convection boiling of water for the design of beam dumps of a high power neutral beam injector for Japan Atomic Energy Research Institute Tokamak-60. These dumps are composed of many circular tubes with two longitudinal fins. The tube was irradiated with nonuniformly distributed hydrogen ion beams of 120 to 200 kW for as long as 10 s. The coolant water was circulated at flow velocities of 3 to 7.5 m/s at exit pressures of 0.4 to 0.9 MPa. The burnout and film-boiling data were obtained at local heat fluxes of 8 to 15 MW/m 2 . These values were as high as 2.5 times larger than those for the circumferentially uniform heat flux case with the same parameters. These data showed insensitivity to local subcooling as well as to pressure, and simple burnout correlations were derived. From these results, the beam dumps have been designed to receive energetic beam fluxes of as high as 5 MW/m 2 with a margin of a factor of 2 for burnout

  20. Heat transfer and pressure drop in flow boiling in microchannels

    CERN Document Server

    Saha, Sujoy Kumar

    2016-01-01

    This Brief addresses the phenomena of heat transfer and pressure drop in flow boiling in micro channels occurring in high heat flux electronic cooling. A companion edition in the Springer Brief Subseries on Thermal Engineering and Applied Science to “Critical Heat Flux in Flow Boiling in Micro channels,” by the same author team, this volume is idea for professionals, researchers and graduate students concerned with electronic cooling.

  1. Measurements of local two-phase flow parameters in a boiling flow channel

    International Nuclear Information System (INIS)

    Yun, Byong Jo; Park, Goon-CherI; Chung, Moon Ki; Song, Chul Hwa

    1998-01-01

    Local two-phase flow parameters were measured lo investigate the internal flow structures of steam-water boiling flow in an annulus channel. Two kinds of measuring methods for local two-phase flow parameters were investigated. These are a two-conductivity probe for local vapor parameters and a Pitot cube for local liquid parameters. Using these probes, the local distribution of phasic velocities, interfacial area concentration (IAC) and void fraction is measured. In this study, the maximum local void fraction in subcooled boiling condition is observed around the heating rod and the local void fraction is smoothly decreased from the surface of a heating rod to the channel center without any wall void peaking, which was observed in air-water experiments. The distributions of local IAC and bubble frequency coincide with those of local void fraction for a given area-averaged void fraction. (author)

  2. Mechanisms and predictions for subcooled flow boiling CHF

    International Nuclear Information System (INIS)

    Liu, Wei; Nariai, Hideki; Inasaka, Fujio

    2000-01-01

    Corresponding to the two kinds of flow pattern reported in literature for subcooled flow boiling, two kinds of CHF triggering mechanism are considered existing with working in different working scope. On the base of a criterion proposed recently by the present authors, subcooled flow boiling data firstly are categorized into two groups by judging whether the first kind or the second kind of flow pattern is established. Possible CHF triggering mechanisms and prediction methods for the two kinds of flow pattern condition are discussed. By considering both the flow pattern development and CHF triggering mechanism, a detailed data categorization is carried out. The corresponding CHF occurrence properties in different data groups are summarized. Parametric trends are reviewed for the first and second kind of data group working condition respectively. Mass flux, pressure, inlet subcooling and inner diameter show almost same effects in the two different working conditions, while the ratio of heated length to diameter's effects on CHF show to be different. Research for the L/D effect on the CHF transverse the interface of the different data groups is carried out. (author)

  3. Subchannel analysis program for boiling water reactor fuel bundles based on five conservation equations of two-phase flow

    International Nuclear Information System (INIS)

    Bessho, Y.; Uchikawa, S.

    1985-01-01

    A subchannel analysis program, MENUETT, is developed for evaluation of thermal-hydraulic characteristics in boiling water reactor fuel bundles. This program is based on five conservation equations of two-phase flow with the drift-flux correlation. The cross flows are calculated separately for liquid and vapor phases from the lateral momentum conservation equation. The effects of turbulent mixing and void drift are accounted for in the program. The conservation equations are implicitly differentiated with the convective terms by the donor-cell method, and are solved iteratively in the axial and lateral directions. Data of the 3 X 3 rod bundle experiments are used for program verification. The lateral distributions of equilibrium quality and mass flow rate at the bundle exit calculated by the program compare satisfactorily with the experimental results

  4. Boiling of subcooled water in forced convection

    International Nuclear Information System (INIS)

    Ricque, R.; Siboul, R.

    1970-01-01

    As a part of a research about water cooled high magnetic field coils, an experimental study of heat transfer and pressure drop is made with the following conditions: local boiling in tubes of small diameters (2 and 4 mm), high heat fluxes (about 1000 W/cm 2 ), high coolant velocities (up to 25 meters/s), low outlet absolute pressures (below a few atmospheres). Wall temperatures are determined with a good accuracy, because very thin tubes are used and heat losses are prevented. Two regimes of boiling are observed: the establishment regime and the established boiling regime and the inception of each regime is correlated. Important delays on boiling inception are also observed. The pressure drop is measured; provided the axial temperature distribution of the fluid and the axial distributions of the wall temperatures, in other words the axial distribution of the heat transfer coefficients under boiling and non boiling conditions, at the same heat flux or the same wall temperatures, are taken in account, then total pressure drop can be correlated, but probably under certain limits of void fraction only. Using the same parameters, it seems possible to correlate the experimental values on critical heat flux obtained previously, which show very important effect of length and hydraulic diameter of the test sections. (authors) [fr

  5. Evaluation of correlations of flow boiling heat transfer of R22 in horizontal channels.

    Science.gov (United States)

    Zhou, Zhanru; Fang, Xiande; Li, Dingkun

    2013-01-01

    The calculation of two-phase flow boiling heat transfer of R22 in channels is required in a variety of applications, such as chemical process cooling systems, refrigeration, and air conditioning. A number of correlations for flow boiling heat transfer in channels have been proposed. This work evaluates the existing correlations for flow boiling heat transfer coefficient with 1669 experimental data points of flow boiling heat transfer of R22 collected from 18 published papers. The top two correlations for R22 are those of Liu and Winterton (1991) and Fang (2013), with the mean absolute deviation of 32.7% and 32.8%, respectively. More studies should be carried out to develop better ones. Effects of channel dimension and vapor quality on heat transfer are analyzed, and the results provide valuable information for further research in the correlation of two-phase flow boiling heat transfer of R22 in channels.

  6. Low-Flow Film Boiling Heat Transfer on Vertical Surfaces

    DEFF Research Database (Denmark)

    Munthe Andersen, J. G.; Dix, G. E.; Leonard, J. E.

    1976-01-01

    The phenomenon of film boiling heat transfer for high wall temperatures has been investigated. Based on the assumption of laminar flow for the film, the continuity, momentum, and energy equations for the vapor film are solved and a Bromley-type analytical expression for the heat transfer...... length, an average film boiling heat transfer coefficient is obtained....

  7. Visualization and void fraction measurement of decompressed boiling flow in a capillary tube

    International Nuclear Information System (INIS)

    Asano, H.; Murakawa, H.; Takenaka, N.; Takiguchi, K.; Okamoto, M.; Tsuchiya, T.; Kitaide, Y.; Maruyama, N.

    2011-01-01

    A capillary tube is often used as a throttle for a refrigerating cycle. Subcooled refrigerant usually flows from a condenser into the capillary tube. Then, the refrigerant is decompressed along the capillary tube. When the static pressure falls below the saturation pressure for the liquid temperature, spontaneous boiling occurs. A vapor-liquid two-phase mixture is discharged from the tube. In designing a capillary tube, it is necessary to calculate the flow rate for given boundary conditions on pressure and temperature at the inlet and exit. Since total pressure loss is dominated by frictional and acceleration losses during two-phase flow, it is first necessary to specify the boiling inception point. However, there will be a delay in boiling inception during decompressed flow. This study aimed to clarify the boiling inception point and two-phase flow characteristics of refrigerant in a capillary tube. Refrigerant flows in a coiled copper capillary tube were visualized by neutron radiography. The one-dimensional distribution of volumetric average void fraction was measured from radiographs through image processing. From the void fraction distribution, the boiling inception point was determined. Moreover, a simplified CT method was successfully applied to a radiograph for cross-sectional measurements. The experimental results show the flow pattern transition from intermittent flow to annular flow that occurred at a void fraction of about 0.45.

  8. Contribution to the development of a Local Predictive Approach of the boiling crisis

    International Nuclear Information System (INIS)

    Montout, M.

    2009-01-01

    EDF aims at developing a 'Local Predictive Approach' of the boiling crisis for PWR core configurations, i.e. an approach resulting in (empirical) critical heat flux predictors based on local parameters provided by NEPTUNE-CFD code (for boiling bubbly flows, only in a first stage). Within this general framework, this PhD work consisted in assess one modelling of NEPTUNE-CFD code selected to simulate boiling bubble flows, then improve it. The latter objective led us to focus on the mechanistic modelling of subcooled nucleate boiling in forced convection. After a literature review, we identified physical improvements to be accounted for, especially with respect to bubble sliding phenomenon along the heated wall. Subsequently, we developed a force balance model in order to provide needed closure laws related to bubble detachment diameter from the nucleation site and lift-off bubble diameter from the wall. A new boiling model including such developments was eventually proposed, and preliminary assessed. (author)

  9. Theoretical modeling of CHF for near-saturated pool boiling and flow boiling from short heaters using the interfacial lift-off criterion

    International Nuclear Information System (INIS)

    Mudawar, I.; Galloway, J.E.; Gersey, C.O.

    1995-01-01

    Pool boiling and flow boiling were examined for near-saturated bulk conditions in order to determine the critical heat flux (CHF) trigger mechanism for each. Photographic studies of the wall region revealed features common to both situations. At fluxes below CHF, the vapor coalesces into a wavy layer which permits wetting only in wetting fronts, the portions of the liquid-vapor interface which contact the wall as a result of the interfacial waviness. Close examination of the interfacial features revealed the waves are generated from the lower edge of the heater in pool boiling and the heater's upstream region in flow boiling. Wavelengths follow predictions based upon the Kelvin-Helmholtz instability criterion. Critical heat flux in both cases occurs when the pressure force exerted upon the interface due to interfacial curvature, which tends to preserve interfacial contact with the wall prior to CHF, is overcome by the momentum of vapor at the site of the first wetting front, causing the interface to lift away from the wall. It is shown this interfacial lift-off criterion facilitates accurate theoretical modeling of CHF in pool boiling and in flow boiling in both straight and curved channels

  10. Theoretical modeling of CHF for near-saturated pool boiling and flow boiling from short heaters using the interfacial lift-off criterion

    Energy Technology Data Exchange (ETDEWEB)

    Mudawar, I.; Galloway, J.E.; Gersey, C.O. [Purdue Univ., West Lafayette, IN (United States)] [and others

    1995-12-31

    Pool boiling and flow boiling were examined for near-saturated bulk conditions in order to determine the critical heat flux (CHF) trigger mechanism for each. Photographic studies of the wall region revealed features common to both situations. At fluxes below CHF, the vapor coalesces into a wavy layer which permits wetting only in wetting fronts, the portions of the liquid-vapor interface which contact the wall as a result of the interfacial waviness. Close examination of the interfacial features revealed the waves are generated from the lower edge of the heater in pool boiling and the heater`s upstream region in flow boiling. Wavelengths follow predictions based upon the Kelvin-Helmholtz instability criterion. Critical heat flux in both cases occurs when the pressure force exerted upon the interface due to interfacial curvature, which tends to preserve interfacial contact with the wall prior to CHF, is overcome by the momentum of vapor at the site of the first wetting front, causing the interface to lift away from the wall. It is shown this interfacial lift-off criterion facilitates accurate theoretical modeling of CHF in pool boiling and in flow boiling in both straight and curved channels.

  11. Theoretical analysis and experimental research on dispersed-flow boiling heat transfer

    International Nuclear Information System (INIS)

    Yu Zhenwan; Jia Dounan; Li Linjiao; Mu Quanhou

    1989-01-01

    Experiment on dispersed-flow boiling heat transfer at low pressure has been done. The hot patch technique has been used to establish post-dryout conditions. The position of the hot patch can be varied along the test section. The superheated vapor temperatures at different elevations after dryout point are obtained. The experimental data are generally in agreement with the models of predictions of existing nonequilibrium film boiling. A heat transfer model for dispersed-flow boiling heat transfer has been developed. And the model can explain the phenomena of heat transfer near the dryout point. (orig./DG)

  12. Gravity Effects in Microgap Flow Boiling

    Science.gov (United States)

    Robinson, Franklin; Bar-Cohen, Avram

    2017-01-01

    Increasing integration density of electronic components has exacerbated the thermal management challenges facing electronic system developers. The high power, heat flux, and volumetric heat generation of emerging devices are driving the transition from remote cooling, which relies on conduction and spreading, to embedded cooling, which facilitates direct contact between the heat-generating device and coolant flow. Microgap coolers employ the forced flow of dielectric fluids undergoing phase change in a heated channel between devices. While two phase microcoolers are used routinely in ground-based systems, the lack of acceptable models and correlations for microgravity operation has limited their use for spacecraft thermal management. Previous research has revealed that gravitational acceleration plays a diminishing role as the channel diameter shrinks, but there is considerable variation among the proposed gravity-insensitive channel dimensions and minimal research on rectangular ducts. Reliable criteria for achieving gravity-insensitive flow boiling performance would enable spaceflight systems to exploit this powerful thermal management technique and reduce development time and costs through reliance on ground-based testing. In the present effort, the authors have studied the effect of evaporator orientation on flow boiling performance of HFE7100 in a 218 m tall by 13.0 mm wide microgap cooler. Similar heat transfer coefficients and critical heat flux were achieved across five evaporator orientations, indicating that the effect of gravity was negligible.

  13. Numerical simulation of falling film flow boiling along a vertical wall

    International Nuclear Information System (INIS)

    Chiaki Kino; Tomoaki Kunugi; Akimi Serizawa

    2005-01-01

    Full text of publication follows: When a dryout occurs in film flows with heating from the wall, the wall surface being cooled is no longer in intimate contact with the liquid film. Consequently, the heat transfer will dramatically reduce and the corresponding wall temperature will rise rapidly up to the melting temperature of the heat transfer plate or pipe. It is very important to investigate the heat transfer characteristics of liquid films flowing along a heating wall and the dryout phenomena of the liquid films associated with increasing heat flux in the high heat flux component devices for chemical and mechanical devices and nuclear reactor systems. Many studies have been conducted on the dryout phenomena and it has been shown that the dryout conditions are influenced by several different flow conditions, for instance, subcooled and saturated liquid films and so on. The dryout process of boiling liquid films is different between them: in the case of subcooled liquid films, the process is caused by the local surface-tension variation along the film. On the contrary, in the case of saturated liquid films the surface temperature of boiling films is maintained at a saturation temperature and there can be no variation of surface tension along the film. The process in the case of saturated liquid films is caused by the reduction of film flow rate due to the flow imbalance. This reduction of film flow rate is promoted by the evaporation and the liquid droplets arising from the film surface due to the burst of vapor bubbles. Therefore, it is very important to predict the sputtering rate of liquid droplets and to understand the behavior of vapor bubbles in film flow boiling. In the present study, numerical simulations based on the MARS (Multi-interface Advection and Reconstruction Solver) developed by one of the authors have been performed in order to understand the dryout of film flow boiling. The film flows along a vertical wall are focused in the present study

  14. Void fraction in horizontal bulk flow boiling at high flow qualities

    Energy Technology Data Exchange (ETDEWEB)

    Collado, Fancisco J.; Monne, Carlos [Dpto. de Ingenieria Mecanica, Universidad de Zaragoza-CPS, Maria de Luna 3, 50018-Zaragoza (Spain); Pascau, Antonio [Dpto. de Ciencia de los Materiales y Fluidos, Universidad de Zaragoza-CPS, Maria de Luna 3, 50018-Zaragoza (Spain)

    2008-04-15

    In this work, a new thermodynamic prediction of the vapor void fraction in bulk flow boiling, which is the core process of many energy conversion systems, is analyzed. The current heat balance is based on the flow quality, which is closely related to the measured void fraction, although some correlation for the vapor-liquid velocity ratio is needed. So here, it is suggested to work with the 'static' or thermodynamic quality, which is directly connected to the void fraction through the densities of the phases. Thus, the relation between heat and the mixture enthalpy (here based on the thermodynamic quality instead of the flow one) should be analyzed in depth. The careful void fraction data taken by Thom during the 'Cambridge project' for horizontal saturated flow boiling with high flow qualities ({<=}0.8) have been used for this analysis. As main results, first, we have found that the applied heat and the increment of the proposed thermodynamic enthalpy mixture throughout the heated duct do not agree, and for closure, a parameter is needed. Second, it has been checked that this parameter is practically equal to the classic velocity ratio or 'slip' ratio, suggesting that it should be included in a true thermodynamic heat balance. Furthermore, it has been clearly possible to improve the 'Cambridge project' correlations for the 'slip' ratio, here based on inlet pressure and water velocity, and heat flux. The calculated void fractions compare quite well with the measured ones. Finally, the equivalence of the suggested new heat balance with the current one through the 'slip' ratio is addressed. Highlighted is the same new energetic relation for saturated flow boiling that has been recently confirmed by the authors for Knights data, also taken during the 'Cambridge project', which include not only horizontal but also vertical upwards flows with moderate outlet flow quality ({<=}0.2). (author)

  15. A forced convective heat transfer model for two-phase hydrogen systems

    International Nuclear Information System (INIS)

    Pasch, J.; Anghaie, S.

    2007-01-01

    A consistent event in the use of hydrogen in nuclear thermal propulsion is film boiling, in which the wall heat is so large that liquid can not exist at the wall. Instead, vapor interfaces with the wall and liquid flows in the core of the duct. To better understand heat transfer under these conditions, a select set of hydrogen test data from these conditions are analyzed. This paper presents the results of an extensive literature search for film boiling heat transfer models. A representative cross-section of these models is then applied to the data. The heat transfer coefficient data were found difficult to predict and highly dependent upon the flow regime. Pre-critical heat flux correlations completely fail to predict the heat transfer of inverted film boiling conditions. Pool boiling models for inverted film boiling also are inappropriate. Current force convection models for inverted film boiling, while far better than the previous two classes of models, still generate large predictive errors. It is recommended that for the inverted annular film boiling flow regime the modified equilibrium bulk Dittus-Boelter model be used. For agitated inverted annular film boiling and dispersed film boiling regimes associated with positive equilibrium qualities, the Hendricks model should be used. (A.C.)

  16. Interfacial area transport of subcooled boiling flow in a vertical annulus

    Energy Technology Data Exchange (ETDEWEB)

    Brooks, Caleb S.; Ozar, Basar; Hibiki, Takashi; Ishii, Mamoru, E-mail: ishii@purdue.edu

    2014-03-15

    Highlights: • Discussion of boiling and wall nucleation dataset obtained in a vertical annulus. • Overview of the interfacial area transport equation modeling in boiling flow. • Comparison of bubble departure diameter and frequency with existing models. • Evaluation of the interfacial area transport equation prediction in boiling flow. - Abstract: In an effort to improve the prediction of void fraction and heat transfer characteristics in two-phase systems, the two-group interfacial area transport equation has been developed for use with the two-group two-fluid model. The two-group approach treats spherical/distorted bubbles as Group-1 and cap/slug/churn-turbulent bubbles as Group-2. Therefore, the interfacial area transport of steam-water two-phase flow in a vertical annulus has been investigated experimentally, including bulk flow parameters and wall nucleation characteristics. The theoretical modeling of interfacial area transport equation with phase change terms is introduced and discussed along with the experimental results. Benchmark of the interfacial area transport equation is performed considering the effects of bubble interaction mechanisms such as bubble break-up and coalescence, as well as, effects of phase change mechanisms such as wall nucleation and condensation for subcooled boiling. From the benchmark, sensitivity in the constitutive relations for Group-1 phase change mechanisms, such as wall nucleation and condensation is clear. The Group-2 interfacial area transport is shown to be dominated by the interfacial heat transfer mechanism causing expansion of Group-1 bubbles into Group-2 bubbles in the boiling flow.

  17. Saturated flow boiling heat transfer in water-heated vertical annulus

    International Nuclear Information System (INIS)

    Sun Licheng; Yan Changqi; Sun Zhonning

    2005-01-01

    This paper describes the saturated flow boiling heat transfer characteristics of water at 1 atm and low velocities in water-heated vertical annuli with equivalent diameters of 10 mm and 6 mm. Test section is consisted of two concentric circular tubes outer of which is made of quartz, so the whole test courses can be visualized. There are three main flow patterns of bubble flow, churn flow and churn-annular flow in the annuli, most important of which is churn flow. Flooding is the mechanism of churn flow and churn can enhance the heat transport between steam and water; Among the three factors of mass flux, inlet subcooling and annulus width, the last one has great effect on heat transport, moderately decreasing the annulus width can enhance the heat transfer; Combined annular flow model with theory of flooding and turbulent Prandtl Number, the numerical value of heat flux is given, the shape of test boiling curve and that of calculated by model is very alike, but there is large discrepancy between test data and calculated results, the most possible reason is that some parameters given by fluid flooding model are based on experimental data of common circular tubes, but not of annuli. Doing more research on flooding in annulus, particularly narrow annulus, is necessary for calculating the saturated boiling in annulus. (authors)

  18. Introduction of image analysis for the quantification of the boiling flow heat transfer

    NARCIS (Netherlands)

    Ferret, C.; Falk, L.; d'Ortona, U.; Chenu, A.; Veenstra, T.T.

    2004-01-01

    Heat transfer performances for non-boiling and boiling flow of a micro-vaporizer have been measured by standard methods (temperatures, flow rates, effective power input). The study was carried out for laminar flow (Re<25) in silicon micro-channels (5 mm×3 cm×200 μm) filled with ordered obstacles to

  19. Critical heat flux and exit film flow rate in a flow boiling system

    International Nuclear Information System (INIS)

    Ueda, Tatsuhiro; Isayama, Yasushi

    1981-01-01

    The critical heat flux in a flowing boiling system is an important problem in the evaporating tubes with high thermal load such as nuclear reactors and boilers, and gives the practical design limit. When the heat flux in uniformly heated evaporating tubes is gradually raised, the tube exit quality increases, and soon, the critical heat flux condition arises, and the wall temperature near tube exit rises rapidly. In the region of low exit quality, the critical heat flux condition is caused by the transition from nucleating boiling, and in the region of high exit quality, it is caused by dry-out. But the demarcation of both regions is not clear. In this study, for the purpose of obtaining the knowledge concerning the critical heat flux condition in a flowing boiling system, the relation between the critical heat flux and exit liquid film flow rate was examined. For the experiment, a uniformly heated vertical tube supplying R 113 liquid was used, and the measurement in the range of higher heating flux and mass velocity than the experiment by Ueda and Kin was carried out. The experimental setup and experimental method, the critical heat flux and exit quality, the liquid film flow rate at heating zone exit, and the relation between the critical heat flux and the liquid film flow rate at exit are described. (Kako, I.)

  20. Contribution to the boiling curve of sodium

    International Nuclear Information System (INIS)

    Schins, H.E.J.

    1975-01-01

    Sodium in a pool was preheated to saturation temperatures at system pressures of 200, 350 and 500 torr. A test section of normal stainless steel was then extra heated by means of the conical fitting condenser zone of a heat pipe. Measurements were made of heat transfer fluxes, q in W/cm 2 , as a function of wall excess temperature above saturation, THETA = Tsub(w) - Tsub(s) in 0 C, both, in natural convection and in boiling regimes. These measurements make it possible to select the Subbotin natural convection and nucleate boiling curves among other variants proposed in literature. Further it is empirically demonstrated on water that the minimum film boiling point corresponds to the homogeneous nucleation temperature calculated by the Doering formula. Assuming that the minimum film boiling point of sodium can be obtained in the same manner, it is then possible to give an appoximate boiling curve of sodium for the use in thermal interaction studies. At 1 atm the heat transfer fluxes q versus wall temperatures THETA are for a point on the natural convection curve 0.3 W/cm 2 and 2 0 C; for start of boiling 1.6 W/cm 2 and 6 0 C; for peak heat flux 360 W/cm 2 and 37 0 C; for minimum film boiling 30 W/cm 2 and 905 0 C and for a point on the film boiling curve 160 W/cm 2 and 2,000 0 C. (orig.) [de

  1. Turbulent subcooled boiling flow visualization experiments through a rectangular channel

    International Nuclear Information System (INIS)

    Estrada-Perez, Carlos E.; Dominguez-Ontiveros, Elvis E.; Hassan, Yassin A.

    2008-01-01

    Full text of publication follows: Proper characterization of subcooled boiling flow is of importance in many applications. It is of exceptional significance in the development of empirical models for the design of nuclear reactors, steam generators, and refrigeration systems. Most of these models are based on experimental studies that share the characteristics of utilizing point measurement probes with high temporal resolution, e.g. Hot Film Anemometry (HFA), Laser Doppler Velocimetry (LDV), and Fiber Optic Probes (FOP). However there appears to be a scarcity of experimental studies that can capture instantaneous whole-field measurements with a fast time response. Particle Tracking Velocimetry (PTV) may be used to overcome the limitations associated with point measurement techniques. PTV is a whole-flow-field technique providing instantaneous velocity vectors capable of high spatial and temporal resolution. PTV is also an exceptional tool for the analysis of boiling flow due to its ability to differentiate between the gas and liquid phases and subsequently deliver independent velocity fields associated with each phase. In this work, using PTV, liquid velocity fields of a turbulent subcooled boiling flow in a rectangular channel were successfully obtained. The present results agree with similar studies that used point measurement probes. However, the present study provides additional information; not only averaged profiles of the velocity components were obtained, instantaneous 2-D velocity fields were also readily available with a high temporal and spatial resolution. Analysis of fluctuating velocities, Reynolds stresses, and higher order statistics of the flow are presented. This work is an attempt to enrich the database already collected on turbulent subcooled boiling flow, with the hope that it will be useful in turbulence modeling efforts. (authors)

  2. Comparative study of heat transfer and pressure drop during flow boiling and flow condensation in minichannels

    Directory of Open Access Journals (Sweden)

    Mikielewicz Dariusz

    2014-09-01

    Full Text Available In the paper a method developed earlier by authors is applied to calculations of pressure drop and heat transfer coefficient for flow boiling and also flow condensation for some recent data collected from literature for such fluids as R404a, R600a, R290, R32,R134a, R1234yf and other. The modification of interface shear stresses between flow boiling and flow condensation in annular flow structure are considered through incorporation of the so called blowing parameter. The shear stress between vapor phase and liquid phase is generally a function of nonisothermal effects. The mechanism of modification of shear stresses at the vapor-liquid interface has been presented in detail. In case of annular flow it contributes to thickening and thinning of the liquid film, which corresponds to condensation and boiling respectively. There is also a different influence of heat flux on the modification of shear stress in the bubbly flow structure, where it affects bubble nucleation. In that case the effect of applied heat flux is considered. As a result a modified form of the two-phase flow multiplier is obtained, in which the nonadiabatic effect is clearly pronounced.

  3. Flow boiling of refrigerant-oil mixtures; Transferts de chaleur dans un melange constitue de fluide frigorigene et d'huile

    Energy Technology Data Exchange (ETDEWEB)

    Feidt, M

    1999-10-13

    The phase out of chlorine containing refrigerants (CFC and HCFC) has led to the introduction of new refrigerants and lubricants to the market. The interest in using HFC fluids as working fluids to replace fluids harmful to the stratospheric ozone layer. The study presents the influence of synthetic oil (POE ISO 68) on flow boiling of refrigerants R134a (pure fluid) and R410A (R32/R125 50%/50%). Local and average heat transfer coefficients and pressure drops have been measured for a smooth horizontal tube. The distribution of the heat transfer coefficient at the inner wall has been obtained from solving the inverse heat conduction problem (IHCP) and resulted in a local combination of nucleate and convective contributions to flow boiling. Local heat transfer coefficients have been averaged and displayed as a function of the vapour quality. For R134a: small amounts of oil (1% to 6%) in the liquid phase increased the heat transfer coefficient at low and intermediate vapour qualities (less than 0.60) compared to pure fluid. However a hugh reduction of the heat transfer has been observed at higher vapour qualities. For R410A : oil dramatically decreases the heat transfer coefficient compared to pure fluid. Pressure drops are also affected by small amounts of lubricant: an important increase has been noted for both fluids. Available design methods for flow boiling heat transfer coefficient (superposition, enhancement, asymptotic) badly predict the experimental results. Nevertheless a new design method accounting for flow patterns has shown good agreements. The influence of the lubricant on the heat transfer is discussed and a new proposition is made to calculate pressure drops. (author)

  4. Flow regimes and mechanistic modeling of critical heat flux under subcooled flow boiling conditions

    Science.gov (United States)

    Le Corre, Jean-Marie

    Thermal performance of heat flux controlled boiling heat exchangers are usually limited by the Critical Heat Flux (CHF) above which the heat transfer degrades quickly, possibly leading to heater overheating and destruction. In an effort to better understand the phenomena, a literature review of CHF experimental visualizations under subcooled flow boiling conditions was performed and systematically analyzed. Three major types of CHF flow regimes were identified (bubbly, vapor clot and slug flow regime) and a CHF flow regime map was developed, based on a dimensional analysis of the phenomena and available data. It was found that for similar geometric characteristics and pressure, a Weber number (We)/thermodynamic quality (x) map can be used to predict the CHF flow regime. Based on the experimental observations and the review of the available CHF mechanistic models under subcooled flow boiling conditions, hypothetical CHF mechanisms were selected for each CHF flow regime, all based on a concept of wall dry spot overheating, rewetting prevention and subsequent dry spot spreading. It is postulated that a high local wall superheat occurs locally in a dry area of the heated wall, due to a cyclical event inherent to the considered CHF two-phase flow regime, preventing rewetting (Leidenfrost effect). The selected modeling concept has the potential to span the CHF conditions from highly subcooled bubbly flow to early stage of annular flow. A numerical model using a two-dimensional transient thermal analysis of the heater undergoing nucleation was developed to mechanistically predict CHF in the case of a bubbly flow regime. In this type of CHF two-phase flow regime, the high local wall superheat occurs underneath a nucleating bubble at the time of bubble departure. The model simulates the spatial and temporal heater temperature variations during nucleation at the wall, accounting for the stochastic nature of the boiling phenomena. The model has also the potential to evaluate

  5. Burnout Conditions for Flow of Boiling Water in Vertical Rod Clusters

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Kurt M

    1962-05-15

    This paper deals with a new concept for predicting burnout conditions for forced convection of boiling water in fuel elements of nuclear boiling reactors. The concept states the importance of considering the ratio of heated channel perimeter to total channel perimeter. The perimeter ratio concept was arrived at from an experimental study of burnout conditions in rod clusters consisting of three rods of 13 mm outside diameter and 970 mm heated length. Data were obtained for pressures between{sub 2}. 5 and 10 kg/cm, surface heat fluxes between 50 and 120 W/cm, mass flow rates between 0.03 and 0.33 kg/sec and steam qualities between 0.01 and 0.52. The rod distances for the experiment were 2 mm and 6 mm. The diameter of the channel was 41.3 mm. Additional runs were also performed after introducing unheated displacement rods in the channel. The rod distance in this case was 6 mm. In the ranges investigated the measured burnout steam qualities at the outlet of the channel decreases with increasing heat flux and decreasing pressure. Furthermore it has been found that the influence of rod distance is, in the range investigated, of small significance for engineering purposes. It has also been observed that the present burnout steam quality data for the rod clusters are much lower than those earlier obtained for round ducts. This may be explained physically by means of the perimeter ratio concept. It has also been found that the surface shear-stress distribution around the channel perimeter and especially the position of maximum shear-stress is of great importance for predicting burnout conditions for flow in channels. Finally the new method has helped us to understand and interpret experimental results which earlier may have seemed inconsistent.

  6. Burnout Conditions for Flow of Boiling Water in Vertical Rod Clusters

    International Nuclear Information System (INIS)

    Becker, Kurt M.

    1962-05-01

    This paper deals with a new concept for predicting burnout conditions for forced convection of boiling water in fuel elements of nuclear boiling reactors. The concept states the importance of considering the ratio of heated channel perimeter to total channel perimeter. The perimeter ratio concept was arrived at from an experimental study of burnout conditions in rod clusters consisting of three rods of 13 mm outside diameter and 970 mm heated length. Data were obtained for pressures between 2 . 5 and 10 kg/cm, surface heat fluxes between 50 and 120 W/cm, mass flow rates between 0.03 and 0.33 kg/sec and steam qualities between 0.01 and 0.52. The rod distances for the experiment were 2 mm and 6 mm. The diameter of the channel was 41.3 mm. Additional runs were also performed after introducing unheated displacement rods in the channel. The rod distance in this case was 6 mm. In the ranges investigated the measured burnout steam qualities at the outlet of the channel decreases with increasing heat flux and decreasing pressure. Furthermore it has been found that the influence of rod distance is, in the range investigated, of small significance for engineering purposes. It has also been observed that the present burnout steam quality data for the rod clusters are much lower than those earlier obtained for round ducts. This may be explained physically by means of the perimeter ratio concept. It has also been found that the surface shear-stress distribution around the channel perimeter and especially the position of maximum shear-stress is of great importance for predicting burnout conditions for flow in channels. Finally the new method has helped us to understand and interpret experimental results which earlier may have seemed inconsistent

  7. Instabilities in parallel channel of forced-convection boiling upflow system, 5

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Aoki, Shigebumi; Inoue, Akira

    1983-01-01

    The density wave instability in a parallel boiling channel system heated electrically has been studied experimentally and analytically by the authors. In our country, the steam generator for LMFBR has been investigated with Power Reactor and Nuclear Fuel Development Corp. as the central figure for its development, and many results of this instability were reported. Their results were different from our ones as regard to the governing factor of the period of flow oscillation in the unstable region and to the effect of the slip ratio on the stability in analysis. A new linear analytical model is proposed in this paper and the analytical results are compared with ones of two-phase analyses based on the same linear method as this model. Subsequently, the effect of the slip ratio on the stability is studied analytically by this model. The parallel boiling channel system is studied experimentally and analytically, using Freon-113 as test fluid heated by hot water as simulation of the SG for LMFBR. The governing factor of the period of flow oscillation is made clear. (author)

  8. Stretched flow of Carreau nanofluid with convective boundary ...

    Indian Academy of Sciences (India)

    journal of. January 2016 physics pp. 3–17. Stretched flow of Carreau nanofluid with ... fluid over a flat plate subjected to convective surface condition. ... the steady laminar boundary layer flow over a permeable plate with a convective boundary.

  9. Characteristics of liquid and boiling sodium flows in heating pin bundles

    International Nuclear Information System (INIS)

    Menant, Bernard

    1976-01-01

    This study is related to cooling accidents which could occur in sodium cooled fast reactors. Thermo-hydraulic aspects of boiling experiments in pin bundles with helical wire-wrap spacer systems, in the case of undamaged geometries, are analyzed. Differences and analogies in the behavior of multi-rod bundle flows and one-dimensional channel flows are studied. A boiling model is developed for bundle geometries, and predictions obtained with the FLICA code using this models are presented. These predictions are compared with experimental results obtained in a water 19-rod bundle. Then, results of sodium boiling experiments through a 19-rod bundle are interpreted. Both cases of high power and reduced power are envisaged. (author) [fr

  10. Results of a photographic study of subcooled forced-convection boiling of high-pressure water and Freon-12

    International Nuclear Information System (INIS)

    Macbeth, R.V.; Wood, R.W.

    1980-06-01

    The use of a 'Freon' to model high-pressure boiling water has been employed successfully in a number of applications. A prerequisite in modelling is that a well tried and proven basis for the modelling exists. This is not entirely the situation with subcooled boiling however, since past work had tended to concentrate on bulk boiling conditions. Since many of the questions that arise in the design of subcooled boiling systems are concerned with two-phase flow structure, it was decided to place emphasis on attempting to match photographs of subcooled two-phase conditions in high-pressure water (at 55.2 and 82.7 bar) with those of Freon-12 at the corresponding pressures (8.13 and 12.75 bar). A special test-section was constructed giving visual access to a vapour forming region and to an unheated region into which vapour bubbles were drawn by the flow of subcooled liquid. The photographs obtained show that close similarity of two-phase flow structure exists in water and in Freon at corresponding conditions as determined by a previously established modelling procedure. (U.K.)

  11. Do tropical wetland plants possess a convective gas flow mechanism?

    DEFF Research Database (Denmark)

    Jensen, Dennis Konnerup; Sorrell, Brian Keith; Brix, Hans

    2011-01-01

    Internal pressurization and convective gas flow, which can aerate wetland plants more efficiently than diffusion, are common in temperate species. Here, we present the first survey of convective flow in a range of tropical plants. The occurrence of pressurization and convective flow was determined...... in 20 common wetland plants from the Mekong Delta in Vietnam. The diel variation in pressurization in culms and the convective flow and gas composition from stubbles were examined for Eleocharis dulcis, Phragmites vallatoria and Hymenachne acutigluma, and related to light, humidity and air temperature....... Nine of the 20 species studied were able to build up a static pressure of >50Pa, and eight species had convective flow rates higher than 1mlmin-1. There was a clear diel variation, with higher pressures and flows during the day than during the night, when pressures and flows were close to zero...

  12. Flow-Boiling Critical Heat Flux Experiments Performed in Reduced Gravity

    Science.gov (United States)

    Hasan, Mohammad M.; Mudawar, Issam

    2005-01-01

    Poor understanding of flow boiling in microgravity has recently emerged as a key obstacle to the development of many types of power generation and advanced life support systems intended for space exploration. The critical heat flux (CHF) is perhaps the most important thermal design parameter for boiling systems involving both heatflux-controlled devices and intense heat removal. Exceeding the CHF limit can lead to permanent damage, including physical burnout of the heat-dissipating device. The importance of the CHF limit creates an urgent need to develop predictive design tools to ensure both the safe and reliable operation of a two-phase thermal management system under the reduced-gravity (like that on the Moon and Mars) and microgravity environments of space. At present, very limited information is available on flow-boiling heat transfer and the CHF under these conditions.

  13. Boiling on a tube bundle: heat transfer, pressure drop and flow patterns

    International Nuclear Information System (INIS)

    Royen Van, E.

    2011-11-01

    The complexity of two-phase flow boiling on a tube bundle presents many challenges to the understanding of the physical phenomena taking place. It is important to quantify these numerous heat flow mechanisms in order to better describe the performance of tube bundles as a function of the operational conditions. In the present study, the bundle boiling facility at the Laboratory of Heat and Mass Transfer (LTCM) was modified to obtain high-speed videos to characterise the two-phase regimes and some bubble dynamics of the boiling process. It was then used to measure heat transfer on single tubes and in bundle boiling conditions. Pressure drop measurements were also made during adiabatic and diabatic bundle conditions. New enhanced boiling tubes from Wolverine Tube Inc. (Turbo-B5) and the Wieland-Werke AG (Gewa-B5) were investigated using R134a and R236fa as test fluids. The tests were carried out at saturation temperatures T sat of 5 °C and 15 °C, mass flow rates from 4 to 35 kg/m 2 s and heat fluxes from 15 to 70 kW/m 2 , typical of actual operating conditions. The flow pattern investigation was conducted using visual observations from a borescope inserted in the middle of the bundle. Measurements of the light attenuation of a laser beam through the intertube two-phase flow and local pressure fluctuations with piezo-electric pressure transducers were also taken to further help in characterising the complex flow. Pressure drop measurements and data reduction procedures were revised and used to develop new, improved frictional pressure drop prediction methods for adiabatic and diabatic two-phase conditions. The physical phenomena governing the enhanced tube evaporation process and their effects on the performance of tube bundles were investigated and insight gained. A new method based on a theoretical analysis of thin film evaporation was used to propose a new correlating parameter. A large new database of local heat transfer coefficients were obtained and then

  14. Numerical study of the bubbly flow regime in micro-channel flow boiling

    Science.gov (United States)

    Bhuvankar, Pramod; Dabiri, Sadegh

    2017-11-01

    Two-phase flow accompanied by boiling in micro-channel heat sinks is an effective means for heat removal from computer chips. We present a numerical study of flow boiling in micro-channels with conjugate heat transfer with a focus on the bubbly flow regime. The bubbles are assumed to nucleate at a pre-determined location and frequency. The Navier Stokes equations are solved using a single fluid formulation with the Front tracking method. Phase change is implemented using the deficit in heat flux across the bubble interface. The analytical solution for bubble growth in a superheated liquid is used as a benchmark to validate the mentioned numerical method. Water and FC-72 are studied as the operating fluids in a micro-channel made of Copper with a focus on hotspot mitigation. The micro-channel of cross-section 231 μm × 1000 μm , is used to study the effects of vertical up-flow, vertical down-flow and horizontal flow of the mentioned fluids on the heat transfer coefficients. A simple film model accounting for mass and energy conservation is applied wherever the bubble approaches closer than a cell width to the wall. The results of the simulation are compared with existing experimental data for bubble growth rates and heat transfer coefficients.

  15. Flow boiling of refrigerant-oil mixtures; Transferts de chaleur dans un melange constitue de fluide frigorigene et d'huile

    Energy Technology Data Exchange (ETDEWEB)

    Feidt, M.

    1999-10-13

    The phase out of chlorine containing refrigerants (CFC and HCFC) has led to the introduction of new refrigerants and lubricants to the market. The interest in using HFC fluids as working fluids to replace fluids harmful to the stratospheric ozone layer. The study presents the influence of synthetic oil (POE ISO 68) on flow boiling of refrigerants R134a (pure fluid) and R410A (R32/R125 50%/50%). Local and average heat transfer coefficients and pressure drops have been measured for a smooth horizontal tube. The distribution of the heat transfer coefficient at the inner wall has been obtained from solving the inverse heat conduction problem (IHCP) and resulted in a local combination of nucleate and convective contributions to flow boiling. Local heat transfer coefficients have been averaged and displayed as a function of the vapour quality. For R134a: small amounts of oil (1% to 6%) in the liquid phase increased the heat transfer coefficient at low and intermediate vapour qualities (less than 0.60) compared to pure fluid. However a hugh reduction of the heat transfer has been observed at higher vapour qualities. For R410A : oil dramatically decreases the heat transfer coefficient compared to pure fluid. Pressure drops are also affected by small amounts of lubricant: an important increase has been noted for both fluids. Available design methods for flow boiling heat transfer coefficient (superposition, enhancement, asymptotic) badly predict the experimental results. Nevertheless a new design method accounting for flow patterns has shown good agreements. The influence of the lubricant on the heat transfer is discussed and a new proposition is made to calculate pressure drops. (author)

  16. Computational multi-fluid dynamics predictions of critical heat flux in boiling flow

    Energy Technology Data Exchange (ETDEWEB)

    Mimouni, S., E-mail: stephane.mimouni@edf.fr; Baudry, C.; Guingo, M.; Lavieville, J.; Merigoux, N.; Mechitoua, N.

    2016-04-01

    Highlights: • A new mechanistic model dedicated to DNB has been implemented in the Neptune-CFD code. • The model has been validated against 150 tests. • Neptune-CFD code is a CFD tool dedicated to boiling flows. - Abstract: Extensive efforts have been made in the last five decades to evaluate the boiling heat transfer coefficient and the critical heat flux in particular. Boiling crisis remains a major limiting phenomenon for the analysis of operation and safety of both nuclear reactors and conventional thermal power systems. As a consequence, models dedicated to boiling flows have being improved. For example, Reynolds Stress Transport Model, polydispersion and two-phase flow wall law have been recently implemented. In a previous work, we have evaluated computational fluid dynamics results against single-phase liquid water tests equipped with a mixing vane and against two-phase boiling cases. The objective of this paper is to propose a new mechanistic model in a computational multi-fluid dynamics tool leading to wall temperature excursion and onset of boiling crisis. Critical heat flux is calculated against 150 tests and the mean relative error between calculations and experimental values is equal to 8.3%. The model tested covers a large physics scope in terms of mass flux, pressure, quality and channel diameter. Water and R12 refrigerant fluid are considered. Furthermore, it was found that the sensitivity to the grid refinement was acceptable.

  17. Computational multi-fluid dynamics predictions of critical heat flux in boiling flow

    International Nuclear Information System (INIS)

    Mimouni, S.; Baudry, C.; Guingo, M.; Lavieville, J.; Merigoux, N.; Mechitoua, N.

    2016-01-01

    Highlights: • A new mechanistic model dedicated to DNB has been implemented in the Neptune_CFD code. • The model has been validated against 150 tests. • Neptune_CFD code is a CFD tool dedicated to boiling flows. - Abstract: Extensive efforts have been made in the last five decades to evaluate the boiling heat transfer coefficient and the critical heat flux in particular. Boiling crisis remains a major limiting phenomenon for the analysis of operation and safety of both nuclear reactors and conventional thermal power systems. As a consequence, models dedicated to boiling flows have being improved. For example, Reynolds Stress Transport Model, polydispersion and two-phase flow wall law have been recently implemented. In a previous work, we have evaluated computational fluid dynamics results against single-phase liquid water tests equipped with a mixing vane and against two-phase boiling cases. The objective of this paper is to propose a new mechanistic model in a computational multi-fluid dynamics tool leading to wall temperature excursion and onset of boiling crisis. Critical heat flux is calculated against 150 tests and the mean relative error between calculations and experimental values is equal to 8.3%. The model tested covers a large physics scope in terms of mass flux, pressure, quality and channel diameter. Water and R12 refrigerant fluid are considered. Furthermore, it was found that the sensitivity to the grid refinement was acceptable.

  18. ASTRID: A 3D Eulerian software for subcooled boiling modelling - comparison with experimental results in tubes and annuli

    International Nuclear Information System (INIS)

    Briere, E.; Larrauri, D.; Olive, J.

    1995-01-01

    For about four years, Electricite de France has been developing a 3-D computer code for the Eulerian simulation of two-phase flows. This code, named ASTRID, is based on the six-equation two-fluid model. Boiling water flows, such as those encountered in nuclear reactors, are among the main applications of ASTRID. In order to provide ASTRID with closure laws and boundary conditions suitable for boiling flows, a boiling model has been developed by EDF and the Institut de Mecanique des Fluides de Toulouse. In the fluid, the heat and mass transfer between a bubble and the liquid is being modelled. At the heating wall, the incipient boiling point is determined according to Hsu's criterion and the boiling heat flux is split into three additive terms: a convective term, a quenching term and a vaporisation term. This model uses several correlations. EDF's program in boiling two-phase flows also includes experimental studies, some of which are performed in collaboration with other laboratories. Refrigerant subcooled boiling both in tubular (DEBORA experiment, CEN Grenoble) and in annular geometry (Arizona State University Experiment) have been computed with ASTRID. The simulations show the satisfactory results already obtained on void fraction and liquid temperature. Ways of improvement of the model are drawn especially on the dynamical part

  19. ASTRID: A 3D Eulerian software for subcooled boiling modelling - comparison with experimental results in tubes and annuli

    Energy Technology Data Exchange (ETDEWEB)

    Briere, E.; Larrauri, D.; Olive, J. [Electricite de France, Chatou (France)

    1995-09-01

    For about four years, Electricite de France has been developing a 3-D computer code for the Eulerian simulation of two-phase flows. This code, named ASTRID, is based on the six-equation two-fluid model. Boiling water flows, such as those encountered in nuclear reactors, are among the main applications of ASTRID. In order to provide ASTRID with closure laws and boundary conditions suitable for boiling flows, a boiling model has been developed by EDF and the Institut de Mecanique des Fluides de Toulouse. In the fluid, the heat and mass transfer between a bubble and the liquid is being modelled. At the heating wall, the incipient boiling point is determined according to Hsu`s criterion and the boiling heat flux is split into three additive terms: a convective term, a quenching term and a vaporisation term. This model uses several correlations. EDF`s program in boiling two-phase flows also includes experimental studies, some of which are performed in collaboration with other laboratories. Refrigerant subcooled boiling both in tubular (DEBORA experiment, CEN Grenoble) and in annular geometry (Arizona State University Experiment) have been computed with ASTRID. The simulations show the satisfactory results already obtained on void fraction and liquid temperature. Ways of improvement of the model are drawn especially on the dynamical part.

  20. A study of vapor bubble departure in subcooled flow boiling at low pressure

    International Nuclear Information System (INIS)

    Donevski, Bozin; Saga, Tetsuo; Kobayashi, Toshio; Segawa, Shigeki

    1999-01-01

    An experimental study of vapor bubble dynamics in sub-cooled flow boiling was conducted using the flow visualization and digital image processing methods. Vapor bubble departure departure in subcooled flow boiling have been experimentally investigated over a range of mass flux G=0.384 (kg/m 2 s), and heat flux q w = 27.2 x 10 4 (W/m 2 ), for the subcooled flow boiling region. It has been observed that once a vapor bubble departs from a nucleation site, it typically slides along the heating surface at sonic finite distance down-stream of nucleation site. The image processing method proposed in this study is based on the detachment and tracing of the edges of the bubbles and their background. The proposed method can be used in various fields of engineering applications. (Original)

  1. Two-phase flow regimes and mechanisms of critical heat flux under subcooled flow boiling conditions

    International Nuclear Information System (INIS)

    Le Corre, Jean-Marie; Yao, Shi-Chune; Amon, Cristina H.

    2010-01-01

    A literature review of critical heat flux (CHF) experimental visualizations under subcooled flow boiling conditions was performed and systematically analyzed. Three major types of CHF flow regimes were identified (bubbly, vapor clot and slug flow regime) and a CHF flow regime map was developed, based on a dimensional analysis of the phenomena and available experimental information. It was found that for similar geometric characteristics and pressure, a Weber number (We)/thermodynamic quality (x) map can be used to predict the CHF flow regime. Based on the experimental observations and the review of the available CHF mechanistic models under subcooled flow boiling conditions, hypothetical CHF mechanisms were selected for each CHF flow regime, all based on a concept of wall dry spot overheating, rewetting prevention and subsequent dry spot spreading. Even though the selected concept has not received much attention (in term or theoretical developments and applications) as compared to other more popular DNB models, its basis have often been cited by experimental investigators and is considered by the authors as the 'most-likely' mechanism based on the literature review and analysis performed in this work. The selected modeling concept has the potential to span the CHF conditions from highly subcooled bubbly flow to early stage of annular flow and has been numerically implemented and validated in bubbly flow and coupled with one- and three-dimensional (CFD) two-phase flow codes, in a companion paper. [Le Corre, J.M., Yao, S.C., Amon, C.H., in this issue. A mechanistic model of critical heat flux under subcooled flow boiling conditions for application to one and three-dimensional computer codes. Nucl. Eng. Des.].

