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Sample records for developed subcooled boiling

  1. Cavitation, subcooled boiling and a measuring method developed at ENEA

    International Nuclear Information System (INIS)

    Tirelli, D.

    1988-01-01

    A brief description of cavitation and subcooled boiling is reported; their effects, measuring methods, operating limits and prescribed standards are described. The whole, to better clarify the usefulness and the importance of a measuring instrument developed at ENEA, to study the above phenomena

  2. 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)

  3. Onset of nucleate boiling and onset of fully developed subcooled boiling detection using pressure transducers signals spectral analysis

    International Nuclear Information System (INIS)

    Maprelian, Eduardo; Castro, Alvaro Alvim de; Ting, Daniel Kao Sun

    1999-01-01

    The experimental technique used for detection of subcooled boiling through analysis of the fluctuation contained in pressure transducers signals is presented. The experimental part of this work was conducted at the Institut fuer Kerntechnik und zertoerungsfreie Pruefverfahren von Hannover (IKPH, Germany) in a thermal-hydraulic circuit with one electrically heated rod with annular geometry test section. Piezo resistive pressure sensors are used for onset of nucleate boiling (ONB) and onset of fully developed boiling (OFDB) detection using spectral analysis/signal correlation techniques. Experimental results are interpreted by phenomenological analysis of these two points and compared with existing correlation. The results allows us to conclude that this technique is adequate for the detection and monitoring of the ONB and OFDB. (author)

  4. 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)

  5. Subcooled boiling effect on dissolved gases behaviour

    International Nuclear Information System (INIS)

    Zmitko, M.; Sinkule, J.; Linek, V.

    1999-01-01

    A model describing dissolved gasses (hydrogen, nitrogen) and ammonia behaviour in subcooled boiling conditions of WWERs was developed. Main objective of the study was to analyse conditions and mechanisms leading to formation of a zone with different concentration of dissolved gases, eg. a zone depleted in dissolved hydrogen in relation to the bulk of coolant. Both, an equilibrium and dynamic approaches were used to describe a depletion of the liquid surrounding a steam bubble in the gas components. The obtained results show that locally different water chemistry conditions can be met in the subcooled boiling conditions, especially, in the developed subcooled boiling regime. For example, a 70% hydrogen depletion in relation to the bulk of coolant takes about 1 ms and concerns a liquid layer of 1 μn surrounding the steam bubble. The locally different concentration of dissolved gases can influence physic-chemical and radiolytic processes in the reactor system, eg. Zr cladding corrosion, radioactivity transport and determination of the critical hydrogen concentration. (author)

  6. THE PREDICTION OF VOID VOLUME IN SUBCOOLED NUCLEATE POOL BOILING

    Energy Technology Data Exchange (ETDEWEB)

    Duke, E. E. [General Dynamics, San Diego, CA (United States)

    1963-11-15

    A three- step equation was developed that adequately describes the average volume of vapor occurring on a horizontal surface due to nucleate pool boiling of subcooled water. Since extensive bubble frequency data are lacking, the data of others were combined with experimental observations to make predictions of void volume at ambient pressure with various degrees of subcooling. (auth)

  7. Direct Numerical Simulation and Visualization of Subcooled Pool Boiling

    Directory of Open Access Journals (Sweden)

    Tomoaki Kunugi

    2014-01-01

    Full Text Available A direct numerical simulation of the boiling phenomena is one of the promising approaches in order to clarify their heat transfer characteristics and discuss the mechanism. During these decades, many DNS procedures have been developed according to the recent high performance computers and computational technologies. In this paper, the state of the art of direct numerical simulation of the pool boiling phenomena during mostly two decades is briefly summarized at first, and then the nonempirical boiling and condensation model proposed by the authors is introduced into the MARS (MultiInterface Advection and Reconstruction Solver developed by the authors. On the other hand, in order to clarify the boiling bubble behaviors under the subcooled conditions, the subcooled pool boiling experiments are also performed by using a high speed and high spatial resolution camera with a highly magnified telescope. Resulting from the numerical simulations of the subcooled pool boiling phenomena, the numerical results obtained by the MARS are validated by being compared to the experimental ones and the existing analytical solutions. The numerical results regarding the time evolution of the boiling bubble departure process under the subcooled conditions show a very good agreement with the experimental results. In conclusion, it can be said that the proposed nonempirical boiling and condensation model combined with the MARS has been validated.

  8. 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)

  9. Influence of subcooled boiling on out-of-phase oscillations in boiling water reactors

    International Nuclear Information System (INIS)

    Munoz-Cobo, J.L.; Chiva, S.; Escriva, A.

    2005-01-01

    In this paper, we develop a reduced order model with modal kinetics for the study of the dynamic behavior of boiling water reactors. This model includes the subcooled boiling in the lower part of the reactor channels. New additional equations have been obtained for the following dynamics magnitudes: the effective inception length for subcooled boiling, the average void fraction in the subcooled boiling region, the average void fraction in the bulk-boiling region, the mass fluxes at the boiling boundary and the channel exit, respectively, and so on. Each channel has three nodes, one of liquid, one with subcooled boiling, and one with bulk boiling. The reduced order model includes also a modal kinetics with the fundamental mode and the first subcritical one, and two channels representing both halves of the reactor core. Also, in this paper, we perform a detailed study of the way to calculate the feedback reactivity parameters. The model displays out-of-phase oscillations when enough feedback gain is provided. The feedback gain that is necessary to self-sustain these oscillations is approximately one-half the gain that is needed when the subcooled boiling node is not included

  10. Subcooled boiling heat transfer on a finned surface

    International Nuclear Information System (INIS)

    Kowalski, J.E.; Tran, V.T.; Mills, P.J.

    1992-01-01

    Experimental and numerical studies have been performed to determine the heat transfer coefficients from a finned cylindrical surface to subcooled boiling water. The heat transfer rates were measured in an annular test section consisting of an electrically heated fuel element simulator (FES) with eight longitudinal, rectangular fins enclosed in a glass tube. A two-dimensional finite-element heat transfer model using the Galerkin method was employed to determine the heat transfer coefficients along the periphery of the FES surface. An empirical correlation was developed to predict the heat transfer coefficients during subcooled boiling. The correlation agrees well with the measured data. (6 figures) (Author)

  11. 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.)

  12. 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)

  13. Calculation of Steam Volume Fraction in Subcooled Boiling

    Energy Technology Data Exchange (ETDEWEB)

    Rouhani, S Z

    1967-06-15

    An analysis of subcooled boiling is presented. It is assumed that heat is removed by vapor generation, heating of the liquid that replaces the detached bubbles, and to some extent by single phase heat transfer. Two regions of subcooled boiling are considered and a criterion is provided for obtaining the limiting value of subcooling between the two regions. Condensation of vapor in the subcooled liquid is analysed and the relative velocity of vapor with respect to the liquid is neglected in these regions. The theoretical arguments result in some equations for the calculation of steam volume fraction and true liquid subcooling.

  14. Development of Bubble Lift-off Diameter Model for Subcooled Boiling Flows

    International Nuclear Information System (INIS)

    Hoang, Nhan Hien; Chu, Incheol; Song Chulhwa; Euh, Dongjin

    2014-01-01

    A lot of models and correlations for predicting the bubble departure/lift-off diameter are available in the literature. Most of them were developed based on a hydrodynamic principle, which balances forces acting on a bubble at the departure/lift-off point. One difficulty of these models is lack of essential information, such as bubble front velocity, liquid velocity, or relative velocity, to estimate the active force elements. Hence, the lift-off bubble diameter predicted by these hydrodynamic-controlled models may be suffered a large uncertainty. In contract to the hydrodynamic approach, there are few models developed based on the heat transfer aspect. By balancing the heat conducted through a microlayer underneath a bubble with the heat taken away by condensation at the upper part of the bubble, Unal derived a heat-controlled model of the bubble lift-off diameter. This model did not consider the role of superheat liquid layer surrounding the bubble as well as the effect of liquid properties on the heat transfer process. Beside these two approaches, several empirical correlations have been proposed based on dimensionless analyses for measured experimental databases. The application of these correlations to different experiments conditions is, of course, questionable because of the lack of physical bases. Regarding the heat transfer accompanied by a vapor bubble, four involved heat transfer regions surrounding this bubble can be defined as in Fig. 1. These are dry region, microlayer, superheated liquid layer (SpLL) and subcooled liquid layer (SbLL). The existing of the microlayer is confirmed by experiments, and it is considered to be very effective in the heat transfer. Sernas and Hoper defined five types of the microlayer and indicated that the microlayer acting as a very thick liquid layer gives a best prediction for the bubble growth. However, beside the microlayer, the SpLL might play an important role in the heat transfer if its effective heat transfer area

  15. 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

  16. 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.

  17. 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)

  18. 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)

  19. 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)

  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. 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)

  2. 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)

  3. 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.)

  4. 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

  5. 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)

  6. 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)

  7. 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

  8. 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)

  9. 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.

  10. Thermal interaction effect on nucleation site distribution in subcooled boiling

    International Nuclear Information System (INIS)

    Zou, Ling; Jones, Barclay

    2012-01-01

    An experimental work on subcooled boiling of refrigerant, R134a, to examine nucleation site distributions on both copper and stainless steel heating surfaces was performed. In order to obtain high fidelity active nucleation site density and distribution data, a high-speed digital camera was utilized to record bubble emission images from a view normal to heating surfaces. Statistical analyses on nucleation site data were done and their statistical distributions were obtained. Those experimentally observed nucleation site distributions were compared to the random spatial Poisson distribution. The comparisons showed that, rather than purely random, active nucleation site distributions on boiling surfaces are relatively more uniform. Experimental results also showed that on the copper heating surface, nucleation site distributions are slightly more uniform than on the stainless steel surface. This was concluded as the results of thermal interactions between nucleation sites with different solid thermal conductivities. A two dimensional thermal interaction model was then developed to quantitatively examine the thermal interactions between nucleation sites. The results give a reasonable explanation to the experimental observation on nucleation site distributions.

  11. Radiolysis effects in sub-cooled nucleate boiling

    International Nuclear Information System (INIS)

    Dickinson, S.; Henshaw, J.; Tuson, A.; Sims, H.E.

    2002-01-01

    A hydrogen depleted region may form in the water during bubble formation when boiling occurs in a PWR. This would arise from stripping of gases into the steam phase. The depleted water may then become oxidising due to radiolysis forming H 2 O 2 . The presence of radiolytic oxidising conditions is one of the mechanisms proposed to explain deposits formed in Axial Offset Anomalies. This work describes a model that has been developed to examine this behaviour. The model deals with bubble growth and material transport as well as the radiolysis chemistry. The model simulates diffusion of species through the gas/liquid boundary layer. The appropriate mass conservation equations for this problem are described and the results of their numerical solution discussed. This model indicates the importance of the assumed boundary conditions on the results of the calculations. These boundary conditions are discussed in detail and the most appropriate ones for the actual reactor situation are outlined. The conclusion of this modelling study is that at normal PWR operating conditions of 40 cc H 2 (STP) kg -1 it is unlikely that radiolysis in a subcooled boiling region would be important. The situation is more ambiguous at the 1 to 5 cc H 2 (STP) kg -1 range. (author)

  12. Radiolysis effects in sub-cooled nucleate boiling

    Energy Technology Data Exchange (ETDEWEB)

    Dickinson, S.; Henshaw, J.; Tuson, A.; Sims, H.E. [AEA Technology (United Kingdom)

    2002-07-01

    A hydrogen depleted region may form in the water during bubble formation when boiling occurs in a PWR. This would arise from stripping of gases into the steam phase. The depleted water may then become oxidising due to radiolysis forming H{sub 2}O{sub 2}. The presence of radiolytic oxidising conditions is one of the mechanisms proposed to explain deposits formed in Axial Offset Anomalies. This work describes a model that has been developed to examine this behaviour. The model deals with bubble growth and material transport as well as the radiolysis chemistry. The model simulates diffusion of species through the gas/liquid boundary layer. The appropriate mass conservation equations for this problem are described and the results of their numerical solution discussed. This model indicates the importance of the assumed boundary conditions on the results of the calculations. These boundary conditions are discussed in detail and the most appropriate ones for the actual reactor situation are outlined. The conclusion of this modelling study is that at normal PWR operating conditions of 40 cc H{sub 2} (STP) kg{sup -1} it is unlikely that radiolysis in a subcooled boiling region would be important. The situation is more ambiguous at the 1 to 5 cc H{sub 2} (STP) kg{sup -1} range. (author)

  13. 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

  14. Transient pool boiling heat transfer due to increasing heat inputs in subcooled water at high pressures

    Energy Technology Data Exchange (ETDEWEB)

    Fukuda, K. [Kobe Univ. of Mercantile Marine (Japan); Shiotsu, M.; Sakurai, A. [Kyoto Univ. (Japan)

    1995-09-01

    Understanding of transient boiling phenomenon caused by increasing heat inputs in subcooled water at high pressures is necessary to predict correctly a severe accident due to a power burst in a water-cooled nuclear reactor. Transient maximum heat fluxes, q{sub max}, on a 1.2 mm diameter horizontal cylinder in a pool of saturated and subcooled water for exponential heat inputs, q{sub o}e{sup t/T}, with periods, {tau}, ranging from about 2 ms to 20 s at pressures from atmospheric up to 2063 kPa for water subcoolings from 0 to about 80 K were measured to obtain the extended data base to investigate the effect of high subcoolings on steady-state and transient maximum heat fluxes, q{sub max}. Two main mechanisms of q{sub max} exist depending on the exponential periods at low subcoolings. One is due to the time lag of the hydrodynamic instability which starts at steady-state maximum heat flux on fully developed nucleate boiling (FDNB), and the other is due to the heterogenous spontaneous nucleations (HSN) in flooded cavities which coexist with vapor bubbles growing up from active cavities. The shortest period corresponding to the maximum q{sub max} for long period range belonging to the former mechanism becomes longer and the q{sub max}mechanism for long period range shifts to that due the HSN on FDNB with the increase of subcooling and pressure. The longest period corresponding to the minimum q{sub max} for the short period range belonging to the latter mechanism becomes shorter with the increase in saturated pressure. On the contrary, the longest period becomes longer with the increase in subcooling at high pressures. Correlations for steady-state and transient maximum heat fluxes were presented for a wide range of pressure and subcooling.