  2. Propagation of Local Bubble Parameters of Subcooled Boiling Flow in a Pressurized Vertical Annulus Channel

    International Nuclear Information System (INIS)

    Chu, In-Cheol; Lee, Seung Jun; Youn, Young Jung; Park, Jong Kuk; Choi, Hae Seob; Euh, Dong Jin

    2015-01-01

    CMFD (Computation Multi-Fluid Dynamics) tools have been being developed to simulate two-phase flow safety problems in nuclear reactor, including the precise prediction of local bubble parameters in subcooled boiling flow. However, a lot of complicated phenomena are encountered in the subcooled boiling flow such as bubble nucleation and departure, interfacial drag of bubbles, lateral migration of bubbles, bubble coalescence and break-up, and condensation of bubbles, and the constitutive models for these phenomena are not yet complete. As a result, it is a difficult task to predict the radial profile of bubble parameters and its propagation along the flow direction. Several experiments were performed to measure the local bubble parameters for the validation of the CMFD code analysis and improvement of the constitutive models of the subcooled boiling flow, and to enhance the fundamental understanding on the subcooled boiling flow. The information on the propagation of the local flow parameters along the flow direction was not provided because the measurements were conducted at the fixed elevation. In SUBO experiments, the radial profiles of local bubble parameters, liquid velocity and temperature were obtained for steam-water subcooled boiling flow in a vertical annulus. The local flow parameters were measured at six elevations along the flow direction. The pressure was in the range of 0.15 to 0.2 MPa. We have launched an experimental program to investigate quantify the local subcooled boiling flow structure under elevated pressure condition in order to provide high precision experimental data for thorough validation of up-to-date CMFD codes. In the present study, the first set of experimental data on the propagation of the radial profile of the bubble parameters was obtained for the subcooled boiling flow of R-134a in a pressurized vertical annulus channel. An experimental program was launched for an in-depth investigation of a subcooled boiling flow in an elevated

  3. The use of Trefftz functions for approximation of measurement data in an inverse problem of flow boiling in a minichannel

    Directory of Open Access Journals (Sweden)

    Hozejowski Leszek

    2012-04-01

    Full Text Available The paper is devoted to a computational problem of predicting a local heat transfer coefficient from experimental temperature data. The experimental part refers to boiling flow of a refrigerant in a minichannel. Heat is dissipated from heating alloy to the flowing liquid due to forced convection. The mathematical model of the problem consists of the governing Poisson equation and the proper boundary conditions. For accurate results it is required to smooth the measurements which was obtained by using Trefftz functions. The measurements were approximated with a linear combination of Trefftz functions. Due to the computational procedure in which the measurement errors are known, it was possible to smooth the data and also to reduce the residuals of approximation on the boundaries.

  4. Mixed convection in a two-phase flow cooling loop

    International Nuclear Information System (INIS)

    Janssens-Maenhout, G.; Daubner, M.; Knebel, J.U.

    2002-03-01

    This report summarizes the numerical simulations using the CFD code CFX4.1 which has additional models for subcooled flow boiling phenomena and the interfacial forces. The improved CFX4.1 code can be applied to the design of boiling induced mixed convection cooling loops in a defined parameter range. The experimental part describes the geysering experiments and the instability effects on the two-phase natural circulation flow. An experimentally validated flow pattern map in the Phase Change Number - Subcooling Number (N PCh - N Sub ) diagram defines the operational range in which flow instabilities such as geysering can be expected. One important perspective of this combined experimental/numerical work, which is in the field of two-phase flow, is its application to the development of accelerator driven systems (ADS). The main objective on an ADS is its potential to transmute minor actinides and long-lived fission products, thus participating in closing the fuel cycle. The development of an ADS is an important issue within the Euratom Fifth FP on Partitioning and Transmutation. One concept of an ADS, which is investigated in more detail within the ''preliminary design study of an experimental ADS'' Project (PDS-XADS) of the Euratom Fifth FP, is the XADS lead-bismuth cooled Experimental ADS of ANSALDO. An essential feature of this concept is the natural circulation of the primary coolant within the reactor pool. The natural circulation, which is driven by the density differences between the blanket and the heat exchanger, is enhanced by the injection of the nitrogen cover gas through spargers located in a riser part just above the blanket. This so-called gas-lift pump system has not been investigated in more detail nor has this gas-lift pump system been numerically/experimentally confirmed. The knowledge gained within the SUCO Programe, i.e. the modelling of the interfacial forces, the experimental work on flow instabilities and the modelling of the interfacial area

  5. Mixed convection in a two-phase flow cooling loop

    Energy Technology Data Exchange (ETDEWEB)

    Janssens-Maenhout, G.; Daubner, M.; Knebel, J.U.

    2002-03-01

    This report summarizes the numerical simulations using the CFD code CFX4.1 which has additional models for subcooled flow boiling phenomena and the interfacial forces. The improved CFX4.1 code can be applied to the design of boiling induced mixed convection cooling loops in a defined parameter range. The experimental part describes the geysering experiments and the instability effects on the two-phase natural circulation flow. An experimentally validated flow pattern map in the Phase Change Number - Subcooling Number (N{sub PCh} - N{sub Sub}) diagram defines the operational range in which flow instabilities such as geysering can be expected. One important perspective of this combined experimental/numerical work, which is in the field of two-phase flow, is its application to the development of accelerator driven systems (ADS). The main objective on an ADS is its potential to transmute minor actinides and long-lived fission products, thus participating in closing the fuel cycle. The development of an ADS is an important issue within the Euratom Fifth FP on Partitioning and Transmutation. One concept of an ADS, which is investigated in more detail within the ''preliminary design study of an experimental ADS'' Project (PDS-XADS) of the Euratom Fifth FP, is the XADS lead-bismuth cooled Experimental ADS of ANSALDO. An essential feature of this concept is the natural circulation of the primary coolant within the reactor pool. The natural circulation, which is driven by the density differences between the blanket and the heat exchanger, is enhanced by the injection of the nitrogen cover gas through spargers located in a riser part just above the blanket. This so-called gas-lift pump system has not been investigated in more detail nor has this gas-lift pump system been numerically/experimentally confirmed. The knowledge gained within the SUCO Programe, i.e. the modelling of the interfacial forces, the experimental work on flow instabilities and the

  6. Study on onset of nucleate boiling and net vapor generation point in subcooled flow boiling

    International Nuclear Information System (INIS)

    Ohtake, Hiroyasu; Wada, Noriyoshi; Koizumi, Yasuo

    2002-01-01

    The onset of nucleate boiling (ONB) and the point of net vapor generation on subcooled flow boiling, focusing on liquid subcooling and liquid velocity were investigated experimentally and analytically. Experiments were conducted using a copper thin-film (35μm) and subcooled water in a range of the liquid velocity from 0.27 to 4.6 m/s at 0.10MPa. The liquid subcoolings were 20, 30 and 40K, respectively. Temperatures at the onset of nucleate boiling obtained in the experiments increased with the liquid subcoolings and the liquid velocities. The increases in the temperature of ONB were represented with the classical stability theory of preexisting nuclei. The measured results of the net vapor generation agreed well with the results of correlation by Saha and Zuber in the range of the present experiments. (J.P.N.)

  7. Heat transfer by natural convection into an horizontal cavity

    International Nuclear Information System (INIS)

    Arevalo J, P.

    1998-01-01

    At this thesis it is studied the heat transfer by natural convection in an horizontal cavity, it is involved a boiling's part that is described the regimes and correlations differences for boiling's curve. It is designed a horizontal cavity for realize the experimental part and it's mention from equipment or instrumentation to succeed in a experimentation that permits to realize the analysis of heat transfer, handling as water fluid at atmospheric pressure and where it's present process from natural convection involving part boiling's subcooled. The system consists of heater zone submerged in a horizontal cavity with water. Once part finished experimental with information to obtained it's proceeded to obtain a correlation, realized starting from analysis dimensionless such as: Jakob, Bond and Grasoft (Boiling) besides of knows in natural convection: Prandtl and Nusselt. The mathematical model explains the behavior for natural convection continued part boiling's subcooled. It is realize analysis graphics too where it's show comparing with Globe Dropkin and Catton equations by natural convection with bottom heating. (Author)

  8. The onset of flow instability for a downward flow of a non-boiling heated liquid

    International Nuclear Information System (INIS)

    Babelli, Ibrahim; Ishii, Mamoru

    1999-01-01

    A procedure for predicting the onset of flow instability (OFI) in downward flows at low-pressure and low-flow conditions without boiling is presented in this paper. It is generally accepted that the onset of significant void in subcooled boiling precedes, and is a precondition to, the occurrence of static flow instability. A detailed analysis of the pressure drop components for a downward flow in a heated channel reveals the possibility of unstable transition from single-phase flow to high-quality two-phase flow, i.e., flow excursion. Low flow rate and high subcooling are the two important conditions for the occurrence of this type of instability. The unstable transition occurs when the resistance to the downward flow caused by local (orifice), frictional, and thermal expansion pressure drops equalizes the driving force of the gravitational pressure drop. The inclusion of the thermal expansion pressure drop is essential to account for this type of transition. Experimental data are yet to be produced to verify the prediction of the present analysis. (author)

  9. A new mechanistic model of critical heat flux in forced-convection subcooled boiling

    International Nuclear Information System (INIS)

    Alajbegovic, A.; Kurul, N.; Podowski, M.Z.; Drew, D.A.; Lahey, R.T. Jr.

    1997-10-01

    Because of its practical importance and various industrial applications, the process of subcooled flow boiling has attracted a lot of attention in the research community in the past. However, the existing models are primarily phenomenological and are based on correlating experimental data rather than on a first-principle analysis of the governing physical phenomena. Even though the mechanisms leading to critical heat flux (CHF) are very complex, the recent progress in the understanding of local phenomena of multiphase flow and heat transfer, combined with the development of mathematical models and advanced Computational Fluid Dynamics (CFD) methods, makes analytical predictions of CHF quite feasible. Various mechanisms leading to CHF in subcooled boiling have been investigated. A new model for the predictions of the onset of CHF has been developed. This new model has been coupled with the overall boiling channel model, numerically implemented in the CFX 4 computer code, tested and validated against the experimental data of Hino and Ueda. The predicted critical heat flux for various channel operating conditions shows good agreement with the measurements using the aforementioned closure laws for the various local phenomena governing nucleation and bubble departure from the wall. The observed differences are consistent with typical uncertainties associated with CHF data

  10. Study of sodium film-boiling heat transfer from a high-temperature sphere

    International Nuclear Information System (INIS)

    Le-Belguet, A.

    2013-01-01

    During a severe accident in a sodium-cooled fast reactor, molten fuel may come into contact with the surrounding liquid sodium, resulting in a so-called Fuel-Coolant Interaction. This work aims at providing a better understanding and knowledge of the associated heat transfer, likely to be in the film-boiling regime and required to study the risks related to a vapor explosion. Scarce literature has been found on sodium film boiling, both from an experimental and a theoretical point of view. Only one experiment has been conducted to investigate sodium pool film-boiling heat transfer. In our analysis of the experiment, two film-boiling regimes have been identified: a stable film boiling regime, without liquid-solid contact, and an unstable film-boiling regime, with contacts. Besides, the only theoretical model dedicated to sodium film boiling has shown some weaknesses. First, a scaling analysis of the problem has been proposed for free and forced convection, considering the two extreme cases of saturated and highly subcooled liquid. This simplified approach, which shows a good agreement with the experimental data, provides the dimensionless numbers which should be used to build correlations. A theoretical model has been developed to describe sodium film-boiling heat transfer from a hot sphere in free and forced convection, whatever the liquid subcooling. It is based on a two-phase laminar boundary layer integral method and includes the inertial and convective terms in the vapor momentum and energy equations, usually neglected. The radiation has been taken into account in the interfacial energy balance and contributes directly to produce vapor. This model enables to predict the heat lost from a hot body within an acceptable error compared to the tests results especially when the experimental uncertainties are considered. The heat partition between liquid heating and vaporization, essential to study the vapor explosion phenomenon, is also estimated. The influence of

  11. Study on Fins' Effect of Boiling Flow in Millimeter Channel Heat Exchanger

    Science.gov (United States)

    Watanabe, Satoshi

    2005-11-01

    Recently, a lot of researches about compact heat exchangers with mini-channels have been carried out with the hope of obtaining a high-efficiency heat transfer, due to the higher ratio of surface area than existing heat exchangers. However, there are many uncertain phenomena in fields such as boiling flow in mini-channels. Thus, in order to understand the boiling flow in mini-channels to design high-efficiency heat exchangers, this work focused on the visualization measurement of boiling flow in a millimeter channel. A transparent acrylic channel (heat exchanger form), high-speed camera (2000 fps at 1024 x 1024 pixels), and halogen lamp (backup light) were used as the visualization system. The channel's depth is 2 mm, width is 30 mm, and length is 400 mm. In preparation for commercial use, two types of channels were experimented on: a fins type and a normal slit type (without fins). The fins are circular cylindrical obstacles (diameter is 5 mm) to promote heat transfer, set in a triangular array (distance between each center point is 10 mm). Especially in this work, boiling flow and heat transfer promotion in the millimeter channel heat exchanger with fins was evaluated using a high-speed camera.

  12. Heat transfer effect of an extended surface in downward-facing subcooled flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Khan, Abdul R., E-mail: khan@vis.t.u-tokyo.ac.jp [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Erkan, Nejdet, E-mail: erkan@vis.t.u-tokyo.ac.jp [Nuclear Professional School, School of Engineering, The University of Tokyo, 2-22 Shirakata, Tokai-mura, Ibaraki, 319-1188 (Japan); Okamoto, Koji, E-mail: okamoto@n.t.u-tokyo.ac.jp [Nuclear Professional School, School of Engineering, The University of Tokyo, 2-22 Shirakata, Tokai-mura, Ibaraki, 319-1188 (Japan)

    2015-12-15

    Highlights: • Compare downward-facing flow boiling results from bare and extended surfaces. • Upstream and downstream temperatures were measured on the extended surface. • Downstream temperatures exceed upstream temperatures for all flow rates. • Bubble accumulation occurs downstream on extended surface. • Extended surface heat transfer lower than bare surface as flow rate reduced. - Abstract: New BWR containment designs are considering cavity flooding as an accident management strategy. Unlike the PWR, the BWR has many Control Rod Guide Tube (CRGT) penetrations in the lower head. During a severe accident scenario with core melt in the lower plenum along with cavity flooding, the penetrations may affect the heat transfer on the ex-vessel surface and disrupt fluid flow during the boiling process. A small-scale experiment was performed to investigate the issues existing in downward-facing boiling phenomenon with an extended surface. The results were compared with a bare (flat) surface. The mass flux of 244 kg/m{sup 2} s, 215 kg/m{sup 2} s, and 177 kg/m{sup 2} s were applied in this study. CHF conditions were observed only for the 177 kg/m{sup 2} s case. The boiling curves for both types of surfaces and all flow rates were obtained. The boiling curves for the highest flow rate showed lower surface temperatures for the extended surface experiments when compared to the bare surface. The downstream location on the extended surface yielded the highest surface temperatures as the flow rate was reduced. The bubble accumulation and low velocity in the wake produced by flow around the extended surface was believed to have caused the elevated temperatures in the downstream location. Although an extended surface may enhance the overall heat transfer, a reduction in the local heat transfer was observed in the current experiments.

  13. A visual study of forced convection boiling. Part 2: Flow patterns and burnout for a round test section

    International Nuclear Information System (INIS)

    Kirby, G.J.; Staniforth, R.; Kinneir, J.H.

    1967-03-01

    The studies of boiling water at 25 p.s.i.a. reported here show the same flow patterns as in earlier tests in that the bubbles formed on the heater regained close to the heated surface to coalesce into large bubbles which eventually spanned the flow channel. Burnout tests were made and it was found there was a change of slope of the heat flux-subcooling curve. Further tests showed that this effect was due to a change in flow regime between burnout with much vapour present and burnout with just nucleate bubbles present. In the latter regime it was found that burnout is dependent only on the conditions local to the burnout point. Photography of the burnout region was practicable only when few bubbles were present but although pictures of the bubble over the burnout point were taken, no clear evidence on the mechanism of formation of the bubble could be gleaned. Some speculation on the cause of burnout in this regime is made in the light of these experiments. (author)

  14. Free convection in a partially submerged fluid loop

    International Nuclear Information System (INIS)

    Britt, T.E.; Wood, D.C.

    1982-01-01

    Several natural convection loop systems are studied in order to determine the operational characteristics for a multiple loop container which is used to cool failed nuclear reactor assemblies. Both analytical and experimental studies were undertaken to examine flow in both circular and rectangular flow loops. It was found that when a circular loop is heated at the bottom and cooled at the top, recirculation cells form at all input power fluxes. At fluxes between 0.1 W/cm 2 and 0.7 W/cm 2 the cells caused flow oscillations and reversals. With the circular loop heated from the side, no recirculation cells were observed at the power fluxes up to 1.5 W/cm. Boiling did not occur in the circular loop. For a rectangular loop heated and cooled on its vertical sides, no recirculation cells or flow reversals were seen. At input power fluxes above 1.2 W/cm 2 , periodic boiling in the heated side caused flow oscillations

  15. Assessment of Nucleation Site Density Models for CFD Simulations of Subcooled Flow Boiling

    International Nuclear Information System (INIS)

    Hoang, N. H.; Chu, I. C.; Euh, D. J.; Song, C. H.

    2015-01-01

    The framework of a CFD simulation of subcooled flow boiling basically includes a block of wall boiling models communicating with governing equations of a two-phase flow via parameters like temperature, rate of phasic change, etc. In the block of wall boiling models, a heat flux partitioning model, which describes how the heat is taken away from a heated surface, is combined with models quantifying boiling parameters, i.e. nucleation site density, and bubble departure diameter and frequency. It is realized that the nucleation site density is an important parameter for predicting the subcooled flow boiling. The number of nucleation sites per unit area decides the influence region of each heat transfer mechanism. The variation of the nucleation site density will mutually change the dynamics of vapor bubbles formed at these sites. In addition, the nucleation site density is needed as one initial and boundary condition to solve the interfacial area transport equation. A lot of effort has been devoted to mathematically formulate the nucleation site density. As a consequence, numerous correlations of the nucleation site density are available in the literature. These correlations are commonly quite different in their mathematical form as well as application range. Some correlations of the nucleation site density have been applied successfully to CFD simulations of several specific subcooled boiling flows, but in combination with different correlations of the bubble departure diameter and frequency. In addition, the values of the nucleation site density, and bubble departure diameter and frequency obtained from simulations for a same problem are relatively different, depending on which models are used, even when global characteristics, e.g., void fraction and mean bubble diameter, agree well with experimental values. It is realized that having a good CFD simulations of the subcooled flow boiling requires a detailed validations of all the models used. Owing to the importance

  16. Subcooled boiling heat transfer to R 12 in an annular vertical channel

    Energy Technology Data Exchange (ETDEWEB)

    Braeuer, H.; Mayinger, F.

    1988-10-01

    Detailed knowledge of the physical phenomena involved in subcooled boiling is of great importance for the design of liquid-cooled heat generating systems with high heat fluxes. Experimental heat transfer data were obtained for forced convective boiling of dichloro-difluoroethane (R 12). The flow is circulated upwards through a concentric annular vertical channel. The inner and outer diameters of the annulus are 0.016 m and 0.03 m respectively. The reduced pressures studied were 0.24 less than or equal to p/p/sub crit/ less than or equal to 0.8, inlet subcooling varied from 10 to 75 K and mass fluxes from 500 to 3000 kg/m/sup 2/s, which corresponds to Re numbers from 30 000 to 300 000. The experiments, described in this study, demonstrate that liquid fluorocarbons show certain unusual boiling characteristics in the subcooled flow, such as hysteresis of the boiling curve. These characteristics are attributed to the properties of the fluid, mainly the Pr number and the very low surface tension. The pronounced boiling curve hysteresis can be explained by the fact that large nucleation sites may have been flooded prior to incipient boiling. A dimensionless regression formula is presented which predicts the onset of subcooled boiling as a function of reduced pressure (p/p/sub crit/), Boiling-(Bo), Reynolds-(Re), and a modified Jacob Number (Ja), over the whole range of parameters studied, with a good accuracy, including water data from literature.

  17. Review of Mixed Convection Flow Regime Map of a Vertical pipe

    International Nuclear Information System (INIS)

    Chae, Myeong-Seon; Chung, Bum-Jin; Kang, Gyeong-Uk

    2015-01-01

    In a vertical pipe, the natural convective force due to buoyancy acts upward only, but forced convective force can be either upward or downward. This determines buoyancy-aided and buoyancy-opposed flows depending on the direction of forced flow with respect to the buoyancy forces. Furthermore, depending on the exchange mechanism, the flow condition is classified into laminar and turbulent. In laminar mixed convection, buoyancy-aided flow presents enhanced heat transfer compared to the pure forced convection and buoyancy-opposed flow shows impaired heat transfer as the flow velocity affected by the buoyancy forces. However, in turbulent mixed convection, buoyancy-aided flow shows an impairment of the heat transfer rate for small buoyancy, and a gradational enhancement for large buoyancy. In this study, the existing flow regime map on mixed convection in a vertical pipe was reviewed through an analysis of literatures. Using the investigated data and heat transfer correlations, the flow regime map was reconstructed independently, and compared with the existing one. This study reviewed the limitations of the classical mixed convection flow regime map. Using the existing data and heat transfer correlations by Martinelli and Boelter and Watzinger and Johnson, the flow regime map was reconstructed independently. The results revealed that the existing map used the data selectively among the experimental and theoretical results, and a detailed description for lines forming mixed convection and transition regime were not given. And the information about uncertainty analysis and the evidentiary data were given insufficiently. The flow regime map and investigator commonly used the diameter as the characteristic length for both Re and Gr in place of the height of the heated wall, though the buoyancy forces are proportional to the third power of the height of heated wall

  18. Flow boiling test of GDP replacement coolants

    International Nuclear Information System (INIS)

    Park, S.H.

    1995-01-01

    The tests were part of the CFC replacement program to identify and test alternate coolants to replace CFC-114 being used in the uranium enrichment plants at Paducah and Portsmouth. The coolants tested, C 4 F 10 and C 4 F 8 , were selected based on their compatibility with the uranium hexafluoride process gas and how well the boiling temperature and vapor pressure matched that of CFC-114. However, the heat of vaporization of both coolants is lower than that of CFC-114 requiring larger coolant mass flow than CFC-114 to remove the same amount of heat. The vapor pressure of these coolants is higher than CFC-114 within the cascade operational range, and each coolant can be used as a replacement coolant with some limitation at 3,300 hp operation. The results of the CFC-114/C 4 F 10 mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C 4 F 10 weight mixture in CFC-114 and the degree of degradation is about 20% from that of CFC-114 boiling heat transfer coefficient. This report consists of the final reports from Cudo Technologies, Ltd

  19. Influence of convective conditions on three dimensional mixed convective hydromagnetic boundary layer flow of Casson nanofluid

    Energy Technology Data Exchange (ETDEWEB)

    Rauf, A., E-mail: raufamar@ciitsahiwal.edu.pk [Department of Mathematics, Comsats Institute of Information Technology, Sahiwal 57000 (Pakistan); Siddiq, M.K. [Centre for Advanced Studies in Pure and Applied Mathematics, Department of Mathematics, Bahauddin Zakariya University, Multan 63000 (Pakistan); Abbasi, F.M. [Department of Mathematics, Comsats Institute of Information Technology, Islamabad 44000 (Pakistan); Meraj, M.A. [Department of Mathematics, Comsats Institute of Information Technology, Sahiwal 57000 (Pakistan); Ashraf, M. [Centre for Advanced Studies in Pure and Applied Mathematics, Department of Mathematics, Bahauddin Zakariya University, Multan 63000 (Pakistan); Shehzad, S.A. [Department of Mathematics, Comsats Institute of Information Technology, Sahiwal 57000 (Pakistan)

    2016-10-15

    The present work deals with the steady laminar three-dimensional mixed convective magnetohydrodynamic (MHD) boundary layer flow of Casson nanofluid over a bidirectional stretching surface. A uniform magnetic field is applied normal to the flow direction. Similarity variables are implemented to convert the non-linear partial differential equations into ordinary ones. Convective boundary conditions are utilized at surface of the sheet. A numerical technique of Runge–Kutta–Fehlberg (RFK45) is used to obtain the results of velocity, temperature and concentration fields. The physical dimensionless parameters are discussed through tables and graphs. - Highlights: • Mixed convective boundary layer flow of Casson nanofluid is taken into account. • Impact of magnetic field is examined. • Convective heat and mass conditions are imposed. • Numerical solutions are presented and discussed.

  20. Reassessment of forced convection heat transfer correlations for refrigerant-12

    International Nuclear Information System (INIS)

    Celata, G.P.; Cuomo, M.; D'Annibale, F.; Farello, G.E.; Setaro, T.

    1986-01-01

    In the frame of a Refrigerant-12 experiment on postulated accidental transients in Pressurized Water Reactors under way at Heat Transfer Laboratory (ENEA Casaccia Research Center), an assessment of the main correlation available in scientific literature, for the different heat transfer regions encountered when a liquid is boiled in a confined heated channel, has been performed. Considering a vertical tube uniformly heated over its length with CHF at the exit, the following heat transfer regimes may be individuated: convective heat transfer to liquid, subcooled boiling, saturated nucleate boiling, forced convective heat transfer through liquid film (annular flow regime) and thermal crisis. From the comparison of computed values with an original ENEA dataset, the best correlations in predicting Refrigerant-12 data have been individuated. In a few cases, though preserving the original structure of the correlations, mainly developed for water, it was necessary to adjust some coefficients by means of best-fit procedures through our experimental data. The work has been performed in the frame of the ENEA Thermal Reactor Department Safety Research Project

  1. Numerical modeling of flow boiling instabilities using TRACE

    International Nuclear Information System (INIS)

    Kommer, Eric M.

    2015-01-01

    Highlights: • TRACE was used to realistically model boiling instabilities in single and parallel channel configurations. • Model parameters were chosen to exactly mimic other author’s work in order to provide for direct comparison of results. • Flow stability maps generated by the model show unstable flow at operating points similar to other authors. • The method of adjudicating when a flow is “unstable” is critical in this type of numerical study. - Abstract: Dynamic flow instabilities in two-phase systems are a vitally important area of study due to their effects on a great number of industrial applications, including heat exchangers in nuclear power plants. Several next generation nuclear reactor designs incorporate once through steam generators which will exhibit boiling flow instabilities if not properly designed or when operated outside design limits. A number of numerical thermal hydraulic codes attempt to model instabilities for initial design and for use in accident analysis. TRACE, the Nuclear Regulatory Commission’s newest thermal hydraulic code is used in this study to investigate flow instabilities in both single and dual parallel channel configurations. The model parameters are selected as to replicate other investigators’ experimental and numerical work in order to provide easy comparison. Particular attention is paid to the similarities between analysis using TRACE Version 5.0 and RELAP5/MOD3.3. Comparison of results is accomplished via flow stability maps non-dimensionalized via the phase change and subcooling numbers. Results of this study show that TRACE does indeed model two phase flow instabilities, with the transient response closely mimicking that seen in experimental studies. When compared to flow stability maps generated using RELAP, TRACE shows similar results with differences likely due to the somewhat qualitative criteria used by various authors to determine when the flow is truly unstable

  2. Surface roughness effects on onset of nucleate boiling and net vapor generation point in subcooled flow boiling

    International Nuclear Information System (INIS)

    Ohtake, Hiroyasu; Wada, Noriyoshi; Koizumi, Yasuo

    2003-01-01

    The ability to predict void formation and void fraction in subcooled flow boiling is of importance to the nuclear reactor technology because the presence of voids affects the steady state and transient response of a reactor. The onset of nucleate boiling and the point of net vapor generation on subcooled flow boiling, focusing on surface roughness, liquid subcooling and liquid velocity were investigated experimentally and analytically. Experiments were conducted using a copper thin-film and subcooled water in a range of the liquid velocity from 0.27 to 4.6 m/s at 0.10MPa; the liquid subcoolings were 20, 30 and 40K, respectively. The surface roughness on the test heater was observed by SEM. Experimental results showed that temperatures at the onset nucleate boiling increased with increasing the liquid subcoolings or the liquid velocities. The trend of increase in the temperature at the ONB was in good agreement with the present analytical result based on the stability theory of preexisting nuclei. The measured results for the net vapor generation point agreed well with the results of correlation by Saha and Zuber in the range of the present experiments. The temperature at the ONB decreased with an increasing size of surface roughness, while the NVG-point was independent on the surface roughness. The dependence on the ONB temperature of the roughness size was also represented well by the present analytical model

  3. Problems of mixed convection flow regime map in a vertical cylinder

    International Nuclear Information System (INIS)

    Kang, Gyeong Uk; Chung, Bum Jin

    2012-01-01

    One of the technical issues by the development of the VHTR is the mixed convection, which is the regime of heat transfer that occurs when the driving forces of both forced and natural convection are of comparable orders of magnitude. In vertical internal flows, the buoyancy force acts upward only, but forced flows can move either upward or downward. Thus, there are two types of mixed convection flows, depending on the direction of the forced flow. When the directions of the forced flow and buoyancy are the same, the flow is a buoyancy aided flow; when they are opposite, the flow is a buoyancy opposed flow. In laminar flows, buoyancy aided flow shows enhanced heat transfer compared to the pure forced convection and buoyancy opposed flow shows impaired heat transfer due to the flow velocity affected by the buoyancy forces. In turbulent flows, however, buoyancy opposed flows shows enhanced heat transfer due to increased turbulence production and buoyancy aided flow shows impaired heat transfer at low buoyancy forces and as the buoyancy increases, the heat transfer restores and at further increases of the buoyancy forces, the heat transfer is enhanced. It is of primary interests to classify which convection regime is mainly dominant. The methods most used to classify between forced, mixed and natural convection have been to refer to the classical flow regime map suggested by Meta is and Eckert. During the course of fundamental literature studies on this topic, it is found that there are some problems on the flow regime map in a vertical cylinder. This paper is to discuss problems identified through reviewing the papers composed in the classical flow regime map. We have tried to reproduce the flow regime map independently using the data obtained from the literatures and compared with the classical flow regime map and finally, the problems on this topic were discussed

  4. Technical and QA plan: Boiling behavior during flow instability

    International Nuclear Information System (INIS)

    Coutts, D.A.

    1991-01-01

    The coolant flow in a nuclear reactor core under normal operating conditions is kept as a subcooled liquid. This coolant is evenly distributed throughout the multiple flow channels with a uniform pressure profile across each coolant flow channel. If the coolant flow is reduced, the flow through individual channels will also decrease. A decrease in coolant flow will result in higher coolant temperatures if the heat flux is not reduced. When flow is significantly decreased, localized boiling may occur. This localized boiling can restrict coolant flow and the ability to transfer heat out of the reactor system. The maximum operating power for the reactor may be limited by how the coolant system reacts to a flow instability. One of the methods to assure safe operation during a reducing flow transient, is to operate at a power level below that necessary to initiate a flow excursion. Several correlations have been used to predict the conditions which will proceed a flow excursion. These correlations rely on the steady state behavior of the coolant and are based on steady-state testing. There are two significant points which this project will try to identify. The first is when vapor first forms on the channel surface. This might be designated as the Nucleate Vapor Transition. (Steady state equivalent is ONB). The second is when the vapor formation rate is large enough to lead to flow instability and thermal excursion. This point might be designated as the Significant Vapor Transition. (Steady state equivalent is OSV). A correlation will be developed to relate established steady state relations with the behavior of transient systems

  5. Assessment of RANS at low Prandtl number and simulation of sodium boiling flows with a CMFD code

    Energy Technology Data Exchange (ETDEWEB)

    Mimouni, S., E-mail: stephane.mimouni@edf.fr; Guingo, M.; Lavieville, J.

    2017-02-15

    Highlights: • Modelling of boiling sodium flows in a multiphase flow solver. • Rod heated with a constant heat flux in a pipe liquid metal flow. • Sodium boiling flow around a rod heated with a constant heat. • Computations in progress in an assembly constituted of 19 pins equipped with a wrapped wire. - Abstract: In France, Sodium-cooled Fast Reactors (SFR) have recently received a renewed interest. In 2006, the decision was taken by the French Government to initiate research in order to build a first Generation IV prototype (called ASTRID) by 2020. The improvement in the safety of SFR is one of the key points in their conception. Accidental sequences may lead to a significant increase of reactivity. This is for instance the case when the sodium coolant is boiling within the fissile zone. As a consequence, incipient boiling superheat of sodium is an important parameter, as it can influence boiling process which may appear during some postulated accidents as the unexpected loss of flow (ULOF). The problem is that despite the reduction in core power, when boiling conditions are reached, the flow decreases progressively and vapour expands into the heating zone. A crucial investigating way is to optimize the design of the fissile assemblies of the core in order to lead to stable boiling during a ULOF accident, without voiding of the fissile zone. Moreover, in order to evaluate nuclear plant design and safety, a CFD tool has been developed at EDF in the framework of the nuclear industry. Advanced models dedicated to boiling flows have been implemented and validated against experimental data for ten years now including a wall law for boiling flows, wall transfer for nucleate boiling, turbulence and polydispersion model. This paper aims at evaluating the generalization of these models to SFR. At least two main issues are encountered. Firstly, at low Prandtl numbers such as those of liquid metal, classical approaches derived for unity or close to unity fail to

  6. DYNAM, Once Through Boiling Flow with Steam Superheat, Laplace Transformation

    International Nuclear Information System (INIS)

    Schlueter, G.; Efferding, L.E.

    1973-01-01

    1 - Description of problem or function: DYNAM performs a dynamic analysis of once-through boiling flow oscillations with steam superheat. The model describing the superheat regime (single- phase, variable density fluid) for subcritical pressure operation is also applicable to the study of once-through operation using supercritical pressure water. 2 - Method of solution: Linearized partial differential conservation equations are solved using Laplace transformation of the temporal terms and integration of the spatial variations. DYNAM is written in complex variable notation. 3 - Restrictions on the complexity of the problem - Maxima of: 30 intervals used to describe the power distribution in the non-boiling and boiling regions, 29 boiling nodes, 7 intervals and corresponding friction multipliers read in per case, 14 exit qualities read in per case, 40 superheat nodes, 10 coefficients read in for the phi 2 vs, x-polynomial fit, 48 frequencies at which open-loop frequency response is desired, 48 frequencies at which signal output is desired

  7. Simulation of boiling flow in evaporator of separate type heat pipe with low heat flux

    International Nuclear Information System (INIS)

    Kuang, Y.W.; Wang, Wen; Zhuan, Rui; Yi, C.C.

    2015-01-01

    Highlights: • A boiling flow model in a separate type heat pipe with 65 mm diameter tube. • Nucleate boiling is the dominant mechanism in large pipes at low mass and heat flux. • The two-phase heat transfer coefficient is less sensitive to the total mass flux. - Abstract: The separate type heat pipe heat exchanger is considered to be a potential selection for developing passive cooling spent fuel pool – for the passive pressurized water reactor. This paper simulates the boiling flow behavior in the evaporator of separate type heat pipe, consisting of a bundle of tubes of inner diameter 65 mm. It displays two-phase characteristic in the evaporation section of the heat pipe working in low heat flux. In this study, the two-phase flow model in the evaporation section of the separate type heat pipe is presented. The volume of fluid (VOF) model is used to consider the interaction between the ammonia gas and liquid. The flow patterns and flow behaviors are studied and the agitated bubbly flow, churn bubbly flow are obtained, the slug bubble is likely to break into churn slug or churn froth flow. In addition, study on the heat transfer coefficients indicates that the nucleate boiling is the dominant mechanism in large pipes at low mass and heat flux, with the heat transfer coefficient being less sensitive to the total mass flux

  8. Theory of boiling-up jump

    International Nuclear Information System (INIS)

    Labuntsov, D.A.; Avdeev, A.A.

    1981-01-01

    Concept of boiling-up jump representing a zone of intense volume boiling-up separating overtaking flow of overheated metastable liquid from an area of equilibrium flow located below along the flow is introduced. It is shown that boiling-up jump is a shock wave of rarefaction. It is concluded that entropy increment occurs on the jump. Characteristics of adiabatic shock wave curve of boiling- up in ''pressure-specific volume'' coordinates have been found and its form has been investigated. Stability of boiling-up jump has been analyzed as well. On the basis of approach developed analysis is carried out on the shock adiobatic curve of condensation. Concept of boiling-up jump may be applied to the analysis of boiling-up processes when flowing liquid through packings during emergency pressure drop etc [ru

  9. RUBI -a Reference mUltiscale Boiling Investigation for the Fluid Science Laboratory

    Science.gov (United States)

    Schweizer, Nils; Stelzer, Marco; Schoele-Schulz, Olaf; Picker, Gerold; Ranebo, Hans; Dettmann, Jan; Minster, Olivier; Toth, Balazs; Winter, Josef; Tadrist, Lounes; Stephan, Peter; Grassi, Walter; di Marco, Paolo; Colin, Catherine; Piero Celata, Gian; Thome, John; Kabov, Oleg

    Boiling is a two-phase heat transfer process where large heat fluxes can be transferred with small driving temperature differences. The high performance of boiling makes the process very interesting for heat transfer applications and it is widely used in industry for example in power plants, refrigeration systems, and electronics cooling. Nevertheless, due to the large number of involved phenomena and their often highly dynamic nature a fundamental understanding and closed theoretical description is not yet accomplished. The design of systems incorporating the process is generally based on empirical correlations, which are commonly accompanied by large uncertainties and, thus, has to be verified by expensive test campaigns. Hence, strong efforts are currently made to develop applicable numerical tools for a reliable prediction of the boiling heat transfer performance and limits. In order to support and validate this development and, in particular as a precondition, to enhance the basic knowledge about boiling the comprehensive multi-scale experiment RUBI (Reference mUlti-scale Boiling Investigation) for the Fluid Science Laboratory on board the ISS is currently in preparation. The scientific objectives and requirements of RUBI have been defined by the members of the ESA topical team "Boiling and Multiphase Flow" and addresses fundamental aspects of boiling phenomena. The main objectives are the measurement of wall temperature and heat flux distribution underneath vapour bubbles with high spatial and tem-poral resolution by means of IR thermography accompanied by the synchronized high-speed observation of the bubble shapes. Furthermore, the fluid temperature in the vicinity and inside of the bubbles will be measured by a micro sensor array. Additional stimuli are the generation of an electric field above the heating surface and a shear flow created by a forced convection loop. The objective of these stimuli is to impose forces on the bubbles and investigate the

  10. Nucleate pool boiling: High gravity to reduced gravity; liquid metals to cryogens

    Science.gov (United States)

    Merte, Herman, Jr.

    1988-01-01

    Requirements for the proper functioning of equipment and personnel in reduced gravity associated with space platforms and future space station modules introduce unique problems in temperature control; power generation; energy dissipation; the storage, transfer, control and conditioning of fluids; and liquid-vapor separation. The phase change of boiling is significant in all of these. Although both pool and flow boiling would be involved, research results to date include only pool boiling because buoyancy effects are maximized for this case. The effective application of forced convection boiling heat transfer in the microgravity of space will require a well grounded and cogent understanding of the mechanisms involved. Experimental results are presented for pool boiling from a single geometrical configuration, a flat surface, covering a wide range of body forces from a/g = 20 to 1 to a/g = 0 to -1 for a cryogenic liquid, and from a/g = 20 to 1 for water and a liquid metal. Similarities in behavior are noted for these three fluids at the higher gravity levels, and may reasonably be expected to continue at reduced gravity levels.

  11. Flow Boiling Critical Heat Flux in Reduced Gravity

    Science.gov (United States)

    Mudawar, Issam; Zhang, Hui; Hasan, Mohammad M.

    2004-01-01

    This study provides systematic method for reducing power consumption in reduced gravity systems by adopting minimum velocity required to provide adequate CHF and preclude detrimental effects of reduced gravity . This study proves it is possible to use existing 1 ge flow boiling and CHF correlations and models to design reduced gravity systems provided minimum velocity criteria are met

  12. Continuous vs. pulsating flow boiling. Part 2: Statistical comparison using response surface methodology

    DEFF Research Database (Denmark)

    Kærn, Martin Ryhl; Elmegaard, Brian; Meyer, Knud Erik

    2016-01-01

    Response surface methodology is used to investigate an active method for flow boiling heat transfer enhancement by means of fluid flow pulsation. The flow pulsations are introduced by a flow modulating expansion device and compared with the baseline continuous flow provided by a stepper...

  13. Experimental investigation of the effect of an electric field on heat transfers at boiling point for a high-resistivity water in forced convection

    International Nuclear Information System (INIS)

    Morin, Henri; Verdier, Jacques

    1964-10-01

    The enhancement of heat exchanges with boiling water in forced convection in an annular duct is studied when applying an electric field between the two walls of the duct. At the local boiling and at saturation temperature, for a water resistivity comprised between 0.5 and 1 M Ω cm, with fields on the cylindrical interior surface of the canal comprised between 4 and 8 kV/cm, significant enhancements of the exchanged heat fluxes are noticed, 2.5 to 10 time superior to exchanges without electric field. When heating, heat fluxes may be increased from two to three times [fr

  14. Modelling of boiling bubbly flows using a polydisperse approach

    International Nuclear Information System (INIS)

    Zaepffel, D.

    2011-01-01

    The objective of this work was to improve the modelling of boiling bubbly flows.We focused on the modelling of the polydisperse aspect of a bubble population, i.e. the fact that bubbles have different sizes and different velocities. The multi-size aspect of a bubble population can originate from various mechanisms. For the bubbly flows we are interested in, bubble coalescence, bubble break-up, phase change kinematics and/or gas compressibility inside the bubbles can be mentioned. Since, bubble velocity depends on bubble size, the bubble size spectrum also leads to a bubble velocity spectrum. An averaged model especially dedicated to dispersed flows is introduced in this thesis. Closure of averaged interphase transfer terms are written in a polydisperse framework, i.e. using a distribution function of the bubble sizes and velocities. A quadratic law and a cubic law are here proposed for the modelling of the size distribution function, whose evolution in space and time is then obtained with the use of the moment method. Our averaged model has been implemented in the NEPTUNE-CFD computation code in order to simulate the DEBORA experiment. The ability of our model to deal with sub-cooled boiling flows has therefore been evaluated. (author) [fr

  15. Burnout in a high heat flux boiling system with forced supply of liquid through a plane jet

    International Nuclear Information System (INIS)

    Katto, Yoshiro; Ishii, Kazunori.

    1978-01-01

    As for pool boiling, the non-dimensional formula for the burnout heat flux of a simple, basic boiling system has been obtained. On the other hand, in forced convection boiling, the studies on the burnout in forced flow boiling in a channel have been continued, but the derivation of a non-dimensional formula applicable generally is far away from the realization because the phenomena are too complex. Accordingly, in this study, the result of the experiment on the burnout of a boiling system to which liquid is supplied by the plane jet flowing out of a thin rectangular nozzle installed near the front edge of a rectangular heating surface is reported. The experimental apparatus is described, and the experiment was carried out in the ranges of two jet thicknesses at the nozzle outlet, two incident angles of jet and from 1.5 to 15 m/s of jet velocity. Burnout occurs under the situation of sufficiently developed nuclear boiling. A part of the liquid supplied from a plane jet is blown apart by the vapor blowing out of the nuclear boiling liquid layer covering the heating surface in the nuclear boiling with sufficiently developed high heat flux. However, the nuclear boiling liquid layer itself continues to exist on the heating surface till burnout occurs. Only the entering velocity of the plane jet affects burnout heat flux. (Kako, I.)

  16. Theoretical investigation of flow regime for boiling water two-phase flow in horizontal rectangular narrow channels

    International Nuclear Information System (INIS)

    Zhang Chunwei; Qiu Suizheng; Yan Mingyu; Wang Bulei; Nie Changhua

    2005-01-01

    The flow regime transition criteria for the boiling water two-phase flow in horizontal rectangular narrow channels (1 x 20 mm, 2 x 20 mm) were theoretically explored. The discernible flow patterns were bubble, intermittent slug, churn, annular and steam-water separation flow. By using two-fluid model, equations of conservation of momentum were established for the two-phase flow. New flow-regime criteria were obtained and agreed well with the experiment data. (authors)

  17. Prediction of subcooled flow boiling characteristics using two-fluid Eulerian CFD model

    Energy Technology Data Exchange (ETDEWEB)

    Braz Filho, Francisco A.; Ribeiro, Guilherme B., E-mail: gbribeiro@ieav.cta.br; Caldeira, Alexandre D.

    2016-11-15

    Highlights: • CFD multiphase model is used to predict subcooled flow boiling characteristics. • Better agreement is achieved for higher saturation pressures. • Onset of nucleate boiling and saturated boiling are well predicted. • CFD multiphase model tends to underestimate the void fraction. • Factors were adjusted in order to improve the void fraction results. - Abstract: The present study concerns a detailed analysis of flow boiling phenomena under high pressure systems using a two-fluid Eulerian approach provided by a Computational Fluid Dynamics (CFD) solver. For this purpose, a vertical heated pipe made of stainless steel with an internal diameter of 15.4 mm was considered as the modeled domain. Two different uniform heat fluxes and three saturation pressures were applied to the channel wall, whereas water mass flux of 900 kg/m{sup 2} s was considered for all simulation cases. The model was validated against a set of experimental data and results have indicated a promising use of the CFD technique for estimation of the wall temperature, the liquid bulk temperature and the location of the departure of nucleate boiling. Changes in factors applied in the modeling of the interfacial heat transfer coefficient and bubble departure frequency were suggested, allowing a better prediction of the void fraction along the heated channel. The commercial CFD solver FLUENT 14.5 was used for the model implementation.

  18. Prediction of subcooled flow boiling characteristics using two-fluid Eulerian CFD model

    International Nuclear Information System (INIS)

    Braz Filho, Francisco A.; Ribeiro, Guilherme B.; Caldeira, Alexandre D.

    2016-01-01

    Highlights: • CFD multiphase model is used to predict subcooled flow boiling characteristics. • Better agreement is achieved for higher saturation pressures. • Onset of nucleate boiling and saturated boiling are well predicted. • CFD multiphase model tends to underestimate the void fraction. • Factors were adjusted in order to improve the void fraction results. - Abstract: The present study concerns a detailed analysis of flow boiling phenomena under high pressure systems using a two-fluid Eulerian approach provided by a Computational Fluid Dynamics (CFD) solver. For this purpose, a vertical heated pipe made of stainless steel with an internal diameter of 15.4 mm was considered as the modeled domain. Two different uniform heat fluxes and three saturation pressures were applied to the channel wall, whereas water mass flux of 900 kg/m"2 s was considered for all simulation cases. The model was validated against a set of experimental data and results have indicated a promising use of the CFD technique for estimation of the wall temperature, the liquid bulk temperature and the location of the departure of nucleate boiling. Changes in factors applied in the modeling of the interfacial heat transfer coefficient and bubble departure frequency were suggested, allowing a better prediction of the void fraction along the heated channel. The commercial CFD solver FLUENT 14.5 was used for the model implementation.

  19. Burnout heat flux in natural flow boiling

    International Nuclear Information System (INIS)

    Helal, M.M.; Darwish, M.A.; Mahmoud, S.I.

    1978-01-01

    Twenty runs of experiments were conducted to determine the critical heat flux for natural flow boiling with water flowing upwards through annuli of centrally heated stainless steel tube. The test section has concentric heated tube of 14mm diameter and heated lengthes of 15 and 25 cm. The outside surface of the annulus was formed by various glass tubes of 17.25, 20 and 25.9mm diameter. System pressure is atmospheric. Inlet subcooling varied from 18 to 5 0 C. Obtained critical heat flux varied from 24.46 to 62.9 watts/cm 2 . A number of parameters having dominant influence on the critical heat flux and hydrodynamic instability (flow and pressure oscillations) preceeding the burnout have been studied. These parameters are mass flow rate, mass velocity, throttling, channel geometry (diameters ratio, length to diameter ratio, and test section length), and inlet subcooling. Flow regimes before and at the moments of burnout were observed, discussed, and compared with the existing physical model of burnout

  20. Feasibility of maintaining natural convection mode core cooling in research reactor power upgrades

    International Nuclear Information System (INIS)

    Ha, J.J.; Belhadj, M.; Aldemir, T.; Christensen, R.N.

    1987-01-01

    Two operational concerns for natural convection coooled research reactors using plate type fuels are: 1) pool top 16 N activity (PTNA), and 2) nucleate boiling in core channels. The feasibility assessment of a power upgrade while maintaining natural convection mode core cooling requires addressing these operational concerns. Previous studies have shown that: a) The conventional technique for reducing PTNA by plume dispersion may not be effective in a large power upgrade of research reactors with small pools. b) Currently used correlations to predict onset of nucleate boiling (ONB) in thin, rectangular core channels are not valid for low-velocity, upward flows such as encountered in natural convection cooling. The PTNA depends on the velocity distribution in the reactor pool. COMMIX-1A code is used to determine the three-dimensional velocity fields in The Ohio State University Research Reactor (OSURR) pool as a function of varying design conditions, following a power upgrade to 500 kW with LEU fuel. It is shown that a sufficiently deep stagnant water layer can be created below the pool top by properly choosing the disperser flow rate. The ONB heat flux is experimentally determined for channel gaps and upward flow velocities in the range 2mm-4mm and 3-16 cm/sec., respectively. Two alternatives to plume dispersion for reducing PTNA and a new correlation to determine the ONB heat flux in thin, rectangular channels under low-velocity, upward flow conditions are proposed. (Author)

  1. Temperature and flow fluctuations under local boiling in a simulated fuel subassembly

    International Nuclear Information System (INIS)

    Inujima, H.; Ogino, T.; Uotani, M.; Yamaguchi, K.

    1980-08-01

    Out-of-pile experiments were carried out with the sodium test loop SIENA in O-arai Engineering Center of PNC, and the feasibility studies had been made on the local boiling detection by use of temperature and flow fluctuations. The studies showed that the temperature fluctuation transferred the information on local boiling toward the end of the bundle, but hardly to the outlet. In addition, it was proved that the anomaly detection method, which used the algorithm of whiteness test method to the residual time series data of autoregressive model, is an effective one for detecting anomaly such as local boiling. (author)

  2. Gravity influence on heat transfer rate in flow boiling

    NARCIS (Netherlands)

    Baltis, C.H.M.; Celata, G.P.; Cumo, M.; Saraceno, L.; Zummo, G.