  15. Transient pool boiling heat transfer due to increasing heat inputs in subcooled water at high pressures

    International Nuclear Information System (INIS)

    Fukuda, K.; Shiotsu, M.; Sakurai, A.

    1995-01-01

    Understanding of transient boiling phenomenon caused by increasing heat inputs in subcooled water at high pressures is necessary to predict correctly a severe accident due to a power burst in a water-cooled nuclear reactor. Transient maximum heat fluxes, q max , on a 1.2 mm diameter horizontal cylinder in a pool of saturated and subcooled water for exponential heat inputs, q o e t/T , with periods, τ, ranging from about 2 ms to 20 s at pressures from atmospheric up to 2063 kPa for water subcoolings from 0 to about 80 K were measured to obtain the extended data base to investigate the effect of high subcoolings on steady-state and transient maximum heat fluxes, q max . Two main mechanisms of q max exist depending on the exponential periods at low subcoolings. One is due to the time lag of the hydrodynamic instability which starts at steady-state maximum heat flux on fully developed nucleate boiling (FDNB), and the other is due to the heterogenous spontaneous nucleations (HSN) in flooded cavities which coexist with vapor bubbles growing up from active cavities. The shortest period corresponding to the maximum q max for long period range belonging to the former mechanism becomes longer and the q max mechanism for long period range shifts to that due the HSN on FDNB with the increase of subcooling and pressure. The longest period corresponding to the minimum q max for the short period range belonging to the latter mechanism becomes shorter with the increase in saturated pressure. On the contrary, the longest period becomes longer with the increase in subcooling at high pressures. Correlations for steady-state and transient maximum heat fluxes were presented for a wide range of pressure and subcooling

  16. Study of the initiation of subcooled boiling during power transients

    International Nuclear Information System (INIS)

    VanVleet, R.J.

    1985-01-01

    An experimental investigation of boiling initiation during power transients has been conducted for horizontal-cylinder heating elements in degassed distilled water. Platinum elements, 0.127 and 0.250 mm in diameter, were internally heated electrically at a controlled superficial heat flux (power applied divided by surface area) increasing linearly with time at rates of 0.035 and 0.35 MW/m 2 s and corresponding test durations of 20 and 2 seconds. Tests were carried out at saturation temperatures from 100 to 195 0 C with bulk fluid subcooling from 0 to 30 K. During the course of a power transient, element temperature and superficial heat flux were measured electrically and the boiling initiation time was determined optically. It was found that the conditions for boiling initiation depended strongly on the pressure-temperature history of the heating element and surround fluid prior to the transient. Boiling initiation times were found to agree qualitatively with predictions of a model based on the contact-angle hysteresis concept. Brief prepressurization prior to a transient was found to increase dramatically the temperature and heat flux required for boiling initiation because of deactivation of boiling initiation sites. However, sites were re-activated during the transient and, in subsequent tests without prepressurization, no elevation in boiling initiation conditions was observed and results were in quantitative agreement with predictions of the model

  17. 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)

  18. 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)

  19. Surface wettability and subcooling on nucleate pool boiling heat transfer

    Science.gov (United States)

    Suroto, Bambang Joko; Kohno, Masamichi; Takata, Yasuyuki

    2018-02-01

    The effect of varying surface wettabilities and subcooling on nucleate pool boiling heat transfer at intermediate heat flux has been examined and investigated. The experiments were performed using pure water as the working fluid and subcooling ranging from 0, 5 and 10 K, respectively. The three types of heat transfer block were used that are bare surface/hydrophilic (polished copper), superhydrophilic/TiO2-coated on copper and hydrophobic/PTFE surface. The experimental results will be examined by the existing model. The results show that the heat transfer performance of surfaces with PTFE coating is better at low heat flux. While for an intermediate heat flux, superhydrophilic surface (TiO2) is superior compared to hydrophilic and hydrophobic surfaces. It is observed that the heat transfer performance is decreasing when the sub cooling degree is increased.

  20. 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

  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. 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.

  3. Analysis of subcooled boiling with the two-fluid particle interaction method

    International Nuclear Information System (INIS)

    Shirakawa, Noriyuki; Horie, Hideki; Yamamoto, Yuichi; Tsunoyama, Shigeaki

    2003-01-01

    A particle interaction method called MPS (the Moving Particle Semi-implicit method), which formulates the differential operators in Navier-Stokes' equation as interactions between particles characterized by a kernel function, has been developed in recent years. We have extended this method to a two-fluid system with a potential-type surface tension in order to analyze the two-phase flow without experimental correlation. This extended method (Two-Fluid MPS: TF-MPS) was successfully applied to a subcooled boiling experiment. The most important element in any effective subcooled boiling model is to be able to accurately calculate where significant void fraction appears, that is, the location of the void departure point. The location of the initial void ejection into the subcooled liquid core can be determined fairly well experimentally and conventionally is given in terms of a critical subcooling. We investigated the relation between Stanton and Peclet numbers at the void departure point in the calculated results with TF-MPS method, varying the inlet water velocity to change Peclet number. (author)

  4. Dependence of bubble behavior in subcooled boiling on surface wettability

    International Nuclear Information System (INIS)

    Harada, Takahiro; Nagakura, Hiroshi; Okawa, Tomio

    2010-01-01

    This paper presents the results of visualization experiments that were carried out to investigate the dynamics of vapor bubbles generated in water pool boiling. In the experiments, vapor bubbles were generated on a vertical circular surface of a copper block containing nine cartridge heaters, and the contact angle of the heated surface was used as a main experimental parameter. The experiments were performed under subcooled as well as nearly saturated conditions. To enable clear observation of individual bubbles with a high speed camera, the heat flux was kept low enough to eliminate significant overlapping of bubbles. When the contact angle was small, the bubbles were lifted-off the vertical heated surface within a short period of time after the nucleation. On the other hand, when the contact angle was large, they slid up the vertical surface for a long distance. When bubbles were lifted-off the heated surface in subcooled liquid, bubble life-time was significantly shortened since bubbles collapsed rapidly due to condensation. It was shown that this distinct difference in bubble dynamics could be attributed to the effects of surface tension force.

  5. 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)

  6. 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)

  7. Use of ultrasonic waves in sub-cooled boiling

    International Nuclear Information System (INIS)

    Bartoli, Carlo; Baffigi, Federica

    2012-01-01

    %. - Highlights: ► We investigated experimentally the heat transfer from a circular cylinder to distilled water in presence of ultrasounds. ► The cylinder was heated up to reach particular temperatures, which are critical for last generation electronic components. ► All tests are carried on in sub-cooled boiling conditions, at atmospheric pressure. ► The tests reported an increase in the heat transfer up to 57%, in the optimal conditions. ► The sub-cooling degree is the only parameter that has the most significant influence on the heat transfer enhancement.

  8. 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)

  9. 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

  10. 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.)

  11. 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.].

  12. Numerical simulation of bubble behavior in subcooled flow boiling under velocity and temperature gradient

    International Nuclear Information System (INIS)

    Bahreini, Mohammad; Ramiar, Abas; Ranjbar, Ali Akbar

    2015-01-01

    Highlights: • Condensing bubble is numerically investigated using VOF model in OpenFOAM package. • Bubble mass reduces as it goes through condensation and achieves higher velocities. • At a certain time the slope of changing bubble diameter with time, varies suddenly. • Larger bubbles experience more lateral migration to higher velocity regions. • Bubbles migrate back to a lower velocity region for higher liquid subcooling rates. - Abstract: In this paper, numerical simulation of the bubble condensation in the subcooled boiling flow is performed. The interface between two-phase is tracked via the volume of fluid (VOF) method with continuous surface force (CSF) model, implemented in the open source OpenFOAM CFD package. In order to simulate the condensing bubble with the OpenFOAM code, the original energy equation and mass transfer model for phase change have been modified and a new solver is developed. The Newtonian flow is solved using the finite volume scheme based on the pressure implicit with splitting of operators (PISO) algorithm. Comparison of the simulation results with previous experimental data revealed that the model predicted well the behavior of the actual condensing bubble. The bubble lifetime is almost proportional to bubble initial size and is prolonged by increasing the system pressure. In addition, the initial bubble size, subcooling of liquid and velocity gradient play an important role in the bubble deformation behavior. Velocity gradient makes the bubble move to the higher velocity region and the subcooling rate makes it to move back to the lower velocity region.

  13. Numerical simulation of bubble behavior in subcooled flow boiling under velocity and temperature gradient

    Energy Technology Data Exchange (ETDEWEB)

    Bahreini, Mohammad, E-mail: m.bahreini1990@gmail.com; Ramiar, Abas, E-mail: aramiar@nit.ac.ir; Ranjbar, Ali Akbar, E-mail: ranjbar@nit.ac.ir

    2015-11-15

    Highlights: • Condensing bubble is numerically investigated using VOF model in OpenFOAM package. • Bubble mass reduces as it goes through condensation and achieves higher velocities. • At a certain time the slope of changing bubble diameter with time, varies suddenly. • Larger bubbles experience more lateral migration to higher velocity regions. • Bubbles migrate back to a lower velocity region for higher liquid subcooling rates. - Abstract: In this paper, numerical simulation of the bubble condensation in the subcooled boiling flow is performed. The interface between two-phase is tracked via the volume of fluid (VOF) method with continuous surface force (CSF) model, implemented in the open source OpenFOAM CFD package. In order to simulate the condensing bubble with the OpenFOAM code, the original energy equation and mass transfer model for phase change have been modified and a new solver is developed. The Newtonian flow is solved using the finite volume scheme based on the pressure implicit with splitting of operators (PISO) algorithm. Comparison of the simulation results with previous experimental data revealed that the model predicted well the behavior of the actual condensing bubble. The bubble lifetime is almost proportional to bubble initial size and is prolonged by increasing the system pressure. In addition, the initial bubble size, subcooling of liquid and velocity gradient play an important role in the bubble deformation behavior. Velocity gradient makes the bubble move to the higher velocity region and the subcooling rate makes it to move back to the lower velocity region.

  14. 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.

  15. Measurement of subcooled boiling pressure drop and local heat transfer coefficient in horizontal tube under LPLF conditions

    International Nuclear Information System (INIS)

    Baburajan, P.K.; Bisht, G.S.; Gupta, S.K.; Prabhu, S.V.

    2013-01-01

    Highlights: ► Measured subcooled boiling pressure drop and local heat transfer coefficient in horizontal tubes. ► Infra-red thermal imaging is used for wall temperature measurement. ► Developed correlations for pressure drop and local heat transfer coefficient. -- Abstract: Horizontal flow is commonly encountered in boiler tubes, refrigerating equipments and nuclear reactor fuel channels of pressurized heavy water reactors (PHWR). Study of horizontal flow under low pressure and low flow (LPLF) conditions is important in understanding the nuclear core behavior during situations like LOCA (loss of coolant accidents). In the present work, local heat transfer coefficient and pressure drop are measured in a horizontal tube under LPLF conditions of subcooled boiling. Geometrical parameters covered in this study are diameter (5.5 mm, 7.5 mm and 9.5 mm) and length (550 mm, 750 mm and 1000 mm). The operating parameters varied are mass flux (450–935 kg/m 2 s) and inlet subcooling (29 °C, 50 °C and 70 °C). Infra-red thermography is used for the measurement of local wall temperature to estimate the heat transfer coefficient in single phase and two phase flows with water as the working medium at atmospheric pressure. Correlation for single phase diabatic pressure drop ratio (diabatic to adiabatic) as a function of viscosity ratio (wall temperature to fluid temperature) is presented. Correlation for pressure drop under subcooled boiling conditions as a function of Boiling number (Bo) and Jakob number (Ja) is obtained. Correlation for single phase heat transfer coefficient in the thermal developing region is presented as a function of Reynolds number (Re), Prandtl number (Pr) and z/d (ratio of axial length of the test section to diameter). Correlation for two phase heat transfer coefficient under subcooled boiling condition is developed as a function of boiling number (Bo), Jakob number (Ja) and Prandtl number (Pr)

  16. 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.)

  17. 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

  18. 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

  19. 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

  20. 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.

  1. 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)

  2. 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

  3. Drift flux formulation of a boiling water reactor channel with subcooled boiling

    International Nuclear Information System (INIS)

    Elias, E.E.; Shak, D.P.; May, R.S.

    1987-01-01

    The channel formulation used in the BWR module of the Modular Modeling System MMS-02 is presented. The purpose of channel model is to accurately predict the transient response of the enthalpy void and flow rate. Accurate prediction of the two-phase enthalpy, and void fraction distributions along the channel is important since they are key input parameters to the neutronic model, and have direct effect on the core and overall reactor response. In order to model the channel response correctly, the physical phenomena had to be realistically represented. The model accounts for subcooled boiling and slip through the use of an empirical subcooled void-quality model. Simplifying assumptions are made so that only one differential equation, the energy equation, is integrated along the channel. A consistent use of semi-empirical correlations enables a complete representation of the channel flow and void fraction with the bulk enthalpy as the only state variable. The differential equation and the constitutive relations of this two-phase flow model are presented. Several numerical examples are given, and finally, come conclusions are presented

  4. 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

  5. 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)

  6. Experimental investigation on subcooled boiling heat transfer in a vertical double-face heated narrow annulus

    International Nuclear Information System (INIS)

    Yan Mingyu; Qiu Suizheng; Jia Dounan

    2005-01-01

    Experimental investigation on the subcooled boiling heat transfer was carried out in a vertical up-flow double narrow annulus with 1.5 mm gap. The working fluid is deionized water. The ranges of parameters as follows: pressure 0.84-6.09 MPa, mass flux 41.9-300.2 kg/(m 2 ·s), heat flux 2.61-114.41kW/m 2 . An empiric correlation used to predict the heat transfer of subcooled boiling in narrow annulus is induced from the experimental data. (author)

  7. Improvement of the RELAP5 subcooled boiling model for low pressure conditions

    International Nuclear Information System (INIS)

    Koncar, B.; Mavko, B.

    2000-01-01

    The RELAP5/MOD3.2.2 Gamma code was assessed against low pressure subcooled boiling experiments performed by Zeitoun and Shoukri [1] in a vertical annulus. The predictions of subcooled boiling bubbly flow showed that the present version of the RELAP5 code underestimates the void fraction growth along the tube. To improve the void fraction prediction at low pressure conditions a set of model changes is proposed, which includes modifications of bubbly-slug transition criterion, drift-flux model, interphase heat transfer coefficient and wall evaporation modeling. The improved experiment predictions with the modified RELAP5 code are presented and analysed. (author)

  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. 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. Measurement of wetted area fraction in subcooled pool boiling of water using infrared thermography

    International Nuclear Information System (INIS)

    Kim, Hyungdae; Park, Youngjae; Buongiorno, Jacopo

    2013-01-01

    The wetted area fraction in subcooled pool boiling of water at atmospheric pressure is measured using the DEPIcT (DEtection of Phase by Infrared Thermography) technique. DEPIcT exploits the contrast in infrared (IR) light emissions between wet and dry areas on the surface of an IR-transparent heater to visualize the instantaneous distribution of the liquid and gas phases in contact with the heater surface. In this paper time-averaged wetted area fraction data in nucleate boiling are reported as functions of heat flux (from 30% up to 100% of the Critical Heat Flux) and subcooling (ΔT sub = 0, 5, 10, 30 and 50 °C). The results show that the wetted area fraction monotonically decreases with increasing heat flux and increases with increasing subcooling: both trends are expected. The range of time-averaged wetted area fractions is from 90%, at low heat flux and high subcooling, to 50% at high heat flux (right before CHF) and low subcooling. It is also shown that the dry areas are periodically rewetted by liquid sloshing on the surface at any subcooling and heat flux; however, the dry areas expand irreversibly at CHF

  11. Development and validation of a new solver based on the interfacial area transport equation for the numerical simulation of sub-cooled boiling with OpenFOAM CFD code for nuclear safety applications

    Energy Technology Data Exchange (ETDEWEB)

    Alali, Abdullah

    2014-02-21

    The one-group interfacial area transport equation has been coupled to a wall heat flux partitioning model in the framework of two-phase Eulerian approach using the OpenFOAM CFD code for better prediction of subcooled boiling phenomena which is essential for safety analysis of nuclear reactors. The interfacial area transport equation has been modified to include the effect of bubble nucleation at the wall and condensation by subcooled liquid in the bulk that governs the non-uniform bubble size distribution.

  12. Development and validation of a new solver based on the interfacial area transport equation for the numerical simulation of sub-cooled boiling with OpenFOAM CFD code for nuclear safety applications

    International Nuclear Information System (INIS)

    Alali, Abdullah

    2014-01-01

    The one-group interfacial area transport equation has been coupled to a wall heat flux partitioning model in the framework of two-phase Eulerian approach using the OpenFOAM CFD code for better prediction of subcooled boiling phenomena which is essential for safety analysis of nuclear reactors. The interfacial area transport equation has been modified to include the effect of bubble nucleation at the wall and condensation by subcooled liquid in the bulk that governs the non-uniform bubble size distribution.

  13. Peak pool boiling heat flux from horizontal cylinders in subcooled liquids

    International Nuclear Information System (INIS)

    Elkassabgi, Y.

    1986-01-01

    The peak pool boiling heat flux is observed on horizontal cylindrical heaters in acetone, Freon-113, methanol, and isopropanol over ranges of subcooling from zero to 120 0 C. Photographs, and the data themselves, reveal that there are three distinct burnout mechanisms at different levels of subcooling. Three interpretive models provide the basis for accurate correlations of the present data, and data from the literature, in each of the three regimes. Burnout is dictated by condensation on the walls of the vapor jets and columns at low subcooling. In the intermediate regime, burnout is limited by natural convection which becomes very effective as vapor near the heater reduces boundary layer resistance. Burnout in the high-subcooling regime is independent of the level of subcoooling and is limited by the process of molecular effusion

  14. 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.

  15. 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

  16. Static flow instability in subcooled flow boiling in parallel channels

    International Nuclear Information System (INIS)

    Siman-Tov, M.; Felde, D.K.; McDuffee, J.L.; Yoder, G.L. Jr.

    1995-01-01

    A series of tests for static flow instability or flow excursion (FE) at conditions applicable to the proposed Advanced Neutron Source reactor was completed in parallel rectangular channels configuration with light water flowing vertically upward at very high velocities. True critical heat flux experiments under similar conditions were also conducted. The FE data reported in this study considerably extend the velocity range of data presently available worldwide. Out of the three correlations compared, the Saha and Zuber correlation had the best fit with the data. However, a modification was necessary to take into account the demonstrated dependence of the Stanton (St) and Nusselt (Nu) numbers on subcooling levels, especially in the low subcooling regime

  17. 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)

  18. 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

  19. Steady-state subcooled nucleate boiling on a downward facing hemispherical surface

    International Nuclear Information System (INIS)

    Haddad, K.H.; Cheung, F.B.

    1996-01-01

    Steady-state nucleate boiling heat transfer experiments in saturated and subcooled water were conducted. The heating surface was a 0.305 m hemispherical aluminum vessel heated from the inside with water boiling on the outside. It was found that subcooling had very little effect on the nucleate boiling curve in the high heat flux regime where latent heat transport dominated. On the other hand, a relatively large effect of subcooling was observed in the low heat flux regime where sensible heat transport was important. Photographic records of the boiling phenomenon and the bubble dynamics indicated that in the high heat flux regime, boiling in the bottom center region of the vessel was cyclic in nature with a liquid heating phase, a bubble nucleation and growth phase, a bubble coalescence phase, and a large vapor mass ejection phase. At the same heat flux level, the size of the vapor masses was found to decrease from the bottom center toward the upper edge of the vessel, which was consistent with the observed increase in the critical heat flux in the flow direction along the curved heating surface

  20. 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

  1. 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)

  2. 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

  3. 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

  4. 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

  5. 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

  6. 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

  7. 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.

  8. 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

  9. 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)

  10. 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)

  11. Phenomenology and thermo-hydraulic stability of the CAREM-25 reactor: Evaluation of subcooled boiling effect

    International Nuclear Information System (INIS)

    Acuna, F.M.; Marcel, C.P.; Zanocco, P.G.; Delmastro, D.F.

    2012-01-01

    In this article the phenomenology present in self/pressurized, integral, natural circulation, low thermodynamic quality nuclear reactors similar to CAREM-25 is investigated. In particular, analytical relations for the mass flow rate, the core mean enthalpy and the location of the two phase boundary are derived in terms of the so-called natural variables of the system: the nuclear power, the condensation power and the system pressure. In addition, some consequences of the flashing phenomenon in the reactor thermal-hydraulics are discussed emphasizing those affecting the reactor stability. The reactor stability performance was studied by using the HUARPE code which is a low diffusive code. The stability results obtained by neglecting the subcooled effect in the system are presented in the so-called the stability maps in which the results are presented for a wide range of conditions. The stability effect caused by the presence of subcooled boiling in the reactor core was also examined. In order to investigate such a consequence, the code was slightly modified such that the predicted vapor profile in the hot leg is similar to that estimated by RELAP system code at steady state conditions. The simple implemented algorithm allows varying a free parameter with which a broad number of cases can be studied. This is important since the subcooled boiling predictions generally have large uncertainties and therefore to cover a large number of situations is desired to derive confident conclusions. The results show the existence of vapor created by means of subcooled boiling enhances the system stability for a wide range of conditions. For this reason from this preliminary investigation, it is concluded neglecting the subcooled effect in CAREM-25 stability studies is a conservative criterion (author))

  12. 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)

  13. Effect of Dissolved gas on bubble behavior of subcooled boiling in narrow channel

    International Nuclear Information System (INIS)

    Li Shaodan; Tan Sichao; Xu Chao; Gao Puzhen; Xu Jianjun

    2013-01-01

    An experimental investigation was performed to study the effect of dissolved gas on bubble behavior in narrow rectangular channel under subcooled boiling condition. A high-speed digital video camera was applied to capture the dynamics of the bubble with or without dissolved gas in a narrow rectangular channel. It is found that the dissolved gas has great influence on bubble behavior in subcooled boiling condition. The dissolved gas slows down the rate of bubble growth and condensation and makes the variation of the bubble diameter present some oscillation characteristics. This phenomenon was discussed in the view of the vapor evaporation and condensation. The existence of the dissolved gas can facilitate the survival of the bubble and promote the aggregation of bubbles, and enhence heat transfer enhancement in some ways. (authors)

  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. Point of net vapor generation and vapor void fraction in subcooled boiling

    International Nuclear Information System (INIS)

    Saha, P.; Zuber, N.

    1974-01-01

    An analysis is presented directed at predicting the point of net vapor generation and vapor void fraction in subcooled boiling. It is shown that the point of net vapor generation depends upon local conditions--thermal and fluid dynamic. Thus, at low mass flow rates the net vapor generation is determined by thermal conditions, whereas at high mass flow rates the phenomenon is hydrodynamically controlled. Simple criteria are derived which can be used to predict these local conditions for net vapor generation. These criteria are used to determine the vapor void fraction is subcooled boiling. Comparison between the results predicted by this analysis and experimental data presently available shows good agreement for wide range of operating conditions, fluids and geometries. (U.S.)

  16. 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)

  17. 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

  18. 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

  19. 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)

  20. 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)

  1. 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)

  2. 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

  3. Subcooled boiling heat transfer correlation to calculate the effects of dissolved gas in a liquid

    International Nuclear Information System (INIS)

    Zarkasi, Amin S.; Chao, W.W.; Kunze, Jay F.

    2004-01-01

    The water coolant in most operating power reactor systems is kept free of dissolved gas, so as to minimize corrosion. However, in most research reactors, which operate at temperatures below 70 deg. C, and between 1 and 5 atm. pressure, the dissolved gas remains present in the water coolant system during operation. This dissolved gas can have a significant effect during accident conditions (i.e. a LOCA), when the fluid quickly reaches boiling, coincident with flow stagnation and subsequent flow reversal. A benchmark experiment was conducted, with an electrically heated, closed loop channel, modeling a research reactor fuel coolant channels (2 mm thick). The results showed 'boiling (bubble) noise' occurring before wall temperatures reached saturation, and a significant increase (up to 50%) in the heat transfer coefficient in the subcooled boiling region when in the presence of dissolved gas, compared to degassed water. Since power reactors do not involve dissolved gas, the RELAP safety analysis code does not include any provisions for the effect of dissolved gas on heat transfer. In this work, the effects of the dissolved gas are evaluated for inclusion in the RELAP code, including provision for initiating 'nucleate boiling' at a lower temperature, and a provision for enhancing the heat transfer coefficient during the subcooled boiling region. Instead of relying on Chen's correlation alone, a modification of the superposition method of Bjorge was adopted. (author)

  4. Subcooled Pool Boiling from Two Tubes of 6 Degree Included Angle in Vertical Alignment

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Myeong-Gie [Andong National University, Andong (Korea, Republic of)

    2015-05-15

    One of the major issues in the design of a heat exchanger is the heat transfer in a tube bundle. The passive condensation heat exchanger (PCHX) adopted in APR+ has U-type tube. The PCHX is submerged in the passive condensation cooling tank (PCCT). The heat exchanging tubes are in vertical alignment and inclined at 3 degrees to prevent water hammer as shown in Fig. 1. For the cases, the upper tube is affected by the lower tube. Therefore, the results for a single tube are not applicable to the design of the PCHX. However, the passive heat exchangers are submerged in the subcooled water under atmospheric pressure. The water temperature in the PCCT rises according to the PAFS actuation and reaches the saturation temperature after more than 2.5 hours. Since this period is very important to maintain reactor integrity, the exact evaluation of heat transfer on the tube bundle is indispensable. Although an experimental study on both subcooled and saturated pool boiling of water was performed to obtain local heat transfer coefficients on a 3 degree inclined tube at atmospheric pressure by Kang, no previous results were treating the bundle effect in the subcooled liquid. The heat transfer on the upper tube is enhanced compared with the single tube. The enhancement of the heat transfer on the upper tube is estimated by the bundle effect ( h{sub r} ). It is defined as the ratio of the heat transfer coefficient ( h{sub b} ) for an upper tube in a bundle with lower tubes activated to that for the same tube activated alone in the bundle. The upper tube within a tube bundle can significantly increase nucleated boiling heat transfer compared to the lower tubes at moderate heat fluxes. Summarizing the published results, it is still necessary to identify effects of liquid subcooling on inclined tubes for application to the PCHX design. Therefore, the present study is aimed to study the variations of pool boiling heat transfer on a tube bundle having a 6 degree included angle in

  5. Experimental measurement of the interfacial heat transfer coefficients of subcooled flow boiling using micro-thermocouple and double directional images

    International Nuclear Information System (INIS)

    Seong-Jin Kim; Goon-Cherl Park

    2005-01-01

    Full text of publication follows: Models or correlations for phase interface are needed to analyze the multi-phase flow. Interfacial heat transfer coefficients are important to constitute energy equation of multi-phase flow, specially. In subcooled boiling flow, bubble condensation at the bubble-liquid interface is a major mechanism of heat transfer within bulk subcooled liquid. Bubble collapse rates and temperatures of each phase are needed to determine the interfacial heat transfer coefficient for bubble condensation. Bubble collapse rates were calculated through image processing in single direction, generally. And in case of liquid bulk temperature, which has been obtained by general temperature sensor such as thermocouple, was used. However, multi-directional images are needed to analyze images due to limitations of single directional image processing. Also, temperature sensor, which has a fast response time, must be used to obtain more accurate interfacial heat transfer coefficient. Low pressure subcooled water flow experiments using micro-thermocouple and double directional image processing with mirrors were conducted to investigate bubble condensation phenomena and to modify interfacial heat transfer correlation. Experiments were performed in a vertical subcooled boiling flow of a rectangular channel. Bubble condensing traces with respect to time were recorded by high speed camera in double direction and bubble collapse rates were calculated by processing recorded digital images. Temperatures were measured by micro-thermocouple, which is a K-type with a 12.7 μm diameter. The liquid temperature was estimated by the developed algorithm to discriminate phases and find each phase temperature in the measured temperature including both liquid and bubble temperature. The interfacial heat transfer coefficient for bubble condensation was calculated from the bubble collapse rates and the estimated liquid temperature, and its correlation was modified. The modified

  6. 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)

  7. 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.