    2012-01-01

    The aim of the present paper is to describe the results of flow boiling heat transfer at low gravity and compare them with those obtained at earth gravity, evaluating possible differences. The experimental campaigns at low gravity have been performed with parabolic flights. The paper will show the

  3. Dispersed flow film boiling

    International Nuclear Information System (INIS)

    Andreani, M.; Yadigaroglu, G.

    1989-12-01

    Dispersed flow film boiling is the heat transfer regime that occurs at high void fractions in a heated channel. The way this transfer mode is modelled in the NRC computer codes (RELAP5 and TRAC) and the validity of the assumption and empirical correlations used is discussed. An extensive review of the theoretical and experimental work related with heat transfer to highly dispersed mixtures reveals the basic deficiencies of these models: the investigation refers mostly to the typical conditions of low rate bottom reflooding, since the simulation of this physical situation by the computer codes has often showed poor results. The alternative models that are available in the literature are reviewed, and their merits and limits are highlighted. The modification that could improve the physics of the models implemented in the codes are identified. (author) 13 figs., 123 refs

  4. A study on the effects of heated surface wettability on nucleation characteristics in subcooled flow boiling

    International Nuclear Information System (INIS)

    Kajihara, Tomoyuki; Kaiho, Kazuhiro; Okawa, Tomio

    2014-01-01

    Subcooled flow boiling plays an important role in boiling water reactors because it influences the heat transfer performance from fuel rods, two-phase flow stabilities, and neutron moderation characteristics. In the present study, flow visualization of water subcooled flow boiling in a vertical heated channel was carried out to investigate the mechanisms of void fraction development. The two surfaces of distinctly different contact angles were used as the heated surface to investigate the effect of the surface wettability. It was observed that with an increase in the wall heat flux, more nucleation sites were activated and larger bubbles were produced at low-frequency. It was considered that formation of these large bubbles primarily contributed to the void fraction development. (author)

  5. Advanced modeling of the size poly-dispersion of boiling flows

    International Nuclear Information System (INIS)

    Ruyer, Pierre; Seiler, Nathalie

    2008-01-01

    Full text of publication follows: This work has been performed within the Institut de Radioprotection et de Surete Nucleaire that leads research programs concerning safety analysis of nuclear power plants. During a LOCA (Loss Of Coolant Accident), in-vessel pressure decreases and temperature increases, leading to the onset of nucleate boiling. The present study focuses on the numerical simulation of the local topology of the boiling flow. There is experimental evidence of a local and statistical large spectra of possible bubble sizes. The relative importance of the correct description of this poly-dispersion in size is due to the dependency of (i) main hydrodynamic forces, like lift, as well as of (ii) transfer area with respect to the individual bubble size. We study the corresponding CFD model in the framework of an ensemble averaged description of the dispersed two-phase flow. The transport equations of the main statistical moment densities of the population size distribution are derived and models for the mass, momentum and heat transfers at the bubble scale as well as for bubble coalescence are achieved. This model introduced within NEPTUNE-CFD code of the NEPTUNE thermal-hydraulic platform, a joint project of CEA, EDF, IRSN and AREVA, has been tested on boiling flows obtained on the DEBORA facility of the CEA at Grenoble. These numerical simulations provide a validation and attest the impact of the proposed model. (authors) [fr

  6. Investigation of Body Force Effects on Flow Boiling Critical Heat Flux

    Science.gov (United States)

    Zhang, Hui; Mudawar, Issam; Hasan, Mohammad M.

    2002-01-01

    The bubble coalescence and interfacial instabilities that are important to modeling critical heat flux (CHF) in reduced-gravity systems can be sensitive to even minute body forces. Understanding these complex phenomena is vital to the design and safe implementation of two-phase thermal management loops proposed for space and planetary-based thermal systems. While reduced gravity conditions cannot be accurately simulated in 1g ground-based experiments, such experiments can help isolate the effects of the various forces (body force, surface tension force and inertia) which influence flow boiling CHF. In this project, the effects of the component of body force perpendicular to a heated wall were examined by conducting 1g flow boiling experiments at different orientations. FC-72 liquid was boiled along one wall of a transparent rectangular flow channel that permitted photographic study of the vapor-liquid interface at conditions approaching CHF. High-speed video imaging was employed to capture dominant CHF mechanisms. Six different CHF regimes were identified: Wavy Vapor Layer, Pool Boiling, Stratification, Vapor Counterflow, Vapor Stagnation, and Separated Concurrent Vapor Flow. CHF showed great sensitivity to orientation for flow velocities below 0.2 m/s, where very small CHF values where measured, especially with downflow and downward-facing heated wall orientations. High flow velocities dampened the effects of orientation considerably. Figure I shows representative images for the different CHF regimes. The Wavy Vapor Layer regime was dominant for all high velocities and most orientations, while all other regimes were encountered at low velocities, in the downflow and/or downward-facing heated wall orientations. The Interfacial Lift-off model was modified to predict the effects of orientation on CHF for the dominant Wavy Vapor Layer regime. The photographic study captured a fairly continuous wavy vapor layer travelling along the heated wall while permitting liquid

  7. Study on Enhancement of Sub-Cooled Flow Boiling Heat Transfer and Critical Heat Flux of Solid-Water Two-Phase Mixture

    International Nuclear Information System (INIS)

    Yasuo Koizumi; Hiroyasu Ohtake; Tomoyuki Suzuki

    2002-01-01

    The influence of particle introduction into a subcooled water flow on boiling heat transfer and critical heat flux (CHF) was examined. When the water velocity was low, the particles crowded on the bottom wall of the flow channel and flowed just like sliding on the wall. When the water velocity was high, the particles were well dispersed in the water flow. In the non-boiling region, the heat transfer was augmented by the introduction of the particles into the water flow. As the introduction of the particles were increased, the augmentation was also increased in the high water flow rate region. However, it was independent upon the particle introduction rate in the low water flow rate region. The onset of boiling was delayed by the particle inclusion. The boiling heat transfer was enhanced by the particles. However, it was rather decreased in the high heat flux fully-developed-boiling region. The CHF was decreased by the particle inclusion in the low water flow region and was not affected in the high water flow region. (authors)

  8. Heat Transfer Characteristics during Boiling of Immiscible Liquids Flowing in Narrow Rectangular Heated Channels

    Directory of Open Access Journals (Sweden)

    Yasuhisa Shinmoto

    2017-11-01

    Full Text Available The use of immiscible liquids for cooling of surfaces with high heat generation density is proposed based on the experimental verification of its superior cooling characteristics in fundamental systems of pool boiling and flow boiling in a tube. For the purpose of practical applications, however, heat transfer characteristics due to flow boiling in narrow rectangular channels with different small gap sizes need to be investigated. The immiscible liquids employed here are FC72 and water, and the gap size is varied as 2, 1, and 0.5 mm between parallel rectangular plates of 30 mm × 175 mm, where one plate is heated. To evaluate the effect of gap size, the heat transfer characteristics are compared at the same inlet velocity. The generation of large flattened bubbles in a narrow gap results in two opposite trends of the heat transfer enhancement due to thin liquid film evaporation and of the deterioration due to the extension of dry patch in the liquid film. The situation is the same as that observed for pure liquids. The latter negative effect is emphasized for extremely small gap sizes if the flow rate ratio of more-volatile liquid to the total is not reduced. The addition of small flow rate of less-volatile liquid can increase the critical heat flux (CHF of pure more-volatile liquid, while the surface temperature increases at the same time and assume the values between those for more-volatile and less-volatile liquids. By the selection of small flow rate ratio of more-volatile liquid, the surface temperature of pure less-volatile liquid can be decreased without reducing high CHF inherent in the less-volatile liquid employed. The trend of heat transfer characteristics for flow boiling of immiscible mixtures in narrow channels is more sensitive to the composition compared to the flow boiling in a round tube.

  9. Experimental study on saturated boiling of two phase natural circulation under low pressure in narrow rectangular channels

    Energy Technology Data Exchange (ETDEWEB)

    Li, Zi-chao; Qi, Shi; Zhou, Tao; Li, Bing; Shahzad, Muhammad Ali [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering; Beijing Key Laboratory of Passive Safety Technology for Nuclear Energy, Beijing (China); Huang, Yan-ping [Nuclear Reactor Thermal Hydraulics Technology, Chengdu (China). CNNC Key Lab.

    2017-12-15

    Saturated boiling of two-phase natural circulation has been experimentally investigated based on a natural circulation device with narrow rectangular channels. When heating power reaches a certain range, it is possible to observe the phenomenon of saturated boiling and flow pattern transition in the system. The results show the heat transfer coefficient of saturated boiling decreases with the increasing of pressure, heating power and size of narrow rectangle channels. The buoyancy force causing mixed convection decreases the heat transfer coefficient. Finally, a dimensionless number is introduced, which reflects length to width ratio of rectangular narrow section and Rayleigh number, in order to revise the presented correlation. All errors fall within the range of ±15%.

  10. Numerical simulation in a subcooled water flow boiling for one-sided high heat flux in reactor divertor

    Energy Technology Data Exchange (ETDEWEB)

    Liu, P., E-mail: pinliu@aust.edu.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); School of Mechanical Engineering, Anhui University of Science and Technology, Huainan 232001 (China); Peng, X.B., E-mail: pengxb@ipp.ac.cn [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); Song, Y.T. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China); University of Science and Technology of China, Hefei 230026 (China); Fang, X.D. [Institute of Air Conditioning and Refrigeration, Nanjing University of Aeronautics and Astronautics, Nanjing 210016 (China); Huang, S.H. [University of Science and Technology of China, Hefei 230026 (China); Mao, X. [Institute of Plasma Physics, Chinese Academy of Sciences, Hefei 230031 (China)

    2016-11-15

    Highlights: • The Eulerian multiphase models coupled with Non-equilibrium Boiling model can effectively simulate the subcooled water flow boiling. • ONB and FDB appear earlier and earlier with the increase of heat fluxes. • The void fraction increases gradually along the flow direction. • The inner CuCrZr tube deteriorates earlier than the outer tungsten layer and the middle OFHC copper layer. - Abstract: In order to remove high heat fluxes for plasma facing components in International Thermonuclear Experimental Reactor (ITER) divertor, a numerical simulation of subcooled water flow boiling heat transfer in a vertically upward smooth tube was conducted in this paper on the condition of one-sided high heat fluxes. The Eulerian multiphase model coupled with Non-equilibrium Boiling model was adopted in numerical simulation of the subcooled boiling two-phase flow. The heat transfer regions, thermodynamic vapor quality (x{sub th}), void fraction and temperatures of three components on the condition of the different heat fluxes were analyzed. Numerical results indicate that the onset of nucleate boiling (ONB) and fully developed boiling (FDB) appear earlier and earlier with increasing heat flux. With the increase of heat fluxes, the inner CuCrZr tube will deteriorate earlier than the outer tungsten layer and the middle oxygen-free high-conductivity (OFHC) copper layer. These results provide a valuable reference for the thermal-hydraulic design of a water-cooled W/Cu divertor.

  11. Multi-scale full-field measurements and near-wall modeling of turbulent subcooled boiling flow using innovative experimental techniques

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin A., E-mail: y-hassan@tamu.edu

    2016-04-01

    Highlights: • Near wall full-field velocity components under subcooled boiling were measured. • Simultaneous shadowgraphy, infrared thermometry wall temperature and particle-tracking velocimetry techniques were combined. • Near wall velocity modifications under subcooling boiling were observed. - Abstract: Multi-phase flows are one of the challenges on which the CFD simulation community has been working extensively with a relatively low success. The phenomena associated behind the momentum and heat transfer mechanisms associated to multi-phase flows are highly complex requiring resolving simultaneously for multiple scales on time and space. Part of the reasons behind the low predictive capability of CFD when studying multi-phase flows, is the scarcity of CFD-grade experimental data for validation. The complexity of the phenomena and its sensitivity to small sources of perturbations makes its measurements a difficult task. Non-intrusive and innovative measuring techniques are required to accurately measure multi-phase flow parameters while at the same time satisfying the high resolution required to validate CFD simulations. In this context, this work explores the feasible implementation of innovative measuring techniques that can provide whole-field and multi-scale measurements of two-phase flow turbulence, heat transfer, and boiling parameters. To this end, three visualization techniques are simultaneously implemented to study subcooled boiling flow through a vertical rectangular channel with a single heated wall. These techniques are listed next and are used as follow: (1) High-speed infrared thermometry (IR-T) is used to study the impact of the boiling level on the heat transfer coefficients at the heated wall, (2) Particle Tracking Velocimetry (PTV) is used to analyze the influence that boiling parameters have on the liquid phase turbulence statistics, (3) High-speed shadowgraphy with LED illumination is used to obtain the gas phase dynamics. To account

  12. Laminar flow and convective transport processes scaling principles and asymptotic analysis

    CERN Document Server

    Brenner, Howard

    1992-01-01

    Laminar Flow and Convective Transport Processes: Scaling Principles and Asymptotic Analysis presents analytic methods for the solution of fluid mechanics and convective transport processes, all in the laminar flow regime. This book brings together the results of almost 30 years of research on the use of nondimensionalization, scaling principles, and asymptotic analysis into a comprehensive form suitable for presentation in a core graduate-level course on fluid mechanics and the convective transport of heat. A considerable amount of material on viscous-dominated flows is covered.A unique feat

  13. Boiling, condensation, and gas-liquid flow

    International Nuclear Information System (INIS)

    Whalley, P.B.

    1987-01-01

    Heat transfer phenomena involving boiling and condensation are an important aspect of engineering in the power and process industries. This book, aimed at advanced first-degree and graduate students in mechanical and chemical engineering, deals with these phenomena in detail. The first part of the book describes gas-liquid two-phase flow, as a necessary preliminary to the later discussion of heat transfer and change of phase. A detailed section on calculation methods shows how theory can be put to practical use, and there are also descriptions of some of the equipment and plant used in the process and power industries

  14. Heat transfer by natural convection into an horizontal cavity; Transferencia de calor por conveccion natural en una cavidad horizontal

    Energy Technology Data Exchange (ETDEWEB)

    Arevalo J, P

    1998-12-31

    At this thesis it is studied the heat transfer by natural convection in an horizontal cavity, it is involved a boiling`s part that is described the regimes and correlations differences for boiling`s curve. It is designed a horizontal cavity for realize the experimental part and it`s mention from equipment or instrumentation to succeed in a experimentation that permits to realize the analysis of heat transfer, handling as water fluid at atmospheric pressure and where it`s present process from natural convection involving part boiling`s subcooled. The system consists of heater zone submerged in a horizontal cavity with water. Once part finished experimental with information to obtained it`s proceeded to obtain a correlation, realized starting from analysis dimensionless such as: Jakob, Bond and Grasoft (Boiling) besides of knows in natural convection: Prandtl and Nusselt. The mathematical model explains the behavior for natural convection continued part boiling`s subcooled. It is realize analysis graphics too where it`s show comparing with Globe Dropkin and Catton equations by natural convection with bottom heating. (Author)

  15. Heat transfer by natural convection into an horizontal cavity; Transferencia de calor por conveccion natural en una cavidad horizontal

    Energy Technology Data Exchange (ETDEWEB)

    Arevalo J, P

    1999-12-31

    At this thesis it is studied the heat transfer by natural convection in an horizontal cavity, it is involved a boiling`s part that is described the regimes and correlations differences for boiling`s curve. It is designed a horizontal cavity for realize the experimental part and it`s mention from equipment or instrumentation to succeed in a experimentation that permits to realize the analysis of heat transfer, handling as water fluid at atmospheric pressure and where it`s present process from natural convection involving part boiling`s subcooled. The system consists of heater zone submerged in a horizontal cavity with water. Once part finished experimental with information to obtained it`s proceeded to obtain a correlation, realized starting from analysis dimensionless such as: Jakob, Bond and Grasoft (Boiling) besides of knows in natural convection: Prandtl and Nusselt. The mathematical model explains the behavior for natural convection continued part boiling`s subcooled. It is realize analysis graphics too where it`s show comparing with Globe Dropkin and Catton equations by natural convection with bottom heating. (Author)

  16. Difficulties in modeling dispersed-flow film boiling

    International Nuclear Information System (INIS)

    Andreani, M.; Yadigaroglu, G.

    1991-01-01

    Dispersed Flow Film Boiling (DFFB) is characterized by important departures from thermal and velocity equilibrium that make it suitable for modeling with two-fluid models. The fundamental limitations and difficulties imposed by the one-dimensional nature of these models are extensively discussed. The validity of the assumptions and empirical laws used to close the system of conservation equations is critically reviewed, in light of the multidimensional aspects of the problem. Modifications that could improve the physics of the models are identified. (orig.) [de

  17. Transient boiling in two-phase helium natural circulation loops

    Science.gov (United States)

    Furci, H.; Baudouy, B.; Four, A.; Meuris, C.

    2014-01-01

    Two-phase helium natural circulation loops are used for cooling large superconducting magnets, as CMS for LHC. During normal operation or in the case of incidents, transients are exerted on the cooling system. Here a cooling system of this type is studied experimentally. Sudden power changes are operated on a vertical-heated-section natural convection loop, simulating a fast increase of heat deposition on magnet cooling pipes. Mass flow rate, heated section wall temperature and pressure drop variations are measured as a function of time, to assess the time behavior concerning the boiling regime according to the values of power injected on the heated section. The boiling curves and critical heat flux (CHF) values have been obtained in steady state. Temperature evolution has been observed in order to explore the operating ranges where heat transfer is deteriorated. Premature film boiling has been observed during transients on the heated section in some power ranges, even at appreciably lower values than the CHF. A way of attenuating these undesired temperature excursions has been identified through the application of high enough initial heating power.

  18. Verification and validation of one-dimensional models used in subcooled flow boiling analysis

    International Nuclear Information System (INIS)

    Braz Filho, Francisco A.; Caldeira, Alexandre D.; Borges, Eduardo M.; Sabundjian, Gaiane

    2009-01-01

    Subcooled flow boiling occurs in many industrial applications and it is characterized by large heat transfer coefficients. However, this efficient heat transfer mechanism is limited by the critical heat flux, where the heat transfer coefficient decreases leading to a fast heater temperature excursion, potentially leading to heater melting and destruction. Subcooled flow boiling is especially important in water-cooled nuclear power reactors, where the presence of vapor bubbles in the core influences the reactor system behavior at operating and accident conditions. With the aim of verifying the subcooled flow boiling calculation models of the most important nuclear reactor thermal-hydraulic computer codes, such as RELAP5, COBRA-EN and COTHA-2tp, the main purpose of this work is to compare experimental data with results from these codes in the pressure range between 15 and 45 bar. For the pressure of 45 bar the results are in good agreement, while for low pressures (15 and 30 bar) the results start to become conflicting. Besides, as a sub-product of this analysis, a comparison among the models is also presented. (author)

  19. Convective boiling in a parallel microchannel heat sink with a diverging cross-section design and artificial nucleation sites

    International Nuclear Information System (INIS)

    Lu, Chun Ting; Pan, Chin

    2009-01-01

    To develop a highly stable boiling heat transfer microchannel heat sink, the three types of diverging microchannels, namely Type-1, Type-2 and Type-3, were designed to explore experimentally the effect of different distribution of artificial nucleation sites on enhancing boiling heat transfer in 10 parallel diverging microchannels with a mean hydraulic diameter of 120 μm. The Type-1 system is with no cavities, Type-2 is with cavities distributed uniformly along the downstream half of the channel, while Type-3 is with cavities distributed uniformly along the whole channel. The artificial nucleation sites are laser-etched pits on the channel bottom wall with a mouth diameter of about 20-22 μm based on the heterogeneous nucleation theory. The results of the present study reveal the presence of the artificial nucleation sites for flow boiling in parallel diverging microchannel significantly reduces the wall superheat and enhances the boiling heat transfer performance. Additionally, the Type-3 design demonstrates the best boiling heat transfer performance. (author)

  20. Industrial aspects of gas-liquid two-phase flow

    International Nuclear Information System (INIS)

    Hewitt, G.F.

    1977-01-01

    The lecture begins by reviewing the various types of plant in which two phase flow occurs. Specifically, boiling plant, condensing plant and pipelines are reviewed, and the various two phase flow problems occurring in them are described. Of course, many other kinds of chemical engineering plant involve two phase flow, but are somewhat outside the scope of this lecture. This would include distillation columns, vapor-liquid separators, absorption towers etc. Other areas of industrial two phase flow which have been omitted for space reasons from this lecture are those concerned with gas/solids, liquid/solid and liquid/liquid flows. There then follows a description of some of the two phase flow processes which are relevant in industrial equipment and where special problems occur. The topics chosen are as follows: (1) pressure drop; (2) horizontal tubes - separation effects non-uniformites in heat transfer coefficient, effect of bends on dryout; (3) multicomponent mixtures - effects in pool boiling, mass transfer effects in condensation and Marangoni effects; (4) flow distribution - manifold problems in single phase flow, separation effects at a single T-junction in two phase flow and distribution in manifolds in two phase flow; (5) instability - oscillatory instability, special forms of instability in cryogenic systems; (6) nucleate boiling - effect of variability of surface, unresolved problems in forced convective nucleate boiling; and (7) shell side flows - flow patterns, cross flow boiling, condensation in cross flow

  1. Parameters, which effect the mass flow in the PRHRS under a natural convection condition

    International Nuclear Information System (INIS)

    Chung, Y. J.; Lee, G. H.; Kim, H. C.; Kim, K. K.; Zee, S. Q.

    2004-01-01

    Small and medium sized integral type reactors for the diverse utilization of nuclear energy are getting much attention from the international nuclear community. They diversify the peaceful uses of nuclear energy in the areas of seawater desalination, district heating, industrial heat-generation process and ship propulsion. The SMART (System integrated Modular Advanced ReacTor) is a small modular integral type pressurized water reactor, which was developed for the dual purposes application of seawater desalination and small-scaled power generation in KOREA. The reactor is designed for a forced convection core cooling during start-up and normal operating conditions and for a natural circulation core cooling during accidental conditions. The main safety objective of the SMART is to increase the degree of inherent safety features by advanced designs such as a passive residual heat removal system (PRHRS). The passive residual heat removal system removes the core decay heat and sensible heat by a natural circulation in the case of emergency conditions. This study focuses on the flow behavior in the passive residual heat removal system of the integral reactor. The system necessitates a hydraulic head to achieve the required natural circulation flow rate, which in turn, may cause a larger two-phase pressure drop and flow oscillation. Also, it is of interest to investigate the complex effects of the boiling and condensation in such low frequency thermo-hydraulic oscillations. Thermal hydraulic analysis for the passive residual heat removal system has been carried out by means of the MARS code for a full range of reactor operating conditions. The MARS code has been developed at the Korea Atomic Energy Research Institute by consolidating and restructuring the RELAP5/MOD3.2 and COBRA-TF which has the capabilities of analyzing the one-dimensional or three-dimensional best estimated thermal-hydraulic system and the fuel responses of the light water reactor transients. A selected

  2. Proposal of experimental setup on boiling two-phase flow on-orbit experiments onboard Japanese experiment module "KIBO"

    Science.gov (United States)

    Baba, S.; Sakai, T.; Sawada, K.; Kubota, C.; Wada, Y.; Shinmoto, Y.; Ohta, H.; Asano, H.; Kawanami, O.; Suzuki, K.; Imai, R.; Kawasaki, H.; Fujii, K.; Takayanagi, M.; Yoda, S.

    2011-12-01

    Boiling is one of the efficient modes of heat transfer due to phase change, and is regarded as promising means to be applied for the thermal management systems handling a large amount of waste heat under high heat flux. However, gravity effects on the two-phase flow phenomena and corresponding heat transfer characteristics have not been clarified in detail. The experiments onboard Japanese Experiment Module "KIBO" in International Space Station on boiling two-phase flow under microgravity conditions are proposed to clarify both of heat transfer and flow characteristics under microgravity conditions. To verify the feasibility of ISS experiments on boiling two-phase flow, the Bread Board Model is assembled and its performance and the function of components installed in a test loop are examined.

  3. Subcooled flow boiling heat transfer of dilute alumina, zinc oxide, and diamond nanofluids at atmospheric pressure

    International Nuclear Information System (INIS)

    Kim, Sung Joong; McKrell, Tom; Buongiorno, Jacopo; Hu Linwen

    2010-01-01

    A nanofluid is a colloidal suspension of nano-scale particles in water, or other base fluids. Previous pool boiling studies have shown that nanofluids can improve the critical heat flux (CHF) by as much as 200%. In a previous paper, we reported on subcooled flow boiling CHF experiments with low concentrations of alumina, zinc oxide, and diamond nanoparticles in water (≤0.1% by volume) at atmospheric pressure, which revealed a substantial CHF enhancement (∼40-50%) at the highest mass flux (G = 2500 kg/m 2 s) and concentration (0.1 vol.%) for all nanoparticle materials (). In this paper, we focus on the flow boiling heat transfer coefficient data collected in the same tests. It was found that for comparable test conditions the values of the nanofluid and water heat transfer coefficient are similar (within ±20%). The heat transfer coefficient increased with mass flux and heat flux for water and nanofluids alike, as expected in flow boiling. A confocal microscopy-based examination of the test section revealed that nanoparticle deposition on the boiling surface occurred during nanofluid boiling. Such deposition changes the number of micro-cavities on the surface, but also changes the surface wettability. A simple model was used to estimate the ensuing nucleation site density changes, but no definitive correlation between the nucleation site density and the heat transfer coefficient data could be found.

  4. An experimental study on micro-scale flow boiling heat transfer

    International Nuclear Information System (INIS)

    Tibirica, Cristiano Bigonha; Ribatski, Gherhardt

    2009-01-01

    In this paper, new experimental flow boiling heat transfer results in micro-scale tubes are presented. The experimental data were obtained in a horizontal 2.32 mm I.D. stainless steel tube with heating length of 464 mm, R134a as working fluid, mass velocities ranging from 50 to 600 kg/m 2 s, heat flux from 5 to 55 kW/m 2 , exit saturation temperatures of 22, 31 and 41 deg C, and vapor qualities from 0.05 to 0.98. Flow pattern characterization was also performed from images obtained by high speed filming. Heat transfer coefficient results from 2 to 14 kW/m 2 K were measured. It was found that the heat transfer coefficient is a strong function of the saturation pressure, heat flux, mass velocity and vapor quality. The experimental data were compared against the following micro-scale flow boiling predictive methods from the literature: Saitoh et al., Kandlikar, Zhang et al. and Thome et al. Comparisons against these methods based on the data segregated according to flow patterns were also performed. Though not satisfactory, Saitoh et al. worked the best and was able of capturing most of the experimental heat transfer trends. (author)

  5. An experimental study on micro-scale flow boiling heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Tibirica, Cristiano Bigonha; Ribatski, Gherhardt [Universidade de Sao Paulo (USP), Sao Carlos, SP (Brazil). Escola de Engenharia. Dept. de Engenharia Mecanica

    2009-07-01

    In this paper, new experimental flow boiling heat transfer results in micro-scale tubes are presented. The experimental data were obtained in a horizontal 2.32 mm I.D. stainless steel tube with heating length of 464 mm, R134a as working fluid, mass velocities ranging from 50 to 600 kg/m{sup 2}s, heat flux from 5 to 55 kW/m{sup 2}, exit saturation temperatures of 22, 31 and 41 deg C, and vapor qualities from 0.05 to 0.98. Flow pattern characterization was also performed from images obtained by high speed filming. Heat transfer coefficient results from 2 to 14 kW/m{sup 2}K were measured. It was found that the heat transfer coefficient is a strong function of the saturation pressure, heat flux, mass velocity and vapor quality. The experimental data were compared against the following micro-scale flow boiling predictive methods from the literature: Saitoh et al., Kandlikar, Zhang et al. and Thome et al. Comparisons against these methods based on the data segregated according to flow patterns were also performed. Though not satisfactory, Saitoh et al. worked the best and was able of capturing most of the experimental heat transfer trends. (author)

  6. Investigation on the heat transfer characteristics during flow boiling of liquefied natural gas in a vertical micro-fin tube

    Science.gov (United States)

    Xu, Bin; Shi, Yumei; Chen, Dongsheng

    2014-03-01

    This paper presents an experimental investigation on the heat transfer characteristics of liquefied natural gas flow boiling in a vertical micro-fin tube. The effect of heat flux, mass flux and inlet pressure on the flow boiling heat transfer coefficients was analyzed. The Kim, Koyama, and two kinds of Wellsandt correlations with different Ftp coefficients were used to predict the flow boiling heat transfer coefficients. The predicted results showed that the Koyama correlation was the most accurate over the range of experimental conditions.

  7. Physical interpretation of geysering phenomena and periodic boiling instability at low flows

    International Nuclear Information System (INIS)

    Duffey, R.B.; Rohatgi, U.S.

    1996-01-01

    Over 30 years ago, Griffith showed that unstable and periodic initial boiling occurred in stagnant liquids in heated pipes coupled to a cooler or condensing plenum volume. This was called ''geysering'', and is a similar phenomenon to the rapid nucleation and voiding observed in tubes filled with superheated liquid. It is also called ''bumping'' when non-uniformly heated water or a chemical suddenly boils in laboratory glassware. In engineering, the stability and predictability has importance to the onset of bulk boiling in a natural and forced circulation loops. The latest available data show the observed stability and periodicity of the onset of boiling flow when there is a plenum, multiple heated channels, and a sustained subcooling in a circulating loop. We examine the available data, both old and new, and develop a new theory to illustrate the simple physics causing the observed periodicity of the flow. We examine the validity of the theory by comparison to all the geysering data, and develop a useful and simple correlation. We illustrate the equivalence of the onset of geysering to the onset of static instability in subcooled boiling. We also derive the stability boundary for geysering, utilizing turbulent transport analysis to determine the effects of pressure and other key parameters. This new result explains the greater stability region observed at higher pressures. The paper builds on the 30 years of quite independent thermal hydraulic work that is still fresh and useful today. We discuss the physical interpretation of geysering onset with a consistent theory, and show where refinements would be useful to the data correlations

  8. Film boiling from spheres in single- and two-phase flow

    International Nuclear Information System (INIS)

    Liu, C.; Theofanous, T.G.; Yuen, W.W.

    1992-01-01

    Experimental data on film boiling heat transfer from single, inductively heated, spheres in single- and two-phase flow (saturated water and steam, respectively) are presented. In the single-phase-flow experiments water velocities ranged from 0.1 to 2.0 m/s; in the two-phase-flow experiments superficial water and steam velocities covered 0.1 to 0.6 m/s and 4 to 10 m/s, respectively. All experiments were run at atmospheric pressure and with sphere temperatures from 900C down to quenching. Limited interpretations of the single-phase- flow data are possible, but the two-phase-flow data are new and unique

  9. Heater rod temperature change at boiling transition under flow oscillation

    International Nuclear Information System (INIS)

    Kasai, Shigeru; Toba, Akio; Takigawa, Yukio; Ebata, Shigeo; Morooka, Shin-ichi; Shirakawa, Ken-etsu; Utsuno, Hideaki.

    1986-01-01

    The experiments were performed to investigate the boiling transition phenomenon under flow oscillation (OSBT) during thermal hydraulic instability. It was found, from the experimental results, that the thermal hydraulic instability did not immediately lead to the boiling transition (BT) and, even when the BT occurred due to a power increase, the change in the heater rod temperature was periodically up and down with a saw-toothed shape and no excursion occurred. To investigate the temperature change characteristics, an analysis was also performed using the transient thermal hydraulics code. The analytical results showed that the shape of the heater rod temperature change was well simulated by presuming a repeat of alternate BT and rewetting. Based on these results, further analysis has been performed with the lumped parameter model to investigate the temperature profile characteristics as well as the effects of the post-BT heat transfer coefficient and the flow oscillation period on the maximum temperature. (author)

  10. Two-phase wall function for modeling of turbulent boundary layer in subcooled boiling flow

    International Nuclear Information System (INIS)

    Bostjan Koncar; Borut Mavko; Yassin A Hassan

    2005-01-01

    Full text of publication follows: The heat transfer and phase-change mechanisms in the subcooled flow boiling are governed mainly by local multidimensional mechanisms near the heated wall, where bubbles are generated. The structure of such 'wall boiling flow' is inherently non-homogeneous and is further influenced by the two-phase flow turbulence, phase-change effects in the bulk, interfacial forces and bubble interactions (collisions, coalescence, break-up). In this work the effect of two-phase flow turbulence on the development of subcooled boiling flow is considered. Recently, the modeling of two-phase flow turbulence has been extensively investigated. A notable progress has been made towards deriving reliable models for description of turbulent behaviour of continuous (liquid) and dispersed phase (bubbles) in the bulk flow. However, there is a lack of investigation considering the modeling of two-phase flow boundary layer. In most Eulerian two-fluid models standard single-phase wall functions are used for description of turbulent boundary layer of continuous phase. That might be a good approximation at adiabatic flows, but their use for boundary layers with high concentration of dispersed phase is questionable. In this work, the turbulent boundary layer near the heated wall will be modeled with the so-called 'two-phase' wall function, which is based on the assumption of additional turbulence due to bubble-induced stirring in the boundary layer. In the two-phase turbulent boundary layer the wall function coefficients strongly depend on the void fraction. Moreover, in the turbulent boundary layer with nucleating bubbles, the bubble size variation also has a significant impact on the liquid phase. As a basis, the wall function of Troshko and Hassan (2001), developed for adiabatic bubbly flows will be used. The simulations will be performed by a general-purpose CFD code CFX-4.4 using additional models provided by authors. The results will be compared to the boiling

  11. Critical heat flux of forced convection boiling in an oscilating acceleration field. Pt. 1

    International Nuclear Information System (INIS)

    Otsuji, T.; Kurosawa, A.

    1982-01-01

    The influence of periodically varying acceleration on critical heat flux (CHF) of Freon-113 flowing upward in a uniformly heated vertical annular channel has been studied experimentally. The freon loop was oscillated vertically to determine the ratio of CHF in the oscillating acceleration field to the corresponding stationary value. The amplitude of inlet flow oscillation induced by variation of acceleration, which causes early CHF, is proportional to the acceleration amplitude. The dependence of inlet flow rate on the oscillating acceleration decreases with increasing inlet subcooling, and no oscillation of inlet flow is observed in the case of negative exit quality (subcooled boiling). Nevertheless the degradation of CHF is more remarkable in the low quality region. This result suggests the necessity to introduce an other mechanism of early CHF than flow oscillation. (orig.)

  12. Effective diffusion in laminar convective flows

    International Nuclear Information System (INIS)

    Rosenbluth, M.N.; Berk, H.L.; Doxas, I.; Horton, W.

    1987-03-01

    The effective diffusion coefficient D* of a passive component, such as test particles, dye, temperature, magnetic flux, etc., is derived for motion in periodic two-dimensional incompressible convective flow with characteristic velocity v and size d in the presence of an intrinsic local diffusivity D. Asymptotic solutions for effective diffusivity D*(P) in the large P limit, with P ∼ vd/D, is shown to be of the form D* = cDP/sup 1/2/ with c being a coefficient that is determined analytically. The constant c depends on the geometry of the convective cell and on an average of the flow speed along the separatrix. The asymptotic method of evaluation applies to both free boundary and rough boundary flow patterns and it is shown that the method can be extended to more complicated patterns such as the flows generated by rotating cylinders, as in the problem considered by Nadim, Cox, and Brenner [J. Fluid Mech., 164: 185 (1986)]. The diffusivity D* is readily calculated for small P, but the evaluation for arbitrary P requires numerical methods. Monte Carlo particle simulation codes are used to evaluate D* at arbitrary P, and thereby describe the transition for D* between the large and small P limits

  13. Minimum heat flux (MHF) point in pool and external-flow boiling

    International Nuclear Information System (INIS)

    Nishio, Shigefumi

    1983-01-01

    As for the boiling phenomena near a minimum heat flux (MHF) point to which attention has been paid recently concerning the safety analysis of LWR cores, the results of research have not been put in order sufficiently. Therefore in this explanation, the object is limited to pool boiling and external flow boiling, and it is attempted to rearrange the present knowledge on the phenomena near a MHF point from the viewpoint of the relation to the state of solid-liquid contact, the effect of various factors on a MHF point and the modeling of a MHF point. The heat transfer characteristics in boiling phenomena are represented by a curve with one maximum and one minimum points. The MHF point is called also minimum film boiling point. In a heat flux-controlled heating surface, temperature jump arises when heat flux is decreased at a MHF point. The phenomena near a MHF point and the technological background when a MHF point becomes a problem are explained. Near a MHF point, only partial, intermittent solid-liquid contact is maintained. The effects of solid-liquid contact mode, the geometry of a heating surface, pressure and others on a MHF point are discussed. (Kako, I.)

  14. A numerical study of boiling flow instability of a reactor thermosyphon system

    International Nuclear Information System (INIS)

    Nayak, A.K.; Lathouwers, D.; Hagen, T.H.J.J. van der; Schrauwen, Frans; Molenaar, Peter; Rogers, Andrew

    2006-01-01

    A numerical study has been carried out to investigate the boiling flow instability of a reactor thermosyphon system. The numerical model solves the conservation equations of mass, momentum and energy applicable to a two-fluid and three-field steam-water system using a finite difference technique. The computer code MONA was used for this purpose. The code was applied to the thermosyphon system of an EO (ethylene oxide) chemical reactor in which the heat released by a catalytic reaction is carried by boiling water under natural circulation conditions. The steady-state characteristics of the reactor thermosyphon system were predicted using the MONA code and conventional two-phase flow models in order to understand the model applicability for this type of thermosyphon system. The two-fluid model was found to predict the flow closest to the measured value of the plant. The stability behaviour of the thermosyphon system was investigated for a wide range of operating conditions. The effects of power, subcooling, riser length and riser diameter on the boiling flow instability were determined. The system was found to be unstable at higher power conditions which is typical for a Type II instability. However, with an increase in riser diameter, oscillations at low power were observed as well. These are classified as Type I instabilities. Stability maps were predicted for both Type I and Type II instabilities. Methods of improving the stability of the system are discussed

  15. A numerical study of boiling flow instability of a reactor thermosyphon system

    Energy Technology Data Exchange (ETDEWEB)

    Nayak, A.K.; Lathouwers, D.; Hagen, T.H.J.J. van der [Interfaculty Reactor Institute, Delft University of Technology, Mekelweg 15, 2629 JB Delft (Netherlands); Schrauwen, Frans; Molenaar, Peter; Rogers, Andrew [Shell Research and Technology Centre, Badhuisweg 3, 1031 CM Amsterdam (Netherlands)

    2006-04-01

    A numerical study has been carried out to investigate the boiling flow instability of a reactor thermosyphon system. The numerical model solves the conservation equations of mass, momentum and energy applicable to a two-fluid and three-field steam-water system using a finite difference technique. The computer code MONA was used for this purpose. The code was applied to the thermosyphon system of an EO (ethylene oxide) chemical reactor in which the heat released by a catalytic reaction is carried by boiling water under natural circulation conditions. The steady-state characteristics of the reactor thermosyphon system were predicted using the MONA code and conventional two-phase flow models in order to understand the model applicability for this type of thermosyphon system. The two-fluid model was found to predict the flow closest to the measured value of the plant. The stability behaviour of the thermosyphon system was investigated for a wide range of operating conditions. The effects of power, subcooling, riser length and riser diameter on the boiling flow instability were determined. The system was found to be unstable at higher power conditions which is typical for a Type II instability. However, with an increase in riser diameter, oscillations at low power were observed as well. These are classified as Type I instabilities. Stability maps were predicted for both Type I and Type II instabilities. Methods of improving the stability of the system are discussed. [Author].

  16. Multi-boiling Heat Transfer Analysis of a Convective Straight Fin with Temperature-Dependent Thermal Properties and Internal Heat Generation

    Directory of Open Access Journals (Sweden)

    Gbeminiyi Sobamowo

    2017-10-01

    Full Text Available In this study, by using the finite volume method, the heat transfer in a convective straight fin with temperature-dependent thermal properties and an internal heat generation under multi-boiling heat transfer modes are analyzed. In this regard, the local heat transfer coefficient is considered to vary within a power-law function of temperature. In the present study, the coexistence of all the boiling modes is taken into consideration. The developed heat transfer models and the corresponding numerical solutions are used to investigate the effects of various thermo-geometric parameters on the thermal performance of the longitudinal rectangular fin. The results shows that the fin temperature distribution, the total heat transfer, and the fin efficiency are significantly affected by the thermo-geometric parameters of the fin and the internal heat generation within the fin. The obtained results can provide a platform for improvements in the design of the fin in the heat transfer equipment.

  17. Heat transport in bubbling turbulent convection.

    Science.gov (United States)

    Lakkaraju, Rajaram; Stevens, Richard J A M; Oresta, Paolo; Verzicco, Roberto; Lohse, Detlef; Prosperetti, Andrea

    2013-06-04

    Boiling is an extremely effective way to promote heat transfer from a hot surface to a liquid due to numerous mechanisms, many of which are not understood in quantitative detail. An important component of the overall process is that the buoyancy of the bubble compounds with that of the liquid to give rise to a much-enhanced natural convection. In this article, we focus specifically on this enhancement and present a numerical study of the resulting two-phase Rayleigh-Bénard convection process in a cylindrical cell with a diameter equal to its height. We make no attempt to model other aspects of the boiling process such as bubble nucleation and detachment. The cell base and top are held at temperatures above and below the boiling point of the liquid, respectively. By keeping this difference constant, we study the effect of the liquid superheat in a Rayleigh number range that, in the absence of boiling, would be between 2 × 10(6) and 5 × 10(9). We find a considerable enhancement of the heat transfer and study its dependence on the number of bubbles, the degree of superheat of the hot cell bottom, and the Rayleigh number. The increased buoyancy provided by the bubbles leads to more energetic hot plumes detaching from the cell bottom, and the strength of the circulation in the cell is significantly increased. Our results are in general agreement with recent experiments on boiling Rayleigh-Bénard convection.

  18. Steady-state nucleate pool boiling mechanism at low heat fluxes

    International Nuclear Information System (INIS)

    Bastos, L.E.G.

    1979-01-01

    Heat is transfered in the steady state to a horizontal cooper disc inmersed in water at saturation temperature. Levels of heat flux are controlled so that convection and the nucleate boiling can be observed. The value of heat flux is determined experimentally and high speed film is used to record bubble growth. In order to explain the phenomenon the oretical model is proposed in which part of the heat is transfered by free convection during nucleate boiling regime. Agreement between the experiments and the theoretical model is good. (Author) [pt

  19. On the occurrence of burnout downstream of a flow obstacle in boiling two-phase upward flow within a vertical annular channel

    International Nuclear Information System (INIS)

    Mori, Shoji; Tominaga, Akira; Fukano, Tohru

    2004-01-01

    If a flow obstruction such as a spacer is set in a boiling two-phase flow within an annular channel, the inner tube of which is used as a heater, the temperature on the surface of the heater tube is severely affected by the existence of the spacer. In some case the spacer has a cooling effect, and in the other case it causes the dryout of the cooling liquid film on the heating surface resulting in the burnout of the tube. But the burnout mechanism near the spacer is not still clear. In the present paper we discus the influence of the flow obstacle on the occurrence of burnout downstream of the flow obstacle in boiling two-phase upward flow within a vertical annular channel. (author)

  20. Volume-heated boiling pool flow behavior and application to transition phase accident conditions

    International Nuclear Information System (INIS)

    Ginsberg, T.; Jones, O.C. Jr.; Chen, J.C.

    1978-01-01

    Observations of two-phase flow fields in volume-heated boiling pools are reported. Photographic observations, together with pool-average void fraction measurements are presented. Flow regime transition criteria derived from the measurements are discussed. The churn-turbulent flow regime was the dominant regime for superficial vapor velocities up to nearly five times the Kutateladze dispersal velocity. Within this range of conditions, a churn-turbulent drift flux model provides a reasonable prediction of the pool-average void fraction data. The results of the experiment and analyses are extrapolated to transition phase conditions. It is shown that intense pool boil-up could occur where the pool-average void fraction would be greater than 0.6 for steel vaporization rates equivalent to power levels greater than one percent of nominal LMFBR power density

  1. The Fractional Step Method Applied to Simulations of Natural Convective Flows

    Science.gov (United States)

    Westra, Douglas G.; Heinrich, Juan C.; Saxon, Jeff (Technical Monitor)

    2002-01-01

    This paper describes research done to apply the Fractional Step Method to finite-element simulations of natural convective flows in pure liquids, permeable media, and in a directionally solidified metal alloy casting. The Fractional Step Method has been applied commonly to high Reynold's number flow simulations, but is less common for low Reynold's number flows, such as natural convection in liquids and in permeable media. The Fractional Step Method offers increased speed and reduced memory requirements by allowing non-coupled solution of the pressure and the velocity components. The Fractional Step Method has particular benefits for predicting flows in a directionally solidified alloy, since other methods presently employed are not very efficient. Previously, the most suitable method for predicting flows in a directionally solidified binary alloy was the penalty method. The penalty method requires direct matrix solvers, due to the penalty term. The Fractional Step Method allows iterative solution of the finite element stiffness matrices, thereby allowing more efficient solution of the matrices. The Fractional Step Method also lends itself to parallel processing, since the velocity component stiffness matrices can be built and solved independently of each other. The finite-element simulations of a directionally solidified casting are used to predict macrosegregation in directionally solidified castings. In particular, the finite-element simulations predict the existence of 'channels' within the processing mushy zone and subsequently 'freckles' within the fully processed solid, which are known to result from macrosegregation, or what is often referred to as thermo-solutal convection. These freckles cause material property non-uniformities in directionally solidified castings; therefore many of these castings are scrapped. The phenomenon of natural convection in an alloy under-going directional solidification, or thermo-solutal convection, will be explained. The

  2. Investigation of bubble flow regimes in nucleate boiling of highly-wetting liquids

    International Nuclear Information System (INIS)

    Tong, W.; Bar-Cohen, A.; Simon, T.W.

    1991-01-01

    This paper describes an investigation of the bubble flow regimes in nucleate boiling of FC-72, a highly-wetting liquid. Theoretically analysis of vapor bubble generation and departure from the heated surface reveals that the heat fluxes required for the merging of consecutive bubbles, for highly-wetting liquids, lie in the upper range of the nucleate boiling heat flux. A visual and photographic study of nucleate boiling from sputtered platinum surfaces has supported the theoretical results and shown that the isolated bubble behavior extends to at least 50-80% of the critical heat flux, considerably higher than observed by others with water. Lateral coalescence of adjacent bubbles has been found to be a more likely cause of the termination of the isolated bubble regime. These findings suggest that thermal transport models which are based on isolated bubble behavior may be applicable to nearly the entire range of nucleate boiling of electronic cooling fluids

  3. ULTRA-SHARP nonoscillatory convection schemes for high-speed steady multidimensional flow

    Science.gov (United States)

    Leonard, B. P.; Mokhtari, Simin

    1990-01-01

    For convection-dominated flows, classical second-order methods are notoriously oscillatory and often unstable. For this reason, many computational fluid dynamicists have adopted various forms of (inherently stable) first-order upwinding over the past few decades. Although it is now well known that first-order convection schemes suffer from serious inaccuracies attributable to artificial viscosity or numerical diffusion under high convection conditions, these methods continue to enjoy widespread popularity for numerical heat transfer calculations, apparently due to a perceived lack of viable high accuracy alternatives. But alternatives are available. For example, nonoscillatory methods used in gasdynamics, including currently popular TVD schemes, can be easily adapted to multidimensional incompressible flow and convective transport. This, in itself, would be a major advance for numerical convective heat transfer, for example. But, as is shown, second-order TVD schemes form only a small, overly restrictive, subclass of a much more universal, and extremely simple, nonoscillatory flux-limiting strategy which can be applied to convection schemes of arbitrarily high order accuracy, while requiring only a simple tridiagonal ADI line-solver, as used in the majority of general purpose iterative codes for incompressible flow and numerical heat transfer. The new universal limiter and associated solution procedures form the so-called ULTRA-SHARP alternative for high resolution nonoscillatory multidimensional steady state high speed convective modelling.

  4. Influence of a flow obstacle on the occurrence of burnout in boiling two-phase upward flow within a vertical annular channel

    Energy Technology Data Exchange (ETDEWEB)

    Mori, S.; Fukano, T. [Kyushu Univ., Fukuoka (Japan)

    2003-07-01

    When a flow obstruction such as a cylindrical spacer is set in a boiling two-phase flow with-in an annular channel, the inner tube of which is used as a heater, the temperature on the surface of the heating tube is severely affected by its existence. In some cases the cylindrical spacer has a cooling effect, and in the other cases it causes the dryout of the cooling water film on the heating surface resulting in the burnout of the heating tube. In the present paper we have focused our attention on the influence of a flow obstacle on the occurrence of burnout of the heating tube in boiling two-phase flow.

  5. Numerical experimentation on convective coolant flow in Ghana ...

    African Journals Online (AJOL)

    Numerical experiments on one dimensional convective coolant flow during steady state operation of the Ghana Research Reactor-1 (GHARR-I) were performed to determine the thermal hydraulic parameters of temperature, density and flow rate. The computational domain was the reactor vessel, including the reactor core.

  6. Experimental investigation on lithium-ion battery thermal management based on flow boiling in mini-channel

    International Nuclear Information System (INIS)

    An, Zhoujian; Jia, Li; Li, Xuejiao; Ding, Yong

    2017-01-01

    Highlights: • A new type of BTM system based on flow boiling in mini-channel are presented. • Uniform temperature and volume distribution of battery module are obtained. • The temperatures of battery cell are maintained around 40 °C. • There exists an appropriate Re number range for boiling heat transfer in mini-channel. - Abstract: In order to guarantee the safety and prolong the lifetime of lithium-ion power battery within electric vehicles, thermal management system is essential. A new type of thermal management system based on flow boiling in mini-channel utilizing dielectric hydrofluoroether liquid which boiling point is 34 °C is proposed. The cooling experiments for battery module are carried out at different discharge rates and flow Re number. The cooling effect and the influence of battery cooling on the electrochemical characteristics are concerned. The experimental results show that the thermal management can efficiently reduce maximum temperature of battery module and surface maximum temperature difference. A relatively uniform temperature and voltage distributions are provided within the battery module at higher discharge rate benefit from the advantage of boiling heat transfer with uniform temperature distribution on cold plate. It is shown that the voltage decreases with the increase of Re number of fluid due to the reducing of temperature. There exist slight fluctuations of voltage distribution because of the non-uniformity of temperature distribution within the battery module at higher discharge rates. For different discharge rate, there also exists an appropriate Re number range during which the mode of heat transfer is mainly in boiling heat transfer mode and the cooling result can be greatly improved.