  8. 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

  9. A theoretical prediction of critical heat flux in subcooled pool boiling during power transients

    International Nuclear Information System (INIS)

    Pasamehmetoglu, K.O.; Nelson, R.A.; Gunnerson, F.S.

    1988-01-01

    Understanding and predicting critical heat flux (CHF) behavior during steady-state and transient conditions are of fundamenatal interest in the design, operation, safety of boiling and two-phase flow devices. This paper discusses the results of a comprehensive theoretical study made specifically to model transient CHF behavior in subcooled pool boiling. This study is based upon a simplified steady-state CHF model in terms of the vapor mass growth period. The results obtained from this theory indicate favorable agreement with the experimental data from cylindrical heaters with small radii. The statistical nature of the vapor mass behavior in transient boiling also is considered and upper and lower limits for the current theory are established. Various factors that affect the discrepancy between the data and the theory are discussed

  10. An appraisal of subcooled boiling and slip ratio from measurements made in Lingen BWR

    International Nuclear Information System (INIS)

    Nash, G.

    1977-08-01

    Measurements of steam bubble velocities and voidage have been made in the relatively small Core B of Lingen BWR. The results of axial scanning in one radial position have produced experimental values of slip ratio, power (from a travelling incore probe), voidage and coolant mean density over the core height for this position. This one set of distributions has enabled us to test current UKAEA models of subcooled boiling and slip ratio against experiment. From the comparisons, it appears that we can predict the onset of voiding well, but the assumption that a constant fraction of the heat flux forms steam in the subcooled region needs modifying. Of four slip options tested, the current one used by HAMBO and JOSHUA III (Bankoff-Jones) predicts too high a slip ratio. A closer fit to experiment comes from the new Bryce flow-dependent slip option. Any changes in the modelling must be checked, however, with coupled thermal hydraulics-neutronics computations. (author)

  11. 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

  12. Influence of test tube material on subcooled flow boiling critical heat flux in short vertical tube

    International Nuclear Information System (INIS)

    Hata, Koichi; Shiotsu, Masahiro; Noda, Nobuaki

    2007-01-01

    The steady state subcooled flow boiling critical heat flux (CHF) for the flow velocities (u=4.0 to 13.3 m/s), the inlet subcoolings (ΔT sub,in =48.6 to 154.7 K), the inlet pressure (P in =735.2 to 969.0 kPa) and the increasing heat input (Q 0 exp(t/τ), τ=10, 20 and 33.3 s) are systematically measured with the experimental water loop. The 304 Stainless Steel (SUS304) test tube of inner diameter (d=6 mm), heated length (L=66 mm) and L/d=11 with the inner surface of rough finished (Surface roughness, Ra=3.18 μm), the Cupro Nickel (Cu-Ni 30%) test tube of d=6 mm, L=60 mm and L/d=10 with Ra=0.18 μm and the Platinum (Pt) test tubes of d=3 and 6 mm, L=66.5 and 69.6 mm, and L/d=22.2 and 11.6 respectively with Ra=0.45 μm are used in this work. The CHF data for the SUS304, Cu-Ni 30% and Pt test tubes were compared with SUS304 ones for the wide ranges of d and L/d previously obtained and the values calculated by the authors' published steady state CHF correlations against outlet and inlet subcoolings. The influence of the test tube material on CHF is investigated into details and the dominant mechanism of subcooled flow boiling critical heat flux is discussed. (author)

  13. Influence of Test Tube Material on Subcooled Flow Boiling Critical Heat Flux in Short Vertical Tube

    International Nuclear Information System (INIS)

    Koichi Hata; Masahiro Shiotsu; Nobuaki Noda

    2006-01-01

    The steady state subcooled flow boiling critical heat flux (CHF) for the flow velocities (u = 4.0 to 13.3 m/s), the inlet subcooling (ΔT sub,in = 48.6 to 154.7 K), the inlet pressure (P in = 735.2 to 969.0 kPa) and the increasing heat input (Q 0 exp(t/t), t = 10, 20 and 33.3 s) are systematically measured with the experimental water loop. The 304 Stainless Steel (SUS304) test tubes of inner diameters (d = 6 mm), heated lengths (L = 66 mm) and L/d = 11 with the inner surface of rough finished (Surface roughness, R a = 3.18 μm), the Cupro Nickel (Cu-Ni 30%) test tubes of d = 6 mm, L = 60 mm and L/d = 10 with R a = 0.18 μm and the Platinum (Pt) test tubes of d = 3 and 6 mm, L = 66.5 and 69.6 mm, and L/d 22.2 and 11.6 respectively with R a = 0.45 μm are used in this work. The CHF data for the SUS304, Cu-Ni 30% and Pt test tubes were compared with SUS304 ones for the wide ranges of d and L/d previously obtained and the values calculated by the authors' published steady state CHF correlations against outlet and inlet subcooling. The influence of the test tube material on CHF is investigated into details and the dominant mechanism of subcooled flow boiling critical heat flux is discussed. (authors)

  14. Experimental investigation of nucleate boiling on heated surfaces under subcooled conditions

    International Nuclear Information System (INIS)

    Schneider, C.; Hampel, R.; Traichel, A.; Hurtado, A.; Meissner, S.; Koch, E.

    2011-01-01

    In case of an accident at pressurized water reactors (PWR), critical boiling conditions can appear at the transition from bubble- to film boiling. During full power operation, heat transfer phenomena of sub cooled nucleate boiling occur on the surface of the fuel rods. To investigate the microscopic processes in nucleate boiling, a test facility with optical measuring methods was constructed. This allows analyzing the effects on a single bubble system at different parameters. For the generation of nucleate boiling, an optically transparent, electrically conductive coating was applied as a heating surface on a borosilicate substrate. The so-called ITO (Indium-Tin-Oxide) coating with a sheet resistance of 20 ohms enables an electrical heating at an optical transparent surface. These properties are prerequisites for the study of microscopic phenomena in the bubble formation with optical coherence tomography (OCT). OCT, generally used in medical diagnostics, is an imaging modality providing cross sectional and volumetric high resolution images. To make sure that the bubble formation takes place at a specific site, artificial nucleation sites in form of micro cavity will be inserted into the surface. Furthermore a small test facility was constructed to dedicate the wall temperature of a heated metal foil during subcooled boiling in non degassed water, which is the content of this paper. (author)

  15. 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.

  16. Numerical Simulation on Subcooled Boiling Heat Transfer Characteristics of Water-Cooled W/Cu Divertors

    Science.gov (United States)

    Han, Le; Chang, Haiping; Zhang, Jingyang; Xu, Tiejun

    2015-04-01

    In order to realize safe and stable operation of a water-cooled W/Cu divertor under high heating condition, the exact knowledge of its subcooled boiling heat transfer characteristics under different design parameters is crucial. In this paper, subcooled boiling heat transfer in a water-cooled W/Cu divertor was numerically investigated based on computational fluid dynamic (CFD). The boiling heat transfer was simulated based on the Euler homogeneous phase model, and local differences of liquid physical properties were considered under one-sided high heating conditions. The calculated wall temperature was in good agreement with experimental results, with the maximum error of 5% only. On this basis, the void fraction distribution, flow field and heat transfer coefficient (HTC) distribution were obtained. The effects of heat flux, inlet velocity and inlet temperature on temperature distribution and pressure drop of a water-cooled W/Cu divertor were also investigated. These results provide a valuable reference for the thermal-hydraulic design of a water-cooled W/Cu divertor. supported by the National Magnetic Confinement Fusion Science Program of China (No. 2010GB104005), Funding of Jiangsu Innovation Program for Graduate Education (CXLX12_0170), the Fundamental Research Funds for the Central Universities of China

  17. 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)

  18. 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)

  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. Pattern recognition of subcooled boiling in core of nuclear reactors

    International Nuclear Information System (INIS)

    Inyushev, V.V.; Sharaevskij, I.G.

    2003-01-01

    The noise signals at the outlet of main nuclear power plant technological parameter gauges (neutron flux, dynamic pressure etc.) contain important information on technical state of the equipment. In the work efficient algorithms of random process identification that after respective spectral transformation are considered as multidimensional random vectors were developed. Automated classification of these vector in the developed algorithms in realized on the base of probability function, especially of Bayes classifier. Application of classifier is based on construction of multidimensional distribution of probabilities in feature space of corresponding dimension, with the help of which random vectors-realizations of corresponding images that are subjected to automated classification are described

  1. 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.

  2. 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.

  3. Subcooled boiling heat transfer and dryout on a constant temperature microheater

    International Nuclear Information System (INIS)

    Chen Tailian; Klausner, James F.; Chung, Jacob N.

    2004-01-01

    An experimental study of single-bubble subcooled boiling heat transfer (ΔT sub =31.5 K) on a small heater with constant wall temperature has been performed to better understand the boiling heat transfer associated with this unique configuration. The heater of 0.27 mm x 0.27 mm is set at different superheats to generate vapor bubbles on the microheater surface. For each superheat, the heater temperature is maintained constant by an electronic feedback control circuit while its power dissipation is measured at a frequency of 4.5 kHz. The single-bubble boiling is characterized by a transient bubble nucleation-departure period and a slow growth period. For the superheat range of 34-114 K in this study, at wall superheats below 84 K, the heater remains partially wetted following bubble departure and subsequent nucleation, and this period is characterized by a heat flux spike. At wall superheats above 90 K, the heater is blanketed with vapor following bubble departure and the heat flux experiences a dip during this period. At all superheats, the slow growth period is characterized by an almost uniform heat flux, and it has been observed that the heater surface is mostly covered by vapor. The unique heat transfer processes associated with boiling on this microheater are considerably different than those typically observed during boiling on a large heater

  4. Effect of subcooling and wall thickness on pool boiling from downward-facing curved surfaces in water

    Energy Technology Data Exchange (ETDEWEB)

    El-Genk, M.S.; Glebov, A.G. [Univ. of New Mexico, Albuquerque, NM (United States)

    1995-09-01

    Quenching experiments were performed to investigate the effects of water subcooling and wall thickness on pool boiling from a downward-facing curved surface. Experiments used three copper sections of the same diameter (50.8 mm) and surface radius (148 mm), but different thickness (12.8, 20 and 30 mm). Local and average pool boiling curves were obtained at saturation and 5 K, 10 K, and 14 K subcooling. Water subcooling increased the maximum heat flux, but decreased the corresponding wall superheat. The minimum film boiling heat flux and the corresponding wall superheat, however, increased with increased subcooling. The maximum and minimum film boiling heat fluxes were independent of wall thickness above 20 mm and Biot Number > 0.8, indicating that boiling curves for the 20 and 30 thick sections were representative of quasi steady-state, but not those for the 12.8 mm thick section. When compared with that for a flat surface section of the same thickness, the data for the 12.8 mm thick section showed significant increases in both the maximum heat flux (from 0.21 to 0.41 MW/m{sup 2}) and the minimum film boiling heat flux (from 2 to 13 kW/m{sup 2}) and about 11.5 K and 60 K increase in the corresponding wall superheats, respectively.

  5. Experiments on the effects of nanoparticles on subcooled nucleate pool boiling

    Science.gov (United States)

    Kangude, Prasad; Bhatt, Dhairya; Srivastava, Atul

    2018-05-01

    The effect of nanoparticles on a single bubble-based nucleate pool boiling phenomenon under subcooled conditions has been studied. Water (as the base fluid) and two different concentrations of water-silica nanofluids (0.005% and 0.01% V/V) have been employed as the working fluids. The boiling experiments have been conducted in a specially designed chamber, wherein an ITO-coated heater substrate has been used to induce single bubble nucleation. Measurements have been performed in a completely non-intrusive manner using one of the refractive index-based diagnostics techniques, namely, rainbow schlieren deflectometry. Thus, the thermal gradients prevailing in the boiling chamber have directly been mapped as a two-dimensional distribution of hue values that are recorded in the form of rainbow schlieren images. The schlieren-based measurements clearly revealed the plausible influence of nanoparticles on the strength of temperature gradients prevailing in the boiling chamber. As compared to the base fluid, the experiments with dilute nanofluids showed that the suspended nanoparticles tend to diffuse (homogenize) the strength of temperature gradients, both in the vicinity of the heated substrate and in the thermal boundary layer enveloping the vapor bubble. An overall reduction in the bubble volume and dynamic contact angle was seen with increasing concentrations of dilute nanofluids. In addition, the vapor bubble was found to assume a more spherical shape at higher concentrations of dilute nanofluids in comparison to its shape with water-based experiments. Clear oscillations of the vapor bubble in the subcooled pool of liquids (water and/or nanofluids) were observed, the frequency of which was found to be significantly reduced as the nanoparticle concentration was increased from 0% (water) to 0.01% (V/V). A force balance analysis has been performed to elucidate the plausible mechanisms explaining the observed trends of the oscillation frequencies of the vapor bubble.

  6. Experimental evaluation of local bubble parameters of subcooled boiling flow in a pressurized vertical annulus channel

    Energy Technology Data Exchange (ETDEWEB)

    Chu, In-Cheol, E-mail: chuic@kaeri.re.kr; Lee, Seung-Jun; Youn, Young Jung; Park, Jong Kuk; Choi, Hae Seob; Euh, Dong-Jin; Song, Chul-Hwa

    2017-02-15

    Experiments were performed to quantify the local bubble parameters such as void fraction, bubble velocity, interfacial area concentration, and Sauter mean diameter for the subcooled boiling flow of a refrigerant R-134a in a pressurized vertical annulus channel. Optical fiber void probe and double pressure boundary visualization windows were installed at four measurement stations with different elevations, thus enabling the quantification of local bubble parameters and observation of global boiling structure. Using high-resolution traverse systems for the optical fiber void probes and the heating tube, the radial profiles of the bubble parameters and their axial propagation can be evaluated at any elevation of the whole heating region. At this first phase of the experiments, three tests were conducted by varying the pressure, heat flux, mass flux, and local liquid subcooling. The radial profiles of the bubble parameters were obtained at seven elevations. The pressure condition of the present experiments covered the normal operating pressure of PWRs according to the similarity criteria. The present experimental data will be useful for thorough validation and improvement of the CMFD (Computation Multi-Fluid Dynamics) codes and constitutive relations.

  7. 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

  8. Physical modeling and numerical simulation of subcooled boiling in one- and three-dimensional representation of bundle geometry

    International Nuclear Information System (INIS)

    Bottoni, M.; Lyczkowski, R.; Ahuja, S.

    1995-01-01

    Numerical simulation of subcooled boiling in one-dimensional geometry with the Homogeneous Equilibrium Model (HEM) may yield difficulties related to the very low sonic velocity associated with the HEM. These difficulties do not arise with subcritical flow. Possible solutions of the problem include introducing a relaxation of the vapor production rate. Three-dimensional simulations of subcooled boiling in bundle geometry typical of fast reactors can be performed by using two systems of conservation equations, one for the HEM and the other for a Separated Phases Model (SPM), with a smooth transition between the two models

  9. Development of Off-take Model, Subcooled Boiling Model, and Radiation Heat Transfer Input Model into the MARS Code for a Regulatory Auditing of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, C.; Rhee, B. W.; Chung, B. D. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahn, S. H.; Kim, M. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2009-05-15

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to a lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use a vendor's code for a regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed the RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of the existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of the RELAP5/MOD3/CANDU code to the MARS code including a quality assurance of the developed models.

  10. Development of Off-take Model, Subcooled Boiling Model, and Radiation Heat Transfer Input Model into the MARS Code for a Regulatory Auditing of CANDU Reactors

    International Nuclear Information System (INIS)

    Yoon, C.; Rhee, B. W.; Chung, B. D.; Ahn, S. H.; Kim, M. W.

    2009-01-01

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to a lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use a vendor's code for a regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed the RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of the existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of the RELAP5/MOD3/CANDU code to the MARS code including a quality assurance of the developed models

  11. 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.

  12. 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.)

  13. Random generation of bubble sizes on the heated wall during subcooled boiling

    International Nuclear Information System (INIS)

    Koncar, B.; Mavko, B.

    2003-01-01

    In subcooled flow boiling, a locally averaged bubble diameter significantly varies in the transverse direction to the flow. From the experimental data of Bartel, a bent crosssectional profile of local bubble diameter with the maximum value shifted away from the heated wall may be observed. In the present paper, the increasing part of the profile (near the heated wall) is explained by a random generation of bubble sizes on the heated wall. The hypothesis was supported by a statistical analysis of different CFD simulations, varying by the size of the generated bubble (normal distribution) and the number of generated bubbles per unit surface. Local averaging of calculated void fraction distributions over different bubble classes was performed. The increasing curve of the locally averaged bubble diameter in the near-wall region was successfully predicted. (author)

  14. Power distribution changes caused by subcooled nucleate boiling at Callaway Nuclear Power Plant

    International Nuclear Information System (INIS)

    Konya, M.J.; Bryant, K.R.; Hopkins, D.L.

    1993-01-01

    This paper reports the results of an evaluation undertaken by Union Electric (UE) and Westinghouse to explain anomalous behavior of the core axial power distribution at the Callaway Nuclear Power Plant. The behavior was characterized by a gradual unexpected power shift toward the bottom of the core and was first detected during cycle 4 at a core average burnup of approximately 7,000 MWD/MTU. Once started, the power shift continued until burnup effects became dominant and caused power to shift back to the top of the core at the end of the cycle. In addition to the anomalous power distribution, UE observed that estimated critical control rod position (ECP) deviations increased to over 500 pcm (0.5%Δk/k) during Cycles 4 and 5. ECPs for plant restarts that occurred early in each cycle agreed well with measured critical conditions. However, this agreement disappeared for restarts that occurred later in core life. After analyzing relevant data, performing scoping calculations and reviewing industry experience, the authors concluded that the power distribution anomaly was most likely caused by subcooled nucleate boiling. Crud deposition on the fuel was believed to enhance the subcooled boiling. The ECP deviations were a secondary effect of the power shift, since void fraction, axial burnup and xenon distributions departed design predictions during a substantial portion of the fuel cycles. Significant evidence supporting these conclusions include incore detector indications of flux depressions between intermediate flow mixing (IFM) and structural grids. In addition, visual exam results show the presence of crud deposits on fuel pins

  15. 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)

  16. 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

  17. 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

  18. Burnout in boiling heat transfer. Part II: subcooled and low quality forced-convection systems

    International Nuclear Information System (INIS)

    Bergles, A.E.

    1977-01-01

    Recent experimental and analytical developments regrading burnout in subcooled and low quality forced-convection systems are reviewed. Much data have been accumulated which clarify the parametric trends and lead to new design correlations for water and a variety of other coolants in both simple and complex geometries. A number of critical experiments and models have been developed to attempt to clarify the burnout mechanism(s) in simpler geometries and power transients

  19. 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.

  20. Calculation of Void Volume Fraction in the Subcooled and Quality Boiling Regions

    Energy Technology Data Exchange (ETDEWEB)

    Rouhani, S Z; Axelsson, E

    1968-10-15

    The complex problem of void calculation in the different regions of flow boiling is divided in two parts. The first part includes only the description of the mechanisms and the calculation of the rates of heat transfer for vapour and liquid. It is assumed that heat is removed by vapour generation, heating of the liquid that replaces the detached bubbles, and in some parts, by single phase heat transfer. By considering the rate of vapour condensation in liquid, an equation for the differential changes in the true steam quality throughout the boiling regions is obtained. Integration of this equation yields the vapour weight fraction at any position. The second part of the problem concerns the determination of the void fractions corresponding to the calculated steam qualities. For this purpose we use the derivations of Zuber and Findlay. This model is compared with data from different geometries including small rectangular channels and large rod bundles. The data covered pressures from 19 to 138 bars, heat fluxes from 18 to 120 W/cm{sup 2} with many different subcoolings and mass velocities. The agreement is generally very good.

  1. Calculation of Void Volume Fraction in the Subcooled and Quality Boiling Regions

    International Nuclear Information System (INIS)

    Rouhani, S.Z.; Axelsson, E.

    1968-10-01

    The complex problem of void calculation in the different regions of flow boiling is divided in two parts. The first part includes only the description of the mechanisms and the calculation of the rates of heat transfer for vapour and liquid. It is assumed that heat is removed by vapour generation, heating of the liquid that replaces the detached bubbles, and in some parts, by single phase heat transfer. By considering the rate of vapour condensation in liquid, an equation for the differential changes in the true steam quality throughout the boiling regions is obtained. Integration of this equation yields the vapour weight fraction at any position. The second part of the problem concerns the determination of the void fractions corresponding to the calculated steam qualities. For this purpose we use the derivations of Zuber and Findlay. This model is compared with data from different geometries including small rectangular channels and large rod bundles. The data covered pressures from 19 to 138 bars, heat fluxes from 18 to 120 W/cm 2 with many different subcoolings and mass velocities. The agreement is generally very good

  2. Characteristics of a single bubble in subcooled boiling region of a narrow rectangular channel under natural circulation

    International Nuclear Information System (INIS)

    Zhou, Tao; Duan, Jun; Hong, Dexun; Liu, Ping; Sheng, Cheng; Huang, Yanping

    2013-01-01

    Highlights: ► We observe the behavior of single bubbles in a narrow vertical rectangular channel. ► We analyze the force characteristics of the single bubble. ► Small bubbles in highly subcooled boiling region stick on the wall or slip slowly. ► The bubbles jumping from the wall are affected by drag force. ► The thermophoretic force makes bubbles jump from the wall strongly. - Abstract: The behavior of bubbles has an important influence on heat transfer during subcooled boiling. By observing the behavior of a single bubble in a narrow vertical rectangular channel, and analyzing the force characteristics of the single bubble, it turns out that small bubbles in the highly subcooled boiling region stick on the wall or slip slowly. The bubbles jumping from the wall are affected by drag force, and move with high speed. Maintaining a certain heating power, at the onset of boiling (ONB) point, the bubbles remain in a stable state. Furthermore, the thermophoretic force is considered in this paper. With increasing the temperature gradient in the fluid, the thermophoretic force causes the bubbles to jump from the wall easier

  3. Effect of transverse power distribution on the ONB location in the subcooled boiling flow

    International Nuclear Information System (INIS)

    Al-Yahia, Omar S.; Lee, Yong Joong; Jo, Daeseong

    2017-01-01

    Highlights: • Effect of transverse power distribution on ONB incipient. • Uniform and non-uniform heat distribution is simulated in a narrow rectangular channel. • Simulations are performed using CFX and TMAP codes. • For uniform heating, ONB incipient by CFX occurs between predictions by TMAP analyses. • For non-uniform heating, ONB incipient by CFX occurs at a higher power than that by TMAP analysis. - Abstract: This study investigates the effect of transverse power distribution on the ONB (Onset of Nucleate Boiling) incipient. For this purpose, a subcooled boiling model with uniform and non-uniform heat flux distribution is simulated in a narrow vertical rectangular channel heated from both sides by applying a wide range of thermal power (8–16 kW). The simulations are performed using the CFX and TMAP codes. The CFX code incorporates both a two-fluid model and RPI wall boiling model to investigate coolant and wall temperature distributions along the heated channel. The TMAP code implements two different sets of heat transfer correlations to evaluate the wall temperature. The results obtained from the TMAP analyses show that the wall temperatures predicted by the Jo et al. heat transfer correlation are higher than the ones predicted by the Dittus and Boelter heat transfer correlation. The wall temperatures predicted by the CFX analyses lie between the predicted wall temperatures obtained by the TMAP analyses. Based on the superheated temperature on the heated surface, the ONB incipient is determined. The axial locations of the ONB incipient are predicted differently by the CFX and TMAP analyses. For uniform heating, the ONB incipient predicted by the CFX analysis occurs between the predictions made by the TMAP analyses. For non-uniform heating, the ONB incipient by the CFX analysis occurs at a higher power than the power required by the TMAP analyses.

  4. Burnout in boiling heat transfer. II. Subcooled and low-quality forced-convection systems

    International Nuclear Information System (INIS)

    Bergles, A.E.

    1977-01-01

    Recent experimental and analytical developments regarding burnout in subcooled and low-quality forced-convection systems are reviewed. Many data have been accumulated which clarify the parametric trends and lead to new design correlations for water and a variety of other coolants in both simple and complex geometries. A number of critical experiments and models have been developed to attempt to clarify the burnout mechanism(s) in simpler geometries. Other topics discussed include burnout with power transients and techniques to augment burnout. 86 references

  5. On the Partitioning of Wall Heat Flux in Subcooled Flow Boiling

    International Nuclear Information System (INIS)

    Chu, In-Cheol; Hoang, Nhan Hien; Euh, Dong-Jin; Song, Chul-Hwa

    2015-01-01

    This region has been treated successfully by two-fluid model coupled with a population balance model or interfacial area transport equation (IATE). The second region is near-wall heat transfer which has been commonly described by a wall heat flux partitioning model coupled with models of nucleation site density (NSD), bubble departure diameter and bubble release frequency. Since the phase change process in the near-wall heat transfer is really complex, comprising different heat transfer mechanisms, bubble dynamics, bubble nucleation and thermal response of heated surface, the modeling of the second region is still a great challenge despite intensive efforts. Numerous models and correlations have been proposed to aim for computing the near-wall heat transfer. The models of nucleation site density, bubble departure diameter and bubble release frequency are used to quantify these components. The models closely related to each other. The heat flux partitioning model controls the wall and liquid temperatures. Then, it turns to control the boiling parameters, i.e. nucleation site density, bubble departure diameter and bubble release frequency. In this study, the partitioning of wall heat flux is taken into account. The existing issues occurred with previous models of the heat flux partitioning are pointed out and then a new model which considers the heat transfer caused by evaporation of superheated liquid at bubble boundary and the actual period of transient conduction term is formulated. The new model is then validated with a collected experimental database. This paper presented a new heat flux partitioning model in which the heat transfer by evaporation of the superheated liquid at the bubble boundary and the active period of the transient conduction were considered. The new model was validated with the experimental data of the subcooled flow boiling of water obtained by Phillips

  6. CHF multiplier of subcooled flow boiling for non-uniform heating conditions in swirl tube

    International Nuclear Information System (INIS)

    Inasaka, F.; Nariai, H.

    1994-01-01

    The high heat flux components of fusion reactors, such as divertor plates and beam dumps of neutral beam injectors, are estimated to be subjected to very high heat loads more than 10 MW/m 2 . Critical heat flux (CHF), which determines the upper limit of heat removal, is one of the most important problems in designing cooling systems. For practical applications in cooling systems, subcooled flow boiling in water combined with swirl-flow in tubes with internal twisted tape is thought to be the most superior for CHF characteristics in fusion reactor components, heat by irradiation comes in from one side of the wall, and cooling channel is then under circumferentially non-uniform heating condition. Authors have conducted the experiments on the CHF with internal twisted tapes under circumferentially non-uniform heating conditions and showed that when the intensity of non-uniformity increased, q cH (peak heat flux at burnout under nonuniform heating condition) in tube with internal twisted tape increased above the q c,unif (CHF under uniform heating condition), though the average qualities were the same for both cases. They also showed that this CHF enhancement was not seen in smooth tubes without tape under the same average qualities

  7. Physical insight in the burnout region of water-subcooled flow boiling

    International Nuclear Information System (INIS)

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

    1998-01-01

    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 characterised by a square cross-section). In order to obtain clear pictures of the flow phenomena, he coolant velocity is in the range 3-9 m.s -1 and the resulting heat flux is in the range 7-13 MW.m -2 . From video images (single frames were taken with a light exposure of 1 μs) 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 sort of elongated bubble called a 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, as well as those of the hot spots, are given as a function of thermal-hydraulic tested conditions. (authors)

  8. Experimental study on two-phase flow parameters of subcooled boiling in inclined annulus

    International Nuclear Information System (INIS)

    Lee, Tae Ho; Kim, Moon Oh; Park, Goon Cherl

    1999-01-01

    Local two-phase flow parameters of subcooled flow boiling in inclined annulus were measured to investigate the effect of inclination on the internal flow structure. Two-conductivity probe technique was applied to measured local gas phasic parameters, including void by fraction, vapor bubble frequency, chord length, vapor bubble velocity and interfacial area concentration. Local liquid velocity was measured by Pitot tube. Experiments were conducted for three angles of inclination: 0 o (vertical), 30 o , 60 o . The system pressure was maintained at atmospheric pressure. The range of average void fraction was up to 10 percent and the average liquid superficial velocities were less than 1.3 m/sec. The results of experiments showed that the distributions of two-phase flow parameters were influenced by the angle of channel inclination. Especially, the void fraction and chord length distributions were strongly affected by the increase of inclination angle, and flow pattern transition to slug flow was observed depending on the flow conditions. The profiles of vapor velocity, liquid velocity and interfacial area concentration were found to be affected by the non-symmetric bubble size distribution in inclined channel. Using the measured distributions of local phasic parameters, an analysis for predicting average void fraction was performed based on the drift flux model and flowing volumetric concentration. And it was demonstrated that the average void fraction can be more appropriately presented in terms of flowing volumetric concentration. (Author). 18 refs., 2 tabs., 18 figs

  9. 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)

  10. Experimental study of flow instability and CHF in a natural circulation system with subcooled boiling

    International Nuclear Information System (INIS)

    Yang, R.C.; Shi, D.Q.; Lu, Z.Q.; Zheng, R.C.; Wang, Y.