  7. Solar-cycle Variations of Meridional Flows in the Solar Convection Zone Using Helioseismic Methods

    Science.gov (United States)

    Lin, Chia-Hsien; Chou, Dean-Yi

    2018-06-01

    The solar meridional flow is an axisymmetric flow in solar meridional planes, extending through the convection zone. Here we study its solar-cycle variations in the convection zone using SOHO/MDI helioseismic data from 1996 to 2010, including two solar minima and one maximum. The travel-time difference between northward and southward acoustic waves is related to the meridional flow along the wave path. Applying the ray approximation and the SOLA inversion method to the travel-time difference measured in a previous study, we obtain the meridional flow distributions in 0.67 ≤ r ≤ 0.96R ⊙ at the minimum and maximum. At the minimum, the flow has a three-layer structure: poleward in the upper convection zone, equatorward in the middle convection zone, and poleward again in the lower convection zone. The flow speed is close to zero within the error bar near the base of the convection zone. The flow distribution changes significantly from the minimum to the maximum. The change above 0.9R ⊙ shows two phenomena: first, the poleward flow speed is reduced at the maximum; second, an additional convergent flow centered at the active latitudes is generated at the maximum. These two phenomena are consistent with the surface meridional flow reported in previous studies. The change in flow extends all the way down to the base of the convection zone, and the pattern of the change below 0.9R ⊙ is more complicated. However, it is clear that the active latitudes play a role in the flow change: the changes in flow speed below and above the active latitudes have opposite signs. This suggests that magnetic fields could be responsible for the flow change.

  8. Heat Transfer in Boiling Dilute Emulsion with Strong Buoyancy

    Science.gov (United States)

    Freeburg, Eric Thomas

    Little attention has been given to the boiling of emulsions compared to that of boiling in pure liquids. The advantages of using emulsions as a heat transfer agent were first discovered in the 1970s and several interesting features have since been studied by few researchers. Early research focuses primarily on pool and flow boiling and looks to determine a mechanism by which the boiling process occurs. This thesis looks at the boiling of dilute emulsions in fluids with strong buoyant forces. The boiling of dilute emulsions presents many favorable characteristics that make it an ideal agent for heat transfer. High heat flux electronics, such as those seen in avionics equipment, produce high heat fluxes of 100 W/cm2 or more, but must be maintained at low temperatures. So far, research on single phase convection and flow boiling in small diameter channels have yet to provide an adequate solution. Emulsions allow the engineer to tailor the solution to the specific problem. The fluid can be customized to retain the high thermal conductivity and specific heat capacity of the continuous phase while enhancing the heat transfer coefficient through boiling of the dispersed phase component. Heat transfer experiments were carried out with FC-72 in water emulsions. FC-72 has a saturation temperature of 56 °C, far below that of water. The parameters were varied as follows: 0% ≤ epsilon ≤ 1% and 1.82 x 1012 ≤ RaH ≤ 4.42 x 1012. Surface temperatures along the heated surface reached temperature that were 20 °C in excess of the dispersed phase saturation temperature. An increase of ˜20% was seen in the average Nusselt numbers at the highest Rayleigh numbers. Holography was used to obtain images of individual and multiple FC-72 droplets in the boundary layer next to the heated surface. The droplet diameters ranged from 0.5 mm to 1.3 mm. The Magnus effect was observed when larger individual droplets were injected into the boundary layer, causing the droplets to be pushed

  9. A sensitivity analysis of the mass balance equation terms in subcooled flow boiling

    International Nuclear Information System (INIS)

    Braz Filho, Francisco A.; Caldeira, Alexandre D.; Borges, Eduardo M.

    2013-01-01

    In a heated vertical channel, the subcooled flow boiling occurs when the fluid temperature reaches the saturation point, actually a small overheating, near the channel wall while the bulk fluid temperature is below this point. In this case, vapor bubbles are generated along the channel resulting in a significant increase in the heat flux between the wall and the fluid. This study is particularly important to the thermal-hydraulics analysis of Pressurized Water Reactors (PWRs). The computational fluid dynamics software FLUENT uses the Eulerian multiphase model to analyze the subcooled flow boiling. In a previous paper, the comparison of the FLUENT results with experimental data for the void fraction presented a good agreement, both at the beginning of boiling as in nucleate boiling at the end of the channel. In the region between these two points the comparison with experimental data was not so good. Thus, a sensitivity analysis of the mass balance equation terms, steam production and condensation, was performed. Factors applied to the terms mentioned above can improve the agreement of the FLUENT results to the experimental data. Void fraction calculations show satisfactory results in relation to the experimental data in pressures values of 15, 30 and 45 bars. (author)

  10. Mixed convection between horizontal plates and consequences for chemical vapor deposition flows

    International Nuclear Information System (INIS)

    Chiu, K.C.

    1986-01-01

    To simulate the fluid dynamics of VD systems, mixed convection between horizontal plates (AR = width/height = 10) heated from below was studied by laser Doppler anemometry in a range 1368 < Ra < 8300 and 15 < R3 < 170. The entrance effects were characterized by two lengths: one for the onset of bouyancy-driven instability, and one for the full development of longitudinal convection rolls. Explicit expressions for both entrance lengths are given in terms of Ra and Re. In addition, unsteady longitudinal convection rolls were observed. These are discussed in terms of the admixture of transverse convection rolls and/or contributions from upstream turbulence. For the fully developed region it is shown analytically that the transverse velocities of the longitudinal convection rolls, v and w, are independent of the forced flow and are identical to those of the two-dimensional Rayleigh-Benard convection rolls. These fundamental results serve as a base for the discussion of horizontal CVD flows. The entrance and sidewall effects are found to have pronounced influences on the flow patterns observed in CVD (AR = 2) reactors

  11. Contribution to the study of natural convection in a boiling medium with power density transfer

    International Nuclear Information System (INIS)

    Bede, M.

    1987-01-01

    This study has been carried out in the framework of fast reactor safety studies and deals with the fuel boiling problem in case of flow blockage at the bottom of a fuel assembly. The experimental part of this study bringss new informations characteristic of a boiling fluid bath (water) simulating in a transient and in a steady regime (pressure, temperature, void fraction, heat flux at the walls). It points out a relation between heat losses through the walls and the importance of the monophase zone of the bath. A model has been developed from the analysis of experimental results. It is based on a quasi-stationary state and allows to find the evolution of the characteristic values in confined transient regime [fr

  12. Thermal and hydrodynamic characteristics of forced and mixed convection flow through vertical rectangular channels

    Directory of Open Access Journals (Sweden)

    Hanafi Abdalla S.

    2008-01-01

    Full Text Available This paper presents experimental and numerical studies for the case of turbulent forced and mixed convection flow of water through narrow vertical rectangular channel. The channel is composed of two parallel plates which are heated at a uniform heat flux, whereas, the other two sides of the channel are thermally insulated. The plates are of 64 mm in width, 800 mm in height, and separated from each other at a narrow gap of 2.7 mm. The Nusselt number distribution along the flow direction normalized by the Nusselt number for the case of turbulent forced convection flow is obtained experimentally with a comparison with the numerical results obtained from a commercial computer code. The quantitative determination of the nor- malized Nusselt number with respect to the dimension-less number Z = (Gr/Re21/8Pr0.5 is presented with a comparison with previous experimental results. Qualitative results are presented for the normalized temperature and velocity profiles in the transverse direction with a comparison between the forced and mixed convection flow for both the cases of upward and downward flow directions. The effect of the axial locations and the parameter Gr/Re on the variation of the normalized temperature profiles in the transverse direction for both the regions of forced and mixed convection and for both of the upward and downward flow directions are obtained. The normalized velocity profiles in the transverse directions are also determined at different inlet velocity and heat fluxes for the previous cases. It is found that the normalized Nusselt number is greater than one in the mixed convection region for both the cases of upward and downward flow and correlated well with the dimension-less parameter Z for both of the forced and mixed convection regions. The temperature profiles increase with increasing the axial location along the flow direction or the parameter Gr/Re for both of the forced and mixed convection regions, but this increase is

  13. On the occurrence of burnout downstream of the flow obstacle in boiling two-phase upward flow within a vertical annular channel

    International Nuclear Information System (INIS)

    Mori, Shoji; Fukano, Tohru

    2003-01-01

    If a flow obstruction such as a spacer is set in a boiling two-phase flow within an annular channel, the inner tube of which is used as a heater, the temperature on the surface of the heater tube is severely affected by the existence of the spacer. In some cases the spacer has a cooling effect, and in the other case it causes the dryout of the cooling liquid film on the heating surface resulting in the burnout of the tube. But the thermo-fluid dynamic mechanism to cause burnout near the spacer is not still clear. In the present paper we discuss the influence of the flow obstacle on the occurrence of burnout downstream of the flow obstacle in boiling two-phase upward flow within a vertical annular channel. (author)

  14. Direct numerical simulations of nucleate boiling flows of binary mixtures

    International Nuclear Information System (INIS)

    Didier Jamet; Celia Fouillet

    2005-01-01

    Full text of publication follows: Better understand the origin and characteristics of boiling crisis is still a scientific challenge despite many years of valuable studies. One of the reasons why boiling crisis is so difficult to understand is that local and coupled physical phenomena are believed to play a key role in the trigger of instabilities which lead to the dry out of large portions of the heated solid phase. Nucleate boiling of a single bubble is fairly well understood compared to boiling crisis. Therefore, the numerical simulation of a single bubble growth during nucleate boiling is a good candidate to evaluate the capabilities of a numerical method to deal with complex liquid-vapor phenomena with phase-change and eventually to tackle the boiling crisis problem. In this paper, we present results of direct numerical simulations of nucleate boiling. The numerical method used is the second gradient method, which is a diffuse interface method dedicated to liquid vapor flows with phase-change. This study is not intended to provide quantitative results, partly because all the simulations are two-dimensional. However, particular attention is paid to the influence of some parameters on the main features of nucleate boiling, i.e. the radius of departure and the frequency of detachment of bubbles. In particular, we show that, as the contact angle increases, the radius of departure increases whereas the frequency of detachment decreases. Moreover, the influence of the existence of quasi non-condensable gas is studied. Numerical results show an important decrease of the heat exchange coefficient when a small amount of a quasi non-condensable gas is added to the pure liquid-vapor water system. This result is in agreement with experimental observations. Beyond these qualitative results, this numerical study allows to get insight into some important physical phenomena and to confirm that during nucleate boiling, large scale quantities are influenced by small scale

  15. Research on boiling and two-phase flow

    International Nuclear Information System (INIS)

    Marinsek, Z.; Gaspersic, B.; Pavselj, D.; Tomsic, M.

    1977-01-01

    Report consists of three contributions. Experimental apparatus with pressure chamber (up to 25 bar and 250 deg C) was constructed including optical bubble detection device, and test measurements of mutual influence of boiling bubbles from two adjacent nucleation sites were performed; for analyses, a computer programme package for coincidence analyses of events was made, including data acquisition hardware. Two-phase pressure drop in subcooled Vertical annular water flow was measured, for pressures up to 10 bar, mass velocity 500 to 760 kg/m 2 s and vapour quality 0 to .01. Results agree fairly well with Martinelli-Nelson model

  16. Critical heat flux of forced flow boiling in a narrow one-side heated rectangular flow channel

    Energy Technology Data Exchange (ETDEWEB)

    Limin, Zheng [Shanghai Nuclear Engineering Research and Design Inst., SH (China); Iguchi, Tadashi; Kureta, Masatoshi; Akimoto, Hajime

    1997-08-01

    The present work deals with the critical heat flux (CHF) under subcooled flow boiling in a narrow one-side uniformly heated rectangular flow channel. The range of interest of parameters such as pressure, flow velocity and subcooling is around 0.1 MPa, 5-15 ms{sup -1} and 50degC, respectively. The rectangular flow channel used is 50 mm long, 12 mm in width and 0.2 to 3 mm in height. Test conditions were selected by combination of the following parameters: Gap=0.2-3.0 mm (D{sub hy}=0.3934-4.8 mm); flow length, 50.0 mm; water mass flux, 4.94-14.82 Mgm{sup -2}s{sup -1} (water flow velocity, 5-15 ms{sup -1}); exit pressure, 0.1 MPa; inlet temperature, 50degC, inlet coolant subcooling, 50degC. Over 40 CHF stable data points were obtained. CHF increased with the gap and flow velocity in a non-linear fashion. HTC increased with flow velocity and decreasing gap. Based on the experimental results, an empirical correlation was developed, indicating the dependence of CHF on the gap and flow velocity. All of data points predicted within {+-}18% error band for the present experimental data. On the other hand, another similitude-based correlation was also developed, indicating the dependence of Boiling number (Bo) on Reynolds number (Re) and the variable of Gap/La, where La is a characteristic length known as Laplace capillary constant. For the limited present experimental data, all of data points were predicted within {+-}16%. (author)

  17. Critical heat flux of forced flow boiling in a narrow one-side heated rectangular flow channel

    International Nuclear Information System (INIS)

    Zheng Limin; Iguchi, Tadashi; Kureta, Masatoshi; Akimoto, Hajime.

    1997-08-01

    The present work deals with the critical heat flux (CHF) under subcooled flow boiling in a narrow one-side uniformly heated rectangular flow channel. The range of interest of parameters such as pressure, flow velocity and subcooling is around 0.1 MPa, 5-15 ms -1 and 50degC, respectively. The rectangular flow channel used is 50 mm long, 12 mm in width and 0.2 to 3 mm in height. Test conditions were selected by combination of the following parameters: Gap=0.2-3.0 mm (D hy =0.3934-4.8 mm); flow length, 50.0 mm; water mass flux, 4.94-14.82 Mgm -2 s -1 (water flow velocity, 5-15 ms -1 ); exit pressure, 0.1 MPa; inlet temperature, 50degC, inlet coolant subcooling, 50degC. Over 40 CHF stable data points were obtained. CHF increased with the gap and flow velocity in a non-linear fashion. HTC increased with flow velocity and decreasing gap. Based on the experimental results, an empirical correlation was developed, indicating the dependence of CHF on the gap and flow velocity. All of data points predicted within ±18% error band for the present experimental data. On the other hand, another similitude-based correlation was also developed, indicating the dependence of Boiling number (Bo) on Reynolds number (Re) and the variable of Gap/La, where La is a characteristic length known as Laplace capillary constant. For the limited present experimental data, all of data points were predicted within ±16%. (author)

  18. Impact of selected parameters on the development of boiling and flow resistance in the minichannel

    Directory of Open Access Journals (Sweden)

    Piasecka Magdalena

    2015-01-01

    Full Text Available The paper presents results of flow boiling in a rectangular minichannel 1 mm deep, 40 mm wide and 360 mm long. The heating element for FC-72 flowing in the minichannel was the thin alloy foil designated as Haynes-230. There was a microstructure on the side of the foil which comes into contact with fluid in the channel. Two types of microstructured heating surfaces: one with micro-recesses distributed evenly and another with mini-recesses distributed unevenly were used. The paper compares the impact of the microstructured heating surface and minichannel positions on the development of boiling and two phase flow pressure drop. The local heat transfer coefficients and flow resistance obtained in experiment using three positions of the minichannel, e.g.: 0°, 90° and 180° were analyzed. The study of the selected thermal and flow parameters (mass flux density and inlet pressure, geometric parameters and type of cooling liquid on the boiling heat transfer was also conducted. The most important factor turned out to be channel orientation. Application of the enhanced heating surface caused the increase of the heat transfer coefficient from several to several tens per cent, in relation to the plain surface.

  19. Meridional flow in the solar convection zone. I. Measurements from gong data

    Energy Technology Data Exchange (ETDEWEB)

    Kholikov, S. [National Solar Observatories, Tucson, AZ 85719 (United States); Serebryanskiy, A. [Ulugh Beg Astronomical Institute, Uzbek Academy of Science, Tashkent 100052 (Uzbekistan); Jackiewicz, J., E-mail: kholikov@noao.edu [Department of Astronomy, New Mexico State University, Las Cruces, NM 88003 (United States)

    2014-04-01

    Large-scale plasma flows in the Sun's convection zone likely play a major role in solar dynamics on decadal timescales. In particular, quantifying meridional motions is a critical ingredient for understanding the solar cycle and the transport of magnetic flux. Because the signal of such features can be quite small in deep solar layers and be buried in systematics or noise, the true meridional velocity profile has remained elusive. We perform time-distance helioseismology measurements on several years worth of Global Oscillation Network Group Doppler data. A spherical harmonic decomposition technique is applied to a subset of acoustic modes to measure travel-time differences to try to obtain signatures of meridional flows throughout the solar convection zone. Center-to-limb systematics are taken into account in an intuitive yet ad hoc manner. Travel-time differences near the surface that are consistent with a poleward flow in each hemisphere and are similar to previous work are measured. Additionally, measurements in deep layers near the base of the convection zone suggest a possible equatorward flow, as well as partial evidence of a sign change in the travel-time differences at mid-convection zone depths. This analysis on an independent data set using different measurement techniques strengthens recent conclusions that the convection zone may have multiple 'cells' of meridional flow. The results may challenge the common understanding of one large conveyor belt operating in the solar convection zone. Further work with helioseismic inversions and a careful study of systematic effects are needed before firm conclusions of these large-scale flow structures can be made.

  20. Experimental and Analytical Study of Lead-Bismuth-Water Direct Contact Boiling Two-Phase Flow

    Science.gov (United States)

    Novitrian; Dostal, Vaclav; Takahashi, Minoru

    The characteristics of lead-bismuth(Pb-Bi)-water boiling two-phase flow were investigated experimentally and analytically using a Pb-Bi-water direct contact boiling two-phase flow loop. Pb-Bi flow rates and void fraction were measured in a vertical circular tube at conditions of system pressure 7MPa, liquid metal temperature 460°C and injected water temperature 220°C. The drift-flux model with the assumption that bubble sizes were dependent on the fluid surface tension and the density ratio of Pb-Bi to steam-water mixture was chosen and modified by the best fit to the measured void fraction. Pb-Bi flow rates were analytically estimated using balance condition between buoyancy force and pressure losses, where the buoyancy force was calculated from void fraction estimated using the modified drift-flux model. The deviation of the analytical results of the flow rates from the experimental ones was less than 10%.

  1. Application of flexibility model in modeling of flow boiling heat transfer

    International Nuclear Information System (INIS)

    Peng Jinfeng; Zhao Fuyu

    2009-01-01

    The mathematical modeling and computer simulation have been widely used in the analysis of system's dynamic characteristics, and often useful for system control. One of the popular methods for this purpose is the lumped parameter method. For flow boiling heat transfer system, the traditional lumped parameter modeling method has a problem that the heat transfer coefficients change suddenly at the boundary of coolant phase change. It can cause error. In this paper, an idea of flexibility model is developed to deal with the boundary problem and to improve the model of flow boiling heat transfer. The segments of coolant phase change's boundary are identified, and the membership functions which are derived from Fuzzy Mathematics are used to derive approximate expressions of heat transfer coefficient in those regions. The continuity of heat transfer coefficient can be described by those expressions. The membership functions are derived from mathematical analysis and transformation. The result shows that this idea is feasible and the conclusion is practicable.

  2. Critical heat flux of subcooled flow boiling in a narrow tube

    International Nuclear Information System (INIS)

    Inasaka, Fujio; Nariai, Hideki; Shimura, Toshiya.

    1986-01-01

    The critical heat flux (CHF) of subcooled flow boiling in a narrow tube was investigated experimentally using water as a coolant. Experiments were conducted at nearly ambient pressure under the following conditions: tube inside diameter: 1 ∼ 3 mm, tube length: 10 ∼ 100 mm, and water mass velocity: 7000 - 20000 kg/(m 2 · s). The critical heat flux increases the shorter the tube length and the smaller the tube inside diameter, at the same water mass velocity and exit quality. Experimental data were compared with empirical correlations, such as the Griffel and Knoebel correlations for subcooled boiling at low pressure, the Tong correlation for subcooled boiling at high pressure, and the Katto correlation. The existence of two parameter regions was confirmed. The first is the low CHF region in which experimental data can be predicted well by Griffel and Knoebel correlations, and the second is the high CHF region in which experimental data are higher than the predictions by the above two correlations. (author)

  3. Analysis of the convective heat transfer of a fluid flow over an ...

    African Journals Online (AJOL)

    Convective heat transfer in a homogeneous fluid flow Reynolds number of order less than 2000 over an immersed axi-symmetrical body with curved surfaces has been investigated. The fluid flow in consideration was unsteady and of constant density .This study analysed the extent to which convective heat transfer has on ...

  4. Modelling of convective heat and mass transfer in rotating flows

    CERN Document Server

    Shevchuk, Igor V

    2016-01-01

     This monograph presents results of the analytical and numerical modeling of convective heat and mass transfer in different rotating flows caused by (i) system rotation, (ii) swirl flows due to swirl generators, and (iii) surface curvature in turns and bends. Volume forces (i.e. centrifugal and Coriolis forces), which influence the flow pattern, emerge in all of these rotating flows. The main part of this work deals with rotating flows caused by system rotation, which includes several rotating-disk configurations and straight pipes rotating about a parallel axis. Swirl flows are studied in some of the configurations mentioned above. Curvilinear flows are investigated in different geometries of two-pass ribbed and smooth channels with 180° bends. The author demonstrates that the complex phenomena of fluid flow and convective heat transfer in rotating flows can be successfully simulated using not only the universal CFD methodology, but in certain cases by means of the integral methods, self-similar and analyt...

  5. A visual study of forced convection boiling. Part I: Results for a flat vertical heater

    International Nuclear Information System (INIS)

    Kirby, G.J.; Staniforth, R.; Kinneir, J.H.

    1965-03-01

    This report presents the first results of a visual study of the hydrodynamics of boiling in channels and of burnout. It was found that the bubbles formed did not diffuse into the main stream at high heat fluxes, but remained close to the heater. Consequently severe coalescence took place, resulting in the formation of large regularly shaped bubbles. An analysis of the forces acting on these bubbles is given; this accounts qualitatively for the observed behaviour. The above bubble formations result from the addition of heat at a wall so that clearly isothermal models, such as those using air-water mixtures, cannot give a true representation of the flow pattern. Attempts to view the heater surface at burnout were frustrated by poor visibility through the boiling mixture. (author)

  6. Thermal performance of cooling system for a laptop computer using a boiling enhancement microstructure

    International Nuclear Information System (INIS)

    Cho, N. H.; Jeong, W. Y.; Park, S. H.

    2008-01-01

    The increasing heat generation rates in CPU of notebook computers motivate a research on cooling technologies with low thermal resistance. This paper develops a closed-loop two-phase cooling system using a micropump to circulate a dielectric liquid(PF5060). The cooling system consists of an evaporator containing a boiling enhancement microstructure connected to a condenser with mini fans providing external forced convection. The cooling system is characterized by a parametric study which determines the effects of volume fill ratio of coolant, existence of a boiling enhancement microstructure and pump flow rates on thermal performance of the closed loop. Experimental data shows the optimal parametric values which can dissipate 33.9W with a film heater maintained at 95 .deg. C

  7. Thermal performance of cooling system for a laptop computer using a boiling enhancement microstructure

    Energy Technology Data Exchange (ETDEWEB)

    Cho, N. H.; Jeong, W. Y.; Park, S. H. [Kumoh National Institute of Technology, Gumi (Korea, Republic of)

    2008-07-01

    The increasing heat generation rates in CPU of notebook computers motivate a research on cooling technologies with low thermal resistance. This paper develops a closed-loop two-phase cooling system using a micropump to circulate a dielectric liquid(PF5060). The cooling system consists of an evaporator containing a boiling enhancement microstructure connected to a condenser with mini fans providing external forced convection. The cooling system is characterized by a parametric study which determines the effects of volume fill ratio of coolant, existence of a boiling enhancement microstructure and pump flow rates on thermal performance of the closed loop. Experimental data shows the optimal parametric values which can dissipate 33.9W with a film heater maintained at 95 .deg. C.

  8. Forced and free convection hydromagnetic flow past a vertical flat plate

    International Nuclear Information System (INIS)

    Abdelkhalek, M.M.

    2004-01-01

    The effects of magnetic field and temperature heat source on the free and forced convection flow past an infinite vertical plate is studied analytically. Solutions of the reduced equation appropriate in the forced convection and free convection regime are obtained using perturbation technique. The expression for the velocity field, skin friction and Nusselt number have been obtained

  9. Modeling the natural convective flow of micropolar nanofluids

    KAUST Repository

    Bourantas, Georgios

    2014-01-01

    A micropolar model for nanofluidic suspensions is proposed in order to investigate theoretically the natural convection of nanofluids. The microrotation of the nanoparticles seems to play a significant role into flow regime and in that manner it possibly can interpret the controversial experimental data and theoretical numerical results over the natural convection of nanofluids. Natural convection of a nanofluid in a square cavity is studied and computations are performed for Rayleigh number values up to 106, for a range of solid volume fractions (0 ≤ φ ≤ 0.2) and, different types of nanoparticles (Cu, Ag, Al2O3 and TiO 2). The theoretical results show that the microrotation of the nanoparticles in suspension in general decreases overall heat transfer from the heated wall and should not therefore be neglected when computing heat and fluid flow of micropolar fluids, as nanofluids. The validity of the proposed model is depicted by comparing the numerical results obtained with available experimental and theoretical data. © 2013 Elsevier Ltd. All rights reserved.

  10. Natural convection heat transfer between vertical channel with flow resistance at the lower end

    International Nuclear Information System (INIS)

    Iwamoto, S.; Nishimura, S.; Ishihara, I.

    2003-01-01

    For natural convection in the geometrically complicated channel, the convection flow is suppressed by flow resistance due to such channel itself and the lopsided flow may take place. This could result in serious influences on the heat transfer in the channel. In order to investigate fundamentally the natural convection flow and heat transfer in such the channel, the vertical channel in which wall was heated with uniform heat flux and the flow resistance was given by small clearance between the lower end of channel and a wide horizontal floor. Flow pattern was observed by illuminating smoke filled in the channel and heat transfer rate was measured. (author)

  11. Heat transfer phenomena related to the boiling crisis

    International Nuclear Information System (INIS)

    Groenveld, D.C.

    1981-03-01

    This report contains a state-of-the-art review of critical heat flux (CHF) and post-CHF heat transfer. Part I reviews the mechanisms controlling the boiling crisis. The observed parametric trends of the CHF in a heat flux controlled system are discussed in detail, paying special attention to parameters pertaining to nuclear fuel. The various methods of predicting the critical power are described. Part II reviews the published information on transition boiling and film boiling heat transfer under forced convective conditions. Transition boiling data were found to be available only within limited ranges of conditions. The data did not permit the derivation of a correlation; however, the parametric trends were isolated from these data. (author)

  12. A review of investigations on flow instabilities in natural circulation boiling loops

    International Nuclear Information System (INIS)

    Gonella V Durga Prasad; Manmohan Pandey; Manjeet S Kalra

    2005-01-01

    Full text of publication follows: Steam generation systems are subjected to flow instabilities due to parametric fluctuations, inlet conditions etc., which may result in mechanical vibrations of components (called flow induced vibrations) and system control problems. Analysis of these instabilities in natural circulation boiling loops is very important for the safety of nuclear reactors and other boiling systems. This paper presents the state of the art in this area by reviewing over 100 contributions made in the past 30 years. A large number of experimental and numerical investigations have been conducted to study and understand the conditions for inception of flow instabilities, parametric effects of instabilities, and the system behavior under such conditions. Work done on instabilities due to channel thermal-hydraulics as well as neutronics-thermohydraulics coupling has been reviewed. Different methods of analysis used by researchers and results obtained by them have been discussed. Various numerical techniques adopted and computer codes developed have also been discussed. The knowledge obtained from the investigations made in the past three decades has been summarized to present the state of the art of the understanding of flow instabilities. (authors)

  13. Mechanism of subcooled water flow boiling critical heat flux in a circular tube at high liquid Reynolds number

    International Nuclear Information System (INIS)

    Hata, K.; Fukuda, K.; Masuzaki, S.

    2014-01-01

    The subcooled boiling heat transfer and the steady state critical heat flux (CHF) in a vertical circular tube for the flow velocities (u=3.95 to 30.80 m/s) are systematically measured by the experimental water loop comprised of a multistage canned-type circulation pump with high pump head. The SUS304 test tube of inner diameter (d=6 mm) and heated length (L=59.5 mm) is used in this work. The outer surface temperatures of the SUS304 test tube with heating are observed by an infrared thermal imaging camera and a video camera. The subcooled boiling heat transfers for SUS304 test tube are compared with the values calculated by other workers' correlations for the subcooled boiling heat transfer. The influence of flow velocity on the subcooled boiling heat transfer and the CHF is investigated into details based on the experimental data. Nucleate boiling surface superheats at the CHF are close to the lower limit of the heterogeneous spontaneous nucleation temperature and the homogeneous spontaneous nucleation temperature. The dominant mechanism of the subcooled flow boiling CHF on the SUS304 circular tube is discussed at high liquid Reynolds number. On the other hand, theoretical equations for k-ε turbulence model in a circular tube of a 3 mm in diameter and a 526 mm long are numerically solved for heating of water on heated section of a 3 mm in diameter and a 67 mm long with various thicknesses of conductive sub-layer by using PHOENICS code under the same conditions as the experimental ones previously obtained considering the temperature dependence of thermo-physical properties concerned. The Platinum (Pt) test tube of inner diameter (d=3 mm) and heated length (L=66.5 mm) was used in this experiment. The thicknesses of conductive sub-layer from non-boiling regime to CHF are clarified. The thicknesses of conductive sub-layer at the CHF point are evaluated for various flow velocities. The experimental values of the CHF are also compared with the corresponding

  14. Influence of the inertia and gravity on the boiling flows stability

    International Nuclear Information System (INIS)

    Delmastro, D.F.; Clausse, A.

    1987-01-01

    A study of boiling flows stability on the basis of a linear analysis is presented. From the homogeneous flows' conservation equations, a distributed parameters model, which allows to deal with the frequency field system, is obtained. The adimensional parameters which characterize the inertia effects and the gravity on the impulse equation, are identified. On the other hand, a mean volumes model which permits to gather analytic criteria helpful for the design and comprehension of the problem is developed. (Author)

  15. IR-thermography-based investigation of critical heat flux in subcooled flow boiling of water at atmospheric and high pressure conditions

    Energy Technology Data Exchange (ETDEWEB)

    Bucci, Matteo [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Seong, Jee H. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Buongiorno, Jdacopo [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Richenderfer, Andrew [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Kossolapov, A. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States)

    2017-11-01

    Here we report on MIT’s THM work in Q4 2016 and Q1 2017. The goal of this project is to design, construct and execute tests of flow boiling critical heat flux (CHF) at high-pressure using high-resolution and high-speed video and infrared (IR) thermometry, to generate unique data to inform the development of and validate mechanistic boiling heat transfer and CHF models. In FY2016, a new test section was designed and fabricated. Data was collected at atmospheric conditions at 10, 25 and 50 K subcoolings, and three mass fluxes, i.e. 500, 750 and 1000 kg/m2/s. Starting in Q4 2016 and continuing forward, new post-processing techniques have been developed to analyze the data collected. These new algorithms analyze the time-dependent temperature and heat flux distributions to calculate nucleation site density, nucleation frequency, growth and wait time, dry area fraction, and the complete heat flux partitioning. In Q1 2017 a new flow boiling loop was designed and constructed to support flow boiling tests up 10 bar pressure and 180 °C. Initial shakedown and testing has been completed. The flow loop and test section are now ready to begin high-pressure flow boiling testing.

  16. CFD for subcooled flow boiling: Simulation of DEBORA experiments

    International Nuclear Information System (INIS)

    Krepper, Eckhard; Rzehak, Roland

    2011-01-01

    Highlights: → In the DEBORA subcooled boiling tests using R12 are investigated. → Radial profiles of void fraction, liquid velocity, temperature and bubble sizes at the end of the heated length were measured. → The theoretical and experimental basis of correlations used in the wall boiling model are reviewed. → An assessment of the necessary recalibrations to describe the DEBORA tests is given. → With increased generated vapour the gas fraction profile changes from wall to core peaking, not captured by the present modelling. - Abstract: In this work we investigate the present capabilities of CFD for wall boiling. The computational model used combines the Euler/Euler two-phase flow description with heat flux partitioning. Very similar modelling was previously applied to boiling water under high pressure conditions relevant to nuclear power systems. Similar conditions in terms of the relevant non-dimensional numbers have been realized in the DEBORA tests using dichlorodifluoromethane (R12) as the working fluid. This facilitated measurements of radial profiles for gas volume fraction, gas velocity, liquid temperature and bubble size. After reviewing the theoretical and experimental basis of correlations used in the model, give a careful assessment of the necessary recalibrations to describe the DEBORA tests. It is then shown that within a certain range of conditions different tests can be simulated with a single set of model parameters. As the subcooling is decreased and the amount of generated vapour increases the gas fraction profile changes from wall to core peaking. This is a major effect not captured by the present modelling. Some quantitative deviations are assessed as well and directions for further model improvement are outlined.

  17. Nucleate boiling heat transfer on horizontal tubes in bundles

    International Nuclear Information System (INIS)

    Fujital, Y.; Ohta, H.; Hidaka, S.; Nishikawa, K.

    1986-01-01

    In order to clarify the heat transfer mechanisms of the flooded type horizontal tube bundle evaporator, heat transfer characteristics of tube bundles of experimental scale which consist both of smooth and enhanced tubes were investigated in detail. The experiments of saturated nucleate boiling were performed by using Freon 113 under pressures 0.1 to 1 MPa, and the effects of various parameters, for example, bundle arrangement, heat flux, pressure on the characteristics of an individual tube are clarified. Experimental data is reproduced well by a proposed heat transfer model in which convective heat transfer coefficients due to rising bubbles are estimated as a function of their volumetric flow rate

  18. Prediction for flow boiling heat transfer in small diameter tube using deep learning

    International Nuclear Information System (INIS)

    Enoki, Koji; Sei, Yuichi; Okawa, Tomio; Saito, Kiyoshi

    2017-01-01

    The applications of Artificial Intelligence ie AI show diversity in any fields. On the other hand, research of the predicting heat transfer regardless of single-phase or two-phase flow is still untouched. Therefore, we have confirmed usefulness using AI's deep learning function on horizontal flow boiling heat transfer in flowing mini-channel that is actively researched. The effect of the surface tension in the mini-channel is large compared with conventional large tubes, and then the heat transfer mechanism is very complicated. For this reason, the numerical correlations of many existing researchers the prediction result is not good. However, the mechanistic correlation based on the visualization experiment, which the authors' research group published several years ago has very high precision. Therefore, in this research paper, we confirmed the effectiveness of using deep learning for predicting of the boiling heat transfer in mini-channel while comparing our correlation. (author)

  19. Influence of a flow obstacle on the occurrence of burnout in boiling two-phase upward flow within a vertical annular channel

    Energy Technology Data Exchange (ETDEWEB)

    Mori, S.; Fukano, T. E-mail: fukanot@mech.kyushu-u.ac.jp

    2003-10-01

    When a flow obstruction such as a cylindrical spacer is set in a boiling two-phase flow within an annular channel, the inner tube of which is used as a heater, the temperature on the surface of the heating tube is severely affected by its existence. In some cases, the cylindrical spacer has a cooling effect, and in the other cases it causes the dryout of the cooling water film on the heating surface resulting in the burnout of the heating tube. In the present paper, we have focused our attention on the influence of a flow obstacle on the occurrence of burnout of the heating tube in boiling two-phase flow. The results are summarized as follows: - When the heat flux approaches the burnout condition, the wall temperature on the heating tube fluctuates with a large amplitude. And once the wall temperature exceeds the Leidenfrost temperature, the burnout occurs without exception. - The trigger of dryout of the water film which causes the burnout is not the nucleate boiling but the evaporation of the base film between disturbance waves. - The burnout never occurs at the downstream side of the spacer. This is because the dryout area downstream of the spacer is rewetted easily by the disturbance waves.

  20. Influence of a flow obstacle on the occurrence of burnout in boiling two-phase upward flow within a vertical annular channel

    International Nuclear Information System (INIS)

    Mori, S.; Fukano, T.

    2003-01-01

    When a flow obstruction such as a cylindrical spacer is set in a boiling two-phase flow within an annular channel, the inner tube of which is used as a heater, the temperature on the surface of the heating tube is severely affected by its existence. In some cases, the cylindrical spacer has a cooling effect, and in the other cases it causes the dryout of the cooling water film on the heating surface resulting in the burnout of the heating tube. In the present paper, we have focused our attention on the influence of a flow obstacle on the occurrence of burnout of the heating tube in boiling two-phase flow. The results are summarized as follows: - When the heat flux approaches the burnout condition, the wall temperature on the heating tube fluctuates with a large amplitude. And once the wall temperature exceeds the Leidenfrost temperature, the burnout occurs without exception. - The trigger of dryout of the water film which causes the burnout is not the nucleate boiling but the evaporation of the base film between disturbance waves. - The burnout never occurs at the downstream side of the spacer. This is because the dryout area downstream of the spacer is rewetted easily by the disturbance waves

  1. Investigation of the transition from forced to natural convection in the research reactor Munich II

    International Nuclear Information System (INIS)

    Skreba, S.; Adamek, J.; Unger, H.

    1999-01-01

    up to limiting amplitudes of 0.1 bar, 0.2 bar and 0.3 bar at the transition from forced to natural convection have been determined. Further tests have been performed to determine minimum heat flux densities leading to boiling processes in the cooling channel and critical heat flux densities causing dry outs of the cooling channel at downwards directed forced flow. During the tests, flow reversals have been observed because of the buoyancy forces in the coolant causing a mixed convection flow. The last test series, which has been finished in March 1999, has been performed in order to determine critical heat flux densities during the transition from forced to natural convection and to measure the occurring pressure amplitudes. All results prove the possibility to remove the decay heat of the FRM-II by natural convection, even in case of a complete loss of the active decay heat removal system. Above this, large safety margins in the FRM-II concerning pressure pulsations, beginning of boiling and dry out could be verified. (author)

  2. Mixed convective flow of immiscible viscous fluids confined between ...

    African Journals Online (AJOL)

    user

    International Journal of Engineering, Science and Technology ... finite difference methods to analyze the problem of natural convection boundary layer flow along a complex vertical surface ... analyzed the flow of two immiscible fluids in a parallel plate channel ... wavy and flat walls are maintained at constant temperatures w.

  3. Improved Performance of Personalized Ventilation by Control of the Convection Flow around Occupant Body

    DEFF Research Database (Denmark)

    Bolashikov, Zhecho Dimitrov; Melikov, Arsen Krikor; Krenek, Miroslav

    2009-01-01

    This paper reports on methods of control of the free convection flow around human body aiming at improvement of inhaled air quality for occupants at workstations with personalized ventilation (PV). Two methods of control were developed and explored: passive - blocking the free convection developm......This paper reports on methods of control of the free convection flow around human body aiming at improvement of inhaled air quality for occupants at workstations with personalized ventilation (PV). Two methods of control were developed and explored: passive - blocking the free convection......-scale test room with background mixing ventilation. Thermal manikin with realistic free convection flow was used. The PV supplied air from front/above towards the face. All measurements were performed under isothermal conditions at 20 °C and 26 °C. The air in the test room was mixed with tracer gas, while...

  4. Combined Lorentz force and ultrasound Doppler velocimetry in a vertical convection liquid metal flow

    Science.gov (United States)

    Zürner, Till; Vogt, Tobias; Resagk, Christian; Eckert, Sven; Schumacher, Jörg

    2017-11-01

    We report experimental studies on turbulent vertical convection flow in the liquid metal alloy gallium-indium-tin. Flow measurements were conducted by a combined use of local Lorentz force velocimetry (LLFV) and ultrasound Doppler velocimetry (UDV). It is known that the forced convection flow in a duct generates a force on the LLFV magnet system, that grows proportional to the flow velocity. We show that for the slower flow of natural convection LLFV retains this linear dependence in the range of micronewtons. Furthermore experimental results on the scaling of heat and momentum transport with the thermal driving are presented. The results cover a range of Rayleigh numbers 3 ×105 Deutsche Forschungsgemeinschaft under Grant No. GRK 1567.

  5. Experimental analysis of refrigerants flow boiling inside small sized microfin tubes

    Science.gov (United States)

    Diani, Andrea; Rossetto, Luisa

    2017-07-01

    The refrigerant charge reduction is one of the most challenging issues that the scientific community has to cope to reduce the anthropic global warming. Recently, mini microfin tubes have been matter of research, since they can reach better thermal performance in small domains, leading to a further refrigerant charge reduction. This paper presents experimental results about R134a flow boiling inside a microfin tube having an internal diameter at the fin tip of 2.4 mm. The mass flux was varied between 375 and 940 kg m-2 s-1, heat flux from 10 to 50 kW m-2, vapor quality from 0.10 to 0.99. The saturation temperature at the inlet of the test section was kept constant and equal to 30 °C. R134a thermal and fluid dynamic performances are presented and compared against those obtained with R1234ze(E) and R1234yf and against values obtained during R134a flow boiling inside a 3.4 mm ID microfin tube.

  6. Measurement on the effect of sound wave in upper plenum of boiling water reactor

    International Nuclear Information System (INIS)

    Kumagai, Kosuke; Someya, Satoshi; Okamoto, Koji

    2009-01-01

    In recent years, the power uprate of Boiling Water Reactors have been conducted at several existing power plants as a way to improve plant economy. In one of the power uprated plants (117.8% uprates) in the United States, the steam dryer breakages due to fatigue fracture occurred. It is conceivable that the increased steam flow passing through the branches caused a self-induced vibration with the propagation of sound wave into the steam-dome. The resonance among the structure, flow and the pressure fluctuation resulted in the breakages. To understand the basic mechanism of the resonance, previous researches were done by a point measurement of the pressure and by a phase averaged measurement of the flow, while it was difficult to detect the interaction among them by the conventional method. In this study, Dynamic Particle Image Velocimetry (PIV) System was applied to investigate the effect of sound on natural convection and forced convection. Especially, when the phases of acoustic sources were different, various acoustic wave effects were checked. (author)

  7. SCDAP/RELAP5 modeling of fluid heat transfer and flow losses through porous debris in a light water reactor

    International Nuclear Information System (INIS)

    Harvego, E. A.; Siefken, L. J.

    2000-01-01

    The SCDAP/RELAP5 code is being developed at the Idaho National Engineering and Environmental Laboratory under the primary sponsorship of the U.S. Nuclear Regulatory Commission (NRC) to provide best-estimate transient simulations of light water reactor coolant systems during severe accidents. This paper describes the modeling approach used in the SCDAP/RELAP5 code to calculate fluid heat transfer and flow losses through porous debris that has accumulated in the vessel lower head and core regions during the latter stages of a severe accident. The implementation of heat transfer and flow loss correlations into the code is discussed, and calculations performed to assess the validity of the modeling approach are described. The different modes of heat transfer in porous debris include: (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, (5) film boiling, and (6) transition from film boiling to convection to vapor. The correlations for flow losses in porous debris include frictional and form losses. The correlations for flow losses were integrated into the momentum equations in the RELAP5 part of the code. Since RELAP5 is a very general non-homogeneous non-equilibrium thermal-hydraulics code, the resulting modeling methodology is applicable to a wide range of debris thermal-hydraulic conditions. Assessment of the SCDAP/RELAP5 debris bed thermal-hydraulic models included comparisons with experimental measurements and other models available in the open literature. The assessment calculations, described in the paper, showed that SCDAP/RELAP5 is capable of calculating the heat transfer and flow losses occurring in porous debris regions that may develop in a light water reactor during a severe accident

  8. Recent developments in the modeling of boiling heat transfer mechanisms

    International Nuclear Information System (INIS)

    Podowski, M.Z.

    2009-01-01

    Due to the importance of boiling for the analysis of operation and safety of nuclear reactors, extensive efforts have been made in the past to develop a variety of methods and tools to study boiling heat transfer for various geometries and operating conditions. Recent progress in the computational multiphase fluid dynamics (CMFD) methods of two- and multiphase flows has already started opening up new exciting possibilities for using complete multidimensional models to predict the operation of boiling systems under both steady-state and transient conditions. However, such models still require closure laws and boundary conditions, the accuracy of which determines the predictive capabilities of the overall models and the associated CMFD simulations. Because of the complexity of the underlying physical phenomena, boiling heat transfer has traditionally been quantified using phenomenological models and correlations obtained by curve-fitting extensive experimental data. Since simple heuristic formulae are not capable of capturing the effect of various specific experimental conditions and the associated wide scattering of data points, most existing correlations are characterized by large uncertainties which are typically hidden behind the 'logarithmic scale' format of plots. Furthermore, such an approach provides only limited insight into the local phenomena of: nucleation, heated surface material properties, temperature fluctuations, and others. The objectives of this paper are two-fold. First, the state of the art is reviewed in the area of modeling concepts for both pool boiling and forced-convection (bulk and subcooled) boiling. Then, new results are shown concerning the development of new mechanistic models and their validation against experimental data. It is shown that a combination of the proposed theoretical approach with advanced computational methods leads to a dramatic improvement in both our understanding of the physics of boiling and the predictive

  9. Optical studies of boiling heat transfer: insights and limitations

    International Nuclear Information System (INIS)

    Kenning, David B.R.

    2004-01-01

    Optical studies provide valuable insights into the complex mechanisms of boiling heat transfer but the large gradients of temperature (and therefore of refractive index) deflect light and multiple reflections at interfaces limit the distance over which observations can be made. Optical measurements are thought of as non-intrusive but this is rarely true. Because they are difficult and time consuming, they constrain the design of boiling experiments and are applied to limited ranges of conditions. There is a risk that deductions from the observations will be applied (not necessarily by the authors) more generally than is justified. These characteristics of optical studies are illustrated by examples from forced convective film boiling on spheres and pool nucleate boiling

  10. Film boiling heat transfer from a hot sphere falling in two-phase pool

    International Nuclear Information System (INIS)

    Bang, K. H.; Kim, K. Y.

    1998-01-01

    The purpose of the present study is to experimentally investigate film boiling heat trasfer from a hot sphere falling in steam-water two-phase pool, which is the key heat transfer mode in molten fuel and coolant mixing. To measure film boiling heat transfer coefficients on a spere falling in two-phase pool, a heated sphere with a thermocouple embedded at the center is dropped in a vertical tube filled with steam-water mixture. The present experiment is unique in making the heated sphere fall through the two-phase pool while the previous experiments were performed with stationary spheres in flowing fluid. The falling speed of the sphere is measured using a set of magnet pickup coils distributed along the tube. The ranges of the experimental conditions are: spere fall speed 0-0.5 m/s, average void fraction 0-25,% steam superficial velocity 0-0.25 m/s. The results show that the forced convection film boiling heat transfer coefficient decrease slightly as the steam superficial velocity (void fraction) is increased

  11. Experimental investigation of onset of nucleate boiling in this rectangular channels

    International Nuclear Information System (INIS)

    Belhadj, M.; Christensen, R.N.; Aldemir, T.

    1988-01-01

    The 10 kW, HEU fueled Ohio State University Research Reactor (OSURR) will be upgraded to operate with plate type LEU U 3 Si 2 , fuel elements in the power range 250-500 kW. The core will be cooled by natural convection and an onset of nucleate boiling (ONB) margin of 1.2 will be maintained in the hot channel under steady-state operation. The validity of the correlations used for predicting ONB in plate type research reactors is not known for low heat flux-low velocity flows. An experiment has been set up at The Ohio State University to investigate ONB for laminar flow in this rectangular channels. The results show that: The Bergles-Rohsenow correlation and the correlation proposed by Ricque and Siboul predict higher and lower ONB fluxes than actual, respectively. The ONB heat flux is flow velocity dependent

  12. Void fraction prediction in saturated flow boiling

    International Nuclear Information System (INIS)

    Francisco J Collado

    2005-01-01

    Full text of publication follows: An essential element in thermal-hydraulics is the accurate prediction of the vapor void fraction, or fraction of the flow cross-sectional area occupied by steam. Recently, the author has suggested to calculate void fraction working exclusively with thermodynamic properties. It is well known that the usual 'flow' quality, merely a mass flow rate ratio, is not at all a thermodynamic property because its expression in function of thermodynamic properties includes the slip ratio, which is a parameter of the process not a function of state. By the other hand, in the classic and well known expression of the void fraction - in function of the true mass fraction of vapor (also called 'static' quality), and the vapor and liquid densities - does not appear the slip ratio. Of course, this would suggest a direct procedure for calculating the void fraction, provided we had an accurate value of the true mass fraction of vapor, clearly from the heat balance. However the classic heat balance is usually stated in function of the 'flow' quality, what sounds really contradictory because this parameter, as we have noted above, is not at all a thermodynamic property. Then we should check against real data the actual relationship between the thermodynamic properties and the applied heat. For saturated flow boiling just from the inlet of the heated tube, and not having into account the kinetic and potential terms, the uniform applied heat per unit mass of inlet water and per unit length (in short, specific linear heat) should be closely related to a (constant) slope of the mixture enthalpy. In this work, we have checked the relation between the specific linear heat and the thermodynamic enthalpy of the liquid-vapor mixture using the actual mass fraction. This true mass fraction is calculated using the accurate measurements of the outlet void fraction taken during the Cambridge project by Knights and Thom in the sixties for vertical and horizontal

  13. A phenomenological model of thermal-hydraulics of convective boiling during the quenching of hot rod bundles

    International Nuclear Information System (INIS)

    Unal, C.; Nelson, R.

    1991-01-01

    After completion of the thermal-hydraulic model developed in a companion paper, the authors performed developmental assessment calculation of the model using steady-state and transient post-critical heat flux (CHF) data. This paper discusses the results of those calculations. The overall interfacial drag model predicted reasonable drag coefficients for both the nucleate boiling and the inverted annular flow (IAF) regimes. The predicted pressure drops agreed reasonably well with the measured data of two transient experiments, CCTF Run 14 and a Lehigh reflood test. The thermal-hydraulic model for post-CHF convective heat transfer predicted the rewetting velocities reasonably well for both experiments. The predicted average slope of the wall temperature traces for these tests showed reasonable agreement with the measured data, indicating that the transient-calculated precursory cooling rates agreed with measured data. The hot-patch model, in conjunction with the other thermal-hydraulic models, was capable of modeling the Winfrith post-CHF hot-patch experiments. The hot-patch model kept the wall temperatures at the specified levels in the hot-patch regions and did not allow any quench-front propagation from either the bottom or the top of the test section. The interfacial heat-transfer model tended to slightly underpredict the vapor temperatures. The maximum difference between calculated and measured vapor temperatures was 20%, with a 10% difference for the remainder of the runs considered. The wall-to-fluid heat transfer was predicted reasonably well, and the predicted wall temperatures were in reasonable agreement with measured data with a maximum relative error of less than 13%

  14. Thermal hydraulic test for reactor safety system; a visualization study on flow boiling and bubble behavior

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Soon Heung; Baek, Won Pil; Ban, In Cheol [Korea Advanced Institute of Science and Technology, Taejeon (Korea)

    2002-03-01

    The project contribute to understand and to clarify the physical mechanism of flow nucleate boiling and CHF phenomena through the visualization experiments. the results are useful in the development of the enhancement device of heat transfer and to enhance nuclear fuel safety 1. Visual experimental facility 2. Application method of visualization Technique 3. Visualization results of flow nucleate boiling regime - Overall Bubble Behavior on the Heated Surface - Bubble Behavior near CHF Condition - Identification of Flow Structure - Three-layer flow structure 4. Quantifying of bubble parameter through a digital image processing - Image Processing Techniques - Classification of objects and measurements of the size - Three dimensional surface plot with using the luminance 5. Development and estimation of a correlation between bubble diameter and flow parameter - The effect of system parameter on bubble diameter - The development of a bubble diameter correlation . 49 refs., 42 figs., 7 tabs. (Author)

  15. Fluid flow and convective transport of solutes within the intervertebral disc.

    Science.gov (United States)

    Ferguson, Stephen J; Ito, Keita; Nolte, Lutz P

    2004-02-01

    Previous experimental and analytical studies of solute transport in the intervertebral disc have demonstrated that for small molecules diffusive transport alone fulfils the nutritional needs of disc cells. It has been often suggested that fluid flow into and within the disc may enhance the transport of larger molecules. The goal of the study was to predict the influence of load-induced interstitial fluid flow on mass transport in the intervertebral disc. An iterative procedure was used to predict the convective transport of physiologically relevant molecules within the disc. An axisymmetric, poroelastic finite-element structural model of the disc was developed. The diurnal loading was divided into discrete time steps. At each time step, the fluid flow within the disc due to compression or swelling was calculated. A sequentially coupled diffusion/convection model was then employed to calculate solute transport, with a constant concentration of solute being provided at the vascularised endplates and outer annulus. Loading was simulated for a complete diurnal cycle, and the relative convective and diffusive transport was compared for solutes with molecular weights ranging from 400 Da to 40 kDa. Consistent with previous studies, fluid flow did not enhance the transport of low-weight solutes. During swelling, interstitial fluid flow increased the unidirectional penetration of large solutes by approximately 100%. Due to the bi-directional temporal nature of disc loading, however, the net effect of convective transport over a full diurnal cycle was more limited (30% increase). Further study is required to determine the significance of large solutes and the timing of their delivery for disc physiology.

  16. Effects of Cooling Fluid Flow Rate on the Critical Heat Flux and Flow Stability in the Plate Fuel Type 2 MW TRIGA Reactor

    OpenAIRE

    H. P. Rahardjo; V. I. Sri Wardhani

    2017-01-01

    The conversion program of the 2 MW TRIGA reactor in Bandung consisted of the replacement of cylindrical fuel (produced by General Atomic) with plate fuel (produced by BATAN). The replacement led into the change of core cooling process from upward natural convection type to downward forced convection type, and resulted in different thermohydraulic safety criteria, such as critical heat flux (CHF) limit, boiling limit, and cooling fluid flow stability. In this paper, a thermohydraulic safety an...

  17. NATO Advanced Study Institute on Buoyant Convection in Geophysical Flows

    CERN Document Server

    Fedorovich, E; Viegas, D; Wyngaard, J

    1998-01-01

    Studies of convection in geophysical flows constitute an advanced and rapidly developing area of research that is relevant to problems of the natural environment. During the last decade, significant progress has been achieved in the field as a result of both experimental studies and numerical modelling. This led to the principal revision of the widely held view on buoyancy-driven turbulent flows comprising an organised mean component with superimposed chaotic turbulence. An intermediate type of motion, represented by coherent structures, has been found to play a key role in geophysical boundary layers and in larger scale atmospheric and hydrospheric circulations driven by buoyant forcing. New aspects of the interaction between convective motions and rotation have recently been discovered and investigated. Extensive experimental data have also been collected on the role of convection in cloud dynamics and microphysics. New theoretical concepts and approaches have been outlined regarding scaling and parameteriz...

  18. Sunspots and the physics of magnetic flux tubes. VI - Convective propulsion. VII - Heat flow in a convective downdraft

    Science.gov (United States)

    Parker, E. N.

    1979-01-01

    The effect of negative aerodynamic drag in an ideal fluid subject to convective instability is considered. It is shown that a cylinder moving in such a fluid is propelled forward in its motion by the convective forces and that the characteristic acceleration time is comparable to the onset time of convective motions in the fluid. It is suggested that convective propulsion plays an important role in the dynamics of flux tubes extending through the surface of the sun. The suppression of the upward heat flow in a Boussinesq convective cell with free upper and lower boundaries by a downdraft is then analyzed. Application to the solar convection zone indicates that downdrafts of 1 to 2 km/s at depths of 1000 to 4000 km beneath the visible surface of the sun are sufficient to reduce the upward heat flux to a small fraction of the ambient value.

  19. A Photographic study of subcooled flow boiling burnout at high heat flux and velocity

    Energy Technology Data Exchange (ETDEWEB)

    Celata, G.P.; Mariani, A.; Zummo, G. [ENEA, National Institute of Thermal-Fluid Dynamics, Rome (Italy); Cumo, M. [University of Rome (Italy); Gallo, D. [University of Palermo (Italy). Department of Nuclear Engineering

    2007-01-15

    The present paper reports the results of a visualization study of the burnout in subcooled flow boiling of water, with square cross section annular geometry (formed by a central heater rod contained in a duct characterized by a square cross section). The coolant velocity is in the range 3-10m/s. High speed movies of flow pattern in subcooled flow boiling of water from the onset of nucleate boiling up to physical burnout of the heater are recorded. From video images (single frames taken with a stroboscope light and an exposure time of 1{mu}s), the following general behaviour of vapour bubbles was observed: when the rate of bubble generation is increasing, with bubbles growing in the superheated layer close to the heating wall, their coalescence produces a type of elongated bubble called vapour blanket. One of the main features of the vapour blanket is that it is rooted to the nucleation site on the heated surface. Bubble dimensions are given as a function of thermal-hydraulic tested conditions for the whole range of velocity until the burnout region. A qualitative analysis of the behaviour of four stainless steel heater wires with different macroscopic surface finishes is also presented, showing the importance of this parameter on the dynamics of the bubbles and on the critical heat flux. (author)

  20. Onset of nuclear boiling in forced convection (Method of detection)

    International Nuclear Information System (INIS)

    Rachedi, M.

    1986-01-01

    Local onset of boiling in any pressure water cooling systems, as a PWR for instance, can mean a possible dangerous mismatch between the produced heat and the cooling capabilities. Its consequences can lead to serious accidental conditions and a reliable technique to detect such a phenomenon is therefore of particular need. Most techniques used up to now rely basically on local measurements and assume therefore usually the previous knowledge of the actual hot or boiling spot. The method proposed here based on externally located accelerometers appears to be sensitive to the global behaviour of the mechanical structure and is therefore not particularly bound to any exact localization of the sensors. The vibrations produced in the mechanical structure of the heated assembly are measured by accelerometers placed on the external surfaces that are easily accessible. The onset of the boiling, the growth and condensation of the bubbles on the heated wall, induces a resonance in the structure and an excitation at its particular eigen frequencies. Distinctive peaks are clearly observed in the spectral density function calculated from the accelerometer signal as soon as bubbles are produced. The technique is shown to be very sensitive even at the earliest phase of boiling and quite independent on sensor position. A complete hydrodynamic analysis of the experimental channels have been performed in order to assess the validity of the method both in steady conditions and during rapid power transients

  1. Can a Wind Model Mimic a Convection-Dominated Accretion Flow Model?

    Science.gov (United States)

    Chang, Heon-Young

    2001-06-01

    In this paper we investigate the properties of advection-dominated accretion flows(ADAFs) in case that outflows carry away infalling matter with its angular momentum and energy. Positive Bernoulli numbers in ADAFs allow a fraction of the gas to be ex-pelled in a form of outflows. The ADAFs are also unstable to convection. We present self-similar solutions for advection-dominated accretion flows in the presence of out-flows from the accretion flows (ADIOS). The axisymmetric flow is treated in variables integrated over polar sections and the effects of outflows on the accretion rlow are parameterized for possible configurations compatible with the one dimensional self-similar ADAF solution. We explicitly derive self-similar solutions of ADAFs in the presence of outflows and show that the strong outflows in the accretion flows result in a flatter density profile, which is similar to that of the convection-dominated accretion flows (CDAFs) in which convection transports the a! ngular momentum inward and the energy outward. These two different versions of the ADAF model should show similar behaviors in X-ray spectrum to some extent. Even though the two models may show similar behaviors, they should be distinguishable due to different physical properties. We suggest that for a central object of which mass is known these two different accretion flows should have different X-ray flux value due to deficient matter in the wind model.

  2. Can a Wind Model Mimic a Convection-Dominated Accretion Flow Model?

    Directory of Open Access Journals (Sweden)

    Heon-Young Chang

    2001-06-01

    Full Text Available In this paper we investigate the properties of advection-dominated accretion flows(ADAFs in case that outflows carry away infalling matter with its angular momentum and energy. Positive Bernoulli numbers in ADAFs allow a fraction of the gas to be ex-pelled in a form of outflows. The ADAFs are also unstable to convection. We present self-similar solutions for advection-dominated accretion flows in the presence of out-flows from the accretion flows (ADIOS. The axisymmetric flow is treated in variables integrated over polar sections and the effects of outflows on the accretion rlow are parameterized for possible configurations compatible with the one dimensional self-similar ADAF solution. We explicitly derive self-similar solutions of ADAFs in the presence of outflows and show that the strong outflows in the accretion flows result in a flatter density profile, which is similar to that of the convection-dominated accretion flows (CDAFs in which convection transports the a! ngular momentum inward and the energy outward. These two different versions of the ADAF model should show similar behaviors in X-ray spectrum to some extent. Even though the two models may show similar behaviors, they should be distinguishable due to different physical properties. We suggest that for a central object of which mass is known these two different accretion flows should have different X-ray flux value due to deficient matter in the wind model.

  3. Identification of flow patterns by neutron noise analysis during actual coolant boiling in thin rectangular channels

    International Nuclear Information System (INIS)

    Kozma, R.; van Dam, H.; Hoogenboom, J.E.

    1992-01-01

    The primary objective of this paper is to introduce results of coolant boiling experiments in a simulated materials test reactor-type fuel assembly with plate fuel in an actual reactor environment. The experiments have been performed in the Hoger Onderwijs Reactor (HOR) research reactor at the Interfaculty Reactor Institute, Delft, The Netherlands. In the analysis, noise signals of self-powered neutron detectors located in the neighborhood of the boiling region and thermocouple in the channel wall and in the coolant are used. Flow patterns in the boiling coolant have been identified by means of analysis of probability density functions and power spectral densities of neutron noise. It is shown that boiling has an oscillating character due to partial channel blockage caused by steam slugs generated periodically between the plates. The observed phenomenon can serve as a basis for a boiling detection method in reactors with plate-type fuels

  4. CFD simulation on critical heat flux of flow boiling in IVR-ERVC of a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Xiang, E-mail: zhangxiang3@snptc.com.cn [State Nuclear Power Technology Research & Development Center, South Area, Future Science and Technology Park, Chang Ping District, Beijing 102209 (China); Hu, Teng [State Nuclear Power Technology Research & Development Center, South Area, Future Science and Technology Park, Chang Ping District, Beijing 102209 (China); Chen, Deqi, E-mail: chendeqi@cqu.edu.cn [Key Laboratory of Low-grade Energy Utilization Technologies and Systems, Chongqing University, 400044 (China); Zhong, Yunke; Gao, Hong [Key Laboratory of Low-grade Energy Utilization Technologies and Systems, Chongqing University, 400044 (China)

    2016-08-01

    Highlights: • CFD simulation on CHF of boiling two-phase flow in ERVC is proposed. • CFD simulation result of CHF agrees well with that of experimental result. • The characteristics of boiling two-phase flow and boiling crisis are analyzed. - Abstract: The effectiveness of in-vessel retention (IVR) by external reactor vessel cooling (ERVC) strongly depends on the critical heat flux (CHF). As long as the local CHF does not exceed the local heat flux, the lower head of the pressure vessel can be cooled sufficiently to prevent from failure. In this paper, a CFD simulation is carried out to investigate the CHF of ERVC. This simulation is performed by a CFD code fluent couple with a boiling model by UDF (User-Defined Function). The experimental CHF of ERVC obtained by State Nuclear Power Technology Research and Development Center (SNPTRD) is used to validate this CFD simulation, and it is found that the simulation result agrees well with the experimental result. Based on the CFD simulation, detailed analysis focusing on the pressure distribution, velocity distribution, void fraction distribution, heating wall temperature distribution are proposed in this paper.

  5. SCDAP/RELAP5 Modeling of Heat Transfer and Flow Losses in Lower Head Porous Debris

    International Nuclear Information System (INIS)

    Siefken, Larry James; Coryell, Eric Wesley; Paik, Seungho; Kuo, Han Hsiung

    1999-01-01

    Designs are described for implementing models for calculating the heat transfer and flow losses in porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonporous debris that results from core material slumping into the lower head. Currently, the COUPLE model has the capability to model convective and radiative heat transfer from the surfaces of nonporous debris in a detailed manner and to model only in a simplistic manner the heat transfer from porous debris. In order to advance beyond the simplistic modeling for porous debris, designs are developed for detailed calculations of heat transfer and flow losses in porous debris. Correlations are identified for convective heat transfer in porous debris for the following modes of heat transfer; (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, and (5) film boiling. Interphase heat transfer is modeled in an approximate manner. Designs are described for models to calculate the flow losses and interphase drag of fluid flowing through the interstices of the porous debris, and to apply these variables in the momentum equations in the RELAP5 part of the code. Since the models for heat transfer and flow losses in porous debris in the lower head are designed for general application, a design is also described for implementation of these models to the analysis of porous debris in the core region. A test matrix is proposed for assessing the capability of the implemented models to calculate the heat transfer and flow losses in porous debris. The implementation of the models described in this report is expected to improve the COUPLE code calculation of the temperature distribution in porous debris and in the lower head that supports the debris. The implementation of these models is also expected to improve the calculation of the temperature and flow distribution in porous debris in the core region

  6. Surface Deformation by Thermo-capillary Convection -Sounding Rocket COMPERE Experiment SOURCE

    Science.gov (United States)

    Fuhrmann, Eckart; Dreyer, Michael E.

    The sounding rocket COMPERE experiment SOURCE was successfully flown on MASER 11, launched in Kiruna (ESRANGE), May 15th, 2008. SOURCE has been intended to partly ful-fill the scientific objectives of the European Space Agency (ESA) Microgravity Applications Program (MAP) project AO-2004-111 (Convective boiling and condensation). Three parties of principle investigators have been involved to design the experiment set-up: ZARM for thermo-capillary flows, IMFT (Toulouse, France) for boiling studies, EADS Astrium (Bremen, Ger-many) for depressurization. The scientific aims are to study the effect of wall heat flux on the contact line of the free liquid surface and to obtain a correlation for a convective heat transfer coefficient. The experiment has been conducted along a predefined time line. A preheating sequence at ground was the first operation to achieve a well defined temperature evolution within the test cell and its environment inside the rocket. Nearly one minute after launch, the pressurized test cell was filled with the test liquid HFE-7000 until a certain fill level was reached. Then the free surface could be observed for 120 s without distortion. Afterwards, the first depressurization was started to induce subcooled boiling, the second one to start saturated boiling. The data from the flight consists of video images and temperature measurements in the liquid, the solid, and the gaseous phase. Data analysis provides the surface shape versus time and the corresponding apparent contact angle. Computational analysis provides information for the determination of the heat transfer coefficient in a compensated gravity environment where a flow is caused by the temperature difference between the hot wall and the cold liquid. Correlations for the effective contact angle and the heat transfer coefficient shall be delivered as a function of the relevant dimensionsless parameters. The data will be used for benchmarking of commercial CFD codes and the tank design

  7. Modeling and numerical simulation of oscillatory two-phase flows, with application to boiling water nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Rosa, M.P. [Instituto de Estudos Avancados - CTA, Sao Paolo (Brazil); Podowski, M.Z. [Rensselaer Polytechnic Institute, Troy, NY (United States)

    1995-09-01

    This paper is concerned with the analysis of dynamics and stability of boiling channels and systems. The specific objectives are two-fold. One of them is to present the results of a study aimed at analyzing the effects of various modeling concepts and numerical approaches on the transient response and stability of parallel boiling channels. The other objective is to investigate the effect of closed-loop feedback on stability of a boiling water reactor (BWR). Various modeling and computational issues for parallel boiling channels are discussed, such as: the impact of the numerical discretization scheme for the node containing the moving boiling boundary on the convergence and accuracy of computations, and the effects of subcooled boiling and other two-phase flow phenomena on the predictions of marginal stability conditions. Furthermore, the effects are analyzed of local loss coefficients around the recirculation loop of a boiling water reactor on stability of the reactor system. An apparent paradox is explained concerning the impact of changing single-phase losses on loop stability. The calculations have been performed using the DYNOBOSS computer code. The results of DYNOBOSS validation against other computer codes and experimental data are shown.

  8. Measurement and analysis of bubble behavior in subcooled nucleate boiling flow field with high fidelity imaging system

    International Nuclear Information System (INIS)

    Wu, W.; Jones, B.G.; Newell, T.A.

    2004-01-01

    Axial offset anomaly (AOA) is an unexpected deviation in the core axial power distribution from the predicted curve. AOA is a current major consideration for reactors operating at increased power levels and is becoming immediate threat to nuclear power's competitiveness in the market. Despite much effort focusing on this topic, a comprehensive understanding is far from being developed. However, previous research indicates first, that a close connection exists between subcooled nucleate boiling occurring in core region and the formation of crud, which directly results in AOA phenomena, secondly, that deposition is greater, and sometimes much greater, on heated than on unheated surfaces. A number of researchers have suggested that boiling promotes deposition, and several observed increased deposition in the subcooled boiling region. Limited detailed information is available on the interaction between heat and mass transfer in subcooled nucleate boiling (SNB) flow. Bubbles formed in SNB region play an important role in helping the formation of crud. This research examines bubble behavior under SNB condition from the dynamic point of view, using a high fidelity digital imaging apparatus. Freon R-134a is chosen as a simulant fluid due to its merit of having smaller surface tension and lower boiling temperature. The apparatus is operated at reduced pressure. Series of images at frame rates up to 4000 frames/s were obtained, showing different characteristics of bubble behavior with varying experimental parameters e.g. flow velocity, fluid subcooled level, etc. Analyses that combine the experimental results with analytical result on flow field in velocity boundary layer are considered. A tentative suggestion is that a rolling movement of a bubble accompanies its sliding along the heating surface in the flow channel. Numerical computations using FLUENT v5.5 have been performed to support this conclusion

  9. Net vapor generation point in boiling flow of trichlorotrifluoroethane at high pressures

    Science.gov (United States)

    Dougall, R. S.; Lippert, T. E.

    1973-01-01

    The conditions at which the void in subcooled boiling starts to undergo a rapid increase were studied experimentally. The experiments were performed in a 12.7 x 9.5 mm rectangular channel. Heating was from a 3.2 mm wide strip embedded in one wall. The pressure ranged from 9.45 to 20.7 bar, mass velocity from 600 to 7000 kg/sq m sec, and subcooling from 16 to 67 C. Photographs were used to determine when detached bubbles first appeared in the bulk flow. Measurements of bubble layer thickness along the wall were also made. Results showed that the point of net vapor generation is close to the occurrence of fully-developed boiling.

  10. Natural Circulation with Boiling

    Energy Technology Data Exchange (ETDEWEB)

    Mathisen, R P

    1967-09-15

    A number of parameters with dominant influence on the power level at hydrodynamic instability in natural circulation, two-phase flow, have been studied experimentally. The geometrical dependent quantities were: the system driving head, the boiling channel and riser dimensions, the single-phase as well as the two phase flow restrictions. The parameters influencing the liquid properties were the system pressure and the test section inlet subcooling. The threshold of instability was determined by plotting the noise characteristics in the mass flow records against power. The flow responses to artificially obtained power disturbances at instability conditions were also measured in order to study the nature of hydrodynamic instability. The results presented give a review over relatively wide ranges of the main parameters, mainly concerning the coolant performance in both single and parallel boiling channel flow. With regard to the power limits the experimental results verified that the single boiling channel performance was intimately related to that of the parallel channels. In the latter case the additional inter-channel factors with attenuating effects were studied. Some optimum values of the parameters were observed.

  11. Nucleate and film pool boiling in R11: the effects of orientation

    International Nuclear Information System (INIS)

    Venart, J.E.S.; Sousa, A.C.M.; Jung, D.S.

    1985-01-01

    In order to understand and model the behaviour of LPG tanks in fires [1] it is necessary to characterize the internal flow and specify its boundary conditions. Tank storage and transport normally utilize horizontal cylinders or spheres and hence the interior fluid sees a variety of surfaces inclinations and heat fluxes. The purpose of this paper is to present results obtained in R11 as a function of heat flux (1-180 kW/m 2 ) and angle (0-80 o ) at pressures from 1 to 2 bars in the free convective, nucleate and film boiling regions. (author)

  12. Study of flow instabilities in double-channel natural circulation boiling systems

    International Nuclear Information System (INIS)

    Durga Prasad, Gonella V.; Pandey, Manmohan; Pradhan, Santosh K.; Gupta, Satish K.

    2008-01-01

    Natural circulation boiling systems consisting of parallel channels can undergo different types of oscillations (in-phase or out-of-phase) depending on the geometric parameters and operating conditions. Disturbances in one channel affect the flow in other channels, which triggers thermal-hydraulic oscillations. In the present work, the modes of oscillation under different operating conditions and channel-to-channel interaction during power fluctuations and on-power refueling in a double-channel natural circulation boiling system are investigated. The system is modeled using a lumped parameter mathematical model and RELAP5/MOD3.4. Parametric studies are carried out for an equal-power double-channel system, at different operating conditions, with both the models, and the results are compared. Instabilities, non-linear oscillations, and effects of recirculation loop dynamics and geometric parameters on the mode of oscillations, are studied using the lumped model. The two channels oscillate out-of-phase in Type-I region, but in Type-II region, both the modes of oscillation are observed under different conditions. Channel-to-channel interaction and on-power refueling studies are carried out using the RELAP model. At high powers, disturbances in one channel significantly affect the stability of the other channel. During on-power refueling, a near-stagnation condition or low-velocity reverse flow can occur, the possibility of reverse flow being higher at lower pressures

  13. Convective flows of colloidal suspension in an inclined closed cell

    Energy Technology Data Exchange (ETDEWEB)

    Smorodin, Boris; Ishutov, Sergey [Department of Physics of Phase Transitions, Perm State University, Perm (Russian Federation); Cherepanov, Ivan, E-mail: bsmorodin@yandex.ru [Department of Radio Electronics and Information Security, Perm State University, Perm (Russian Federation)

    2016-12-15

    The nonlinear spatiotemporal evolution of convective flows is numerically investigated in the case of colloidal suspension filling an inclined closed cell heated from below. The bifurcation diagram (the dependency of the Nusselt number on the Rayleigh number) is obtained. The characteristics of the wave and steady patterns are investigated depending on heat intensity. The travelling wave changing travel direction and the non-regular oscillatory flow are found to be stable solutions within a certain interval of the Rayleigh number. Temporal Fourier decomposition is used together with other diagnostic tools to analyse the complex bifurcation and spatiotemporal properties caused by the interplay of the gravity-induced gradient of concentration and convective mixing of the fluid. It is shown that a more complex flow structure exists at a lower heating intensity (Rayleigh number). (paper)

  14. Validation of a multidimensional computational fluid dynamics model for subcooled flow boiling analysis

    Energy Technology Data Exchange (ETDEWEB)

    Braz Filho, Francisco A.; Caldeira, Alexandre D.; Borges, Eduardo M., E-mail: fbraz@ieav.cta.b, E-mail: alexdc@ieav.cta.b, E-mail: eduardo@ieav.cta.b [Instituto de Estudos Avancados (IEAv/CTA), Sao Jose dos Campos, SP (Brazil). Div. de Energia Nuclear

    2011-07-01

    In a heated vertical channel, the subcooled flow boiling regime occurs when the bulk fluid temperature is lower than the saturation temperature, but the fluid temperature reaches the saturation point near the channel wall. This phenomenon produces a significant increase in heat flux, limited by the critical heat flux. This study is particularly important to the thermal-hydraulics analysis of pressurized water reactors. The purpose of this work is the validation of a multidimensional model to analyze the subcooled flow boiling comparing the results with experimental data found in literature. The computational fluid dynamics code FLUENT was used with Eulerian multiphase model option. The calculated values of wall temperature in the liquid-solid interface presented an excellent agreement when compared to the experimental data. Void fraction calculations presented satisfactory results in relation to the experimental data in pressures of 15, 30 and 45 bars. (author)

  15. Validation of a multidimensional computational fluid dynamics model for subcooled flow boiling analysis

    International Nuclear Information System (INIS)

    Braz Filho, Francisco A.; Caldeira, Alexandre D.; Borges, Eduardo M.

    2011-01-01

    In a heated vertical channel, the subcooled flow boiling regime occurs when the bulk fluid temperature is lower than the saturation temperature, but the fluid temperature reaches the saturation point near the channel wall. This phenomenon produces a significant increase in heat flux, limited by the critical heat flux. This study is particularly important to the thermal-hydraulics analysis of pressurized water reactors. The purpose of this work is the validation of a multidimensional model to analyze the subcooled flow boiling comparing the results with experimental data found in literature. The computational fluid dynamics code FLUENT was used with Eulerian multiphase model option. The calculated values of wall temperature in the liquid-solid interface presented an excellent agreement when compared to the experimental data. Void fraction calculations presented satisfactory results in relation to the experimental data in pressures of 15, 30 and 45 bars. (author)

  16. Developing the technique of image processing for the study of bubble dynamics in subcooled flow boiling

    International Nuclear Information System (INIS)

    Donevski, Bozin; Saga, Tetsuo; Kobayashi, Toshio; Segawa, Shigeki

    1998-01-01

    This study presents the development of an image processing technique for studying the dynamic behavior of vapor bubbles in a two-phase bubbly flow. It focuses on the quantitative assessment of some basic parameters such as a local bubble size and size distribution in the range of void fraction between 0.03 < a < 0.07. The image processing methodology is based upon the computer evaluation of high speed motion pictures obtained from the flow field in the region of underdeveloped subcooled flow boiling for a variety of experimental conditions. This technique has the advantage of providing computer measurements and extracting the bubbles of the two-phase bubbly flow. This method appears to be promising for determining the governing mechanisms in subcooled flow boiling, particularly near the point of net vapor generation. The data collected by the image analysis software can be incorporated into the new models and computer codes currently under development which are aimed at incorporating the effect of vapor generation and condensation separately. (author)

  17. Finite bandwidth, nonlinear convective flow in a mushy layer

    Energy Technology Data Exchange (ETDEWEB)

    Riahi, D N, E-mail: daniel.riahi@utrgv.edu [School of Mathematical and Statistical Sciences, University of Texas Rio Grande Valley, One West University Boulevard, Brownsville, TX 78520 (United States)

    2017-04-15

    Finite amplitude convection with a continuous finite bandwidth of modes in a horizontal mushy layer during the solidification of binary alloys is investigated. We analyze the nonlinear convection for values of the Rayleigh number close to its critical value by using multiple scales and perturbation techniques. Applying a combined temporal and spatial evolution approach, we determine a set of three coupled differential equations for the amplitude functions of the convective modes. A large number of new subcritical or supercritical stable solutions to these equations in the form of steady rolls and hexagons with different horizontal length scales are detected. We find, in particular, that depending on the parameter values and on the magnitude and direction of the wave number vectors for the amplitude functions, hexagons with down-flow or up-flow at the cells’ centers or rolls can be stable. Rolls or hexagons with longer horizontal wave length can be stable at higher amplitudes, and there are cases where hexagons are unstable for any value of the Rayleigh number, while rolls are stable only for the values of the Rayleigh number beyond some value. We also detected new stable and irregular flow patterns with two different horizontal scales in the form of superposition of either two sets of hexagons or two sets of inclined rolls. (paper)

  18. Convection Study by PIV Method Within Horizontal Liquid Layer Evaporating Into Inert Gas Flow

    Directory of Open Access Journals (Sweden)

    Kreta Aleksei

    2016-01-01

    Full Text Available The paper is devoted to the experimental study of convection in a horizontal evaporating liquid layer (ethanol of limited size under the action of gas flow (air. The two-dimensional velocity field in the liquid layer is obtained using the PIV method. The existence of a vortex convective flow within a liquid layer directed towards the gas flow has been revealed.

  19. Development of measurement method of void fraction distribution on subcooled flow boiling using neutron radiography

    International Nuclear Information System (INIS)

    Kureta, Masatoshi; Matsubayashi, Masahito; Akimoto, Hajime

    1999-03-01

    In relation to the development of a solid target of high intensity neutron source, plasma-facing components of fusion reactor and so forth, it is indispensable to estimate the void fraction for high-heat-load subcooled flow boiling of water. Since the existing prediction method of void fraction is based on the database for tubes, it is necessary to investigate extendibility of the existing prediction method to narrow-gap rectangular channels that is used in the high-heat-load devices. However, measurement method of void fraction in the narrow-gap rectangular channel has not been established yet because of the difficulty of measurement. The objectives of this investigation are development of a new system for bubble visualization and void fraction measurement on subcooled flow boiling in narrow-gap rectangular channels using the neutron radiography, and establishment of void fraction database by using this measurement system. This report describes the void fraction measurement method by the neutron radiography technique, and summarizes the measured void fraction data in one-side heated narrow-gap rectangular channels at subcooled boiling condition. (author)

  20. An assessment of in-tube flow boiling correlations for ammonia-water mixtures and their influence on heat exchanger size

    DEFF Research Database (Denmark)

    Kærn, Martin Ryhl; Modi, Anish; Jensen, Jonas Kjær

    2016-01-01

    on the required heat exchanger size (surface area)is investigated during numerical design. For this purpose, two case studies related to the use of the Kalina cycle are considered: a flue gas based heat recovery boiler for acombined cycle power plant and a hot oil based boiler for a solar thermal power plant......Heat transfer correlations for pool and flow boiling are indispensable for boiler design. The correlations for predicting in-tube flow boiling heat transfer ofammonia-water mixtures are not well established in the open literature and there is a lack of experimental measurements for the full range...... of composition, vapor qualities, fluid conditions, etc. This paper presents a comparison of several flow boiling heat transfer prediction methods (correlations) for ammonia-water mixtures. Firstly, these methods are reviewed and compared at various fluid conditions. The methods include: (1) the ammonia...

  1. Assessment of interfacial heat transfer models under subcooled flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Ribeiro, Guilherme B.; Braz Filho, Francisco A., E-mail: gbribeiro@ieav.cta.br, E-mail: fbraz@ieav.cta.br [Instituto de Estudos Avançados (DCTA/IEAv), São José dos Campos, SP (Brazil). Div. de Energia Nuclear

    2017-07-01

    The present study concerns a detailed analysis of subcooled flow boiling characteristics under high pressure systems using a two-fluid Eulerian approach provided by a Computational Fluid Dynamics (CFD) solver. For this purpose, a vertical heated pipe made of stainless steel with an internal diameter of 15.4 mm was considered as the modeled domain. An uniform heat flux of 570 kW/m2 and saturation pressure of 4.5 MPa were applied to the channel wall, whereas water mass flux of 900 kg/m2s was considered for all simulation cases. The model was validated against a set of experimental data and results have indicated a promising use of CFD technique for the estimation of wall temperature, the liquid bulk temperature and the location of the departure of nucleate boiling. Different sub-models of interfacial heat transfer coefficient were applied and compared, allowing a better prediction of void fraction along the heated channel. (author)

  2. SULTAN test facility for large-scale vessel coolability in natural convection at low pressure

    International Nuclear Information System (INIS)

    Rouge, S.

    1997-01-01

    The SULTAN facility (France/CEA/CENG) was designed to study large-scale structure coolability by water in boiling natural convection. The objectives are to measure the main characteristics of two-dimensional, two-phase flow, in order to evaluate the recirculation mass flow in large systems, and the limits of the critical heat flux (CHF) for a wide range of thermo-hydraulic (pressure, 0.1-0.5 MPa; inlet temperature, 50-150 C; mass flow velocity, 5-4400 kg s -1 m -2 ; flux, 100-1000 kW m -2 ) and geometric (gap, 3-15 cm; inclination, 0-90 ) parameters. This paper makes available the experimental data obtained during the first two campaigns (90 , 3 cm; 10 , 15 cm): pressure drop differential pressure (DP) = f(G), CHF limits, local profiles of temperature and void fraction in the gap, visualizations. Other campaigns should confirm these first results, indicating a favourable possibility of the coolability of large surfaces under natural convection. (orig.)

  3. SCDAP/RELAP5 Modeling of Heat Transfer and Flow Losses in Lower Head Porous Debris

    International Nuclear Information System (INIS)

    Coryell, E.W.; Siefken, L.J.; Paik, S.

    1998-01-01

    Designs are described for implementing models for calculating the heat transfer and flow losses in porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and non-porous debris that results from core material slumping into the lower head. Currently, the COUPLE model has the capability to model convective and radiative heat transfer from the surfaces of non-porous debris in a detailed manner and to model only in a simplistic manner the heat transfer from porous debris. In order to advance beyond the simplistic modeling for porous debris, designs are developed for detailed calculations of heat transfer and flow losses in porous debris. Correlations are identified for convective heat transfer in porous debris for the following modes of heat transfer; (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, and (5) film boiling. Interphase heat transfer is modeled in an approximate manner. A design is also described for implementing a model of heat transfer by radiation from debris to the interstitial fluid. A design is described for implementation of models for flow losses and interphase drag in porous debris. Since the models for heat transfer and flow losses in porous debris in the lower head are designed for general application, a design is also described for implementation of these models to the analysis of porous debris in the core region. A test matrix is proposed for assessing the capability of the implemented models to calculate the heat transfer and flow losses in porous debris. The implementation of the models described in this report is expected to improve the COUPLE code calculation of the temperature distribution in porous debris and in the lower head that supports the debris. The implementation of these models is also expected to improve the calculation of the temperature and flow distribution in porous debris in the core region

  4. MHD free convection flow of a non-Newtonian power-law fluid over ...

    African Journals Online (AJOL)

    ... flow have been presented for various parameters such as Prandtl number, flow behavior index (n), porous plate parameter and magnetic parameter. The local Nusselt number and skin friction coefficient is also presented graphically. Keywords: Magnetohydrodynamic flow; free convection flow; Non-Newtonian power-law

  5. Spectrally-consistent regularization modeling of turbulent natural convection flows

    International Nuclear Information System (INIS)

    Trias, F Xavier; Gorobets, Andrey; Oliva, Assensi; Verstappen, Roel

    2012-01-01

    The incompressible Navier-Stokes equations constitute an excellent mathematical modelization of turbulence. Unfortunately, attempts at performing direct simulations are limited to relatively low-Reynolds numbers because of the almost numberless small scales produced by the non-linear convective term. Alternatively, a dynamically less complex formulation is proposed here. Namely, regularizations of the Navier-Stokes equations that preserve the symmetry and conservation properties exactly. To do so, both convective and diffusive terms are altered in the same vein. In this way, the convective production of small scales is effectively restrained whereas the modified diffusive term introduces a hyperviscosity effect and consequently enhances the destruction of small scales. In practice, the only additional ingredient is a self-adjoint linear filter whose local filter length is determined from the requirement that vortex-stretching must stop at the smallest grid scale. In the present work, the performance of the above-mentioned recent improvements is assessed through application to turbulent natural convection flows by means of comparison with DNS reference data.

  6. SCDAP/RELAP5 modeling of heat transfer and flow losses in lower head porous debris. Rev. 1

    International Nuclear Information System (INIS)

    Siefken, L.J.; Coryell, E.W.; Paik, S.; Kuo, H.

    1999-01-01

    Designs are described for implementing models for calculating the heat transfer and flow losses in porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonporous debris that results from core material slumping into the lower head. Currently, the COUPLE model has the capability to model convective and radiative heat transfer from the surfaces of nonporous debris in a detailed manner and to model only in a simplistic manner the heat transfer from porous debris. In order to advance beyond the simplistic modeling for porous debris, designs are developed for detailed calculations of heat transfer and flow losses in porous debris. Correlations are identified for convective heat transfer in porous debris for the following modes of heat transfer; (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, and (5) film boiling. Interphase heat transfer is modeled in an approximate ma nner. Designs are described for models to calculate the flow losses and interphase drag of fluid flowing through the interstices of the porous debris, and to apply these variables in the momentum equations in the RELAP5 part of the code. Since the models for heat transfer and flow losses in porous debris in the lower head are designed for general application, a design is also described for implementation of these models to the analysis of porous debris in the core region. A test matrix is proposed for assessing the capability of the implemented models to calculate the heat transfer and flow losses in porous debris. The implementation of the models described in this report is expected to improve the COUPLE code calculation of the temperature distribution in porous debris and in the lower head that supports the debris. The implementation of these models is also expected to improve the calculation of the temperature and flow distribution in porous debris in the core region

  7. Assessment of correlations and models for prediction of CHF in subcooled flow boiling

    International Nuclear Information System (INIS)

    Celata, G.P.; Mariani, A.; Cumo, M.

    1992-01-01

    This paper provides an analysis of available correlations and models for the prediction of Critical Heat Flux (CHF) in subcooled flow boiling in the ranges of interest of fusion reactor thermal-hydraulic conditions, i.e., high inlet liquid subcooling and velocity and small channel diameter and length. The aim of the study was to establish the limits of validity of present predictive tools (most of them were proposed with reference to LWR thermal-hydraulic studies) in the above conditions. The reference data-set represents most of available data covering wide ranges of operating conditions in the framework of present interest (0.1 s ub, in < 230 K). Among the tens of predictive tools available in literature, four correlations (Levy, Westinghouse, modified-Tong and Tong-75) and three models (Weisman and Ileslamlou Lee and Mudawar and Katto) were selected. The modified-Tong correlation and the Katto model seem to be reliable predictive tools for the calculation of the CHF in subcooled flow boiling

  8. Intelligent information data base of flow boiling characteristics in once-through steam generator for integrated type marine water reactor

    International Nuclear Information System (INIS)

    Inasaka, Fujio; Nariai, Hideki

    1998-01-01

    Valuable experimental knowledge with flow boiling characteristics of the helical-coil type once-through steam generator was converted into an intelligent information data base program. The program was created as a windows application using the Visual Basic. Main functions of the program are as follows: (1) steady state flow boiling analysis of any helical-coil type once-through steam generator, (2) analysis and comparison with the experimental data, (3) reference and graph display of the steady state experimental data, (4) reference of the flow instability experimental data and display of the instability threshold correlated by each parameter, (5) summary of the experimental apparatus. (6) menu bar such as a help and print. In the steady state analysis, the region lengths of subcooled boiling, saturated boiling, and super-heating, and the temperature and pressure distributions etc. for secondary water calculated. Steady state analysis results agreed well with the experimental data, with the exception of the pressure drop at high mass velocity. The program will be useful for the design of not only the future integrated type marine water reactor but also the small sized water reactor with helical-coil type steam generator

  9. Convection flow structure in the central polar cap

    Science.gov (United States)

    Bristow, W. A.

    2017-12-01

    A previous study of spatially averaged flow velocity in the central polar cap [Bristow et al., 2015] observed under steady IMF conditions found that it was extremely rare for the average to exceed 850 m/s (less than 0.2 % of the time). Anecdotally, however it is not uncommon to observe line-of-sight velocities in excess of 100 m/s in the McMurdo radar field of view directly over the magnetic pole. This discrepancy motivated this study, which examines the conditions under which high-velocity flows are observed at latitudes greater than 80° magnetic latitude. It was found that highly structured flows are common in the central polar cap, which leads to the flow within regions to have significant deviation from the average. In addition, the high-speed flow regions are usually directed away from the earth-sun line. No specific set of driving conditions was identified to be associated with high-speed flows. The study did conclude that 1)Polar cap velocities are generally highly structured. 2)Flow patterns typically illustrate narrow channels, vortical flow regions, and propagating features. 3) Persistent waves are a regular occurrence. 3)Features are observed to propagate from day side to night side, and from night side to day side.. 4)Convection often exhibits significant difference between the two hemispheres. And 5)About 10% of the time the velocity somewhere in the cap exceeds 1 Km/s The presentation will conclude with a discussion of the physical reasons for the flow structure. Bristow, W. A., E. Amata, J. Spaleta, and M. F. Marcucci (2015), Observations of the relationship between ionospheric central polar cap and dayside throat convection velocities, and solar wind/IMF driving, J. Geophys. Res. Space Physics, 120, doi:10.1002/2015JA021199.

  10. Transient heat transfer for forced convection flow of helium gas

    International Nuclear Information System (INIS)

    Liu, Qiusheng; Fukuda, Katsuya; Sasaki, Kenji; Yamamoto, Manabu

    1999-01-01

    Transient heat transfer coefficients for forced convection flow of helium gas over a horizontal cylinder were measured using a forced convection test loop. The platinum heater with a diameter of 1.0 mm was heated by electric current with an exponential increase of Q 0 exp(t/τ). It was clarified that the heat transfer coefficient approaches the steady-state one for the period τ over 1 s, and it becomes higher for the period of τ shorter than 1 s. The transient heat transfer shows less dependent on the gas flowing velocity when the period becomes very shorter. Semi-empirical correlations for steady-state and transient heat transfer were developed based on the experimental data. (author)

  11. Visualization of boiling two-phase flow in a small diameter tube using neutron radiography

    International Nuclear Information System (INIS)

    Hibiki, Takashi; Mishima, Kaichiro; Yoneda, Kenji; Fujine, Shigenori; Kanda, Keiji; Nishihara, Hideaki

    1991-01-01

    The characteristics of boiling two-phase flow in a small diameter tube are very important for cooling the blanket in a nuclear fusion reactor or a high performance electronic device. For all these subjects, it is necessary to visualize the flow in a tube as a starting point of the study. However, when an optical method cannot be used for the visualization, it is expected that neutron radiography is useful. In this study, the feasibility of visualization of boiling two-phase flow in a small diameter tube was investigated by using various facilities of neutron radiography as the first step. The basic concept of neutron radiography and the block diagram of a neutron television system are shown. The neutron beam attenuated by water in the test section makes a scintillator emit visible light, and produces an image of two-phase flow, which is taken with a TV camera. Thus the image can be observed at real time. Three kinds of the experiments were performed with the facilities of KUR, NSRR and JRR-3. The experimental methods and the results are reported. The images obtained were sufficiently clear. (K.I.)

  12. Time-Dependent Natural Convection Couette Flow of Heat ...

    African Journals Online (AJOL)

    Time-Dependent Natural Convection Couette Flow of Heat Generating/Absorbing Fluid between Vertical Parallel Plates Filled With Porous Material. ... The numerical simulation conducted for some saturated liquids reveled that at t ≥ Pr the steady and unsteady state velocities (as well as the temperature of the fluid) ...

  13. An Experimental Study on the Pool Boiling Heat Transfer on a Square Surface

    International Nuclear Information System (INIS)

    Kim, Jae Kwang

    2000-02-01

    An experimental study was carried out to identify the various regimes of natural convective boiling and to determine the Critical Heat Flux (CHF) on a square surface. The basic knowledge on the boiling heat transfer and CHF on the square surface is necessary for various engineering problems, such as the design of compact heat exchangers, cooling of CPU chips, and design of the external cooling mechanism for the reactor during the severe accidents in the nuclear power plants. The heater block made of copper with cartridge heaters in it is submerged in a water tank with windows for visualization. The heater surface has dimension of 70mm x 70mm and the maximum heat flux capacity is about 1.8MW/m 2 . The boiling heat transfer coefficient for the various flow regimes up to CHF has been measured for upward facing surface, vertical surface, and nearly horizontal downward facing surfaces. The temperatures of the heater block are measured by the thermocouples imbedded in the heater block. As the heat flux increases from 100kW/m 2 to 1.0MW/m 2 , the heat-transfer regime changes from the nucleate boiling to the CHF. Near 1.0MW/m 2 , the heat transfer regime suddenly changed from nucleate boiling to film boiling and it resulted in a rapid heat up of the heater block. The various boiling patterns on the vertical surface, upward facing surface, and downward facing surface are observed by a high speed video camera whose frame rate is 1000fps. An explosive vapor generation on the heated surface, whose size and frequency are characterized by the heat flux and inclination angle, is observed

  14. A research of vapour-film characteristics of inverted-annular flow film boiling by visual method

    International Nuclear Information System (INIS)

    Xu Jijun; Guo Zhichao; Yan An; Bi Haoran

    1988-01-01

    The vapour-film characteristics are an interesting topic in inverted-annular flow film boiling. A practical set of experimental rig has been designed and constructed for visual observation. Photographic method is adopted for obtaining number of photographs in the conditions of steady state. For references at hands, photographs under steady conditions of water flow film boiling have not been published yet. This paper discusses the typical vapour film characteristics and regards Elias' two-region model summarized from transient visual experiment as reasonable. In addition, under heated conditions, at least, three types of vapour-water interfaces have been observed. They are asymmetric sine waves, symmetic varicose waves, and roll waves offered by Jarlais from an adiabatic simulation. In diabatic conditions a transition of flow pattern to slug flow is usually caused by hydrodynamic instability and/or by thermodynamic instability. The effects of mass velocity, inlet subcooling, heat flux input, initial quality and pressure to vapour-film characteristics are described. An empirical correlation is fitted to 23 sets of tests of discussion

  15. Prediction of flow recirculation in a blanket assembly under worst-case natural-convection conditions

    International Nuclear Information System (INIS)

    Khan, E.U.; Rector, D.R.

    1982-01-01

    Reactor fuel and blanket assemblies within a Liquid Metal Fast Breeder Reactor (LMFBR) can be subjected to severe radial heat flux gradients. At low-flow conditions, with power-to-flow ratios of nearly the same magnitude as design conditions, buoyancy forces cause flow redistribution to the side of a bundle with the higher heat generation rate. Recirculation of fluid within a rod bundle can occur during a natural convection transient because of the combined effect of flow coastdown and buoyancy-induced redistribution. An important concern is whether recirculation leads to high coolant temperatures. For this reason, the COBRA-WC code was developed with the capability of modeling recirculating flows. Experiments have been conducted in a 2 x 6 rod bundle for flow and power transients to study recirculation in the mixed-convection (forced cooled) and natural-convection regimes. The data base developed was used to validate the recirculation module in the COBRA-WC code. COBRA-WC code calculations were made to predict flow and temperature distributions in a typical LMFBR blanket assembly for the worst-case, natural-circulation transient

  16. Theoretical prediction method of subcooled flow boiling CHF

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Young Min; Chang, Soon Heung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1999-12-31

    A theoretical critical heat flux (CHF ) model, based on lateral bubble coalescence on the heated wall, is proposed to predict the subcooled flow boiling CHF in a uniformly heated vertical tube. The model is based on the concept that a single layer of bubbles contacted to the heated wall prevents a bulk liquid from reaching the wall at near CHF condition. Comparisons between the model predictions and experimental data result in satisfactory agreement within less than 9.73% root-mean-square error by the appropriate choice of the critical void fraction in the bubbly layer. The present model shows comparable performance with the CHF look-up table of Groeneveld et al.. 28 refs., 11 figs., 1 tab. (Author)

  17. Theoretical prediction method of subcooled flow boiling CHF

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Young Min; Chang, Soon Heung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1998-12-31

    A theoretical critical heat flux (CHF ) model, based on lateral bubble coalescence on the heated wall, is proposed to predict the subcooled flow boiling CHF in a uniformly heated vertical tube. The model is based on the concept that a single layer of bubbles contacted to the heated wall prevents a bulk liquid from reaching the wall at near CHF condition. Comparisons between the model predictions and experimental data result in satisfactory agreement within less than 9.73% root-mean-square error by the appropriate choice of the critical void fraction in the bubbly layer. The present model shows comparable performance with the CHF look-up table of Groeneveld et al.. 28 refs., 11 figs., 1 tab. (Author)

  18. Prediction of bubble detachment diameter in flow boiling based on force analysis

    International Nuclear Information System (INIS)

    Chen Deqi; Pan Liangming; Ren Song

    2012-01-01

    Highlights: ► All the forces acting on the growing bubbles are taken into account in the model. ► The bubble contact diameter has significant effect on bubble detachment. ► Bubble growth force and surface tension are more significant in narrow channel. ► A good agreement between the predicted and the measured results is achieved. - Abstract: Bubble detachment diameter is one of the key parameters in the study of bubble dynamics and boiling heat transfer, and it is hard to be measured in a boiling system. In order to predict the bubble detachment diameter, a theoretical model is proposed based on forces analysis in this paper. All the forces acting on a bubble are taken into account to establish a model for different flow boiling configurations, including narrow and conventional channels, upward, downward and horizontal flows. A correlation of bubble contact circle diameter is adopted in this study, and it is found that the bubble contact circle diameter has significant effect on bubble detachment. A new correlation taking the bubble contact circle diameter into account for the evaluation of bubble growth force is proposed in this study, and it is found that the bubble growth force and surface tension force are more significant in narrow channel when comparing with that in conventional channel. A visual experiment was carried out in order to verify present model; and the experimental data from published literature are used also. A good agreement between predicted and measured results is achieved.