    1996-01-01

    Experimental study has been performed to investigate flow instability and critical heat flux (CHF) in a natural circulation system with subcooled boiling. In the experiments three kinds of heated sections were used. Freon-12 was used as the working medium. The experiments show which one of the two phenomena, flow instability and CHF condition, may first occur in the system depends on not only the heat input power to the heated section and the parameters of the working medium, but also the construction of the heated section. The occurrence of the flow instability mainly depends on the total heat input power to the heated section and the CHF condition is mainly caused by the local heat flux of the heated section. In the experiments two kinds of flow instability, flow instability with high frequency and flow instability with low frequency, were found. But they all belong to density wave instability. The influence of the parameters of the working medium on the onset of the flow instability and CHF condition in the system were investigated. The stability boundaries were determined through the experiments. By means of dimensional analysis of integral equations, a common correlation describing the threshold condition of onset of the flow instability was obtained

  11. Development of an experimental apparatus for nucleate boiling analysis

    International Nuclear Information System (INIS)

    Castro, A.J.A. de.

    1984-01-01

    An experimental apparatus is developed for the study of the parameters that affect nucleate boiling. The experimental set up is tested for nucleate boiling in an annular test section with subcooled water flow. The following parameters are analysed: pressure, fluid velocity and the fluid temperature at the test section entrance. The performance of the experimental apparatus is analysed by the results and by the problems raised by the operation of the setup. (Author) [pt

  12. Modeling and Thermal Performance Evaluation of Porous Curd Layers in Sub-Cooled Boiling Region of PWRs and Effects of Sub-Cooled Nucleate Boiling on Anomalous Porous Crud Deposition on Fuel Pin Surfaces

    International Nuclear Information System (INIS)

    Barclay Jones

    2005-01-01

    A significant number of current PWRs around the world are experiencing anomalous crud deposition in the sub-cooled region of the core, resulting in an axial power shift or Axial Offset Anomaly (AOA), a condition that continues to elude prediction of occurrence and thermal/neutronic performance. This creates an operational difficulty of not being able to accurately determine power safety margin. In some cases this condition has required power ''down rating'' by as much as thirty percent and the concomitant considerable loss of revenue for the utility. This study examines two aspects of the issue: thermal performance of crud layer and effect of sub-cooled nucleate boiling on the solute concentration and its influence on initiation of crud deposition/formation on fuel pin surface

  13. Experimental study of CHF enhancement using Fe{sub 3}O{sub 4} nanofluids in the subcooled boiling region

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Young Jae; Kam, Dong Hoon; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-10-15

    This study may give overall trends of CHF enhancement in the subcooled boiling region. In our experiment, subcooled flow boiling CHF enhancement phenomena in water and nano-coated surface was investigated in mass flux from 1000 to 5000 kg/m{sup 2}s. CHF enhancement of nanoparticles coated tube in DI water increased as exit quality get bigger at same massflux. Various methods to improve CHF characteristics are introduced, especially nanofluids are used for enhancing the CHF. Nanofluids is a colloidal suspension that nanoparticles are mixed with basic fluid. Normally the use of nanofluids as working fluid improves the flow boiling CHF characteristics. Lee et al. already researched the CHF characteristics using nanofluids. As exit quality increased from 0.07 to 0.74, CHF enhancement gradually decreased and approached zero. CHF enhancement was observed when exit quality was low and a DNB-like thermal crisis occurred. But CHF enhancement didn't occur for high exit quality, but LFD-type thermal crisis occurred. Because LFD phenomena are nearly unaffected by the surface conditions, CHF enhancement is not expected for annular flow with high exit quality. Kim et al. performed flow boiling CHF enhancement at subcooled region using alumina-water, zinc-oxide-water and diamond-water nanofluids. The CHF was enhanced by increasing wettability from nanoparticle deposition. CHF enhancement occurred in high mass flux (2000-2500 kg/m{sup 2}s), but CHF enhancement didn't occur in low mass flux (1500 kg/m{sup 2}s). The amount of nanoparticle deposition on each tube can be different during experiments by the several conditions such as deposition time, mass flux and heat flux. So, before the nanofluid experiment conducted, all tube are deposited in same condition of heat flux, concentration and time.

  14. 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.

  15. 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)

  16. Study on model of onset of nucleate boiling in natural circulation with subcooled boiling using unascertained mathematics

    Energy Technology Data Exchange (ETDEWEB)

    Zhou Tao [Department of Thermal Engineering, Tsinghua University, Beijing 100084 (China)]. E-mail: zhoutao@mail.tsinghua.edu.cn; Wang Zenghui [Department of Engineering Mechanics, Tsinghua University, Beijing 100084 (China); Yang Ruichang [Department of Thermal Engineering, Tsinghua University, Beijing 100084 (China)

    2005-10-01

    Experiment data got from onset of nucleate boiling (ONB) in natural circulation is analyzed using unascertained mathematics. Unitary mathematics model of the relation between the temperature and onset of nucleate boiling is built up to analysis ONB. Multiple unascertained mathematics models are also built up with the onset of natural circulation boiling equation based on the experiment. Unascertained mathematics makes that affirmative results are a range of numbers that reflect the fluctuation of experiment data more truly. The fluctuating value with the distribution function F(x) is the feature of unascertained mathematics model and can express fluctuating experimental data. Real status can be actually described through using unascertained mathematics. Thus, for calculation of ONB point, the description of unascertained mathematics model is more precise than common mathematics model. Based on the unascertained mathematics, a new ONB model is developed, which is important for advanced reactor safety analysis. It is conceivable that the unascertained mathematics could be applied to many other two-phase measurements as well.

  17. Study on model of onset of nucleate boiling in natural circulation with subcooled boiling using unascertained mathematics

    International Nuclear Information System (INIS)

    Zhou Tao; Wang Zenghui; Yang Ruichang

    2005-01-01

    Experiment data got from onset of nucleate boiling (ONB) in natural circulation is analyzed using unascertained mathematics. Unitary mathematics model of the relation between the temperature and onset of nucleate boiling is built up to analysis ONB. Multiple unascertained mathematics models are also built up with the onset of natural circulation boiling equation based on the experiment. Unascertained mathematics makes that affirmative results are a range of numbers that reflect the fluctuation of experiment data more truly. The fluctuating value with the distribution function F(x) is the feature of unascertained mathematics model and can express fluctuating experimental data. Real status can be actually described through using unascertained mathematics. Thus, for calculation of ONB point, the description of unascertained mathematics model is more precise than common mathematics model. Based on the unascertained mathematics, a new ONB model is developed, which is important for advanced reactor safety analysis. It is conceivable that the unascertained mathematics could be applied to many other two-phase measurements as well

  18. A study on bubble detachment and the impact of heated surface structure in subcooled nucleate boiling flows

    International Nuclear Information System (INIS)

    Wu Wen; Chen Peipei; Jones, Barclay G.; Newell, Ty A.

    2008-01-01

    This study examines the bubble detachment phenomena under subcooled nucleate boiling conditions, in order to obtain a better understanding of the bubble dynamics on horizontal flat heat exchangers. Refrigerant R134a is chosen as a simulant fluid due to its merits of having smaller surface tension, reduced latent heat, and lower boiling temperature than water. Experiments are run with varying experimental parameters, e.g. pressure, inlet subcooled level, flow rate, etc. Digital images are obtained at frame rates up to 4000 frames/s, showing the characteristics of bubble movements. Bubble departure and bubble lift-off, which are described as bubbles detaching from the original nucleation sites and bubbles detaching from the horizontal heated surface respectively, are both considered and measured. Results are compared against the model proposed by Klausner et al. for the prediction of bubble detachment sizes. While good overall agreement is shown, it is suggested that finite rather than zero bubble contact area should be assumed, which improves the model prediction at the pressure range of 300-500 kPa while playing no significant role at a lower pressure of 150 kPa where the model was originally benchmarked. The impact of heated surface structure is studied whose results provide support to the above assumption

  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. Experimental study of average void fraction in low-flow subcooled boiling

    International Nuclear Information System (INIS)

    Sun Qi; Wang Xiaojun; Xi Zhao; Zhao Hua; Yang Ruichang

    2005-01-01

    Low-flow subcooled void fraction in medium pressure was investigated using high-temperature high-pressure single-sensor optical probe in this paper. And then average void fraction was obtained through the integral calculation of local void fraction in the cross-section. The experimental data were compared with the void fraction model proposed in advance. The results show that the predictions of this model agree with the data quite well. The comparisons of Saha and Levy models with low-flow subcooled data show that Saha model overestimates the experimental data distinctively, and Levy model also gets relatively higher predictions although it is better than Saha model. (author)

  1. 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)

  2. Subcooled film boiling heat transfer on a high temperature sphere in very dilute Al2O3 nano-fluids

    International Nuclear Information System (INIS)

    Hyun Sun Park; Dereje Shiferaw; Bal Raj Sehgal

    2005-01-01

    found in similar experiments with distilled water. The experiments have also shown that if some nano-particles stick to the surface of the hot sphere (in the event that the surface is not washed in-between the experiments), film boiling practically disappears and the quench proceeds very rapidly. In this study, a series of quenching experiments with a high temperature sphere in subcooled Al 2 O 3 nano-fluids with the very dilute concentrations of 0.01 to 0.05 g/liter are conducted to investigate the characteristics of film boiling and compare to those in pure water tests. One stainless steel sphere of 10 mm in diameter is heated to the initial sphere temperature up to 1100 K in the induction furnace and plunged by a pneumatic actuator into a pool of the dilute nano-fluid at the liquid subcooling of about 20 K by a pneumatic actuator. The transient center temperature is continually recorded and monitored with a data acquisition system. The center temperature data is analyzed to obtain sphere wall heat fluxes and the corresponding heat transfer coefficients. The test results are carefully compared with our previous experimental results for nano-fluids with the higher concentrations of nano-particles in water. (authors)

  3. Subcooled boiling heat transfer in a short vertical SUS304-tube at liquid Reynolds number range 5.19 x 104 to 7.43 x 105

    International Nuclear Information System (INIS)

    Hata, Koichi; Masuzaki, Suguru

    2009-01-01

    The subcooled boiling heat transfer and the steady-state critical heat fluxes (CHFs) in a short vertical SUS304-tube for the flow velocities (u = 17.28-40.20 m/s), the inlet liquid temperatures (T in = 293.30-362.49 K), the inlet pressures (P in = 842.90-1467.93 kPa) and the exponentially increasing heat input (Q = Q 0 exp(t/τ), τ = 8.5 s) are systematically measured by the experimental water loop comprised of a multistage canned-type circulation pump with high pump head. The SUS304 test tubes of inner diameters (d = 3 and 6 mm), heated lengths (L = 33 and 59.5 mm), effective lengths (L eff = 23.3 and 49.1 mm), L/d (=11 and 9.92), L eff /d (=7.77 and 8.18), and wall thickness (δ = 0.5 mm) with average surface roughness (Ra = 3.18 μm) are used in this work. The inner surface temperature and the heat flux from non-boiling to CHF are clarified. The subcooled boiling heat transfer for SUS304 test tube is compared with our Platinum test tube data and 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 and the widely and precisely predictable correlation of the subcooled boiling heat transfer for turbulent flow of water in a short vertical SUS304-tube is given based on the experimental data. The correlation can describe the subcooled boiling heat transfer obtained in this work within 15% difference. Nucleate boiling surface superheats for the SUS304 test tube become very high. Those at the high flow velocity are close to the lower limit of Heterogeneous Spontaneous Nucleation Temperature. The dominant mechanisms of the flow boiling CHF in a short vertical SUS304-tube are discussed.

  4. Void Measurements in the Regions of Sub-Cooled and Low-Quality Boiling. Part 1. Low Mass Velocities

    Energy Technology Data Exchange (ETDEWEB)

    Rouhani, S Z

    1966-07-15

    By the application of the ({gamma}, n) reaction to boiling heavy water, void volume fractions have been measured in a vertical annular channel with 25 mm O.D. and 12 mm I.D. at a heated length of 1090 mm. The experiments covered pressures from 10 to 50 bars, mass velocities from 50 to 1450 kg/m-sec, heat fluxes from 30 to 90 W/cm{sup 2}, sub coolings from 30 to 0 C, and steam qualities from 0 to 15 %. The results indicate noticeable effects of pressure, heat flux and even mass velocity upon the variations of void with subcooling and steam quality. A novel explanation of the mechanism of their effects has been found and proved by qualitative analysis.

  5. Experimental Investigation on the Effects of Coolant Concentration on Sub-Cooled Boiling and Crud Deposition on Reactor Cladding at Prototypical PWR Operating Conditions

    Energy Technology Data Exchange (ETDEWEB)

    Schultis, J., Kenneth; Fenton, Donald, L.