  19. Spatial distribution of nanoparticles in PWR nanofluid coolant subjected to local nucleate boiling

    Energy Technology Data Exchange (ETDEWEB)

    Mirghaffari, Reza; Jahanfarnia, Gholamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering

    2016-12-15

    Nanofluids have shown to be promising as an alternative for a PWR reactor coolant or as a safety system coolant to cover the core in the event of a loss of coolant accident. The nanoparticles distribution and neutronic parameters are intensively affected by the local boiling of nanofluid coolant. The main goal of this study was the physical-mathematical modeling of the nanoparticles distribution in the nucleate boiling of nanofluids within the viscous sublayer. Nanoparticles concentration, especially near the heat transfer surfaces, plays a significant role in the enhancement of thermal conductivity of nanofluids and prediction of CHF, Hide Out and Return phenomena. By solving the equation of convection-diffusion for the liquid phase near the heating surface and the bulk stream, the effect of heat flux on the distribution of nanoparticles was studied. The steady state mass conservation equations for liquids, vapors and nanoparticles were written for the flow boiling within the viscous sublayer adjacent the fuel cladding surface. The derived differential equations were discretized by the finite difference method and were solved numerically. It was found out that by increasing the surface heat flux, the concentration of nanoparticles increased.

  20. Automated high-speed video analysis of the bubble dynamics in subcooled flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Maurus, Reinhold; Ilchenko, Volodymyr; Sattelmayer, Thomas [Technische Univ. Muenchen, Lehrstuhl fuer Thermodynamik, Garching (Germany)

    2004-04-01

    Subcooled flow boiling is a commonly applied technique for achieving efficient heat transfer. In the study, an experimental investigation in the nucleate boiling regime was performed for water circulating in a closed loop at atmospheric pressure. The test-section consists of a rectangular channel with a one side heated copper strip and a very good optical access. For the optical observation of the bubble behaviour the high-speed cinematography is used. Automated image processing and analysis algorithms developed by the authors were applied for a wide range of mass flow rates and heat fluxes in order to extract characteristic length and time scales of the bubbly layer during the boiling process. Using this methodology, a huge number of bubble cycles could be analysed. The structure of the developed algorithms for the detection of the bubble diameter, the bubble lifetime, the lifetime after the detachment process and the waiting time between two bubble cycles is described. Subsequently, the results from using these automated procedures are presented. A remarkable novelty is the presentation of all results as distribution functions. This is of physical importance because the commonly applied spatial and temporal averaging leads to a loss of information and, moreover, to an unjustified deterministic view of the boiling process, which exhibits in reality a very wide spread of bubble sizes and characteristic times. The results show that the mass flux dominates the temporal bubble behaviour. An increase of the liquid mass flux reveals a strong decrease of the bubble life - and waiting time. In contrast, the variation of the heat flux has a much smaller impact. It is shown in addition that the investigation of the bubble history using automated algorithms delivers novel information with respect to the bubble lift-off probability. (Author)

  1. Automated high-speed video analysis of the bubble dynamics in subcooled flow boiling

    International Nuclear Information System (INIS)

    Maurus, Reinhold; Ilchenko, Volodymyr; Sattelmayer, Thomas

    2004-01-01

    Subcooled flow boiling is a commonly applied technique for achieving efficient heat transfer. In the study, an experimental investigation in the nucleate boiling regime was performed for water circulating in a closed loop at atmospheric pressure. The test-section consists of a rectangular channel with a one side heated copper strip and a very good optical access. For the optical observation of the bubble behaviour the high-speed cinematography is used. Automated image processing and analysis algorithms developed by the authors were applied for a wide range of mass flow rates and heat fluxes in order to extract characteristic length and time scales of the bubbly layer during the boiling process. Using this methodology, a huge number of bubble cycles could be analysed. The structure of the developed algorithms for the detection of the bubble diameter, the bubble lifetime, the lifetime after the detachment process and the waiting time between two bubble cycles is described. Subsequently, the results from using these automated procedures are presented. A remarkable novelty is the presentation of all results as distribution functions. This is of physical importance because the commonly applied spatial and temporal averaging leads to a loss of information and, moreover, to an unjustified deterministic view of the boiling process, which exhibits in reality a very wide spread of bubble sizes and characteristic times. The results show that the mass flux dominates the temporal bubble behaviour. An increase of the liquid mass flux reveals a strong decrease of the bubble life- and waiting time. In contrast, the variation of the heat flux has a much smaller impact. It is shown in addition that the investigation of the bubble history using automated algorithms delivers novel information with respect to the bubble lift-off probability

  2. RELAP5 analysis of subcooled boiling appearance and disappearance in downward flow

    International Nuclear Information System (INIS)

    Ristevski, R.; Parzer, I.; Spasojevic, D.

    1999-01-01

    The presented paper will mainly consider heat and mass transfer phenomenology in the subcooled boiling regime of downward liquid flow at low velocities. More specifically, it will focus on the effects of appearance and disappearance of two-phase flow at low liquid velocities, in the area where gravity force has significant influence. Two among a series of tests performed on a high-pressure circulation loop, installed in Vinca, will be analyzed. The experimental findings and theoretical consideration of these processes and phenomena will be compared with RELAP5/MOD3.2.2 predictions.(author)

  3. Experimental Study of Flow Boiling Heat Transfer in a Horizontal Microfin Tube

    OpenAIRE

    Yu, Jian; Koyama, Shigeru; Momoki, Satoru

    1995-01-01

    An experimental study on flow boiling heat transfer in a horizontal microfin tube is conducted with pure refrigerants HFC134a, HCFC123 and HCFC22 using a water-heated double-tube type test section. The test microfin tube is a copper tube having the following dimensions: 8.37mm mean inside diameter, 0.168mm fin height, 60fin number and 18 degree of helix angle. The local heat transfer coefficients for both counter and parallel flows are measured in a range of heat flux of 1 to 93W/m^2, mass ve...

  4. Experimental study on transient boiling heat transfer

    International Nuclear Information System (INIS)

    Visentini, R.

    2012-01-01

    well. A flexible power supply that can generate a free-shape signal, allows to get to a wall-temperature increase rate up to 2500 K/s but also to obtain lower rates, which permits to study weaker transients and steady state conditions. The thermal measurements are realised by means of an infra-red camera and a high-speed camera is employed in order to see the boiling phenomena at the same time. From the voltage and current measurements the heat flux that is passed to the fluid is known. It is possible to underline some of the main results of this work. We found that, even when the boiling onset occurs soon because of the high power, transient conduction is always coupled with transient convection. The boiling onset occurs when the wall superheat is between 10 K et 30 K. This value corresponds to the activation of the smallest nucleation sites at the wall. The literature correlations well fit the nucleate boiling data in steady-state conditions. When the wall-temperature increase rate leads to transient boiling, the heat flux is higher than in steady state. This is consistent with what was found in previous studies. The nucleate boiling phase may last only a few milliseconds when the power is really high and the wall temperature increases really rapidly (500-2000 K/s). The experiments in transient boiling also point out that the heat flux is larger than in steady state conditions for the other regimes: Critical heat flux and also film boiling. The experimental set-up allows to investigate a large range of parameters (wall-temperature increase rate, flow rate, fluid temperature) by means of accurate temperature measurements and visualisations. Some modeling of the heat transfer are also proposed. (author)

  5. A photographic study on flow boiling of R-134a in a vertical channel

    International Nuclear Information System (INIS)

    Bang, In Cheol; Baek, Won Pil; Chang, Soon Heung

    2002-01-01

    The behavior of near-wall bubbles in subcooled flow boiling has been investigated photographically for R134a flow in vertical, one-side heated and rectangular channels at mass fluxes of 0, 190, 1000 and 2000 kg/m 2 s and inlet subcooling condition of 8 .deg. C under 7 bar(Tsat 27 .deg. C). Digital photographic techniques and high-speed camera are used for the visualization, which have significantly advanced for recent decades. Primary attention is given to the bubble coalescence phenomena and the structure of the near-wall bubble layer. At subcooled and low-quality conditions, discrete attached bubbles, sliding bubbles, small coalesced bubbles and large coalesced bubbles or vapor clots are observed on the heated surface as the heat flux is increased from a low value. Particularly in beginning of vapor formation, vapor remnants below discrete bubble on the heating surface are clearly observed. Nucleation site density increases with the increases in heat flux and channel-averaged enthalpy, while discrete bubbles coalesce and form large bubbles, resulting in large vapor clots. Waves formed on the surface of the vapor clots are closely related to Helmholtz instability. At CHF occurrence it is also observed that wall bubble layer beneath large vapor clots is removed and large film boiling occurs. Through the present visual test, it is observed that wall bubble layer begins to develop with the onset of nucleate boiling(ONB) and to extinguish with the occurrence of the CHF. It could be considered that this layer made an important role of CHF mechanism macroscopically. However, there may be another structure beneath wall bubbles which supplies specific information on CHF from viewpoint of microstructure based upon the observation of the liquid sublayer beneath coalesced bubbles. Through this microscopic visualization, it may be suggested that the following flow structures characterize the flow boiling phenomena : (a) vapor remnants as a continuous source of bubbles, (b

  6. Large-Scale Flows and Magnetic Fields Produced by Rotating Convection in a Quasi-Geostrophic Model of Planetary Cores

    Science.gov (United States)

    Guervilly, C.; Cardin, P.

    2017-12-01

    Convection is the main heat transport process in the liquid cores of planets. The convective flows are thought to be turbulent and constrained by rotation (corresponding to high Reynolds numbers Re and low Rossby numbers Ro). Under these conditions, and in the absence of magnetic fields, the convective flows can produce coherent Reynolds stresses that drive persistent large-scale zonal flows. The formation of large-scale flows has crucial implications for the thermal evolution of planets and the generation of large-scale magnetic fields. In this work, we explore this problem with numerical simulations using a quasi-geostrophic approximation to model convective and zonal flows at Re 104 and Ro 10-4 for Prandtl numbers relevant for liquid metals (Pr 0.1). The formation of intense multiple zonal jets strongly affects the convective heat transport, leading to the formation of a mean temperature staircase. We also study the generation of magnetic fields by the quasi-geostrophic flows at low magnetic Prandtl numbers.

  7. An analytical model for annular flow boiling heat transfer in microchannel heat sinks

    International Nuclear Information System (INIS)

    Megahed, A.; Hassan, I.

    2009-01-01

    An analytical model has been developed to predict flow boiling heat transfer coefficient in microchannel heat sinks. The new analytical model is proposed to predict the two-phase heat transfer coefficient during annular flow regime based on the separated model. Opposing to the majority of annular flow heat transfer models, the model is based on fundamental conservation principles. The model considers the characteristics of microchannel heat sink during annular flow and eliminates using any empirical closure relations. Comparison with limited experimental data was found to validate the usefulness of this analytical model. The model predicts the experimental data with a mean absolute error 8%. (author)

  8. Mixed convection flow past a horizontal plate

    Directory of Open Access Journals (Sweden)

    Savić Lj.

    2005-01-01

    Full Text Available The mixed convection flow past a horizontal plate being aligned through a small angle of attack to a uniform free stream will be considered in the limit of large Reynolds number and small Richardson number. Even a small angle of inclination of the wake is sufficient for the buoyancy force to accelerate the flow in the wake which causes a velocity overshoot in the wake. Moreover a hydrostatic pressure difference across the wake induces a correction to the potential flow which influences the inclination of the wake. Thus the wake and the correction of the potential flow have to be determined simultaneously. However, it turns out that solutions exist only if the angle of attack is sufficiently large. Solutions are computed numerically and the influence of the buoyancy on the lift coefficient is determined.

  9. Research progresses and future directions on pool boiling heat transfer

    Directory of Open Access Journals (Sweden)

    M. Kumar

    2015-12-01

    Full Text Available This paper reviews the previous work carried on pool boiling heat transfer during heating of various liquids and commodities categorized as refrigerants and dielectric fluids, pure liquids, nanofluids, hydrocarbons and additive mixtures, as well as natural and synthetic colloidal solutions. Nucleate pool boiling is an efficient and effective method of boiling because high heat fluxes are possible with moderate temperature differences. It is characterized by the growth of bubbles on a heated surface. It occurs during boiling of liquids for excess temperature ranging from 5 to 30 °C in various processes related to high vaporization of liquid for specific purposes like sugarcane juice heating for jaggery making, milk heating for khoa making, steam generation, cooling of electronic equipments, refrigeration and etcetera. In this review paper, pool boiling method during heating of liquids for specific purpose is depicted. It is inferred that enhancement in pool boiling heat transfer is a challenging and complex task. Also, recent research and use of various correlations for natural convection pool boiling is reviewed.

  10. A new correlation for nucleate pool boiling of aqueous mixtures

    International Nuclear Information System (INIS)

    Thome, J.R.; Shakir, S.

    1987-01-01

    A new mixture boiling correlation was developed for nucleate pool boiling of aqueous mixtures on plain, smooth tubes. The semi-empirical correlation models the rise in the local bubble point temperature in a mixture caused by the preferential evaporation of the more volatile component during bubble growth. This rise varies from zero at low heat fluxes (where only single-phase natural convection is present) up to nearly the entire boiling range at the peak heat flux (where latent heat transport is dominant). The boiling range, which is the temperature difference between the dew point and bubble point of a mixture, is used to characterize phase equilibrium effects. An exponential term models the rise in the local bubble point temperature as a function of heat flux. The correlation was compared against binary mixture boiling data for ethanol-water, methanol-water, n-propanol-water, and acetone-water. The majority of the data was predicted to within 20%. Further experimental research is currently underway to obtain multicomponent boiling data for aqueous mixtures with up to five components and for wider boiling ranges

  11. The ionospheric signature of transient dayside reconnection and the associated pulsed convection return flow

    Directory of Open Access Journals (Sweden)

    S. E. Milan

    Full Text Available Three SuperDARN coherent HF radars are employed to investigate the excitation of convection in the dayside high-latitude ionosphere in response to transient reconnection occurring in the cusp region. This study demonstrates the existence of transient antisunward-propagating backscatter features at the expected location of the ionospheric footprint of the cusp region, which have a repetition rate near 10 min. These are interpreted as the ionospheric signature of flux transfer events. Moreover, transient sunward-propagating regions of backscatter are observed in the convection return flow regions of both the pre- and post-noon sectors. These patches are observed to propagate towards the noon sector from at least as far around the auroral zone as 07 MLT in the pre-noon sector and 17 MLT in the post-noon sector, travelling with a velocity of approximately 1.5 to 2 km s-1. These return flow patches have a repetition rate similar to that of the transient features observed at local noon. While providing supporting evidence for the impulsive nature of convection flow, the observation of sunward-propagating features in the return flow region is not consistent with current conceptual models of the excitation of convection.

    Key words. Ionosphere (plasma convection · Magnetospheric physics (magnetopause · cusp · and boundary layers; magnetosphere-ionosphere interactions

  12. R134a Flow Boiling Analysis with Modified Thermodynamic Property File of MARS Code

    Energy Technology Data Exchange (ETDEWEB)

    Son, Gyumin; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2016-10-15

    Previous study showed application of RELAP5 code to solar energy facility with molten salt (60% NaNO3 and 40% KNO3) as working fluid. Based on external experimental correlations, thermodynamic properties of molten salt were evaluated as a function of pressure and temperature and those equations were used to generate tpf. To validate external tpf, experimental values were compared with RELAP5 analysis. In nuclear field, utilization of other fluid is also important since many thermal-hydraulic experiments used various fluids such as FC-72, R123, and R134a. Theses refrigerants have been used to simulate the high pressure environment of nuclear power plants due to their low boiling point, and density ratio between vapor and liquid. Thus, this study aims for tpf generation of R134a and verification by analyzing real case. R134a is selected as a fluid to be implemented and analyzed because it is currently used in refrigerator and frequently used in flow boiling experiment related with heat transfer coefficient and CHF measurement. R134a property file were generated with fitted equation using temperature and pressure as variables, originated from external data source. For validation, flow boiling experiment case were made into simplified input. Analysis with tpfr134a showed that application of Gnielinksi correlation could enhance single phase flow accuracy. Large error of HTC from two phase analysis requires parameter study. Future work aims for more specified experimental case comparison and correlation enhancement for two phase analysis.

  13. Experimental Investigation of Pressure Drop and Pressure Distribution Along a Heated Channel in Subcooled Flow Boiling

    International Nuclear Information System (INIS)

    Aharon, Y.; Hochbaum, I.; Shai, I.

    2002-01-01

    The state of knowledge relating to pressure drop in subcooled boiling region is very unsatisfactory. That pressure drop is an important factor in considering the design of nuclear reactors because of the possibility of flow excursion during a two phase flow in the channels. In operational systems with multiple flow channels, an increase in pressure drop in one flow channel, can cause the flow to be diverted to other channels. A burnout can occur in the unstable channel

  14. Convective heat flow probe

    Science.gov (United States)

    Dunn, James C.; Hardee, Harry C.; Striker, Richard P.

    1985-01-01

    A convective heat flow probe device is provided which measures heat flow and fluid flow magnitude in the formation surrounding a borehole. The probe comprises an elongate housing adapted to be lowered down into the borehole; a plurality of heaters extending along the probe for heating the formation surrounding the borehole; a plurality of temperature sensors arranged around the periphery of the probe for measuring the temperature of the surrounding formation after heating thereof by the heater elements. The temperature sensors and heater elements are mounted in a plurality of separate heater pads which are supported by the housing and which are adapted to be radially expanded into firm engagement with the walls of the borehole. The heat supplied by the heater elements and the temperatures measured by the temperature sensors are monitored and used in providing the desired measurements. The outer peripheral surfaces of the heater pads are configured as segments of a cylinder and form a full cylinder when taken together. A plurality of temperature sensors are located on each pad so as to extend along the length and across the width thereof, with a heating element being located in each pad beneath the temperature sensors. An expansion mechanism driven by a clamping motor provides expansion and retraction of the heater pads and expandable packer-type seals are provided along the probe above and below the heater pads.

  15. Analysis of flow boiling heat transfer in narrow annular gaps applying the design of experiments method

    Directory of Open Access Journals (Sweden)

    Gunar Boye

    2015-06-01

    Full Text Available The axial heat transfer coefficient during flow boiling of n-hexane was measured using infrared thermography to determine the axial wall temperature in three geometrically similar annular gaps with different widths (s = 1.5 mm, s = 1 mm, s = 0.5 mm. During the design and evaluation process, the methods of statistical experimental design were applied. The following factors/parameters were varied: the heat flux q · = 30 − 190 kW / m 2 , the mass flux m · = 30 − 700 kg / m 2 s , the vapor quality x · = 0 . 2 − 0 . 7 , and the subcooled inlet temperature T U = 20 − 60 K . The test sections with gap widths of s = 1.5 mm and s = 1 mm had very similar heat transfer characteristics. The heat transfer coefficient increases significantly in the range of subcooled boiling, and after reaching a maximum at the transition to the saturated flow boiling, it drops almost monotonically with increasing vapor quality. With a gap width of 0.5 mm, however, the heat transfer coefficient in the range of saturated flow boiling first has a downward trend and then increases at higher vapor qualities. For each test section, two correlations between the heat transfer coefficient and the operating parameters have been created. The comparison also shows a clear trend of an increasing heat transfer coefficient with increasing heat flux for test sections s = 1.5 mm and s = 1.0 mm, but with increasing vapor quality, this trend is reversed for test section 0.5 mm.

  16. The verification of subcooled boiling models in CFX-4.2 at low pressure in annulus channel flow

    International Nuclear Information System (INIS)

    Kim, Seong-Jin; Kim, Moon-Oh; Park, Goon-Cherl

    2003-01-01

    Heat transfer in subcooled boiling is an important issue to increase the effectiveness of design and safety in operation of engineering system such as nuclear plant. The subcooled boiling, which may occur in the hot channel of reactor in normal state and in decreased pressure condition in transient state, can cause multi-dimensional and complicated respects. The variation of local heat transfer phenomena is created by changing of liquid and vapor velocity, by simultaneous bubble break-ups and coalescences, and by corresponding to bubble evaporation and condensation, and that can affect the stability of the system. The established researches have carried out not a point of local distributions of two-phase variables, but a point of systematical distributions, mostly. Although the subcooled boiling models have been used to numerical analysis using CFX-4.2, there are few verification of subcooled boiling models. This paper demonstrated locally and systematically the validation of subcooled boiling model in numerical calculations using CFX-4.2 especially, in annulus channel flow condition in subcooled boiling at low pressure with respect to subcooled boiling models such as mean bubble diameter model, bubble departure diameter model or wall heat flux model and models related with phase interface. (author)

  17. The effect of diameter on vertical and horizontal flow boiling crisis in a tube cooled by Freon-12

    International Nuclear Information System (INIS)

    Merilo, M.; Ahmad, S.Y.

    1979-03-01

    The influence of test section orientation and diameter on flow boiling crisis occurring in tubes has been studied experimentally using Freon-12 as a coolant. At low mass flux the critical heat flux (CHF) was lower in horizontal flow than in vertical. As either the liquid or vapour velocity, or both, were increased the vertical and horizontal CHF results converged. Above a mass flux of 4 Mg.m -2 .s -1 the results were essentially identical. The effect of tube diameter on boiling crisis in general depends crucially on the parameters which are maintained constant when the comparison is made. (author)

  18. Combined free and forced convection flow in a rotating channel with ...

    African Journals Online (AJOL)

    user

    free and forced convection flow of a viscous incompressible electrically conducting fluid in a .... The boundary conditions (10) and (11), in dimensionless form, become ...... On hydromagnetic Flow and heat transfer in a rotating fluid past an infinite porous ... Electrically Conducting Fluid in Non-Rotating and Rotating Media”.

  19. Vapor bubble behavior in subcooled flow boiling in annuli heated by water

    International Nuclear Information System (INIS)

    Licheng Sun; Zhongning Sun; Changqi Yan

    2005-01-01

    Full text of publication follows: This paper describes experimental and theoretical work conducted on vapor bubble behavior in subcooled flow boiling at atmospheric pressure. The test section is mainly consisted of two concentrically installed circular tubes, the outside tube is made of quartz and therefore all test courses can be visualized. Water is forced to flow through annuli with gap sizes of 3 mm and 5 mm, and is heated by high temperature water in the inner tube. The main objective is to visually study the bubble behavior of subcooled flow boiling water in the condition of surface heated by water. The results show that bubbles depart from wall directly or slide a certain distance before departure, this is same as that heated by electricity. There exists a bubble layer near the wall, most bubbles move and disappear in the layer after departure, the bubble sliding behavior is not very obvious in 5 mm annulus, however, we found that most bubbles in 3 mm annulus will slide a long distance before departure and their growth courses are different from usual experimental results. The bubbles are not always growing, but shrinking a little quickly after growing for some time, and then the course will repeat for some times till they depart from wall or disappeared, the collision and coalescence of bubbles is very common and makes the bubbles depart from wall more easily in 3 mm annulus. At last, the forces on bubbles growing and detaching in flow along the wall are analyzed to comprehend these phenomena more accurately. (authors)

  20. Bubble nucleation of R134A refrigerant in a pressurized flow boiling system

    Energy Technology Data Exchange (ETDEWEB)

    Murshed, S.M. Sohel; Vereen, Keon; Kumar, Ranganathan [University of Central Florida, Orlando, FL (United States). Dept. of Mechanical, Materials and Aerospace Engineering], e-mail: rnkumar@mail.ucf.edu

    2009-07-01

    The effect of heat flux and pressure on bubble nucleation of R134a refrigerant in a flow boiling system is experimentally studied. An experimental facility was built and an innovative concept of thermochromic liquid crystal (TLC) technique was introduced for the high resolution and accurate measurement of the overall heater surface temperature. The visualization and image recording process is performed by employing two synchronized high resolution and high speed cameras which simultaneously capture colored TLC images as well as bubble nucleation activities at high frame rates. Experiments were conducted at different high pressures ranging from 690 to 830 kPa and at different heat flux conditions in order to identify their influence on flow boiling performance specially bubbling event. Present results demonstrate that both the heat flux and pressure influence the bubble generation rate and size. For example, bubble generation frequency and size are found to increase with heat flux. An increase in pressure of 137 kPa (from 690 to 827 kPa) increased the bubble frequency and size about 32 Hz and 20 {mu}m, respectively. (author)

  1. Bubble nucleation of R134A refrigerant in a pressurized flow boiling system

    International Nuclear Information System (INIS)

    Murshed, S.M. Sohel; Vereen, Keon; Kumar, Ranganathan

    2009-01-01

    The effect of heat flux and pressure on bubble nucleation of R134a refrigerant in a flow boiling system is experimentally studied. An experimental facility was built and an innovative concept of thermochromic liquid crystal (TLC) technique was introduced for the high resolution and accurate measurement of the overall heater surface temperature. The visualization and image recording process is performed by employing two synchronized high resolution and high speed cameras which simultaneously capture colored TLC images as well as bubble nucleation activities at high frame rates. Experiments were conducted at different high pressures ranging from 690 to 830 kPa and at different heat flux conditions in order to identify their influence on flow boiling performance specially bubbling event. Present results demonstrate that both the heat flux and pressure influence the bubble generation rate and size. For example, bubble generation frequency and size are found to increase with heat flux. An increase in pressure of 137 kPa (from 690 to 827 kPa) increased the bubble frequency and size about 32 Hz and 20 μm, respectively. (author)

  2. Axisymmetric free convection boundary-layer flow past slender bodies

    NARCIS (Netherlands)

    Kuiken, H.K.

    1968-01-01

    Radial curvature effects on axisymmetric free convection boundary-layer flow are investigated for vertical cylinders and cones for some special non-uniform temperature differences between the surface and the ambient fluid. The solution is given as a power series expansion, the first term being equal

  3. A New Computational Tool for Simulation of 3-D Flow and Heat Transfer in Boiling Water Reactors

    International Nuclear Information System (INIS)

    Chen, Hudong

    2002-01-01

    This Phase I work has developed a novel hybrid Lattice Boltzmann Model for the simulation of nonideal fluid thermal dynamics and demonstrated that this model can be used to simulate fundamental two-phase flow processes including boiling initiation, bubble formation and coalescency, and flow-regime formation

  4. Heat Transfer by Thermo-capillary Convection -Sounding Rocket COMPERE Experiment SOURCE

    Science.gov (United States)

    Dreyer, Michael; Fuhrmann, Eckart

    The sounding rocket COMPERE experiment SOURCE was successfully flown on MASER 11, launched in Kiruna (ESRANGE), May 15th, 2008. SOURCE has been intended to partly ful-fill the scientific objectives of the European Space Agency (ESA) Microgravity Applications Program (MAP) project AO-2004-111 (Convective boiling and condensation). Three parties of principle investigators have been involved to design the experiment set-up: ZARM for thermo-capillary flows, IMFT (Toulouse, France) for boiling studies, EADS Astrium (Bremen, Ger-many) for depressurization. The topic of this paper is to study the effect of wall heat flux on the contact line of the free liquid surface and to obtain a correlation for a convective heat trans-fer coefficient. The experiment has been conducted along a predefined time line. A preheating sequence at ground was the first operation to achieve a well defined temperature evolution within the test cell and its environment inside the rocket. Nearly one minute after launch, the pressurized test cell was filled with the test liquid HFE-7000 until a certain fill level was reached. Then the free surface could be observed for 120 s without distortion. Afterwards, the first depressurization was started to induce subcooled boiling, the second one to start saturated boiling. The data from the flight consists of video images and temperature measurements in the liquid, the solid, and the gaseous phase. Data analysis provides the surface shape versus time and the corresponding apparent contact angle. Computational analysis provides information for the determination of the heat transfer coefficient in a compensated gravity environment where a flow is caused by the temperature difference between the hot wall and the cold liquid. The paper will deliver correlations for the effective contact angle and the heat transfer coefficient as a function of the relevant dimensionsless parameters as well as physical explanations for the observed behavior. The data will be used

  5. Boiling Heat Transfer to Halogenated Hydrocarbon Refrigerants

    Science.gov (United States)

    Yoshida, Suguru; Fujita, Yasunobu

    The current state of knowledge on heat transfer to boiling refrigerants (halogenated hydrocarbons) in a pool and flowing inside a horizontal tube is reviewed with an emphasis on information relevant to the design of refrigerant evaporators, and some recommendations are made for future research. The review covers two-phase flow pattern, heat transfer characteristics, correlation of heat transfer coefficient, influence of oil, heat transfer augmentation, boiling from tube-bundle, influence of return bend, burnout heat flux, film boiling, dryout and post-dryout heat transfer.

  6. Combined convective heat transfer of liquid sodium flowing across tube banks

    International Nuclear Information System (INIS)

    Ma, Ying; Sugiyama, Ken-ichiro; Ishiguro, Ryoji

    1989-01-01

    In order to clarify the heat transfer characteristics of combined convection of liquid sodium, a numerical analysis is performed for liquid sodium which flows through a single horizontal row of tubes in the direction of gravity. The correlation of heat transfer characteristics between liquid sodium and ordinary fluids is also discussed. The heat transfer characteristics at large Reynolds numbers are improved when the Richardson number is increased, and the improvement rate is enlarged with increase in p/d value, since convection effect is relatively large. However heat transfer coefficients do not differ from those of forced convection at small Reynolds numbers even when the Richardson number reaches a high value because of conduction effect. A good consistence of heat transfer characteristics of combined convection between liquid sodium and air is obtained at the same Peclet number and Richardson number. This means that the fundamental heat transfer characteristics of combined convection of liquid sodium can be investigated with ordinary fluids. (author)

  7. Modelling of void formation in the subcooled boiling regime in the ATHLET code to simulate flow instability for research reactors

    International Nuclear Information System (INIS)

    Hainoun, A.

    1996-01-01

    The ATHLET thermohydraulic code was developed at the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Society for Plant and Reactor Safety) to analyse leaks and transients for power reactors. In order to extend the code's range of application to the safety analysis of research reactors, a model was implemented permitting a description of the thermodynamic non-equilibrium effects in the subcooled boiling regime. The aim of the extension is, on one hand, to cover the thermohydraulic instability which is particularly characteristic of research reactors owing to their high power densities and low system pressures and, on the other hand, to provide a consideration of the influence of the steam formed in this boiling regime on the neutron balance. The model developed takes into consideration the competing evaporation and condensation effects in the subcooled boiling regime. It describes the bubble production rate at the superheated heating surfaces as well as the subsequent condensation of the bubbles in the subcooled core flow. The installed model is validated by the recalculation of two extensive series of experiments. In the first series the McMaster experiments on axial void distribution in the subcooled boiling regime are recalculated. The recalculation shows that the extended programme is capable of calculating the axial void distribution in the subcooled boiling regime with good agreement with the data. The second series deals with KFA experiments on thermohydraulic instability (flow excursion) in the subcooled boiling regime, comprising a broad parameter range of heat flow density, inlet temperature and channel width. Recalculation of this experimental series shows that the programme extension ensures simulation of thermohydraulic instability. (orig.)

  8. Measurement of multi-dimensional flow structure for flow boiling in a tube

    International Nuclear Information System (INIS)

    Adachi, Yu; Ito, Daisuke; Saito, Yasushi

    2014-01-01

    With an aim of the measurement of multi-dimensional flow structure of in-tube boiling two-phase flow, the authors built their own wire mesh measurement system based on electrical conductivity measurement, and examined the relationship between the electrical conductivity obtained by the wire mesh sensor and the void fraction. In addition, the authors measured the void fraction using neutron radiography, and compared the result with the measured value using the wire mesh sensor. From the comparison with neutron radiography, it was found that the new method underestimated the void fraction in the flow in the vicinity of the void fraction of 0.2-0.5, similarly to the conventional result. In addition, since the wire mesh sensor cannot measure dispersed droplets, it tends to overestimate the void fraction in the high void fraction region, such as churn flow accompanied by droplet generation. In the electrical conductivity wire-mesh sensor method, it is necessary to correctly take into account the effect of liquid film or droplets. The authors also built a measurement system based on the capacitance wire mesh sensor method using the difference in dielectric constant, performed the confirmation of transmission and reception signals using deionized water as a medium, and showed the validity of the system. As for the dispersed droplets, the capacitance method has a potential to be able to measure them. (A.O.)

  9. Research progress on microgravity boiling heat transfer

    International Nuclear Information System (INIS)

    Xiao Zejun; Chen Bingde

    2003-01-01

    Microgravity boiling heat transfer is one of the most basic research topics in aerospace technology, which is important for both scientific research and engineering application. Research progress on microgravity boiling heat transfer is presented, including terrestrial simulation technique, terrestrial simulation experiment, microgravity experiment, and flow boiling heat transfer

  10. Performance Evaluation of the International Space Station Flow Boiling and Condensation Experiment (FBCE) Test Facility

    Science.gov (United States)

    Hasan, Mohammad; Balasubramaniam, R.; Nahra, Henry; Mackey, Jeff; Hall, Nancy; Frankenfield, Bruce; Harpster, George; May, Rochelle; Mudawar, Issam; Kharangate, Chirag R.; hide

    2016-01-01

    A ground-based experimental facility to perform flow boiling and condensation experiments is built in support of the development of the long duration Flow Boiling and Condensation Experiment (FBCE) destined for operation on board of the International Space Station (ISS) Fluid Integrated Rack (FIR). We performed tests with the condensation test module oriented horizontally and vertically. Using FC-72 as the test fluid and water as the cooling fluid, we evaluated the operational characteristics of the condensation module and generated ground based data encompassing the range of parameters of interest to the condensation experiment to be performed on the ISS. During this testing, we also evaluated the pressure drop profile across different components of the fluid subsystem, heater performance, on-orbit degassing subsystem, and the heat loss from different components. In this presentation, we discuss representative results of performance testing of the FBCE flow loop. These results will be used in the refinement of the flight system design and build-up of the FBCE which is scheduled for flight in 2019.

  11. Mixed convection flow and heat transfer in a vertical wavy channel ...

    African Journals Online (AJOL)

    Mixed convection flow and heat transfer in a vertical wavy channel filled with porous and fluid layers is studied analytically. The flow in the porous medium is modeled using Darcy-Brinkman equation. The coupled non-linear partial differential equations describing the conservation of mass, momentum and energy are solved ...

  12. Basic study on an energy conversion system using boiling two-phase flows of temperature-sensitive magnetic fluid. Theoretical analysis based on thermal nonequilibrium model and flow visualization using ultrasonic echo

    International Nuclear Information System (INIS)

    Ishimoto, Jun; Kamiyama, Shinichi; Okubo, Masaaki.

    1995-01-01

    Effects of magnetic field on the characteristics of boiling two-phase pipe flow of temperature-sensitive magnetic fluid are clarified in detail both theoretically and experimentally. Firstly, governing equations of two-phase magnetic fluid flow based on the thermal nonequilibrium two-fluid model are presented and numerically solved considering evaporation and condensation between gas- and liquid-phases. Next, behaviour of vapor bubbles is visualized with ultrasonic echo in the region of nonuniform magnetic field. This is recorded and processed with an image processor. As a result, the distributions of void fraction in the two-phase flow are obtained. Furthermore, detailed characteristics of the two-phase magnetic fluid flow are investigated using a small test loop of the new energy conversion system. From the numerical and experimental results, it is known that the precise control of the boiling two-phase flow and bubble generation is possible by using the nonuniform magnetic field effectively. These fundamental studies on the characteristics of two-phase magnetic fluid flow will contribute to the development of the new energy conversion system using a gas-liquid boiling two-phase flow of magnetic fluid. (author)

  13. Thermal and solutal stratification in mixed convection three-dimensional flow of an Oldroyd-B nanofluid

    Directory of Open Access Journals (Sweden)

    Tasawar Hayat

    Full Text Available This paper investigates the double stratified effects in mixed convection three-dimensional flow of an Oldroyd-B nanofluid. The flow is due to a bidirectional stretching surface. Mathematical analysis is carried out using the temperature and concentration stratification effects. Brownian motion, thermophoresis and chemical reaction effects are also considered. The governing nonlinear boundary layer equations are first converted into the dimensionless ordinary differential equations and then solved for the convergent series solutions of velocity, temperature and nanoparticles concentration. Convergence analysis of the obtained series solutions is also checked and verified. Effects of various emerging parameters are studied in details. Numerical values of local Nusselt and Sherwood numbers are tabulated and analyzed. It is noticed that the impact of mixed convection parameter on temperature and nanoparticles concentration is quite similar. Both temperature and nanoparticles concentration are reduced for larger mixed convection parameter. Keywords: Three-dimensional flow, Oldroyd-B fluid, Nanoparticles, Mixed convection, Thermal and solutal stratification, Chemically reactive species

  14. Convectively driven flow past an infinite moving vertical cylinder with ...

    Indian Academy of Sciences (India)

    2013-10-01

    Oct 1, 2013 ... tical cylinder with combined effects of heat and mass transfer is an ... presented a numerical study of free convective flow of a viscous ... models. The simultaneous effects of thermal and mass stratifications have application.

  15. Numerical simulation of double-diffusive mixed convective flow in rectangular enclosure with insulated moving lid

    Energy Technology Data Exchange (ETDEWEB)

    Teamah, M.A. [Faculty of Engineering, Alexandria University, Mech. Eng. Dept, Alexandria (Egypt); El-Maghlany, W.M. [Faculty of Engineering, Suez Canal University, Ismailia (Egypt)

    2010-09-15

    The present study is concerned with the mixed convection in a rectangular lid-driven cavity under the combined buoyancy effects of thermal and mass diffusion. Double-diffusive convective flow in a rectangular enclosure with moving upper surface is studied numerically. Both upper and lower surfaces are being insulated and impermeable. Constant different temperatures and concentration are imposed along the vertical walls of the enclosure, steady state laminar regime is considered. The transport equations for continuity, momentum, energy and spices transfer are solved. The numerical results are reported for the effect of Richardson number, Lewis number, and buoyancy ratio on the iso-contours of stream line, temperature, and concentration. In addition, the predicted results for both local and average Nusselt and Sherwood numbers are presented and discussed for various parametric conditions. This study was done for 0.1 <= Le <= 50 and Prandtl number Pr = 0.7. Through out the study the Grashof number and aspect ratio are kept constant at 10{sup 4} and 2 respectively and -10 <= N <= 10, while Richardson number has been varied from 0.01 to 10 to simulate forced convection dominated flow, mixed convection and natural convection dominated flow. (authors)

  16. Application of the Proper Orthogonal Decomposition to Turbulent Czochralski Convective Flows

    International Nuclear Information System (INIS)

    Rahal, S; Cerisier, P; Azuma, H

    2007-01-01

    The aim of this work is to study the general aspects of the convective flow instabilities in a simulated Czochralski system. We considered the influence of the buoyancy and crystal rotation. Velocity fields, obtained by an ultrasonic technique, the corresponding 2D Fourier spectra and a correlation function, have been used. Steady, quasi-periodic and turbulent flows, are successively recognized, as the Reynolds number was increased, for a fixed Rayleigh number. The orthogonal decomposition method was applied and the numbers of modes, involved in the dynamics of turbulent flows, calculated. As far as we know, this method has been used for the first time to study the Czochralski convective flows. This method provides also information on the most important modes and allows simple theoretical models to be established. The large rotation rates of the crystal were found to stabilize the flow, and conversely the temperature gradients destabilize the flow. Indeed, the increase of the rotation effects reduces the number of involved modes and oscillations, and conversely, as expected, the increase of the buoyancy effects induces more modes to be involved in the dynamics. Thus, the flow oscillations can be reduced either by increasing the crystal rotation rate to the adequate value, as shown in this study or by imposing a magnetic field

  17. On the definition of dominant force regimes for flow boiling heat transfer by using single mini-tubes

    Science.gov (United States)

    Baba, Soumei; Sawada, Kenichiro; Kubota, Chisato; Kawanami, Osamu; Asano, Hitoshi; Inoue, Koichi; Ohta, Haruhiko

    Recent increase in the size of space platforms requires the management of larger amount of waste heat under high heat flux conditions and the transportation of it along a long distance to the radiator. Flow boiling applied to the thermal management system in space attracts much attention as promising means to realize high-performance heat transfer and transport because of large latent heat of vaporization. In microgravity two-phase flow phenomena are quite different from those under 1-g condition because buoyancy effects are significantly reduced and surface tension becomes dominant. By the similar reason, flow boiling characteristics in mini channels are not the same as those in channels of normal sizes. In the present stage, however, the boundary between the regimes of body force dominated and of surface tension dominated is not clear. The design of space thermal devices, operated under the conditions where no effect of gravity is expected, will improve the reliability of their ground tests, provided that the boundaries of dominant force regimes are clarified quantitatively in advance. In flow boiling in mini channels or in parallel channels, back flow could be occurred because of rapid growth of bubbles in a confined space, resulting flow rate fluctuation. Flow boiling heat transfer characteristics in mini channels can be changed considerably by the existence of inlet flow rate fluctuation. It is important to pay attention to experimental accuracy and to use a single circular mini-tube to compare heat transfer characteristics with those of normal size tubes. In the present paper, effects of tube orientations, i.e. vertical upward flow, vertical downward flow and horizontal flow, on flow boiling heat transfer characteristics is investigated for FC72 flowing in single mini-tubes with inner diameters of 0.13 and 0.51 mm to establish a reliable dominant force regime map. If the regime map is described by using dimensionless groups of Bond, Weber and Froude numbers

  18. Implementation of a phenomenological DNB prediction model based on macroscale boiling flow processes in PWR fuel bundles

    International Nuclear Information System (INIS)

    Mohitpour, Maryam; Jahanfarnia, Gholamreza; Shams, Mehrzad

    2014-01-01

    Highlights: • A numerical framework was developed to mechanistically predict DNB in PWR bundles. • The DNB evaluation module was incorporated into the two-phase flow solver module. • Three-dimensional two-fluid model was the basis of two-phase flow solver module. • Liquid sublayer dryout model was adapted as CHF-triggering mechanism in DNB module. • Ability of DNB modeling approach was studied based on PSBT DNB tests in rod bundle. - Abstract: In this study, a numerical framework, comprising of a two-phase flow subchannel solver module and a Departure from Nucleate Boiling (DNB) evaluation module, was developed to mechanistically predict DNB in rod bundles of Pressurized Water Reactor (PWR). In this regard, the liquid sublayer dryout model was adapted as the Critical Heat Flux (CHF) triggering mechanism to reduce the dependency of the model on empirical correlations in the DNB evaluation module. To predict local flow boiling processes, a three-dimensional two-fluid formalism coupled with heat conduction was selected as the basic tool for the development of the two-phase flow subchannel analysis solver. Evaluation of the DNB modeling approach was performed against OECD/NRC NUPEC PWR Bundle tests (PSBT Benchmark) which supplied an extensive database for the development of truly mechanistic and consistent models for boiling transition and CHF. The results of the analyses demonstrated the need for additional assessment of the subcooled boiling model and the bulk condensation model implemented in the two-phase flow solver module. The proposed model slightly under-predicts the DNB power in comparison with the ones obtained from steady-state benchmark measurements. However, this prediction is acceptable compared with other codes. Another point about the DNB prediction model is that it has a conservative behavior. Examination of the axial and radial position of the first detected DNB using code-to-code comparisons on the basis of PSBT data indicated that the our

  19. Experimental Study on Flow Boiling of Carbon Dioxide in a Horizontal Microfin Tube

    Science.gov (United States)

    Kuwahara, Ken; Ikeda, Soshi; Koyama, Shigeru

    This paper deals with the experimental study on flow boiling heat transfer of carbon dioxide in a micro-fin tube. The geometrical parameters of micro-fin tube used in this study are 6.07 mm in outer diameter, 5.24 mm in average inner diameter, 0.256 mm in fin height, 20.4 in helix angle, 52 in number of grooves and 2.35 in area expansion ratio. Flow patterns and heat transfer coefficients were measured at 3-5 MPa in pressure, 300-540 kg/(m2s) in mass velocity and -5 to 15 °C in CO2 temperature. Flow patterns of wavy flow, slug flow and annular flow were observed. The measured heat transfer coefficients of micro-fin tube were 10-40 kW/(m2K). Heat transfer coefficients were strongly influenced by pressure.

  20. Two dimensional heat transfer problem in flow boiling in a rectangular minichannel

    Directory of Open Access Journals (Sweden)

    Hożejowska Sylwia

    2015-01-01

    Full Text Available The paper presents mathematical modelling of flow boiling heat transfer in a rectangular minichannel asymmetrically heated by a thin and one-sided enhanced foil. Both surfaces are available for observations due to the openings covered with glass sheets. Thus, changes in the colour of the plain foil surface can be registered and then processed. Plain side of the heating foil is covered with a base coat and liquid crystal paint. Observation of the opposite, enhanced surface of the minichannel allows for identification of the gas-liquid two-phase flow patterns and vapour quality. A two-dimensional mathematical model of heat transfer in three subsequent layers (sheet glass, heating foil, liquid was proposed. Heat transfer in all these layers was described with the respective equations: Laplace equation, Poisson equation and energy equation, subject to boundary conditions corresponding to the observed physical process. The solutions (temperature distributions in all three layers were obtained by Trefftz method. Additionally, the temperature of the boiling liquid was obtained by homotopy perturbation method (HPM combined with Trefftz method. The heat transfer coefficient, derived from Robin boundary condition, was estimated in both approaches. In comparison, the results by both methods show very good agreement especially when restricted to the thermal sublayer.

  1. Doubly stratified mixed convection flow of Maxwell nanofluid with heat generation/absorption

    Energy Technology Data Exchange (ETDEWEB)

    Abbasi, F.M., E-mail: abbasisarkar@gmail.com [Department of Mathematics, Comsats Institute of Information Technology, Islamabad 44000 (Pakistan); Shehzad, S.A. [Department of Mathematics, Comsats Institute of Information Technology, Sahiwal 57000 (Pakistan); Hayat, T. [Department of Mathematics, Quaid-i-Azam University, 45320, Islamabad 44000 (Pakistan); NAAM Research Group, Department of Mathematics, Faculty of Science, King Abdulaziz University, Jeddah 21589 (Saudi Arabia); Ahmad, B. [NAAM Research Group, Department of Mathematics, Faculty of Science, King Abdulaziz University, Jeddah 21589 (Saudi Arabia)

    2016-04-15

    Magnetohydrodynamic (MHD) doubly stratified flow of Maxwell nanofluid in presence of mixed convection is analyzed in this article. Effects of thermophoresis, Brownian motion and heat generation/absorption are present. The flow is induced due to linear stretching of sheet. Mathematical formulation is made under boundary layer approach. Expressions of velocity, temperature and nanoparticles concentration are developed. The obtained results are plotted and discussed to examine the variations in temperature and nanoparticles concentration due to different physical parameters. Numerical computations are made to obtain the values of local Nusselt and Sherwood numbers. Impact of sundry parameters on the flow quantities is analyzed graphically. - Highlights: • Double stratified flow of Maxwell nanofluid with mixed convection is modeled. • Thermophoresis and Brownian motion effects are encountered. • Computations are made to obtain the solution expressions. • Numerical values of local Nusselt and Sherwood numbers are computed and examined.

  2. Experimental analysis of R134a flow boiling inside a 5 PPI copper foam

    Science.gov (United States)

    Diani, A.; Mancin, S.; Rossetto, L.

    2014-04-01

    Heat dissipation is one of the most important issues for the reliability of electronic equipment. Boiling can be a very efficient heat transfer mechanism when used to face with the electronic technology needs of efficient and compact heat sinks. Recently, cellular structured materials both stochastic and periodic, particularly open cell metal foams, have been proposed as possible enhanced surfaces to lower the junction temperatures at high heat fluxes. Up today, most of the research on metal foams only regards single phase flow, whereas the two phase flow is still almost unexplored. This paper presents an experimental study on the heat transfer of R134a during flow boiling inside a 5 PPI (Pores Per linear Inch) copper foam, which is 5 mm high, 10 mm wide and 200 mm long, and it is brazed on a 10 mm thick copper plate. The experimental measurements were carried out by imposing three different heat fluxes (50, 75, and 100 kW m-2) and by varying the refrigerant mass velocity between 50 and 200 kg m-2 s-1 and the vapour quality from 0.2 to 0.90, at constant saturation temperature (30°C). The effects of the refrigerant mass flow rate, heat flux and vapour quality on the heat transfer coefficient, dry out phenomenon, and pressure drop are studied.

  3. CHF Enhancement in Flow Boiling using Al2O3 Nano-Fluid and Al2O3 Nano-Particle Deposited Tube

    International Nuclear Information System (INIS)

    Kim, Tae Il; Chun, T. H.; Chang, S. H.

    2010-01-01

    Nano-fluids are considered to have strong ability to enhance CHF. Most CHF experiments using nano-fluids were conducted in pool boiling conditions. However there are very few CHF experiments with nano-fluids in flow boiling condition. In the present study, flow boiling CHF experiments using bare round tube with Al 2 O 3 nano-fluid and Al 2 O 3 nano-particle deposited tube with DI water were conducted under atmospheric pressure. CHFs were enhanced up to ∼ 80% with Al 2 O 3 nano-fluid and CHFs with Al 2 O 3 nano-particle deposited tube were also enhanced up to ∼ 80%. Inner surface of test section tube were observed by SEM and AFM after CHF experiments

  4. An experimental study of flow boiling chf with porous surface coatings and surfactant solutions

    International Nuclear Information System (INIS)

    Sarwar, Mohammad Sohail

    2007-02-01

    The boiling crisis or critical heat flux (CHF) phenomenon is an enormously studied topic of the boiling heat transfer. The great interest in the CHF is due to practical motives, since it is desirable to design an equipment (heat exchanger or boiler, etc) to operate at as high a heat flux as possible with optimum heat transfer rates but without the risk of physical burnout. This study consists of two parts of flow boiling CHF experiment: with porous surface coated tubes and by using surfactant solutions as working fluid. In first part, the effect of micro- and nano-porous inside surface coated vertical tubes on the CHF was determined for flow boiling of water in vertical round tubes at atmospheric pressure. CHF was measured for a smooth and three different coated tubes, at mass fluxes of 100∼300 kg/m 2 s and two inlet subcooling temperatures (50 .deg. C and 75 .deg. C). Greater CHF enhancement was found with microporous coatings. Al 2 O 3 microporous coatings with particle size <10 μm and coating thickness of 50 μm showed the best CHF enhancement. The maximum increase in the CHF was about 25% for microporous Al 2 O 3 . A wettability test was performed to study the physical mechanism of increase of CHF with microporous coated surfaces and contact angle was measured for smooth and coated surfaces. Pressure drop measurements were also performed across the coated tubes using the DP-cell apparatus. In second part, surfactant effect on the CHF was determined for water flow boiling at atmospheric pressure in a closed loop filled with solution of tri-sodium phosphate (TSP, Na 3 PO 4 ·12H 2 O). The TSP is usually added to the containment sump water to adjust pH level during accident in nuclear power plants. The CHF was measured for four different surfactant solutions of water in vertical tubes, at different mass fluxes (100 ∼ 500 kg/m 2 s) and two inlet subcooling temperatures (50 .deg. C and 75 .deg. C). Surfactant solutions in the range of 0.05%∼0.2% at low mass

  5. DETERMINATION OF CONVECTIVE HEAT TRANSFER COEFFICIENT AT THE OUTER SURFACE OF A CRYOVIAL BEING PLUNGED INTO LIQUID NITROGEN.