    2006-10-20

    Increasing demand for energy necessitates nuclear power units to increase power limits. This implies significant changes in the design of the core of the nuclear power units, therefore providing better performance and safety in operations. A major hindrance to the increase of nuclear reactor performance especially in Pressurized Deionized water Reactors (PWR) is Axial Offset Anomaly (AOA)--the unexpected change in the core axial power distribution during operation from the predicted distribution. This problem is thought to be occur because of precipitation and deposition of lithiated compounds like boric acid (H{sub 2}BO{sub 3}) and lithium metaborate (LiBO{sub 2}) on the fuel rod cladding. Deposited boron absorbs neutrons thereby affecting the total power distribution inside the reactor. AOA is thought to occur when there is sufficient build-up of crud deposits on the cladding during subcooled nucleate boiling. Predicting AOA is difficult as there is very little information regarding the heat and mass transfer during subcooled nucleate boiling. An experimental investigation was conducted to study the heat transfer characteristics during subcooled nucleate boiling at prototypical PWR conditions. Pool boiling tests were conducted with varying concentrations of lithium metaborate (LiBO{sub 2}) and boric acid (H{sub 2}BO{sub 3}) solutions in deionized water. The experimental data collected includes the effect of coolant concentration, subcooling, system pressure and heat flux on pool the boiling heat transfer coefficient. The analysis of particulate deposits formed on the fuel cladding surface during subcooled nucleate boiling was also performed. The results indicate that the pool boiling heat transfer coefficient degrades in the presence of boric acid and lithium metaborate compared to pure deionized water due to lesser nucleation. The pool boiling heat transfer coefficients decreased by about 24% for 5000 ppm concentrated boric acid solution and by 27% for 5000 ppm

  6. Critical heat flux with subcooled boiling of water at low pressure

    International Nuclear Information System (INIS)

    Chen Yuzhou; Zhou Runbin; Hao Laomi; Chen Haiyan

    1997-01-01

    The critical heat flux experiment has been performed in round tubes of 10 and 16 mm in diameter with different heating length, covering the range of pressure 1.5-16.7 bar, velocity 1.4-15.4 m/s and exit subcooling 30-136 K. The experimental data and empirical correlations are presented. Based on the results an evaluation of some correlations and 1995 CHF look-up table is made. For the conditions tested the effect of diameter on the critical heat flux is found to be related to the liquid velocity. (author)

  7. 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

  8. Interferometric and numerical study of the temperature field in the boundary layer and heat transfer in subcooled flow boiling

    Energy Technology Data Exchange (ETDEWEB)

    Lucic, Anita; Emans, Maximilian; Mayinger, Franz; Zenger, Christoph

    2004-04-01

    An interferometric study and a numerical simulation are presented of the combined process of the bulk turbulent convection and the dynamic of a vapor bubble which is formed in the superheated boundary layer of a subcooled flowing liquid, in order to determine the heat transfer to the flowing subcooled liquid. In this investigation focus has been given on a single vapor bubble at a defined cavity site to provide reproducible conditions. In the experimental study single bubbles were generated at a single artificial cavity by means of a CO{sub 2}-laser as a spot heater at a uniformly heated wall of a vertical rectangular channel with water as the test fluid. The experiments were performed at various degrees of subcooling and mass flow rates. The bubble growth and the temporal decrease of the bubble volume were captured by means of the high-speed cinematography. The thermal boundary layer and the temperature field at the phase-interface between fluid and bubble were visualized by means of the optical measurement method holographic interferometry with a high temporal and spatial resolution, and thus the local and temporal heat transfer could be quantified. The experimental results form a significant data basis for the description of the mean as well as the local heat transfer as a function of the flow conditions. According to the experimental configuration and the obtained data the numerical simulations were performed. A numerical method has been developed to simulate the influence of single bubbles on the surrounding fluid which is based on a Lagrangian approach to describe the motion of the bubbles. The method is coupled to a large-eddy simulations by the body force term which is locally evaluated based on the density field. The obtained experimental data correspond well with the numerical predictions, both of which demonstrate the thermo- and fluiddynamic characteristics of the interaction between the vapor bubble and the subcooled liquid.

  9. Interferometric and numerical study of the temperature field in the boundary layer and heat transfer in subcooled flow boiling

    International Nuclear Information System (INIS)

    Lucic, Anita; Emans, Maximilian; Mayinger, Franz; Zenger, Christoph

    2004-01-01

    An interferometric study and a numerical simulation are presented of the combined process of the bulk turbulent convection and the dynamic of a vapor bubble which is formed in the superheated boundary layer of a subcooled flowing liquid, in order to determine the heat transfer to the flowing subcooled liquid. In this investigation focus has been given on a single vapor bubble at a defined cavity site to provide reproducible conditions. In the experimental study single bubbles were generated at a single artificial cavity by means of a CO 2 -laser as a spot heater at a uniformly heated wall of a vertical rectangular channel with water as the test fluid. The experiments were performed at various degrees of subcooling and mass flow rates. The bubble growth and the temporal decrease of the bubble volume were captured by means of the high-speed cinematography. The thermal boundary layer and the temperature field at the phase-interface between fluid and bubble were visualized by means of the optical measurement method holographic interferometry with a high temporal and spatial resolution, and thus the local and temporal heat transfer could be quantified. The experimental results form a significant data basis for the description of the mean as well as the local heat transfer as a function of the flow conditions. According to the experimental configuration and the obtained data the numerical simulations were performed. A numerical method has been developed to simulate the influence of single bubbles on the surrounding fluid which is based on a Lagrangian approach to describe the motion of the bubbles. The method is coupled to a large-eddy simulations by the body force term which is locally evaluated based on the density field. The obtained experimental data correspond well with the numerical predictions, both of which demonstrate the thermo- and fluiddynamic characteristics of the interaction between the vapor bubble and the subcooled liquid

  10. Experimental study of static flow instability in subcooled flow boiling in parallel channels

    International Nuclear Information System (INIS)

    Siman-Tov, M.; Felde, D.K.; McDuffee, J.L.; Yoder, G.L.

    1995-01-01

    Experimental data for static flow instability or flow excursion (FE) at conditions applicable to the Advanced Neutron Source Reactor are very limited. A series of FE tests with light water flowing vertically upward was completed covering a local exit heat flux range of 0.7--18 MW/m 2 , exit velocity range of 2.8--28.4 m/s, exit pressure range of 0.117--1.7 MPa, and inlet temperature range of 40-- 50 degrees C. Most of the tests were performed in a ''stiff'' (constant flow) system where the instability threshold was detected through the minimum of the pressure-drop curve. A few tests were also conducted using as ''soft'' (constant pressure drop) a system as possible to secure a true FE phenomenon (actual secondary burnout). True critical heat flux experiments under similar conditions were also conducted using a stiff system. The FE data reported in this study considerably extend the velocity range of data presently available worldwide, most of which were obtained at velocities below 10 m/s. The Saha and Zuber correlation had the best fit with the data out of the three correlations compared. However, a modification was necessary to take into account the demonstrated dependence of the St and Nu numbers on subcooling levels, especially in the low subcooling regime. Comparison of Thermal Hydraulic Test Loop (THTL) data, as well as extensive data from other investigators, led to a proposed modification to the Saha and Zuber correlation for onset of significant void, applied to FE prediction. The mean and standard deviation of the THTL data were 0.95 and 15%, respectively, when comparing the THTL data with the original Saha and Zuber correlation, and 0.93 and 10% when comparing them with the modification. Comparison with the worldwide database showed a mean and standard deviation of 1.37 and 53%, respectively, for the original Saha and Zuber correlation and 1.0 and 27% for the modification

  11. 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.

  12. A model for steady-state and transient determination of subcooled boiling for calculations coupling a thermohydraulic and a neutron physics calculation program for reactor core calculation

    International Nuclear Information System (INIS)

    Mueller, R.G.

    1987-06-01

    Due to the strong influence of vapour bubbles on the nuclear chain reaction, an exact calculation of neutron physics and thermal hydraulics in light water reactors requires consideration of subcooled boiling. To this purpose, in the present study a dynamic model is derived from the time-dependent conservation equations. It contains new methods for the time-dependent determination of evaporation and condensation heat flow and for the heat transfer coefficient in subcooled boiling. Furthermore, it enables the complete two-phase flow region to be treated in a consistent manner. The calculation model was verified using measured data of experiments covering a wide range of thermodynamic boundary conditions. In all cases very good agreement was reached. The results from the coupling of the new calculation model with a neutron kinetics program proved its suitability for the steady-state and transient calculation of reactor cores. (orig.) [de

  13. Comparison of FLOWTRAN predictions of onset of significant voiding (OSV) to Savannah River Heat Transfer Laboratory subcooled boiling flow instability measurements, Part 1

    International Nuclear Information System (INIS)

    Laurinat, J.E.

    1988-10-01

    The onset of flow instability (OFI) was measured in the first of a scheduled series of subcooled boiling tests at the Savannah River Heat Transfer Laboratory (HTL). This report summarizes the benchmarking of predictions of the onset of significant voiding (OSV) using Version 16 of the FLOWTRANΩ reactor limits code against the HTL measurements. This study confirms that, for this series of HTL subcooled boiling tests, the Saha-Zuber OSV correlation was a conservative indicator of OFI for Peclet numbers between 30,000 and 80,000. The Saha-Zuber correlation was not a conservative indicator of OFI for Peclet numbers below 30,000. A conservative bound to the Saha-Zuber correlation (the Saha-Zuber constant Stanton number criterion -- 30%) was agreed to at a meeting of SRL, DOE, and the DOE EH and DP review panels. This bound was a conservative indicator of OFI for all measurements in this study

  14. 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)

  15. Study on subcooled-forced flow boiling heat transfer and critical heat flux of solid particle-water two-phase mixture

    International Nuclear Information System (INIS)

    Koizumi, Yasuo; Mochizuki, Manabu; Ohtake, Hiroyasu

    1999-01-01

    The effect of solid particle introduction on forced flow boiling and the critical heat flux was examined for the mixture of subcooled-water and 0.6 mm glass beads. When the particles were introduced, the growth on of a superheated layer near a wall seemed to be suppressed and the onset of nucleate boiling was delayed. The particles tempted for bubbles to condense at nucleation sites, and then the initiation of net vapor generation was also delayed and sifted to a high wall-superheat region. The nucleate boiling heat transfer was augmented by the particles, which considered to be caused by the combination of the suppression of the superheated layer growth and the promotion of the condensation and dissipation of the bubbles. The wall superheat at the critical heat flux condition was sifted to a high wall superheat region and the critical heat flux itself was also elevated a little. (author)

  16. Local Heat Transfer and CHF for Subcooled Flow Boiling - Annual Report 1996

    International Nuclear Information System (INIS)

    Boyd, Ronald D.

    2000-01-01

    For the past decade, efforts have been growing in the development of high heat flux (HHF) components for many applications, including fusion and fission reactor components, advanced electronic components, synchrotrons and optical components, and other advanced HHF engineering applications. From a thermal prospective, work in the fusion reactor development arena has been underway in a number of areas including: (1) Plasma thermal, and electro-magnetics, and particle transport, (2) Fusion material, rheology, development, and expansion and selection; (3) High heat flux removal; and (4) Energy production and efficiency

  17. 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

  18. Theoretical and experimental studies on critical heat flux in subcooled boiling and vertical flow geometry

    International Nuclear Information System (INIS)

    Staron, E.

    1996-01-01

    Critical Heat Flux is a very important subject of interest due to design, operation and safety analysis of nuclear power plants. Every new design of the core must be thoroughly checked. Experimental studies have been performed using freon as a working fluid. The possibility of transferring of results into water equivalents has been proved. The experimental study covers vertical flow, annular geometry over a wide range of pressure, mass flow and temperature at inlet of test section. Theoretical models of Critical Heat Flux have been presented but only those which cover DNB. Computer programs allowing for numerical calculations using theoretical models have been developed. A validation of the theoretical models has been performed in accordance with experimental results. (author). 83 refs, 32 figs, 4 tabs

  19. 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)

  20. On the Application of Image Processing Methods for Bubble Recognition to the Study of Subcooled Flow Boiling of Water in Rectangular Channels

    Directory of Open Access Journals (Sweden)

    Concepción Paz

    2017-06-01

    Full Text Available This work introduces the use of machine vision in the massive bubble recognition process, which supports the validation of boiling models involving bubble dynamics, as well as nucleation frequency, active site density and size of the bubbles. The two algorithms presented are meant to be run employing quite standard images of the bubbling process, recorded in general-purpose boiling facilities. The recognition routines are easily adaptable to other facilities if a minimum number of precautions are taken in the setup and in the treatment of the information. Both the side and front projections of subcooled flow-boiling phenomenon over a plain plate are covered. Once all of the intended bubbles have been located in space and time, the proper post-process of the recorded data become capable of tracking each of the recognized bubbles, sketching their trajectories and size evolution, locating the nucleation sites, computing their diameters, and so on. After validating the algorithm’s output against the human eye and data from other researchers, machine vision systems have been demonstrated to be a very valuable option to successfully perform the recognition process, even though the optical analysis of bubbles has not been set as the main goal of the experimental facility.

  1. Effect of heated length on the Critical Heat Flux of subcooled flow boiling. 2. Effective heated length under axially nonuniform heating condition

    International Nuclear Information System (INIS)

    Kinoshita, Hidetaka; Yoshida, Takuya; Nariai, Hideki; Inasaka, Fujio

    1998-01-01

    Effect of heated length on the Critical Heat Flux (CHF) of subcooled flow boiling with water was experimentally investigated by using direct current heated tube made of stainless steel a part of whose wall thickness was axially cut for realizing nonuniform heat flux condition. The higher enhancement of the CHF was derived for shorter tube length. The effective heated length was determined for the tube under axially nonuniform heat flux condition. When the lower heat flux part below the Net Vapor Generation (NVG) heat flux exists at the middle of tube length, then the effective heated length becomes the tube length downstream the lower heat flux parts. However, when the lower heat flux part is above the NVG, then the effective heated length is full tube length. (author)

  2. 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

  3. 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

  4. Void Measurements in the Regions of Sub-Cooled and Low-Quality Boiling. Part 2. Higher Mass Velocities

    Energy Technology Data Exchange (ETDEWEB)

    Rouhani, S Z

    1966-07-15

    This report consists mostly of tables of experimental data obtained in void measurements. It is a continuation and the completing part of a previous report with the same title. The data are from the measurements in a vertical annular channel with 25 mm O.D. and 12 mm I.D. at a heated length of 1090 mm. These experiments covered pressures from 10 to 50 bars, mass velocities from 650 to 1450 kg/m -sec., heat fluxes from 60 to 120 W/cm{sup 2}, sub-coolings from 30 to 0 C, and steam qualities from 0 to 12 %. The tables include the inlet temperatures and measured wall super-heat.