    Science.gov (United States)

    Wang, T; Zhao, G; Tang, H Y; Jiang, Z D

    2015-01-01

    Cell survival upon cryopreservation is affected by the cooling rate. However, it is difficult to model the heat transfer process or to predict the cooling curve of a cryoprotective agent (CPA) solution due to the uncertainty of its convective heat transfer coefficient (h). To measure the h and to better understand the heat transfer process of cryovials filled with CPA solution being plunged in liquid nitrogen. The temperatures at three locations of the CPA solution in a cryovial were measured. Different h values were selected after the cooling process was modeled as natural convection heat transfer, the film boiling and the nucleate boiling, respectively. And the temperatures of the selected points are simulated based on the selected h values. h was determined when the simulated temperature best fitted the experimental temperature. When the experimental results were best fitted, according to natural convection heat transfer model, h(1) = 120 W/(m(2)·K) while due to film boiling and nucleate boiling regimes h(f) = 5 W/(m(2)·K) followed by h(n) = 245 W/(m(2)·K). These values were verified by the differential cooling rates at the three locations of a cryovial. The heat transfer process during cooling in liquid nitrogen is better modeled as film boiling followed by nucleate boiling.

  6. Experimental comparison and visualization of in-tube continuous and pulsating flow boiling

    DEFF Research Database (Denmark)

    Kærn, Martin Ryhl; Markussen, Wiebke Brix; Meyer, Knud Erik

    2018-01-01

    This experimental study investigated the application of fluid flow pulsations for in-tube flow boiling heat transfer enhancement in an 8 mm smooth round tube made of copper. The fluid flow pulsations were introduced by a flow modulating expansion device and were compared with continuous flow...... cycle time (7 s) reduced the time-averaged heat transfer coefficients by 1.8% and 2.3% for the low and high subcooling, respectively, due to significant dry-out when the flow-modulating expansion valve was closed. Furthermore, the flow pulsations were visualized by high-speed camera to assist...... generated by a stepper-motor expansion valve in terms of the time-averaged heat transfer coefficient. The cycle time ranged from 1 s to 7 s for the pulsations, the time-averaged refrigerant mass flux ranged from 50 kg m−2 s−1 to 194 kg m−2 s−1 and the time-averaged heat flux ranged from 1.1 kW m−2 to 30.6 k...

  7. Study of turbulent natural-circulation flow and low-Prandtl-number forced-convection flow

    International Nuclear Information System (INIS)

    Chung, K.S.; Thompson, D.H.

    1980-01-01

    Calculational methods and results are discussed for the coupled energy and momentum equations of turbulent natural circulation flow and low Prandtl number forced convection flow. The objective of this paper is to develop a calculational method for the study of the thermal-hydraulic behavior of coolant flowing in a liquid metal fast breeder reactor channel under natural circulation conditions. The two-equation turbulence model is used to evaluate the turbulent momentum transport property. Because the analogy between momentum transfer and heat transfer does not generally hold for low Prandtl number fluid and natural circulation flow conditions, the turbulent thermal conductivity is calculated independently using equations similar to the two-equation turbulence model. The numerical technique used in the calculation is the finite element method

  8. Boiling transition and the possibility of spontaneous nucleation under high subcooling and high mass flux density flow in a tube

    International Nuclear Information System (INIS)

    Fukuyama, Y.; Kuriyama, T.; Hirata, M.

    1986-01-01

    Boiling transition and inverted annular heat transfer for R-113 have been investigated experimentally in a horizontal tube of 1.2 X 10/sup -3/ meter inner diameter with heating length over inner diameter ratio of 50. Experiments cover a high mass flux density range, a high local subcooling range and a wide local pressure range. Heat transfer characteristics were obtained by using heat flux control steady-state apparatus. Film boiling treated here is limited to the case of inverted annular heat transfer with very thin vapor film, on the order of 10/sup -6/ meter. Moreover, film boiling region is always limited to a certain downstream part, since the system has a pressure gradient along the flow direction. Discussions are presented on the parametric trends of boiling heat transfer characteristic curves and characteristic points. The possible existence is suggested of a spontaneous nucleation control surface boiling phenomena. And boiling transition heat flux and inverted annular heat transfer were correlated

  9. The current status of theoretically based approaches to the prediction of the critical heat flux in flow boiling

    International Nuclear Information System (INIS)

    Weisman, J.

    1991-01-01

    This paper reports on the phenomena governing the critical heat flux in flow boiling. Inducts which vary with the flow pattern. Separate models are needed for dryout in annular flow, wall overheating in plug or slug flow and formation of a vapor blanket in dispersed flow. The major theories and their current status are described for the annular and dispersed regions. The need for development of the theoretical approach in the plug and slug flow region is indicated

  10. Development of nuclear thermal hydraulic verification test and evaluation technology; study on 3-dimension measurement of two-phase flow parameters in subcooled boiling flow

    Energy Technology Data Exchange (ETDEWEB)

    Park, Goon Cherl; Kim, Moon Oh; Cho, Hyung Kyoo; Kim, Seong Jin [Seoul National University, Seoul (Korea)

    2002-04-01

    In this study, the experiments were conducted at different levels of inlet subcooling, flow rate and heat flux in a vertical concentric annulus channel located heater at the center with subcooled boiling conditions of atmosphere pressure and superficial velocity under 1.5m/s. The profiles of void fraction, vapor size, vapor frequency, vapor velocity and IAC were measured by 2 sensor conductivity probe in axially 3 points (L/D{sub h}=90.5,80.1,71.4) and those of liquid velocity by pitot tube. Based on the experiment data subcooled boiling models in MARS and multidimensional code, CFX-4.2 were evaluated was verified for analysis ability of these codes in subcooled boiling. 61 refs., 41 figs., 11 tabs. (Author)

  11. Combined convective heat and airborne pollutant removals in a slot vented enclosure under different flow schemes: Parametric investigations and non unique flow solutions

    International Nuclear Information System (INIS)

    Ren, Xiu-Hong; Hu, Jiang-Tao; Liu, Di; Zhao, Fu-Yun; Li, Xiao-Hong; Wang, Han-Qing

    2016-01-01

    Highlights: • Combined convective heat and airborne transports under different flow schemes. • Natural and forced convection dominated regimes were identified with transition. • Dual solution branches were sustained for the transitional mixing flow scheme. • Rest solutions evolving from motionless flows coincided with other solution branch. • Heat and species lines were presented to delineate heat and mass transport structures. - Abstract: This paper reports a numerical study of mixed convection on a heated and polluted strip within a slot ventilated enclosure in which the displacement and mixing flow schemes are considered. Contours of streamfunction, heatfunction, and massfunction are presented to clearly scrutinize the mechanism of heat and airborne pollutant transports. For the displacement flow scheme, thermal Nusselt and pollutant Sherwood numbers under different Reynolds numbers remain almost constant as the value of Gr/Re 2 decreases down to the regime of forced convection dominated. However, as Ar increases up to the regime of natural convection dominated, both Nu and Sh increase sharply with Ar (Gr/Re 2 ). Similar trends could be observed for the situation of mixing ventilated flow scheme. In the mixing scheme, non unique steady flow solutions could be observed for the range of transitional flow regime. Upward solutions, downward solutions and rest solutions have been exemplified with varying Gr/Re 2 . Dual solution branches could be sustained at the range of 39.0 ≤ Gr/Re 2  ≤ 6.0 × 10 3 , while the rest solutions obtained from rest states were completely coinciding with former continuous solutions. The present work could be significant for the natural optimization and passive control of heat and pollutant removals from the electronic boxes or building enclosures.

  12. Power measurement in the boiling capsules in R2 using delayed neutron detector

    International Nuclear Information System (INIS)

    Roennberg, G.

    1979-03-01

    LWR fuel testing is performed in the R2 reactor by irradiation in both loops and so-called boiling capsules. The loops have forced cooling, and the power can be measured calorimetrically by conventional instrumentation. The boiling capsules have convection cooling, and it has therefore been necessary to develop a special technique for power measurement, the delayed neutron detector (DND). The DND is a pneumatic rabbit system, which activates small uranium samples in the boiling capsules and counts the delayed neutrons for determination of the fission rate. This report describes the equipment used, the procedure of measurement, and the method of evaluation. (atuhor)

  13. Human convective boundary layer and its interaction with room ventilation flow.

    Science.gov (United States)

    Licina, D; Melikov, A; Sekhar, C; Tham, K W

    2015-02-01

    This study investigates the interaction between the human convective boundary layer (CBL) and uniform airflow with different velocity and from different directions. Human body is resembled by a thermal manikin with complex body shape and surface temperature distribution as the skin temperature of an average person. Particle image velocimetry (PIV) and pseudocolor visualization (PCV) are applied to identify the flow around the manikin's body. The findings show that the direction and magnitude of the surrounding airflows considerably influence the airflow distribution around the human body. Downward flow with velocity of 0.175 m/s does not influence the convective flow in the breathing zone, while flow at 0.30 m/s collides with the CBL at the nose level reducing the peak velocity from 0.185 to 0.10 m/s. Transverse horizontal flow disturbs the CBL at the breathing zone even at 0.175 m/s. A sitting manikin exposed to airflow from below with velocity of 0.30 and 0.425 m/s assisting the CBL reduces the peak velocity in the breathing zone and changes the flow pattern around the body, compared to the assisting flow of 0.175 m/s or quiescent conditions. In this case, the airflow interaction is strongly affected by the presence of the chair. © 2014 John Wiley & Sons A/S. Published by John Wiley & Sons Ltd.

  14. Thermal convection in a toroidal duct of a liquid metal blanket. Part II. Effect of axial mean flow

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Xuan, E-mail: xuanz@umich.edu; Zikanov, Oleg

    2017-03-15

    Highlights: • 2D convection flow develops with internal heating and strong axial magnetic field. • The flow is strongly modified by the buoyancy force associated with growing T{sub m}. • Thermal convection is suppressed at high Gr. • High temperature difference between top and bottom walls is expected at high Gr. - Abstract: The work continues the exploration of the effect of thermal convection on flows in toroidal ducts of a liquid metal blanket. This time we consider the effect of the mean flow along the duct and of the associated heat transfer diverting the heat deposited by captured neutrons. Numerical simulations are conducted for a model system with two-dimensional (streamwise-uniform) fully developed flow, purely toroidal magnetic field, and perfectly electrically and thermally insulating walls. Realistically high Grashof (up to 10{sup 11}) and Reynolds (up to 10{sup 6}) numbers are used. It is found that the flow develops thermal convection in the transverse plane at moderate Grashof numbers. At large Grashof numbers, the flow is dominated by the top-bottom asymmetry of the streamwise velocity and stable stratification of temperature, which are caused by the buoyancy force due to the mean temperature growing along the duct. This leads to suppression of thermal convection, weak mixing, and substantial gradients of wall temperature. Further analysis based on more realistic models is suggested.

  15. Derivation of a well-posed and multidimensional drift-flux model for boiling flows

    International Nuclear Information System (INIS)

    Gregoire, O.; Martin, M.

    2005-01-01

    In this note, we derive a multidimensional drift-flux model for boiling flows. Within this framework, the distribution parameter is no longer a scalar but a tensor that might account for the medium anisotropy and the flow regime. A new model for the drift-velocity vector is also derived. It intrinsically takes into account the effect of the friction pressure loss on the buoyancy force. On the other hand, we show that most drift-flux models might exhibit a singularity for large void fraction. In order to avoid this singularity, a remedy based on a simplified three field approach is proposed. (authors)

  16. The Unsteady Variable – Viscosity Free Convection Flow on a ...

    African Journals Online (AJOL)

    The unsteady variable-viscosity free convection flow of a viscous incompressible fluid near an infinite vertical plate (or wall) is investigated under an arbitrary timedependent heating of the plates, and the governing equations of motion and energy transformed into ordinary differential equations. Employing asymptotic ...

  17. Flow Reversal of Fully-Developed Mixed MHD Convection in Vertical Channels

    International Nuclear Information System (INIS)

    Saleh, H.; Hashim, I.

    2010-01-01

    The present analysis is concerned with flow reversal phenomena of the fully-developed laminar combined free and forced MHD convection in a vertical parallel-plate channel. The effect of viscous dissipation is taken into account. Flow reversal adjacent to the cold (or hot) wall is found to exist within the channel as Gr/Re is above (or below) a threshold value. Parameter zones for the occurrence of reversed flow are presented. (fundamental areas of phenomenology(including applications))

  18. Analogy of convective heat transfer between developing laminar secondary flows in pipes

    OpenAIRE

    Ishigaki, Hiroshi; 石垣 博

    1998-01-01

    Analogy of convective heat transfer between developing laminar flows in curved pipes and orthogonally rotating pipes is described through similarity arguments and numerical computation. Governing parameters and a dimensionless axial distance are properly used for the respective flows. When the second parameter is large in each flow, it is shown that the temperature profiles and the Nusselt numbers of the two flows are approximately similar for the same values of the governing parameter, Prand...

  19. Hydrodynamical wind in magnetized accretion flows with convection

    International Nuclear Information System (INIS)

    Abbassi, Shahram; Mosallanezhad, Amin

    2012-01-01

    The existence of outflow and magnetic fields in the inner region of hot accretion flows has been confirmed by observations and numerical magnetohydrodynamic (MHD) simulations. We present self-similar solutions for radiatively inefficient accretion flows (RIAFs) around black holes in the presence of outflow and a global magnetic field. The influence of outflow is taken into account by adopting a radius that depends on mass accretion rate M-dot = M-dot 0 (r/r 0 ) s with s > 0. We also consider convection through a mixing length formula to calculate convection parameter α con . Moreover we consider the additional magnetic field parameters β r,φ,z [ = c 2 r,φ,z /(2c 2 s )], where c 2 r,φ,z are the Alfvén sound speeds in three directions of cylindrical coordinates. Our numerical results show that by increasing all components of the magnetic field, the surface density and rotational velocity increase, but the sound speed and radial infall velocity of the disk decrease. We have also found that the existence of wind will lead to reduction of surface density as well as rotational velocity. Moreover, the radial velocity, sound speed, advection parameter and the vertical thickness of the disk will increase when outflow becomes important in the RIAF. (research papers)

  20. Spontaneous generation and reversals of mean flows in a convectively-generated internal gravity wave field

    Science.gov (United States)

    Couston, Louis-Alexandre; Lecoanet, Daniel; Favier, Benjamin; Le Bars, Michael

    2017-11-01

    We investigate via direct numerical simulations the spontaneous generation and reversals of mean zonal flows in a stably-stratified fluid layer lying above a turbulent convective fluid. Contrary to the leading idealized theories of mean flow generation by self-interacting internal waves, the emergence of a mean flow in a convectively-generated internal gravity wave field is not always possible because nonlinear interactions of waves of different frequencies can disrupt the mean flow generation mechanism. Strong mean flows thus emerge when the divergence of the Reynolds stress resulting from the nonlinear interactions of internal waves produces a strong enough anti-diffusive acceleration for the mean flow, which, as we will demonstrate, is the case when the Prandtl number is sufficiently low, or when the energy input into the internal wavefield by the convection and density stratification are sufficiently large. Implications for mean zonal flow production as observed in the equatorial stratospheres of the Earth, Saturn and Jupiter, and possibly occurring in other geophysical systems such as planetary and stellar interiors will be briefly discussed. Funding provided by the European Research Council (ERC) under the European Union's Horizon 2020 research and innovation program through Grant Agreement No. 681835-FLUDYCO-ERC-2015-CoG.

  1. Void fraction and flow regime determination by optical probe for boiling two-phase flow in a tube subchannel

    International Nuclear Information System (INIS)

    Cheng Huiping; Wu Hongtao; Ba Changxi; Yan Xiaoming; Huang Suyi

    1995-12-01

    In view of the need to determine void fraction and flow regime of vapor-liquid two-phase flow in the steam generator test model, domestic made optical probe was applied on a small-scale freon two-phase flow test rig. Optical probe signals were collected at a sampling rate up to 500 Hz and converted into digital form. Both the time signal, and the amplitude probability density function and FFT spectrum function calculated thereof were analysed in the time and frequency domains respectively. The threshold characterizing vapor or liquid contact with the probe tip was determined from the air-water two-phase flow pressure drop test results. Then, the boiling freon two-phase flow void fraction was determined by single threshold method, and compared with numerical heat transfer computation. Typical patterns which were revealed by the above-mentioned time signal and the functions were found corresponding to distinct flow regimes, as corroborated by visual observation. The experiment shows that the optical probe was a promising technique for two-phase flow void fraction measurement and flow regime identification (3 refs., 15 figs., 1 tab.)

  2. Soret and Hall effects on unsteady MHD free convection flow of ...

    African Journals Online (AJOL)

    International Journal of Engineering, Science and Technology ... effects on unsteady MHD free convection flow of radiating and chemically reactive fluid ... Expressions for shear stress, rate of heat transfer and rate of mass transfer at the plate ...

  3. Mixed convection flow of nanofluid in a square enclosure with an intruded rectangular fin

    International Nuclear Information System (INIS)

    Cong, Ran; Zhou, Xuanyu; De Souza Machado, Bruno; Das, Prodip K.

    2016-01-01

    Mixed convection flow in enclosures has been a subject of interest for many years due to their ever increasing applications in solar collectors, electronic cooling, lubrication technologies, food processing, and nuclear reactors. In comparison, little effort has been given to the problem of mixed convection in enclosures filled with nanofluids, while the addition of nanoparticles in a fluid base to alter specific material properties is considered a feasible solution for many heat transfer problems. Mixed convection of nanofluids is a challenging problem as the addition of nanoparticles changes the fluid’s thermo-physical properties as well as due to the complex interactions among inertia, viscous, and buoyancy forces. In this study, a two-dimensional steady-state numerical model has been developed to investigate mixed convection flow of nanofluids in a square enclosure with an intruded rectangular fin and to optimize the fin geometry for maximizing the heat transfer using the Constructal design. The model has been developed using ANSYS-FLUENT for various fin geometries. Flow fields, temperature fields, and heat transfer rates are examined for different values of Rayleigh and Reynolds numbers for several geometries of the fin with the aim of maximizing the heat transfer from the fin to the surrounding flow. Outcome of this study provides important insight into the heat transfer behavior of nanofluids, which will help in developing novel geometries with enhanced and controlled heat transfer for solar collectors and electronic devices.

  4. Mixed convection flow of nanofluid in a square enclosure with an intruded rectangular fin

    Energy Technology Data Exchange (ETDEWEB)

    Cong, Ran; Zhou, Xuanyu; De Souza Machado, Bruno; Das, Prodip K., E-mail: prodip.das@ncl.ac.uk [School of Mechanical and Systems Engineering Newcastle University Newcastle upon Tyne, NE1 7RU United Kingdom (United Kingdom)

    2016-07-12

    Mixed convection flow in enclosures has been a subject of interest for many years due to their ever increasing applications in solar collectors, electronic cooling, lubrication technologies, food processing, and nuclear reactors. In comparison, little effort has been given to the problem of mixed convection in enclosures filled with nanofluids, while the addition of nanoparticles in a fluid base to alter specific material properties is considered a feasible solution for many heat transfer problems. Mixed convection of nanofluids is a challenging problem as the addition of nanoparticles changes the fluid’s thermo-physical properties as well as due to the complex interactions among inertia, viscous, and buoyancy forces. In this study, a two-dimensional steady-state numerical model has been developed to investigate mixed convection flow of nanofluids in a square enclosure with an intruded rectangular fin and to optimize the fin geometry for maximizing the heat transfer using the Constructal design. The model has been developed using ANSYS-FLUENT for various fin geometries. Flow fields, temperature fields, and heat transfer rates are examined for different values of Rayleigh and Reynolds numbers for several geometries of the fin with the aim of maximizing the heat transfer from the fin to the surrounding flow. Outcome of this study provides important insight into the heat transfer behavior of nanofluids, which will help in developing novel geometries with enhanced and controlled heat transfer for solar collectors and electronic devices.

  5. Mixed convection flow of nanofluid in a square enclosure with an intruded rectangular fin

    Science.gov (United States)

    Cong, Ran; Zhou, Xuanyu; De Souza Machado, Bruno; Das, Prodip K.

    2016-07-01

    Mixed convection flow in enclosures has been a subject of interest for many years due to their ever increasing applications in solar collectors, electronic cooling, lubrication technologies, food processing, and nuclear reactors. In comparison, little effort has been given to the problem of mixed convection in enclosures filled with nanofluids, while the addition of nanoparticles in a fluid base to alter specific material properties is considered a feasible solution for many heat transfer problems. Mixed convection of nanofluids is a challenging problem as the addition of nanoparticles changes the fluid's thermo-physical properties as well as due to the complex interactions among inertia, viscous, and buoyancy forces. In this study, a two-dimensional steady-state numerical model has been developed to investigate mixed convection flow of nanofluids in a square enclosure with an intruded rectangular fin and to optimize the fin geometry for maximizing the heat transfer using the Constructal design. The model has been developed using ANSYS-FLUENT for various fin geometries. Flow fields, temperature fields, and heat transfer rates are examined for different values of Rayleigh and Reynolds numbers for several geometries of the fin with the aim of maximizing the heat transfer from the fin to the surrounding flow. Outcome of this study provides important insight into the heat transfer behavior of nanofluids, which will help in developing novel geometries with enhanced and controlled heat transfer for solar collectors and electronic devices.

  6. Impact of chemical reaction in fully developed radiated mixed convective flow between two rotating disk

    Science.gov (United States)

    Hayat, T.; Khan, M. Waleed Ahmed; Khan, M. Ijaz; Waqas, M.; Alsaedi, A.

    2018-06-01

    Flow of magnetohydrodynamic (MHD) viscous fluid between two rotating disks is modeled. Angular velocities of two disks are different. Flow is investigated for nonlinear mixed convection. Heat transfer is analyzed for nonlinear thermal radiation and heat generation/absorption. Chemical reaction is also implemented. Convective conditions of heat and mass transfer are studied. Transformations used lead to reduction of PDEs into the ODEs. The impacts of important physical variables like Prandtl number, Reynold number, Hartman number, mixed convection parameter, chemical reaction and Schmidt number on velocities, temperature and concentration are elaborated. In addition velocity and temperature gradients are physically interpreted. Our obtained results indicate that radial, axial and tangential velocities decrease for higher estimation of Hartman number.

  7. Local Heat Transfer and CHF for Subcooled Flow Boiling - Annual Report 1993

    International Nuclear Information System (INIS)

    Boyd, Ronald D.

    2000-01-01

    Subcooled flow boiling in heated coolant channels is an important heat transfer enhancement technique in the development of fusion reactor components, where high heat fluxes must be accommodated. As energy fluxes increase in magnitude, additional emphasis must be devoted to enhancing techniques such as sub cooling and enhanced surfaces. In addition to subcooling, other high heat flux alternatives such as high velocity helium and liquid metal cooling have been considered as serious contenders. Each technique has its advantages and disadvantages [1], which must be weighed as to reliability and reduced cost of fusion reactor components. Previous studies [2] have set the stage for the present work, which will concentrate on fundamental thermal hydraulic issues associated with the h-international Thermonuclear Experimental Reactor (ITER) and the Engineering Design Activity (EDA). This proposed work is intended to increase our understanding of high heat flux removal alternatives as well as our present capabilities by: (1) including single-side heating effects in models for local predictions of heat transfer and critical heat flux; (2) inspection of the US, Japanese, and other possible data sources for single-side heating, with the aim of exploring possible correlations for both CHF and local heat transfer; and (3) assessing the viability of various high heat flux removal techniques. The latter task includes: (a) sub-cooled water flow boiling with enhancements such as twisted tapes, and hypervapotrons, (b) high velocity helium cooling, and (c) other potential techniques such as liquid metal cooling. This assessment will increase our understanding of: (1) hypervapotron heat transfer via fins, flow recirculation, and flow oscillation, and (2) swirl flow. This progress report contains selective examples of ongoing work. Section II contains an extended abstract, which is part of and evolving technical paper on single-side f heating. Section III describes additional details

  8. Convective boiling performance of refrigerant R-134a in herringbone and microfin copper tubes

    Energy Technology Data Exchange (ETDEWEB)

    Bandarra Filho, Enio P [Faculdade de Engenharia Mecanica, Universidade Federal de Uberlandia, Av. Joao Naves de Avila, 2160 Bloco 1M, Santa Monica, Uberlandia, MG (Brazil); Saiz Jabardo, Jose M [Escuela Politecnica Superior, Universidad de la Coruna, Calle Mendizabal s/n, 15403 Ferrol, Espana (Spain)

    2006-01-01

    This paper reports an experimental investigation of convective boiling heat transfer and pressure drop of refrigerant R-134a in smooth, standard microfin and herringbone copper tubes of 9.52mm external diameter. Tests have been conducted under the following conditions: inlet saturation temperature of 5{sup o}C, qualities from 5 to 90%, mass velocity from 100 to 500kgs{sup -1}m{sup -2}, and a heat flux of 5kWm{sup -2}. Experimental results indicate that the herringbone tube has a distinct heat transfer performance over the mass velocity range considered in the present study. Thermal performance of the herringbone tube has been found better than that of the standard microfin in the high range of mass velocities, and worst for the smallest mass velocity (G=100kgs{sup -1}m{sup -2}) at qualities higher than 50%. The herringbone tube pressure drop is higher than that of the standard microfin tube over the whole range of mass velocities and qualities. The enhancement parameter is higher than one for both tubes for mass velocities lower than 200kgs{sup -1}m{sup -2}. Values lower than one have been obtained for both tubes in the mass velocity upper range as a result of a significant pressure drop increment not followed by a correspondent increment in the heat transfer coefficient. (author)

  9. Development of thermohydraulic codes for modeling liquid metal boiling in LMR fuel subassemblies

    International Nuclear Information System (INIS)

    Sorokin, G.A.; Avdeev, E.F.; Zhukov, A.V.; Bogoslovskaya, G.P.; Sorokin, A.P.

    2000-01-01

    An investigation into the reactor core accident cooling, which are associated with the power grow up or switch off circulation pumps in the event of the protective equipment comes into action, results in the problem of liquid metal boiling heat transfer. Considerable study has been given over the last 30 years to alkaline metal boiling including researches of heat transfer, boiling patterns, hydraulic resistance, crisis of heat transfer, initial heating up, boiling onset and instability of boiling. The results of these investigations have shown that the process of liquid metal boiling has substantial features in comparison with water boiling. Mathematical modeling of two phase flows in fast reactor fuel subassemblies have been developed intensively. Significant success has been achieved in formulation of two phase flow through the pin bundle and in their numerical realization. Currently a set of codes for thermohydraulic analysis of two phase flows in fast reactor subassembly have been developed with 3D macrotransfer governing equations. These codes are used for analysis of boiling onset and liquid metals boiling in fuel subassemblies during loss-of-coolant accidents, of warming up of reactor core, of blockage of some part of flow cross section in fuel subassembly. (author)

  10. Hydrogen transfer in Pb–Li forced convection flow with permeable wall

    Energy Technology Data Exchange (ETDEWEB)

    Fukada, Satoshi, E-mail: sfukada@nucl.kyushu-u.ac.jp; Muneoka, Taiki; Kinjyo, Mao; Yoshimura, Rhosuke; Katayama, Kazunari

    2015-10-15

    Highlights: • The paper presents experimental and analytical results of Pb–Li eutectic alloy forced convection flow. • Analytical results are in good agreement with ones of hydrogen permeation in Pb–Li forced convection flow. • The results are useful for the design of liquid blanket of fusion reactors. - Abstract: Transient- or steady-state hydrogen permeation from a primary fluid of Li{sub 17}Pb{sub 83} (Pb–Li) through a permeable tube of Inconel-625 alloy to a secondary Ar purge is investigated experimentally under a forced convection flow in a dual cylindrical tube system. Results of the overall hydrogen permeation flux are correlated in terms of diffusivity, solubility and an average axial velocity of Pb–Li and diffusivity and solubility of the solid wall. Analytical solutions under proper assumptions are derived to simulate the transient- and steady-state rates of the overall hydrogen permeation, and close agreement is obtained between experiment and analysis. Two things are clarified from the comparison: (i) how the steady-state permeation rate is affected by the mass-transfer properties and the average velocity of Pb–Li and the properties of Inconel-625, and (ii) how its transient behavior is done by the diffusivity of the two materials. The results obtained here will give important information to estimate or to analyze the tritium transfer rate in fluidized Pb–Li blankets of DEMO or the future commercial fusion reactors.

  11. Flow and Convective Heat Transfer of Cylinder Misaligned from Aerodynamic Axis of Cyclone Flow

    Directory of Open Access Journals (Sweden)

    I. L. Leukhin

    2008-01-01

    Full Text Available The paper provides and analyzes results of experimental investigations on physical specific features of hydrodynamics and convective heat transfer of a cyclone flow with a group of round cylinders located symmetrically relative to its aerodynamic axis, calculative equations for average and local heat transfer factors at characteristic sections of cylinder surface.

  12. Numerical simulation of turbulent convective flow over wavy terrain

    Science.gov (United States)

    Dörnbrack, A.; Schumann, U.

    1993-09-01

    By means of a large-eddy simulation, the convective boundary layer is investigated for flows over wavy terrain. The lower surface varies sinusoidally in the downstream direction while remaining constant in the other. Several cases are considered with amplitude δ up to 0.15 H and wavelength λ of H to 8 H, where H is the mean fluid-layer height. At the lower surface, the vertical heat flux is prescribed to be constant and the momentum flux is determined locally from the Monin-Obukhov relationship with a roughness length z o=10-4 H. The mean wind is varied between zero and 5 w *, where w * is the convective velocity scale. After rather long times, the flow structure shows horizontal scales up to 4 H, with a pattern similar to that over flat surfaces at corresponding shear friction. Weak mean wind destroys regular spatial structures induced by the surface undulation at zero mean wind. The surface heating suppresses mean-flow recirculation-regions even for steep surface waves. Short surface waves cause strong drag due to hydrostatic and dynamic pressure forces in addition to frictional drag. The pressure drag increases slowly with the mean velocity, and strongly with δ/ H. The turbulence variances increase mainly in the lower half of the mixed layer for U/w *>2.

  13. A stability analysis of ventilated boiling channels

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.; Podowski, M.Z.; Lahey, R.T. Jr.

    1986-01-01

    A mathematical model has been developed for the linear stability analysis of a system of ventilated parallel boiling channels. This model accounts for subcooled boiling, an arbitrary heat flux distribution, distributed and local hydraulic losses, heated wall dynamics, slip flow, turbulent mixing and arbitrary flow paths for transverse ventilation. The digital computer program MAZDA-NF was written for numerical evaluation of the mathematical model. Comparison of MAZDA-NF results with those obtained form both a closed form analytical solution and experiment, showed good agreement. A parametric study revealed that such phenomena as subcooled boiling, the transverse coupling between channels (due to cross-flow and mixing) and power skewing can have a significant impact on predicted stability margins. An analysis of an advanced BWR fuel, of the ASEA-ATOM SVEA design, has indicated that transverse ventilation may considerably improve channel stability. (orig.)

  14. Flow visualization and critical heat flux measurement of a boundary layer pool boiling process

    International Nuclear Information System (INIS)

    Cheung, F.B.; Haddad, K.H.; Liu, Y.C.; Shiah, S.W.

    1998-01-01

    As part of the effort to evaluate the concept of external passive cooling of core melt by cavity flooding under severe accident conditions, a subscale boundary layer boiling (SBLB) facility, consisting of a pressurized water tank with a condenser unit, a heated hemispherical test vessel, and a data acquisition/photographic system, was developed to simulate the boiling process on the external bottom surface of a fully submerged reactor vessel. Transient quenching and steady-state boiling experiments were conducted in the facility to measure the local critical heat flux (CHF) and observe the underlying mechanisms under well controlled saturated and subcooled conditions. Large elongated vapor slugs were observed in the bottom region of the vessel which gave rise to strong upstream influences in the resulting two-phase liquid-vapor boundary layer flow along the vessel outer surface. The local CHF values deduced from the transient quenching data appeared to be very close to those obtained in the steady-state boiling experiments. Comparison of the SBLB data was made with available 2-D full-scale data and the differences were found to be rather small except in a region near the bottom center of the vessel. The angular position of the vessel outer surface and the degree of subcooling of water had dominant effects on the local critical heat flux. They totally dwarfed the effect of the physical dimensions of the test vessels. (author)

  15. Boiling heat transfer on horizontal tube bundles

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    Nucleate boiling heat transfer characteristics for a tube in a bundle differ from that for a single tube in a pool and this difference is known as 'tube bundle effect.' There exist two bundle effects, positive and negative. The positive bundle effect enhances heat transfer due to convective flow induced by rising bubbles generated from the lower tubes, while the negative bundle effect deteriorates heat transfer due to vapor blanketing caused by accumulation of bubbles. Staggered tube bundles tested and found that the upper tubes in bundles have higher heat transfer coefficients than the lower tubes. The effects of various parameters such as pressure, tube geometry and oil contamination on heat transfer have been examined. Some workers attempted to clarify the mechanism of occurrence of 'bundle effect' by testing tube arrangements of small scale. All reported only enhancement in heat transfer but results showed the symptom of heat transfer deterioration at higher heat fluxes. As mentioned above, it has not been clarified so far even whether the 'tube bundle effect' should serve as enhancement or deterioration of heat transfer in nucleate boiling. In this study, experiments are performed in detail by using bundles of small scale, and effects of heat flux distribution, pressure and tube location are clarified. Furthermore, some consideration on the mechanisms of occurrence of 'tube bundle effect' is made and a method for prediction of heat transfer rate is proposed

  16. Unstable fluid flow in a water-cooled heating channel

    International Nuclear Information System (INIS)

    Delayre, R.; Saunier, J.P.

    1961-01-01

    Experimental investigations of the instable behavior of a pressurized water flow in forced convection in a heating channel, with subcooled or bulk boiling have been carried. Tests were conducted at 1140, 850 and 570 psi. The test section was 35 in. high, surmounted by a 25.4 in. riser, these sections were by-passed by a pipe where the flow was between 1 and 4 times the flow in the test section. The water velocity (in the test section) was between 1.6 and 6.6 ft/s. Under certain conditions oscillations with a period of several seconds and perfectly stable have been observed. A mathematical model has been defined and a good agreement obtained for the main characteristics of the oscillations. It seems that the dimensions of the riser have a determining effect: the inception of bulk boiling gives an important variation of the driving head which can generate oscillations due to the non-zero delay for the system to reach its equilibrium. (author) [fr

  17. Unsteady MHD free convection flow and heat transfer along an ...

    African Journals Online (AJOL)

    Steady free convection flow of an electrically conducting fluid along an infinite vertical porous plate under Arrhenius kinetics are investigated in the presence of strong transverse magnetic field imposed perpendicularly to the plate .A similarity parameter length scale (h) as a function of time and the suction velocity are ...

  18. Convection flows driven by laser heating of a liquid layer

    OpenAIRE

    Rivière , David; Selva , Bertrand; Chraibi , Hamza; Delabre , Ulysse; Delville , Jean-Pierre

    2016-01-01

    International audience; When a fluid is heated by the absorption of a continuous laser wave, the fluid density decreases in the heated area. This induces a pressure gradient that generates internal motion of the fluid. Due to mass conservation, convection eddies emerge in the sample. To investigate these laser-driven bulk flows at the microscopic scale, we built a setup to perform temperature measurements with a fluorescent-sensitive dye on the one hand, and measured the flow pattern at diffe...

  19. Local pressure gradients due to incipience of boiling in subcooled flows

    Energy Technology Data Exchange (ETDEWEB)

    Ruggles, A.E.; McDuffee, J.L. [Univ. of Tennessee, Knoxville, TN (United States)

    1995-09-01

    Models for vapor bubble behavior and nucleation site density during subcooled boiling are integrated with boundary layer theory in order to predict the local pressure gradient and heat transfer coefficient. Models for bubble growth rate and bubble departure diameter are used to scale the movement of displaced liquid in the laminar sublayer. An added shear stress, analogous to a turbulent shear stress, is derived by considering the liquid movement normal to the heated surface. The resulting mechanistic model has plausible functional dependence on wall superheat, mass flow, and heat flux and agrees well with data available in the literature.

  20. Pool Boiling Characteristics on the Microstructure surfaces with Both Rectangular Cavities and Channels

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Dong Eok; Myung, Byung-Soo [Kyungpook Nat’l Univ., Daegu (Korea, Republic of); Park, Su Cheong; Yu, Dong In [POSTECH, Pohang (Korea, Republic of); Kim, Moo Hwan [Korea Institute of Nuclear Safety (KINS), Daejeon (Korea, Republic of); Ahn, Ho Seon [Incheon Nat’l Univ., Incheon (Korea, Republic of)

    2016-06-15

    Based on a surface design with rectangular cavities and channels, we investigated the effects of gravity and capillary pressure on pool-boiling Critical Heat Flux (CHF). The microcavity structures could prevent liquid flow by the capillary pressure effect. In addition, the microchannel structures contributed to induce one-dimensional liquid flow on the boiling surface. The relationship between the CHF and capillary flow was clearly established. The driving potentials for the liquid supply into a boiling surface can be generated by the gravitational head and capillary pressure. Through an analysis of pool boiling and visualization data, we reveal that the liquid supplement to maintain the nucleate boiling condition on a boiling surface is closely related to the gravitational pressure head and capillary pressure effect.

  1. Prediction of void fraction in subcooled flow boiling

    International Nuclear Information System (INIS)

    Petelin, S.; Koncar, B.

    1998-01-01

    The information on heat transfer and especially on the void fraction in the reactor core under subcooled conditions is very important for the water-cooled nuclear reactors, because of its influence upon the reactivity of the systems. This paper gives a short overview of subcooled boiling phenomenon and indicates the simplifications made by the RELAP5 model of subcooled boiling. RELAP5/MOD3.2 calculations were compared with simple one-dimensional models and with high-pressure Bartolomey experiments.(author)

  2. Investigations on the propagation of free surface boiling in a vertical superheated liquid column

    International Nuclear Information System (INIS)

    Das, P.K.; Bhat, G.S.; Arakeri, V.H.

    1987-01-01

    Some experimental studies on boiling propagation in a suddenly depressurized superheated vertical liquid column are reported. The propagation velocity of this phase change has been measured using an optical method. This velocity is strongly dependent on liquid superheat, liquid purity and test section size. The measured velocities of less than 5 m s -1 are significantly lower than the sonic velocity. Present observations suggest that the dominant mechanism for boiling propagation is convection. (author)

  3. Investigations on the propagation of free surface boiling in a vertical superheated liquid column

    Energy Technology Data Exchange (ETDEWEB)

    Das, P.K.; Bhat, G.S.; Arakeri, V.H.

    1987-04-01

    Some experimental studies on boiling propagation in a suddenly depressurized superheated vertical liquid column are reported. The propagation velocity of this phase change has been measured using an optical method. This velocity is strongly dependent on liquid superheat, liquid purity and test section size. The measured velocities of less than 5 m s/sup -1/ are significantly lower than the sonic velocity. Present observations suggest that the dominant mechanism for boiling propagation is convection.

  4. The determination of the initial point of net vapor generation in flow subcooled boiling

    International Nuclear Information System (INIS)

    Yan Changqi; Sun Zhongning

    2000-01-01

    The experimental results for the initial point of net vapor generation in up-flow subcooled boiling in an internally-heated annulus are given. The characteristics of the initial point of net vapor generation and the problem on gamma ray attenuation measurement are discussed. The comparison between the data and a calculation model is given, it is showed that the data agree well with the model

  5. Nonlinear Multiplicative Schwarz Preconditioning in Natural Convection Cavity Flow

    KAUST Repository

    Liu, Lulu

    2017-03-17

    A natural convection cavity flow problem is solved using nonlinear multiplicative Schwarz preconditioners, as a Gauss-Seidel-like variant of additive Schwarz preconditioned inexact Newton (ASPIN). The nonlinear preconditioning extends the domain of convergence of Newton’s method to high Rayleigh numbers. Convergence performance varies widely with respect to different groupings of the fields of this multicomponent problem, and with respect to different orderings of the groupings.

  6. Nonlinear Multiplicative Schwarz Preconditioning in Natural Convection Cavity Flow

    KAUST Repository

    Liu, Lulu; Zhang, Wei; Keyes, David E.

    2017-01-01

    A natural convection cavity flow problem is solved using nonlinear multiplicative Schwarz preconditioners, as a Gauss-Seidel-like variant of additive Schwarz preconditioned inexact Newton (ASPIN). The nonlinear preconditioning extends the domain of convergence of Newton’s method to high Rayleigh numbers. Convergence performance varies widely with respect to different groupings of the fields of this multicomponent problem, and with respect to different orderings of the groupings.

  7. Grid dependency of wall heat transfer for simulation of natural convection flow problems

    NARCIS (Netherlands)

    Loomans, M.G.L.C.; Seppänen, O.; Säteri, J.

    2007-01-01

    In the indoor environment natural convection is a well known air flow phenomenon. In numerical simulations applying the CFD technique it is also known as a flow problem that is difficult to solve. Alternatives are available to overcome the limitations of the default approach (standard k-e model with

  8. Effect of liquid density differences on boiling two-phase flow stability

    International Nuclear Information System (INIS)

    Furuya, Masahiro; Manera, Annalisa; Bragt, David D.B.; Hagen, Tim H.J.J. van der; Kruijf, Willy J.M.de

    2002-01-01

    In order to investigate the effect of considering liquid density dependence on local fluid temperature in the thermal-hydraulic stability, a linear stability analysis is performed for a boiling natural circulation loop with an adiabatic riser. Type-I and Type-II instabilities were to investigate according to Fukuda-Kobori's classification. Type-I instability is dominant when the flow quality is low, while Type-II instability is relevant at high flow quality. Type-II instability is well known as the typical density wave oscillation. Neglecting liquid density differences yields estimates of Type-II instability margins that are too small, due to both a change in system-dynamics features and in the operational point. On the other hand, neglecting liquid density differences yields estimates of Type-I stability margins that are too large, especially due to a change in the operational point. Neglecting density differences is thus non-conservative in this case. Therefore, it is highly recommended to include liquid density dependence on the fluid subcooling in the stability analysis if a flow loop with an adiabatic rise is operated under the condition of low flow quality. (author)

  9. Heat transfer properties of organic coolants containing high boiling residues

    International Nuclear Information System (INIS)

    Debbage, A.G.; Driver, M.; Waller, P.R.

    1964-01-01

    Heat transfer measurements were made in forced convection with Santowax R, mixtures of Santowax R and pyrolytic high boiling residue, mixtures of Santowax R and CMRE Radiolytic high boiling residue, and OMRE coolant, in the range of Reynolds number 10 4 to 10 5 . The data was correlated with the equation Nu = 0.015 Re b 0.85 Pr b 0.4 with an r.m.s. error of ± 8.5%. The total maximum error arising from the experimental method and inherent errors in the physical property data has been estimated to be less than ± 8.5%. From the correlation and physical property data, the decrease in heat transfer coefficient with increasing high boiling residue concentration has been determined. It has been shown that subcooled boiling in organic coolants containing high boiling residues is a complex phenomenon and the advantages to be gained by operating a reactor in this region may be marginal. Gas bearing pumps used initially in these experiments were found to be unsuitable; a re-designed ball bearing system lubricated with a terphenyl mixture was found to operate successfully. (author)

  10. Group Analysis of Free Convection Flow of a Magnetic Nanofluid with Chemical Reaction

    Directory of Open Access Journals (Sweden)

    Md. Jashim Uddin

    2015-01-01

    Full Text Available A theoretical study of two-dimensional magnetohydrodynamics viscous incompressible free convective boundary layer flow of an electrically conducting, chemically reacting nanofluid from a convectively heated permeable vertical surface is presented. Scaling group of transformations is used in the governing equations and the boundary conditions to determine absolute invariants. A third-order ordinary differential equation which corresponds to momentum conservation and two second-order ordinary differential equations which correspond to energy and nanoparticle volume fraction (species conservation are derived. Our (group analysis indicates that, for the similarity solution, the convective heat transfer coefficient and mass transfer velocity are proportional to x-1/4 whilst the reaction rate is proportional to x-1/2, where x is the axial distance from the leading edge of the plate. The effects of the relevant controlling parameters on the dimensionless velocity, temperature, and nanoparticle volume fraction are examined. The accuracy of the technique we have used was tested by performing comparisons with the results of published work and the results were found to be in good agreement. The present computations indicate that the flow is accelerated and temperature enhanced whereas nanoparticle volume fractions are decreased with increasing order of chemical reaction. Furthermore the flow is strongly decelerated, whereas the nanoparticle volume fraction and temperature are enhanced with increasing magnetic field parameter. Increasing convection-conduction parameter increases velocity and temperatures but has a weak influence on nanoparticle volume fraction distribution. The present study demonstrates the thermal enhancement achieved with nanofluids and also magnetic fields and is of relevance to nanomaterials processing.

  11. Numerical simulation of the electro convective onset and complex flows of dielectric liquid in an annulus

    Energy Technology Data Exchange (ETDEWEB)

    Fernandes, Dolfred Vijay; Lee, Heon Deok; Alapati, Suresh; Suh, Yong Kweon [Dong A Univ., Busan (Korea, Republic of)

    2012-12-15

    We conducted a numerical study on the onset of electro-convection as well as the complex flow phenomena of dielectric liquid subjected to unipolar autonomous charge injection in the annular gap between two concentric circular cylindrical electrodes. The Nernst Planck equations governing the charge density transport, the Poisson equation for the electric potential and the Navier Stokes equations for the fluid flow are solved numerically using the finite volume method. The developed code is validated by comparing the critical stability parameter values for the onset of electro convection with those obtained from the linear stability analysis. We identify in a parameter space the stable hydrostatic state and the electro convection state. The electro convection is again divided into three regimes: stationary, oscillatory and chaotic. For inner cylinder radius 1.0, i r {>=} we observed an increase in the number of charged plumes and vortex pairs with stability parameter T before the electro convection becomes chaotic. For outer injection, although the onset of electroconvection starts at T higher than the inner injection, the onset of chaotic motion occurs at lower T.

  12. The Natural Convection Heat Transfer inside Vertical Pipe: Characteristic of Pipe Flow according to the Boundary layer

    Energy Technology Data Exchange (ETDEWEB)

    Ohk, Seung Min; Chung, Bum Jin [Kyunghee University, Yongin (Korea, Republic of)

    2016-05-15

    The Passive Cooling System (PCS) driven by natural forces drew research attention since Fukushima nuclear power plant accident. This study investigated the natural convection heat transfer inside of vertical pipe with emphasis on the phenomena regarding the boundary layer interaction. Numerical calculations were carried out using FLUENT 6.3. Experiments were performed for the parts of the cases to explore the accuracy of calculation. Based on the analogy, heat transfer experiment is replaced by mass transfer experiment using sulfuric acid copper sulfate (CuSO{sub 4}. H{sub 2}SO{sub 4}) electroplating system. The natural convection heat transfer inside a vertical pipe is studied experimentally and numerically. Experiments were carried out using sulfuric acid-copper sulfate (H{sub 2}SO{sub 4}-CuSO{sub 4}) based on the analogy concept between heat and mass transfer system. Numerical analysis was carried out using FLUENT 6.3. It is concluded that the boundary layer interaction along the flow passage influences the heat transfer, which is affected by the length, diameter, and Prandtl number. For the large diameter and high Prandtl number cases, where the thermal boundary layers do not interfered along the pipe, the heat transfer agreed with vertical flat plate for laminar and turbulent natural convection correlation within 8%. When the flow becomes steady state, the forced convective flow appears in the bottom of the vertical pipe and natural convection flow appears near the exit. It is different behavior from the flow on the parallel vertical flat plates. Nevertheless, the heat transfer was not different greatly compared with those of vertical plate.

  13. Encyclopedia of two-phase heat transfer and flow III macro and micro flow boiling and numerical modeling fundamentals

    CERN Document Server

    2018-01-01

    Set III of this encyclopedia is a new addition to the previous Sets I and II. It contains 26 invited chapters from international specialists on the topics of numerical modeling of two-phase flows and evaporation, fundamentals of evaporation and condensation in microchannels and macrochannels, development and testing of micro two-phase cooling systems for electronics, and various special topics (surface wetting effects, microfin tubes, two-phase flow vibration across tube bundles). The chapters are written both by renowned university researchers and by well-known engineers from leading corporate research laboratories. Numerous "must read" chapters cover the fundamentals of research and engineering practice on boiling, condensation and two-phase flows, two-phase heat transfer equipment, electronics cooling systems, case studies and so forth. Set III constitutes a "must have" reference together with Sets I and II for thermal engineering researchers and practitioners.

  14. Analysis of natural convection heat transfer and flows in internally heated stratified liquid pools

    International Nuclear Information System (INIS)

    Gubaidullin, A.A. Jr.; Dinh, T.N.; Sehgal, B.R.

    1999-01-01

    In this paper, natural convection flows and heat transfer in a liquid pool, with two superposed immiscible fluid layers, are analyzed. The objective of the study is to examine the effect of interfacial hydrodynamics and to develop a method which enables energy splitting to be evaluated in a stratified liquid pool. The thermal convection, with and without an internal heat source, in a rectangular cavity with different pairs of fluids was numerically simulated by a CFD code FLOW-3D. It was found that the code performs very well for prediction of heat transfer coefficients for different conditions. The hydrodynamic coupling between immiscible layers was found to have minor, if any, impact on the natural convection heat transfer for the conditions examined. Calculated results were used to develop, and validate, a new correlation for energy splitting and for heat transfer in stratified liquid pools

  15. The boiling crisis in a subcooled liquid flowing in a vertical annular channel

    International Nuclear Information System (INIS)

    Passos, J.C.

    1989-01-01

    Experimental results concerning the critical heat flux density for a variety of forced flow conditions of Freon 113 in a circular annular channel of 3 mm width and 107 mm length when the inside wall is heated are presented. The flow configurations were also visualized prior and during the boiling crisis. For inlet liquid velocities equal or larger than 0.041 m/s, the correlated dimensionless data extends the range of validity of those of Katto for relatively much longer tubes. A simple balance of forces over a bubble attached to the wall shows that, for smaller velocities, the gravity effect has to be taken into account in the establishment of a more general correlation. (author)

  16. Hydromagnetic natural convection flow between vertical parallel plates with time-periodic boundary conditions

    International Nuclear Information System (INIS)

    Adesanya, S.O.; Oluwadare, E.O.; Falade, J.A.; Makinde, O.D.

    2015-01-01

    In this paper, the free convective flow of magnetohydrodynamic fluid through a channel with time periodic boundary condition is investigated by taking the effects of Joule dissipation into consideration. Based on simplifying assumptions, the coupled governing equations are reduced to a set of nonlinear boundary valued problem. Approximate solutions are obtained by using semi-analytical Adomian decomposition method. The effect of pertinent parameters on the fluid velocity, temperature distribution, Nusselt number and skin friction are presented graphically and discussed. The result of the computation shows that an increase in the magnetic field intensity has significant influence on the fluid flow. - Highlights: • The influence of magnetic field on the free convective fluid flow is considered. • The coupled equations are solved by using Adomian decomposition method. • The Adomian series solution agreed with previously obtained result. • Magnetic field decreases the velocity maximum but enhances temperature field

  17. Impact of magnetic field in three-dimensional flow of Sisko nanofluid with convective condition

    Energy Technology Data Exchange (ETDEWEB)

    Hayat, T. [Department of Mathematics, Quaid-I-Azam University, Islamabad 44000 (Pakistan); Nonlinear Analysis and Applied Mathematics (NAAM) Research Group, Department of Mathematics, Faculty of Science, King Abdulaziz University, P.O. Box 80203, Jeddah 21589 (Saudi Arabia); Muhammad, Taseer, E-mail: taseer_qau@yahoo.com [Department of Mathematics, Quaid-I-Azam University, Islamabad 44000 (Pakistan); Ahmad, B. [Nonlinear Analysis and Applied Mathematics (NAAM) Research Group, Department of Mathematics, Faculty of Science, King Abdulaziz University, P.O. Box 80203, Jeddah 21589 (Saudi Arabia); Shehzad, S.A. [Department of Mathematics, Comsats Institute of Information Technology, Sahiwal 57000 (Pakistan)

    2016-09-01

    This communication addresses the magnetohydrodynamic (MHD) three dimensional flow of Sisko nanofluid bounded by a surface stretched bidirectionally. Nanofluid model includes the Brownian motion and thermophoresis. Heat transfer through convective condition is discussed. Developed condition with the zero nanoparticles mass flux at the surface is implemented. The governing problems subject to boundary layer approximations are computed for the convergent series solutions. Effects of interesting flow parameters on the temperature and nanoparticles concentration distributions are studied and discussed. Skin friction coefficients and the local Nusselt number are computed and analyzed. - Highlights: • Three-dimensional flow of Sisko nanofluid is modeled. • Uniform applied magnetic field is adopted. • Brownian motion and thermophoresis effects are accounted. • Heat transfer convective condition is utilized. • Recently constructed condition with zero nanoparticles mass flux is implemented.

  18. Numerical investigation of boiling heat transfer on hydrocarbon mixture refrigerant in vertical rectangular minichannel

    Directory of Open Access Journals (Sweden)

    Huixing Li

    2016-05-01

    Full Text Available In order to investigate the characteristics of boiling heat transfer for hydrocarbon mixture refrigerant in plate-fin heat exchanger which is used in the petrochemical industry field, a model was established on boiling heat transfer in vertical rectangular channel. The simulated results were compared with the experimental data from literature. The results show that the deviation between the simulated results and experimental data is within ±15%. Meanwhile, the characteristic of boiling heat transfer was investigated in vertical rectangular minichannel of plate-fin heat exchanger. The results show that the boiling heat transfer coefficient increases with the increase in quality and mass flux and is slightly impacted by the heat flux. This is because that the main boiling mechanism is forced convective boiling while the contribution of nucleate boiling is slight. The correlation of Liu and Winterton is in good agreement with the simulation results. The deviation between correlation calculations and simulation results is mostly less than ±15%. These results will provide some constructive instructions for the understanding of saturated boiling mechanism in a vertical rectangular minichannel and the prediction of heat transfer performance in plate-fin heat exchanger.

  19. Investigation of combined free and forced convection in a 2 x 6 rod bundle during controlled flow transients

    International Nuclear Information System (INIS)

    Bates, J.M.; Khan, E.U.

    1980-10-01

    An experimental study was performed to obtain local fluid velocity and temperature measurements in the mixed (combined free and forced) convection regime for specific flow coastdown transients. A brief investigation of steady-state flows for the purely free-convection regime was also completed. The study was performed using an electrically heated 2 x 6 rod bundle contained in a flow housing. In addition a transient data base was obtained for evaluating the COBRA-WC thermal-hydraulic computer program

  20. Laminar mixed convection heat transfer in a vertical circular tube under buoyancy-assisted and opposed flows

    International Nuclear Information System (INIS)

    Mohammed, Hussein A.

    2008-01-01

    Laminar mixed convection heat transfer for assisted and opposed air flows in the entrance region of a vertical circular tube with the using of a uniform wall heat flux boundary condition has been experimentally investigated. The experimental setup was designed for determining the effect of flow direction and the effect of tube inclination on the surface temperature, local and average Nusselt numbers with Reynolds number ranged from 400 to 1600 and Grashof number from 2.0 x 10 5 to 6.2 x 10 6 . It was found that the circumferential surface temperature along the dimensionless tube length for opposed flow would be higher than that both of assisted flow and horizontal tube [Mohammed HA, Salman YK. Experimental investigation of combined convection heat transfer for thermally developing flow in a horizontal circular cylinder. Appl Therm Eng 2007;27(8-9):1522-33] due to the stronger free convective currents within the cross-section. The Nusselt number values would be lower for opposed flow than that for assisted flow. It was inferred that the behaviour of Nu x for opposed flow to be strongly dependent on the combination of Re and Gr numbers. Empirical equations expressing the average Nusselt numbers in terms of Grashof and Reynolds numbers were proposed for both assisted and opposed flow cases. The average heat transfer results were compared with previous literature and showed similar trend and satisfactory agreement

  1. Return to nucleate boiling

    International Nuclear Information System (INIS)

    Shumway, R.W.

    1985-01-01

    This paper presents a collection of TMIN (temperature of return to nucleate boiling) correlations, evaluates them under several conditions, and compares them with a wide range of data. Purpose is to obtain the best one for use in a water reactor safety computer simulator known as TRAC-B. Return to nucleate boiling can occur in a reactor accident at either high or low pressure and flow rates. Most of the correlations yield unrealistic results under some conditions. A new correlation is proposed which overcomes many of the deficiencies

  2. Experimental determination of heat transfer critical conditions in water forced convection at low pressure in a circular channel

    International Nuclear Information System (INIS)

    Fernandes, M.P.

    1973-02-01

    An experimental determination was made of heat transfer critical conditions in a circular channel, uniformly heated, and internally cooled by water in ascending forced convection, under a pressure slightly above atmospheric pressure. Measurements were made of water flow, pressure, electric power temperature and heating, and a systematic analysis was made of the system's parameters. The values obtained for the heat critical flux are circa 50% lower than those predicted by Becker and Biasi and this is accounted to flowing instabilities of thermo-hydrodynamic nature. It is suggested that the flowing channels of circuits aiming at the study of the boiling crisis phenomenon be expanded in its upper extremity, and that the coolant circulation be kept through a pump with a pressure X flow characteristic as vertical as possible

  3. Analysis of the Onset of Flow Instability in rectangular heated channel using drift flux model

    International Nuclear Information System (INIS)

    El-Hadjen, H.; Balistrou, M.; Hamidouche, T.; Bousbia-Salah, A.

    2005-01-01

    Two-phase flow excursion (Ledinegg) instability in boiling channels is of great concern in the design and operation of numerous practical systems especially in Research Reactors. Such instability can lead to significant reduction in channel flow, thereby causing premature burnout of the heated channel before the CHF point. The present work focuses on a simulation of pressure drop in forced convection boiling in vertical narrow and parallel uniformly heated channels. The objective is to determine the point of Onset of Flow Instability (OFI) by varying input flow rate. The axial void distribution is also provided. The numerical model is based on the finite difference method which transforms the partial differential conservation equation of mass, momentum and energy, in algebraic equations. Closure relationships based upon the drift flux model and other constitutive equations are considered to determine the channel pressure drop under steady state boiling conditions. The model validation is performed by confronting the calculations with the Oak Ridge National Laboratory thermal Hydraulic Test Loop (THTL) experimental data set. Further verification of this model is performed by code- to code verification using the results of RELAP5/Mod 3.2 code. (author)

  4. Mixed convection boundary-layer flow from a horizontal circular cylinder with a constant surface heat flux

    Energy Technology Data Exchange (ETDEWEB)

    Nazar, R.; Amin, N. [Department of Mathematics, Universiti Teknologi Malaysia, 81310 Johor Bahru, Johor (Malaysia); Pop, I. [Faculty of Mathematics, University of Cluj, R-3400 Cluj, CP 253 (Romania)

    2004-02-01

    The laminar mixed convection boundary-layer flow of a viscous and incompressible fluid past a horizontal circular cylinder, which is maintained at a constant heat flux and is placed in a stream flowing vertically upward has been theoretically studied in this paper. The solutions for the flow and heat transfer characteristics are evaluated numerically for different values of the mixed convection parameter {lambda} with the Prandtl number Pr = 1 and 7, respectively. It is found, as for the case of a heated or cooled cylinder, considered by Merkin [5], that assisting flow delays separation of the boundary-layer and can, if the assisting flow is strong enough, suppress it completely. The opposing flow, on the other side, brings the separation point nearer to the lower stagnation point and for sufficiently strong opposing flows there will not be a boundary-layer on the cylinder. (orig.)

  5. A look-up table for fully developed film-boiling heat transfer

    International Nuclear Information System (INIS)

    Groeneveld, D.C.; Leung, L.K.H.; Vasic, A.Z.; Guo, Y.J.; Cheng, S.C.

    2003-01-01

    An improved look-up table for film-boiling heat-transfer coefficients has been derived for steam-water flow inside vertical tubes. Compared to earlier versions of the look-up table, the following improvements were made: - The database has been expanded significantly. The present database contains 77,234 film-boiling data points obtained from 36 sources. - The upper limit of the thermodynamic quality range was increased from 1.2 to 2.0. The wider range was needed as non-equilibrium effects at low flows can extend well beyond the point where the thermodynamic quality equals unity. - The surface heat flux has been replaced by the surface temperature as an independent parameter. - The new look-up table is based only on fully developed film-boiling data. - The table entries at flow conditions for which no data are available is based on the best of five different film-boiling prediction methods. The new film-boiling look-up table predicts the database for fully developed film-boiling data with an overall rms error in heat-transfer coefficient of 10.56% and an average error of 1.71%. A comparison of the prediction accuracy of the look-up table with other leading film-boiling prediction methods shows that the look-up table results in a significant improvement in prediction accuracy

  6. A Ghost Fluid/Level Set Method for boiling flows and liquid evaporation: Application to the Leidenfrost effect

    International Nuclear Information System (INIS)

    Rueda Villegas, Lucia; Alis, Romain; Lepilliez, Mathieu; Tanguy, Sébastien

    2016-01-01

    The development of numerical methods for the direct numerical simulation of two-phase flows with phase change, in the framework of interface capturing or interface tracking methods, is the main topic of this study. We propose a novel numerical method, which allows dealing with both evaporation and boiling at the interface between a liquid and a gas. Indeed, in some specific situations involving very heterogeneous thermodynamic conditions at the interface, the distinction between boiling and evaporation is not always possible. For instance, it can occur for a Leidenfrost droplet; a water drop levitating above a hot plate whose temperature is much higher than the boiling temperature. In this case, boiling occurs in the film of saturated vapor which is entrapped between the bottom of the drop and the plate, whereas the top of the water droplet evaporates in contact of ambient air. The situation can also be ambiguous for a superheated droplet or at the contact line between a liquid and a hot wall whose temperature is higher than the saturation temperature of the liquid. In these situations, the interface temperature can locally reach the saturation temperature (boiling point), for instance near a contact line, and be cooler in other places. Thus, boiling and evaporation can occur simultaneously on different regions of the same liquid interface or occur successively at different times of the history of an evaporating droplet. Standard numerical methods are not able to perform computations in these transient regimes, therefore, we propose in this paper a novel numerical method to achieve this challenging task. Finally, we present several accuracy validations against theoretical solutions and experimental results to strengthen the relevance of this new method.

  7. Effect of viscous dissipation on mixed convection flow in a vertical ...

    African Journals Online (AJOL)

    International Journal of Engineering, Science and Technology .... third kind for flow over a flat plate and in the thermal entrance region of a rectangular channel. ... on mixed convection in a vertical channel using Robin boundary conditions was ... Hajmohammadi and Nourazar (2014) studied the effect of a thin gas layer in ...

  8. Local interfacial structure of subcooled boiling flow in a heated annulus

    International Nuclear Information System (INIS)

    Lee, Tae-Ho; Kim, Seong-O; Yun, Byong-Jo; Park, Goon-Cherl; Hibiki, Takashi

    2008-01-01

    Local measurements of flow parameters were performed for vertical upward subcooled boiling flows in an internally heated annulus. The annulus channel consisted of an inner heater rod with a diameter of 19.0 mm and an outer round tube with an inner diameter of 37.5 mm, and the hydraulic equivalent diameter was 18.5 mm. The double-sensor conductivity probe method was used for measuring the local void fraction, interfacial area concentration, bubble Sauter mean diameter and gas velocity, whereas the miniature Pitot tube was used for measuring the local liquid velocity. A total of 32 data sets were acquired consisting of various combinations of heat flux, 88.1-350.9 kW/m 2 , mass flux, 469.7-1061.4kg(m 2 s) and inlet liquid temperature, 83.8-100.5degC. Six existing drift-flux models, six exiting correlations of the interfacial area concentration and bubble layer thickness model were evaluated using the data obtained in the experiment. (author)

  9. Radial basis functions in mathematical modelling of flow boiling in minichannels

    Directory of Open Access Journals (Sweden)

    Hożejowska Sylwia

    2017-01-01

    Full Text Available The paper addresses heat transfer processes in flow boiling in a vertical minichannel of 1.7 mm depth with a smooth heated surface contacting fluid. The heated element for FC-72 flowing in a minichannel was a 0.45 mm thick plate made of Haynes-230 alloy. An infrared camera positioned opposite the central, axially symmetric part of the channel measured the plate temperature. K-type thermocouples and pressure converters were installed at the inlet and outlet of the minichannel. In the study radial basis functions were used to solve a problem concerning heat transfer in a heated plate supplied with the controlled direct current. According to the model assumptions, the problem is treated as twodimensional and governed by the Poisson equation. The aim of the study lies in determining the temperature field and the heat transfer coefficient. The results were verified by comparing them with those obtained by the Trefftz method.

  10. Mixed convection around calandria tubes in a ¼ scale CANDU-6 moderator circulation tank

    Energy Technology Data Exchange (ETDEWEB)

    Atkins, M.D.; Rossouw, D.J.; Boer, M. [Nuclear Science Division, School of Mechanical and Aeronautical Engineering, University of the Witwatersrand, Johannesburg (South Africa); Kim, T., E-mail: tong.kim@wits.ac.za [Nuclear Science Division, School of Mechanical and Aeronautical Engineering, University of the Witwatersrand, Johannesburg (South Africa); Rhee, B.W.; Kim, H.T. [Severe Accident and PHWR Safety Research Division, Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2017-05-15

    Highlights: • A secondary jet is formed at a stagnation region and is directed towards the center of the MCT. • The secondary jet undergoes the significant dissipation and mixing due to calandria tubes (CTs). • Its cooling effectiveness is reduced on the CTs in the bottom of the MCT. • With forced convection dominance, peak heat transfer is on the upper CT surface. • With natural convection dominance, peak heat transfer is on the lower CT surface. - Abstract: This study experimentally characterizes mixed convection around calandria tubes (CTs) in a ¼ scale CANDU-6 moderator circulation tank (MCT) that uses air as the working fluid. In a full scale CANDU-6 reactor that undergoes a postulated dual failure with a loss-of-coolant accident without the emergency core cooling system available, mixed convection heat transfer occurs around the CTs. The cooling effectiveness of the moderator is diminished as an emergency heat sink if overheating eventually leads to film boiling. To prevent the onset of film boiling, local sub-cooling margins of the moderator needs to be maintained or else the critical heat flux should be increased. Circulating the moderator which interacts with the overheated CTs increases the heat transfer into the moderator which may suppress film boiling. The present experimental results demonstrate that the cooling effectiveness of the circulating moderator, in particular the secondary jet, is attenuated substantially as it is convected away from the inner wall towards the center of the MCT. The momentum of the secondary jet is diffused through the CTs. At a low jet Reynolds number, the secondary jet becomes ineffective so that some overheated CTs positioned in the other half of the MCT are cooled only by natural convection.

  11. R245fa Flow Boiling inside a 4.2 mm ID Microfin Tube

    Science.gov (United States)

    Longo, G. A.; Mancin, S.; Righetti, G.; Zilio, C.

    2017-11-01

    This paper presents the R245fa flow boiling heat transfer and pressure drop measurements inside a mini microfin tube with internal diameter at the fin tip of 4.2 mm, having 40 fins, 0.15 mm high with a helix angle of 18°. The tube was brazed inside a copper plate and electrically heated from the bottom. Sixteen T-type thermocouples are located in the copper plate to monitor the wall temperature. The experimental measurements were carried out at constant mean saturation temperature of 30 °C, by varying the refrigerant mass velocity between 100 kg m-2 s-1 and 300 kg m-2 s-1, the vapour quality from 0.15 to 0.95, at two different heat fluxes: 30 and 60 kW m-2. The experimental results are presented in terms of two-phase heat transfer coefficient, onset dryout vapour quality, and frictional pressure drop. Moreover, the experimental measurements are compared against the most updated models for boiling heat transfer coefficient and frictional pressure drop estimations available in the open literature for microfin tubes.

  12. Modeling of Multisize Bubbly Flow and Application to the Simulation of Boiling Flows with the Neptune_CFD Code

    Directory of Open Access Journals (Sweden)

    Christophe Morel

    2009-01-01

    Full Text Available This paper describes the modeling of boiling multisize bubbly flows and its application to the simulation of the DEBORA experiment. We follow the method proposed originally by Kamp, assuming a given mathematical expression for the bubble diameter pdf. The original model is completed by the addition of some new terms for vapor compressibility and phase change. The liquid-to-interface heat transfer term, which essentially determines the bubbles condensation rate in the DEBORA experiment, is also modeled with care. First numerical results realized with the Neptune_CFD code are presented and discussed.

  13. An acoustic-convective splitting-based approach for the Kapila two-phase flow model

    Energy Technology Data Exchange (ETDEWEB)

    Eikelder, M.F.P. ten, E-mail: m.f.p.teneikelder@tudelft.nl [EDF R& D, AMA, 7 boulevard Gaspard Monge, 91120 Palaiseau (France); Eindhoven University of Technology, Department of Mathematics and Computer Science, P.O. Box 513, 5600 MB Eindhoven (Netherlands); Daude, F. [EDF R& D, AMA, 7 boulevard Gaspard Monge, 91120 Palaiseau (France); IMSIA, UMR EDF-CNRS-CEA-ENSTA 9219, Université Paris Saclay, 828 Boulevard des Maréchaux, 91762 Palaiseau (France); Koren, B.; Tijsseling, A.S. [Eindhoven University of Technology, Department of Mathematics and Computer Science, P.O. Box 513, 5600 MB Eindhoven (Netherlands)

    2017-02-15

    In this paper we propose a new acoustic-convective splitting-based numerical scheme for the Kapila five-equation two-phase flow model. The splitting operator decouples the acoustic waves and convective waves. The resulting two submodels are alternately numerically solved to approximate the solution of the entire model. The Lagrangian form of the acoustic submodel is numerically solved using an HLLC-type Riemann solver whereas the convective part is approximated with an upwind scheme. The result is a simple method which allows for a general equation of state. Numerical computations are performed for standard two-phase shock tube problems. A comparison is made with a non-splitting approach. The results are in good agreement with reference results and exact solutions.

  14. Experimental study of the hydrodynamic instabilities occurring in boiling-water reactors

    International Nuclear Information System (INIS)

    Fabreca, S.

    1964-10-01

    The subjects is an experimental out-of pile loop study of the hydrodynamic oscillations occurring in boiling-water reactors. The study was carried out at atmospheric pressure and at pressure of about 8 atmospheres, in channels heated electrically by a constant and uniform specified current. In the test at 8 atmospheres the channel was a round tube of approximately 6 mm interior diameter. At 1 atmosphere a ring-section channel was used, 10 * 20 mm in diameter, with an inner heating tube and an outer tube of pyrex. It was possible to operate with natural convection and also with forced convection with test-channel by-pass. The study consists of 3 parts: 1. Preliminary determination of the laws governing pressure-drop during boiling. 2. Determination of the fronts at which oscillation appears, within a wide range of the parameters involved. 3. A descriptive study of the oscillations and measurement of the periods. The report gives the oscillation fronts with natural and forced convection for various values of the singular pressure drop at the channel inlet and for various riser lengths. The results are presented in non-dimensional form, which is available, in first approximation, for all geometric scales and for all fluids. Besides the following points were observed: - the wall (nature and thickness) can be an important factor ; - oscillation can occur in a horizontal channel. (author) [fr

  15. Numerical solution of one-dimensional transient, two-phase flows with temporal fully implicit high order schemes: Subcooled boiling in pipes

    Energy Technology Data Exchange (ETDEWEB)

    López, R., E-mail: ralope1@ing.uc3m.es; Lecuona, A., E-mail: lecuona@ing.uc3m.es; Nogueira, J., E-mail: goriba@ing.uc3m.es; Vereda, C., E-mail: cvereda@ing.uc3m.es

    2017-03-15

    Highlights: • A two-phase flows numerical algorithm with high order temporal schemes is proposed. • Transient solutions route depends on the temporal high order scheme employed. • ESDIRK scheme for two-phase flows events exhibits high computational performance. • Computational implementation of the ESDIRK scheme can be done in a very easy manner. - Abstract: An extension for 1-D transient two-phase flows of the SIMPLE-ESDIRK method, initially developed for incompressible viscous flows by Ijaz is presented. This extension is motivated by the high temporal order of accuracy demanded to cope with fast phase change events. This methodology is suitable for boiling heat exchangers, solar thermal receivers, etc. The methodology of the solution consist in a finite volume staggered grid discretization of the governing equations in which the transient terms are treated with the explicit first stage singly diagonally implicit Runge-Kutta (ESDIRK) method. It is suitable for stiff differential equations, present in instant boiling or condensation processes. It is combined with the semi-implicit pressure linked equations algorithm (SIMPLE) for the calculation of the pressure field. The case of study consists of the numerical reproduction of the Bartolomei upward boiling pipe flow experiment. The steady-state validation of the numerical algorithm is made against these experimental results and well known numerical results for that experiment. In addition, a detailed study reveals the benefits over the first order Euler Backward method when applying 3rd and 4th order schemes, making emphasis in the behaviour when the system is subjected to periodic square wave wall heat function disturbances, concluding that the use of the ESDIRK method in two-phase calculations presents remarkable accuracy and computational advantages.

  16. Simplified numerical model for predicting onset of flow instability in parallel heated channels

    International Nuclear Information System (INIS)

    Noura Rassoul; El-Khider Si-Ahmed; Tewfik Hamidouche; Anis Bousbia-Salah

    2005-01-01

    Full text of publication follows: Flow instabilities are undesirable phenomena in heated channels since change in flow rate affects the local heat transfer characteristics and may results in premature burnout. For instance, two-phase flow excursion (Ledinegg) instability in boiling channels is of great concern in the design and operation of numerous practical systems especially the MTR fuel type Research Reactors. For heated parallel channels, the negative-sloped segment of the pressure drop-flow rate characteristics (demand curve) of a boiling channel becomes negative. Such instability can lead to significant reduction in channel flow, thereby causing premature burnout of the heated channel before the CHF point. Furthermore, as a consequence of this flow decrease, different types of flow instabilities that may appear can also induce (density wave) flow oscillations of constant amplitude or diverging amplitude. The present work focuses on a numerical simulation of pressure drop in forced convection boiling in vertical narrow and parallel uniformly heated channels. The objective is to determine the point of Onset of flow instability by varying input flow rate without any consideration to density wave oscillations. By the way, the axial void distribution is provided. The numerical model is based on the finite difference method which transform the partial differential conservation equations of Mass, Momentum and Energy, in algebraic equations. Closure relationships as the drift flux model and other constitutive equations are considered to determine the channel pressure drop under steady state boiling conditions. The model validation is performed by confronting the calculations with the Oak Ridge National Laboratory Thermal Hydraulic Test Loop (THTL) experimental data set. Further verification of this model is performed by code-to code verification using the results of RELAP5/Mod 3.2 code. (authors)

  17. Numerical simulation of nucleate boiling and heat transfer using MPL-MAFL

    Energy Technology Data Exchange (ETDEWEB)

    Han Young Yoon, Hee Cheol Kim [Korea Atomic Energy Research Inst., Taejon (Korea, Republic of); Koshizuka, Seiichi; Oka, Yoshiaki [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab

    2000-10-01

    A mesh-free numerical method is presented for direct calculation of bubble growth. It is a combination of particle and gridless methods where the terms, 'particle' and 'gridless', refer to Lagrangian and Eulerian schemes respectively. Thus, an arbitrary-Lagrangian-Eulerian calculation is possible, in this method, with a cloud of computing points that are equivalent to the computing cells in mesh-based methods. The moving interface is traced through the Lagrangian motion of the computing points using a particle method and, at the fixed computing points, convection is calculated using a gridless method. The particle interaction model of the moving-particle semi-implicit (MPS) method is applied to the differential operators and the meshless-advection using a flow-directional local-grid (MAFL) scheme is utilized for the gridless method. A complex moving interface problems can be effectively analyzed by MPS-MAFL since the mesh is no longer used. The present method is applied to the calculation of gas-liquid two-phase flow with and without the phase change in two dimensions. The pressure and temperature gradients are ignored for the vapor region and the phase interface is treated as a free boundary. As an isothermal flow, a gas bubble rising in viscous liquids is simulated numerically and the results are compared with the empirical correlation. The energy equation is coupled with the equation of motion for the calculation of nucleate pool boiling. The numerical results are provided for the bubble growth rate, departure radius, and the heat transfer rate, which show good agreement with the experimental observations. The heat transfer mechanism associated with nucleate pool boiling is quantitatively evaluated and discussed with previous empirical studies. (author)

  18. Identification of dominant flow structures in rapidly rotating convection of liquid metals using Dynamic Mode Decomposition

    Science.gov (United States)

    Horn, S.; Schmid, P. J.; Aurnou, J. M.

    2016-12-01

    The Earth's metal core acts as a dynamo whose efficiency in generating and maintaining the magnetic field is essentially determined by the rotation rate and the convective motions occurring in its outer liquid part. For the description of the primary physics in the outer core the idealized system of rotating Rayleigh-Bénard convection is often invoked, with the majority of studies considering only working fluids with Prandtl numbers of Pr ≳ 1. However, liquid metals are characterized by distinctly smaller Prandtl numbers which in turn result in an inherently different type of convection. Here, we will present results from direct numerical simulations of rapidly rotating convection in a fluid with Pr ≈ 0.025 in cylindrical containers and Ekman numbers as low as 5 × 10-6. In this system, the Coriolis force is the source of two types of inertial modes, the so-called wall modes, that also exist at moderate Prandtl numbers, and cylinder-filling oscillatory modes, that are a unique feature of small Prandtl number convection. The obtained flow fields were analyzed using the Dynamic Mode Decomposition (DMD). This technique allows to extract and identify the structures that govern the dynamics of the system as well as their corresponding frequencies. We have investigated both the regime where the flow is purely oscillatory and the regime where wall modes and oscillatory modes co-exist. In the purely oscillatory regime, high and low frequency oscillatory modes characterize the flow. When both types of modes are present, the DMD reveals that the wall-attached modes dominate the flow dynamics. They precess with a relatively low frequency in retrograde direction. Nonetheless, also in this case, high frequency oscillations have a significant contribution.

  19. Neural network modeling of chaotic dynamics in nuclear reactor flows

    International Nuclear Information System (INIS)

    Welstead, S.T.

    1992-01-01

    Neural networks have many scientific applications in areas such as pattern classification and time series prediction. The universal approximation property of these networks, however, can also be exploited to provide researchers with tool for modeling observed nonlinear phenomena. It has been shown that multilayer feed forward networks can capture important global nonlinear properties, such as chaotic dynamics, merely by training the network on a finite set of observed data. The network itself then provides a model of the process that generated the data. Characterizations such as the existence and general shape of a strange attractor and the sign of the largest Lyapunov exponent can then be extracted from the neural network model. In this paper, the author applies this idea to data generated from a nonlinear process that is representative of convective flows that can arise in nuclear reactor applications. Such flows play a role in forced convection heat removal from pressurized water reactors and boiling water reactors, and decay heat removal from liquid-metal-cooled reactors, either by natural convection or by thermosyphons

  20. Computational fluid dynamics modeling of mixed convection flows in buildings enclosures

    Energy Technology Data Exchange (ETDEWEB)

    Kayne, Alexander; Agarwal, Ramesh K. [Department of Mechanical Engineering and Materials Science, Washington University, St. Louis, MO 63130 (United States)

    2013-07-01

    In recent years Computational Fluid Dynamics (CFD) simulations are increasingly used to model the air circulation and temperature environment inside the rooms of residential and office buildings to gain insight into the relative energy consumptions of various HVAC systems for cooling/heating for climate control and thermal comfort. This requires accurate simulation of turbulent flow and heat transfer for various types of ventilation systems using the Reynolds-Averaged Navier-Stokes (RANS) equations of fluid dynamics. Large Eddy Simulation (LES) or Direct Numerical Simulation (DNS) of Navier-Stokes equations is computationally intensive and expensive for simulations of this kind. As a result, vast majority of CFD simulations employ RANS equations in conjunction with a turbulence model. In order to assess the modeling requirements (mesh, numerical algorithm, turbulence model etc.) for accurate simulations, it is critical to validate the calculations against the experimental data. For this purpose, we use three well known benchmark validation cases, one for natural convection in 2D closed vertical cavity, second for forced convection in a 2D rectangular cavity and the third for mixed convection in a 2D square cavity. The simulations are performed on a number of meshes of different density using a number of turbulence models. It is found that k-epsilon two-equation turbulence model with a second-order algorithm on a reasonable mesh gives the best results. This information is then used to determine the modeling requirements (mesh, numerical algorithm, turbulence model etc.) for flows in 3D enclosures with different ventilation systems. In particular two cases are considered for which the experimental data is available. These cases are (1) air flow and heat transfer in a naturally ventilated room and (2) airflow and temperature distribution in an atrium. Good agreement with the experimental data and computations of other investigators is obtained.

  1. Electric fields effect on the rise of single bubbles during boiling

    International Nuclear Information System (INIS)

    Siedel, Samuel; Cioulachtjian, Serge; Bonjour, Jocelyn

    2009-01-01

    An experimental study of saturated pool boiling on a single artificial nucleation site without and with the application of an electric field on the boiling surface has been conducted. N-pentane is boiling on a copper surface and is recorded with a high speed camera providing high quality pictures and movies. The accuracy of the visualization allowed establishing an experimental bubble growth law from a large number of experiments. This law shows that the evaporation rate is decreasing during the bubble growth, and underlines the importance of liquid motion induced by the preceding bubble. Bubble rise is therefore studied: once detached, bubbles accelerate vertically until reaching a maximum velocity in good agreement with a correlation from literature. The bubbles then turn to another direction. The effect of applying an electric field on the boiling surface in finally studied. In addition to changes of the bubble shape, changes are also shown in the liquid plume and the convective structures above the surface. Lower maximum rising velocities were measured in the presence of electric fields, especially with a negative polarity. (author)

  2. Forced convection boiling of sodium. Study carried out in the framework of fast neutrons reactors safety

    International Nuclear Information System (INIS)

    Charlety, Paul

    1971-01-01

    Within the framework of the safety of fast neutron reactors, this research thesis reports the study of sodium boiling in order to assess safety margins, and to predict the consequences of some accidents. The author thus addresses issues related to sodium boiling by notably focussing on boiling physics. He first defines these issues and presents the adopted approach for this research, and then describes the experimental installation. He reports the experimental study which comprised different types of tests, and presents experimental results. He reports the development of a calculation model which could report phenomena which have been experimentally noticed [fr

  3. Effect of coolant flow rate on the power at onset of nucleate boiling in a swimming pool type research reactor

    International Nuclear Information System (INIS)

    Khan, L.A.; Ahmad, N.; Ahmad, S.

    1998-01-01

    The effect of flow rate of coolant on power of Onset Nucleate Boiling (ONB) in a reference core of a swimming pool type research reactor has been studied using a as standard computer code PARET. It has been found that the decrease in the coolant flow rate results in a corresponding decrease in power at ONB. (author)

  4. CFD analysis of bubble microlayer and growth in subcooled flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Owoeye, Eyitayo James, E-mail: msgenius10@ufl.edu; Schubring, DuWanye, E-mail: dlschubring@ufl.edu

    2016-08-01

    Highlights: • A new LES-microlayer model is introduced. • Analogous to the unresolved SGS in LES, analysis of bubble microlayer was performed. • The thickness of bubble microlayer was computed at both steady and transient states. • The macroscale two-phase behavior was captured with VOF coupled with AMR. • Numerical validations were performed for both the micro- and macro-region analyses. - Abstract: A numerical study of single bubble growth in turbulent subcooled flow boiling was carried out. The macro- and micro-regions of the bubble were analyzed by introducing a LES-microlayer model. Analogous to the unresolved sub-grid scale (SGS) in LES, a microlayer analysis was performed to capture the unresolved thermal scales for the micro-region heat transfer by deriving equations for the microlayer thickness at steady and transient states. The phase change at the macro-region was based on Volume-of-Fluid (VOF) interface tracking method coupled with adaptive mesh refinement (AMR). Large Eddy Simulation (LES) was used to model the turbulence characteristics. The numerical model was validated with multiple experimental data from the open literature. This study includes parametric variations that cover the operating conditions of boiling water reactor (BWR) and pressurized water reactor (PWR). The numerical model was used to study the microlayer thickness, growth rate, dynamics, and distortion of the bubble.

  5. The relative contributions of thermo-solutal Marangoni convections on flow patterns in a liquid bridge

    Science.gov (United States)

    Minakuchi, H.; Takagi, Y.; Okano, Y.; Gima, S.; Dost, S.

    2014-01-01

    A numerical simulation study was carried out to investigate the relative contributions of thermal and solutal Marangoni convections on transport structures in a liquid bridge under zero gravity. The liquid bridge in the model represents a three dimensional half-zone configuration of the Floating Zone (FZ) growth system. Three dimensional field equations of the liquid zone, i.e. continuity, momentum, energy, and diffusion equations, were solved by the PISO algorithm. Computations were performed using the open source software OpenFOAM. The numerical simulation results show that the flow field becomes three-dimensional and time-depended when the solutal Marangoni number is larger than the critical value. It was also shown that not only flow patterns but also the azimuthal wave number (m) changes due to the competing contributions of thermal and solutal Marangoni convective flows.

  6. Rotating turbulent Rayleigh-Bénard convection subject to harmonically forced flow reversals

    NARCIS (Netherlands)

    Geurts, B.J.; Kunnen, R.P.J.

    2014-01-01

    The characteristics of turbulent flow in a cylindrical Rayleigh–B´enard convection cell which can be modified considerably in case rotation is included in the dynamics. By incorporating the additional effects of an Euler force, i.e., effects induced by nonconstant rotation rates, a remarkably strong

  7. Rotating turbulent Rayleigh–Bénard convection subject to harmonically forced flow reversals

    NARCIS (Netherlands)

    Geurts, Bernardus J.; Kunnen, Rudie P.J.

    2014-01-01

    The characteristics of turbulent flow in a cylindrical Rayleigh–Bénard convection cell which can be modified considerably in case rotation is included in the dynamics. By incorporating the additional effects of an Euler force, i.e., effects induced by non-constant rotation rates, a remarkably strong

  8. Elevator mode convection in flows with strong magnetic fields

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Li; Zikanov, Oleg, E-mail: zikanov@umich.edu [Department of Mechanical Engineering, University of Michigan-Dearborn, 48128-1491 Michigan (United States)

    2015-04-15

    Instability modes in the form of axially uniform vertical jets, also called “elevator modes,” are known to be the solutions of thermal convection problems for vertically unbounded systems. Typically, their relevance to the actual flow state is limited by three-dimensional breakdown caused by rapid growth of secondary instabilities. We consider a flow of a liquid metal in a vertical duct with a heated wall and strong transverse magnetic field and find elevator modes that are stable and, thus, not just relevant, but a dominant feature of the flow. We then explore the hypothesis suggested by recent experimental data that an analogous instability to modes of slow axial variation develops in finite-length ducts, where it causes large-amplitude fluctuations of temperature. The implications for liquid metal blankets for tokamak fusion reactors that potentially invalidate some of the currently pursued design concepts are discussed.

  9. Natural convection cooling of LEU cores for Pakistan research reactor-1

    International Nuclear Information System (INIS)

    Khan, L.A.; Bokhari, I.H.; Akhtar, K.M.

    1991-08-01

    The first high power and equilibrium LEU cores of PARR-1 have been analysed to assess the maximum operating power based on natural convection cooling, need for forced cooling to remove the decay heat and to estimate safety margins that commensurate with the predetermined power limit. Computer code NATCON and standard correlations have been used for the analysis. The parameters studied includes coolant velocity, temperature distribution in the core, heat fluxes at onset of nucleate boiling, pulsed boiling and burnup. (author)

  10. Assessment of correlations and models for the prediction of CHF in water subcooled flow boiling

    Science.gov (United States)

    Celata, G. P.; Cumo, M.; Mariani, A.

    1994-01-01

    The present paper provides an analysis of available correlations and models for the prediction of Critical Heat Flux (CHF) in subcooled flow boiling in the range of interest of fusion reactors thermal-hydraulic conditions, i.e. high inlet liquid subcooling and velocity and small channel diameter and length. The aim of the study was to establish the limits of validity of present predictive tools (most of them were proposed with reference to light water reactors (LWR) thermal-hydraulic studies) in the above conditions. The reference dataset represents almost all available data (1865 data points) covering wide ranges of operating conditions in the frame of present interest (0.1 less than p less than 8.4 MPa; 0.3 less than D less than 25.4 mm; 0.1 less than L less than 0.61 m; 2 less than G less than 90.0 Mg/sq m/s; 90 less than delta T(sub sub,in) less than 230 K). Among the tens of predictive tools available in literature four correlations (Levy, Westinghouse, modified-Tong and Tong-75) and three models (Weisman and Ileslamlou, Lee and Mudawar and Katto) were selected. The modified-Tong correlation and the Katto model seem to be reliable predictive tools for the calculation of the CHF in subcooled flow boiling.

  11. Interfacial energies of aqueous mixtures and porous coverings for enhancing pool boiling heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Melendez, Elva [CIICAp, Universidad Autonoma del Estado de Morelos, 62210 (Mexico); Reyes, Rene [Departamento de Ingenieria Quimica y Alimentos, Universidad de las Americas Puebla, Santa Catarina Martir Cholula, Puebla 72820 (Mexico)

    2006-08-15

    The interfacial energies effects on pool boiling were measured for combinations of aqueous ethanol mixtures and cationic surfactants. The mixture with 16% ethanol by weight had the lowest contact angle (associated to the highest wettability) and produced the highest convective heat transfer coefficient, h, among the aqueous ethanol mixtures. The surfactant sodium-lauryl-sulfate added at 100 ppm (its calculated critical micelle concentration CMC) to the 16% ethanol aqueous mixture produced an additional increment of the wettability of the mixture and of the h values; other concentrations of the surfactant reduced de contact angle and h values. The effect of these interfacial energies represents a mass-transfer contribution to pool boiling and the proposal of mixture effects both as increased spreadability and as micelle states. Several randomly constructed porous coverings, contributing to the breakage of vapor slugs around the heater, were tested; produced the highest h values for average pore diameters of 0.5 mm, and covering thickness of 0.972 mm. The synergistic effect on h of the interfacial energies of mixtures at their critical micelle concentration, and porous coverings was measured. Therefore, the independent driving forces combined in this study for increasing pool boiling heat transfer are (a) spreadability of the liquid on the solid; (b) the bubble's size reduction, achieved by micelle states; and (c) the bubble's breakage, induced by the porous coverings, for vapor flow not under pressure drop control. (author)

  12. Experimental and analytical investigations of granular materials: Shear flow and convective heat transfer

    Science.gov (United States)

    Ahn, Hojin

    1989-12-01

    Granular materials flowing down an inclined chute were studied experimentally and analytically. Characteristics of convective heat transfer to granular flows were also investigated experimentally and numerically. Experiments on continuous, steady flows of granular materials in an inclined chute were conducted with the objectives of understanding the characteristics of chute flows and of acquiring information on the rheological behavior of granular material flow. Existing constitutive equations and governing equations were used to solve for fully developed chute flows of granular materials, and thus the boundary value problem was formulated with two parameters (the coefficient of restitution between particles, and the chute inclination) and three boundary values at the chute base wall (the values of solid fraction, granular temperature, and mean velocity at the wall). The boundary value problem was numerically solved by the shooting method. These analytical results were also compared with the present experimental values and with the computer simulations by other investigators in their literature. Experiments on heat transfer to granular flows over a flat heating plate were conducted with three sizes of glass beads, polystyrene beads, and mustard seeds. A modification on the existing model for the convective heat transfer was made using the effective Nusselt number and the effective Peclet number, which include the effects of solid fraction variations. The slightly modified model could describe the heat transfer characteristics of both fast and slow flows (supercritical and subcritical). A numerical analysis of the transfer to granular flows was also performed. The results were compared with the present experimental data, and reasonable agreement was found in the comparison.

  13. Cryostatic stability equation

    International Nuclear Information System (INIS)

    Sydoriak, S.G.

    1976-01-01

    Although criteria for cryostatic stability of superconducting magnets cooled by pool boiling of liquid helium have been widely discussed the same cannot be said for magnets cooled by natural convection or forced flow boiling in channels. Boiling in narrow channels is shown to be qualitatively superior to pool boiling because the recovery heat flux equals the breakaway flux for narrow channels, whereas the two are markedly different in pool boiling. A second advantage of channel boiling is that it is well understood and calculable; pool peak nucleate boiling heat flux has been adequately measured only for boiling from the top of an immersed heated body. Peak boiling from the bottom is much less and (probably) depends strongly on the extent of the bottom surface. Equations are presented by which one can calculate the critical boiling heat flux for parallel wall vertical channels subject to either natural convection or forced flow boiling, with one or both walls heated. The one-heated-wall forced flow equation is discussed with regard to design of a spiral wound solenoid (pancake magnet) having a slippery insulating tape between the windings

  14. Heat Transfer and Flows of Thermal Convection in a Fluid-Saturated Rotating Porous Medium

    Directory of Open Access Journals (Sweden)

    Jianhong Kang

    2015-01-01

    Full Text Available Thermal convection at the steady state for high Rayleigh number in a rotating porous half space is investigated. Taking into account the effect of rotation, Darcy equation is extended to incorporate the Coriolis force term in a rotating reference frame. The velocity and temperature fields of thermal convection are obtained by using the homotopy analysis method. The influences of Taylor number and Rayleigh number on the Nusselt number, velocity profile, and temperature distribution are discussed in detail. It is found that the Nusselt number decreases rapidly with the increase of Taylor number but tends to have an asymptotic value. Besides, the rotation can give rise to downward flow in contrast with the upward thermal convection.

  15. Regimes of Axisymmetric Flow and Scaling Laws in a Rotating Annulus with Local Convective Forcing

    Directory of Open Access Journals (Sweden)

    Susie Wright

    2017-07-01

    Full Text Available We present a numerical study of axisymmetric flow in a rotating annulus in which local thermal forcing, via a heated annular ring on the outside of the base and a cooled circular disk in the centre of the top surface, drives convection. This new configuration is a variant of the classical thermally-driven annulus, where uniform heating and cooling are applied through the outer and inner sidewalls respectively. The annulus provides an analogue to a planetary circulation and the new configuration, with its more relaxed vertical thermal boundary conditions, is expected to better emulate vigorous convection in the tropics and polar regions as well as baroclinic instability in the mid-latitude baroclinic zone. Using the Met Office/Oxford Rotating Annulus Laboratory (MORALS code, we have investigated a series of equilibrated, two dimensional axisymmetric flows across a large region of parameter space. These are characterized in terms of their velocity and temperature fields. When rotation is applied several distinct flow regimes may be identified for different rotation rates and strengths of differential heating. These regimes are defined as a function of the ratio of the horizontal Ekman layer thickness to the non-rotating thermal boundary layer thickness and are found to be similar to those identified in previous annulus experiments. Convection without rotation is also considered and the scaling of the heat transport with Rayleigh number is calculated. This is then compared with existing work on the classical annulus as well as horizontal and Rayleigh-Bénard convection. As with previous studies on both rotating and non-rotating convection the system’s behaviour is found to be aspect ratio dependent. This dependence is seen in the scaling of the non-rotating Nusselt number and in transitions between regimes in the rotating case although further investigation is required to fully explain these observations.

  16. Effects of Al{sub 2}O{sub 3} nanoparticles deposition on critical heat flux of R-123 in flow boiling heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Seok Bin; Bang, In Cheol [School of Mechanical and Nuclear Engineering, Ulsan National Institute of Science and Technology (UNIST), Ulsan (Korea, Republic of)

    2015-06-15

    In this study, R-123 flow boiling experiments were carried out to investigate the effects of nanoparticle deposition on heater surfaces on flow critical heat flux (CHF) and boiling heat transfer. It is known that CHF enhancement by nanoparticles results from porous structures that are very similar to layers of Chalk River unidentified deposit formed on nuclear fuel rod surfaces during the reactor operation period. Although previous studies have investigated the surface effects through surface modifications, most studies are limited to pool boiling conditions, and therefore, the effects of porous surfaces on flow boiling heat transfer are still unclear. In addition, there have been only few reports on suppression of wetting for decoupled approaches of reasoning. In this study, bare and Al{sub 2}O{sub 3} nanoparticle-coated surfaces were prepared for the study experiments. The CHF of each surface was measured with different mass fluxes of 1,600 kg/m{sup 2}s, 1,800 kg/m{sup 2}s, 2,100 kg/m{sup 2}s, 2,400 kg/m{sup 2}s, and 2,600 kg/m{sup 2}s. The nanoparticle-coated tube showed CHF enhancement up to 17% at a mass flux of 2,400 kg/m{sup 2}s compared with the bare tube. The factors for CHF enhancement are related to the enhanced rewetting process derived from capillary action through porous structures built-up by nanoparticles while suppressing relative wettability effects between two sample surfaces as a highly wettable R-123 refrigerant was used as a working fluid.

  17. MHD free convection flow past an oscillating plate in the presence of ...

    African Journals Online (AJOL)

    The study of unsteady magnetohydrodynamic heat and mass transfer in MHD flow past an infinite vertical oscillating plate through porous medium, taking account of the presence of free convection and mass transfer. The energy and chemical species equations are solved in closed form by Laplace-transform technique and ...

  18. Microchannel boiling mechanisms leading to burnout

    International Nuclear Information System (INIS)

    Landram, C.S.; Riddle, R.A.

    1994-01-01

    The authors are analyzing the thermal performance of microchannel heat sinks to extend their applied heat loads beyond coolant single-phase limits. This is the first investigation of boiling in the narrow (50-μm) microchannels having typically high-aspect-ratio (of order 10/1) flow cross-sections. The prescription of local, wall-coolant, interfacial, two-phase correlations first required development of a validated, approximate, thermal-model accounting for conjugate heat transfer. The strongest mechanism for heat transfer in two-phase microchannel flow was found to be saturated boiling in a channel region near the heated base. When this region dried out, burnout occurred, both in the computations and in the experiment

  19. Little low-power boiling never hurt anybody

    International Nuclear Information System (INIS)

    Dunn, F.E.

    1985-01-01

    Failures in the shutdown heat removal system of an LMFBR might lead to flow stagnation and coolant boiling in the reactor core. At normal operating power, the onset of sodium boiling will lead to film dryout and melting of the cladding and fuel within a few seconds. On the other hand, both calculations and currently available experimental data indicate that at heat fluxes corresponding to decay heat power levels, boiling leads to improved heat removal; and it limits the temperature rise in the fuel pins. Therefore, when setting safety criteria for decay heat removal systems, there is no reason to preclude sodium boiling per se because of heat removal considerations. As an example that illustrates the beneficial impact of coolant boiling, a case involving temporary loss of feedwater and staggered pump failures in a hypothetical, 1000-MWe loop-type reactor was run in the SASSYS-1 code

  20. Encyclopedia of two-phase heat transfer and flow IV modeling methodologies, boiling of CO₂, and micro-two-phase cooling

    CERN Document Server

    2018-01-01

    Set IV is a new addition to the previous Sets I, II and III. It contains 23 invited chapters from international specialists on the topics of numerical modeling of pulsating heat pipes and of slug flows with evaporation; lattice Boltzmann modeling of pool boiling; fundamentals of boiling in microchannels and microfin tubes, CO2 and nanofluids; testing and modeling of micro-two-phase cooling systems for electronics; and various special topics (flow separation in microfluidics, two-phase sensors, wetting of anisotropic surfaces, ultra-compact heat exchangers, etc.). The invited authors are leading university researchers and well-known engineers from leading corporate research laboratories (ABB, IBM, Nokia Bell Labs). Numerous "must read" chapters are also included here for the two-phase community. Set IV constitutes a "must have" engineering and research reference together with previous Sets I, II and III for thermal engineering researchers and practitioners.