  5. 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)

  6. 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

  7. Boiling detection using signals of self-powered neutron detectors and thermocouples

    International Nuclear Information System (INIS)

    Kozma, R.

    1989-01-01

    A specially-equipped simulated fuel assembly has been placed into the core of the 2 MW research reactor of the IRI, Delft. In this paper the recent results concerning the detection of coolant boiling in the simulated fuel assembly are introduced. Applying the theory of boiling temperature noise, different stages of boiling, i.e. one-phase flow, subcooled boiling, volume boiling, were identified in the measurements using the low-frequency noise components of the thermocouple signals. It has been ascertained that neutron noise spectra remained unchanged when subcooled boiling appeared, and that they changed reasonably only when developed volume boiling took place in the channels. At certain neutron detector positions neutron spectra did not vary at all, although developed volume boiling occurred at a distance of 3-4 cm from these neutron detectors. This phenomenon was applied in studying the field-of-view of neutron detectors

  8. 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.

  9. Boiling heat transfer modern developments and advances

    CERN Document Server

    Lahey, Jr, RT

    2013-01-01

    This volume covers the modern developments in boiling heat transfer and two-phase flow, and is intended to provide industrial, government and academic researchers with state-of-the-art research findings in the area of multiphase flow and heat transfer technology. Special attention is given to technology transfer, indicating how recent significant results may be used for practical applications. The chapters give detailed technical material that will be useful to engineers and scientists who work in the field of multiphase flow and heat transfer. The authors of all chapters are members of the

  10. A development of two-fluid multifield model for low-quality boiling transition simulations

    International Nuclear Information System (INIS)

    Park, J.W.; Choi, H.B.

    1998-09-01

    A three-dimensional two-fluid model has been developed using ensemble-averaging techniques. The two-fluid model was closed for two-phase bubbly flows using cell averaging which accounted for the dispersed phase distribution in the region of the averaging volume. The phasic interfacial momentum exchange includes the surface stress developed on the interface which is induced by the relative motion of the phases. Since no direct mean for validating the interfacial pressure model is available, the void wae data has been used. Since the presented model has been rigorously constitute for the bubbly two-phase flow of spherical bubbles, dilute two-phase flow situations, such as the subcooled boiling, can be realistically simulated by the presented local instantaneous form of the average equations. Finally, this model should be able to predict local thermal-hydraulic conditions under which the critical heat flux occurs. (author). 25 refs., 6 figs

  11. In-situ Monitoring of Sub-cooled Nucleate Boiling on Fuel Cladding Surface in Water at 1 bar and 130 bars using Acoustic Emission Method

    Energy Technology Data Exchange (ETDEWEB)

    Baek, Seung Heon; Wu, Kaige; Shim, Hee-Sang; Lee, Deok Hyun; Hur, Do Haeng [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Crud deposition increases through a sufficient corrosion product supply around the steam-liquid interface of a boiling bubble. Therefore, the understanding of this SNB phenomenon is important for effective and safe operation of nuclear plants. The experimental SNB studies have been performed in visible conditions at a low pressure using a high speed video camera. Meanwhile, an acoustic emission (AE) method is an on-line non-destructive evaluation method to sense transient elastic wave resulting from a rapid release of energy within a dynamic process. Some researchers have investigated boiling phenomena using the AE method. However, their works were performed at atmospheric pressure conditions. Therefore, the objective of this work is for the first time to detect and monitor SNB on fuel cladding surface in simulated PWR primary water at 325 .deg. C and 130 bars using an AE technique. We successfully observed the boiling AE signals in primary water at 1 bar and 130 bars using AE technique. Visualization test was performed effectively to identify a correlation between water boiling phenomenon and AE signals in a transparent glass cell at 1 bar, and the boiling AE signals were in good agreement with the boiling behavior. Based on the obtained correlations at 1 bar, the AE signals obtained at 130 bars were analyzed. The boiling density and size of the AE signals at 130 bars were decreased by the flow parameters. However, overall AE signals showed characteristics and a trend similar to the AE signals at 1 bar. This indicates that boiling AE signals are detected successfully at 130 bars, and the AE technique can be effectively implemented in non-visualized condition at high pressures.

  12. Development of an experimental apparatus for boiling analysis

    International Nuclear Information System (INIS)

    Castro, A.J.A. de.

    1984-04-01

    The nucleate boiling is the most interesting boiling regime for practical appliccations, including nuclear reactor engineering. such regime is characterized by very high heat transfer rates with only small surface superheating. An experimental apparatus is developed for studying parameters which affect nucleate boiling. The following parameters are analysed: pressure, fluid velocity and the fluid temperature at the test section entrance. The performance of experimental apparatus is analysed by results and by problems raised by the oeration of setup. (Author) [pt

  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. Correlations for developing film boiling effect in tubes

    International Nuclear Information System (INIS)

    Guo, Y.; Leung, L.K.H.

    2005-01-01

    Full text of publication follows: Reducing uncertainties in predicting film-boiling heat transfer can provide improved margins in reactor safety analysis, hence improved operating margins in nuclear power plants. Most reactor safety codes employed the tube-based prediction method for the fully developed film-boiling heat transfer coefficient. This approach tends to underpredict the heat-transfer coefficient and over-predict the sheath temperature at post-dryout conditions close to the CHF point. The under-prediction is due mainly to the droplet impingement on the heated surface and vapour superheating. This heat-transfer regime is referred to as the developing film boiling, which is associated with an enhancement in heat transfer compared to the fully developed film boiling. An improvement in the prediction accuracy is achievable by accounting for the effect of vapour-film development on film boiling heat transfer. In addition to system safety analyses, the prediction of developing film boiling heat transfer is required in subchannel analyses for fuel bundles. A tube-data-based prediction method is particularly relevant for subchannel applications. The objective of this study is to derive a correlation for the developing film boiling effect in tubes. The current CANDU R . system safety and subchannel analyses codes apply the look-up table approach to predict the film boiling heat transfer. The post-dryout look-up table provides the fully developed film boiling heat transfer in an 8-mm vertical tube, and has been extended to other tube sizes using a diameter modification factor. In this study, a modification factor has been developed to account for the developing film-boiling effect, and is expressed in the following non-dimensional form: K = (h FB - h FD )/(h NB - h FD ) = f ((T W - T sat )/T CHF - T sat )) where h FB is the film boiling heat transfer coefficient, h FD is the fully developed film-boiling heat transfer coefficient, which is evaluated using the film-boiling

  15. Condensation of vapor bubble in subcooled pool

    Science.gov (United States)

    Horiuchi, K.; Koiwa, Y.; Kaneko, T.; Ueno, I.

    2017-02-01

    We focus on condensation process of vapor bubble exposed to a pooled liquid of subcooled conditions. Two different geometries are employed in the present research; one is the evaporation on the heated surface, that is, subcooled pool boiling, and the other the injection of vapor into the subcooled pool. The test fluid is water, and all series of the experiments are conducted under the atmospheric pressure condition. The degree of subcooling is ranged from 10 to 40 K. Through the boiling experiment, unique phenomenon known as microbubble emission boiling (MEB) is introduced; this phenomenon realizes heat flux about 10 times higher than the critical heat flux. Condensation of the vapor bubble is the key phenomenon to supply ambient cold liquid to the heated surface. In order to understand the condensing process in the MEB, we prepare vapor in the vapor generator instead of the evaporation on the heated surface, and inject the vapor to expose the vapor bubble to the subcooled liquid. Special attention is paid to the dynamics of the vapor bubble detected by the high-speed video camera, and on the enhancement of the heat transfer due to the variation of interface area driven by the condensation.

  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. Experiments on nucleate boiling heat transfer with a highly-wetting dielectric fluid

    International Nuclear Information System (INIS)

    You, S.M.; Simon, T.W.; Bar-Cohen, A.

    1990-01-01

    This paper reports on experiments on pool boiling heat transfer in an electronic cooling fluid (Fluorinert, FC-72) that were conducted using a 0.51 mm diameter cylindrical heater. The effects of pressure, subcooling and dissolved gas content on nucleate boiling heat transfer are investigated. When boiling with dissolved gas in the bulk fluid, the fluid in the vicinity of the heating element appears to be liberated of dissolved gas by boiling. Thus, boiling under these conditions appears to be similar to subcooled boiling without dissolved gas. Nucleate boiling hysteresis is observed for subcooled and gassy-subcooled situations

  18. Vapor Explosions with Subcooled Freon

    International Nuclear Information System (INIS)

    Henry, R.E.; Fauske, Hans K.; McUmber, L.M.

    1976-01-01

    Explosive vapor formation accompanied by destructive shock waves, can be produced when two liquids, at much different temperatures, are brought into intimate contact. A proposed analytical model states that the interface temperature upon contact between the two liquid systems, gust be greater than or equal to the spontaneous nucleation temperature of that liquid-liquid system and that the thermal boundary layer must be sufficiently developed to support a critical size cavity. For time scales greater than 10-12 sec, the interface temperature upon contact of two semi-infinite masses, with constant thermal properties, can be related to the initial liquid temperatures. The spontaneous nucleation behavior at the interface can either be heterogeneous or homogeneous in nature. In either case, the critical size cavities, which initiate the vaporization process, are produced by local density fluctuations within the cold liquid. For homogeneous conditions, the two liquids present a well-wetted system and the vapor embryos are produced entirely within the cold liquid. For heterogeneous conditions, which result from poor, or imperfect wetting, at the liquid-liquid interface, the critical sized cavities are created at the interface at somewhat lower temperatures. A sequence of experiments, using Freon-22 and water, Freon-22 and mineral oil, and Freon-12 and mineral oil have been performed to test this spontaneous nucleation premise. For Freon-22 at its normal boiling point, the interface temperature of the water must be at least 77 deg. C before the interface temperature equals or exceeds the minimum homogeneous nucleation value of 54 deg. C and 84 deg. C before the interface temperature equals 60 deg. C where the homogeneous nucleation rate becomes truly explosive. The Freon-water test demonstrated explosive interactions for water temperatures considerably lower than this value and this was attributed to the heterogeneous nucleation characteristics of that particular system

  19. Computational fluid dynamics and population balance modelling of nucleate boiling of cryogenic liquids: Theoretical developments

    Directory of Open Access Journals (Sweden)

    Guan Heng Yeoh

    2016-12-01

    Full Text Available The main focus in the analysis of pool or flow boiling in saturated or subcooled conditions is the basic understanding of the phase change process through the heat transfer and wall heat flux partitioning at the heated wall and the two-phase bubble behaviours in the bulk liquid as they migrate away from the heated wall. This paper reviews the work in this rapid developing area with special reference to modelling nucleate boiling of cryogenic liquids in the context of computational fluid dynamics and associated theoretical developments. The partitioning of the wall heat flux at the heated wall into three components – single-phase convection, transient conduction and evaporation – remains the most popular mechanistic approach in predicting the heat transfer process during boiling. Nevertheless, the respective wall heat flux components generally require the determination of the active nucleation site density, bubble departure diameter and nucleation frequency, which are crucial to the proper prediction of the heat transfer process. Numerous empirical correlations presented in this paper have been developed to ascertain these three important parameters with some degree of success. Albeit the simplicity of empirical correlations, they remain applicable to only a narrow range of flow conditions. In order to extend the wall heat flux partitioning approach to a wider range of flow conditions, the fractal model proposed for the active nucleation site density, force balance model for bubble departing from the cavity and bubble lifting off from the heated wall and evaluation of nucleation frequency based on fundamental theory depict the many enhancements that can improve the mechanistic model predictions. The macroscopic consideration of the two-phase boiling in the bulk liquid via the two-fluid model represents the most effective continuum approach in predicting the volume fraction and velocity distributions of each phase. Nevertheless, the

  20. Cavitational boiling of liquids

    International Nuclear Information System (INIS)

    Kostyuk, V.V.; Berlin, I.I.; Borisov, N.N.; Karpyshev, A.V.

    1986-01-01

    Transition boiling is a term usually denoting the segment of boiling curve 1-2, where the heat flux, q, decreases as the temperature head, ΔT/sub w/=T/sub w/-T/sub s/, increases. Transition boiling is the subject of numerous papers. Whereas most researchers have studied transition boiling of saturated liquids the authors studied for many years transition boiling of liquids subcooled to the saturation temperature. At high values of subcooling, ΔT/sub sub/=T/sub s/-T/sub 1/, an anomalous dependence of the heat flux density on the temperature head was detected. Unlike a conventional boiling curve, where a single heat flux maximum occurs, another maximum is seen in the transition boiling segment, the boiling being accompanied by strong noise. The authors refer to this kind of boiling as cavitational. This process is largely similar to noisy boiling of helium-II. This article reports experimental findings for cavitational boiling of water, ethanol, freon-113 and noisy boiling of helium-II

  1. Development, implementation and assessment of specific, two-fluid closure laws for inverted-annular film-boiling

    Energy Technology Data Exchange (ETDEWEB)

    Cachard, F. de [Laboratory for Thermal Hydraulics, Villigen (Switzerland)

    1995-09-01

    Inverted-Annular Film-Boiling (IAFB) is one of the post-burnout heat transfer modes taking place during the reflooding phase of the loss-of-coolant accident, when the liquid at the quench front is subcooled. Under IAFB conditions, a continuous, liquid core is separated from the wall by a superheated vapour film. the heat transfer rate in IAFB is influenced by the flooding rate, liquid subcooling, pressure, and the wall geometry and temperature. These influences can be accounted by a two-fluid model with physically sound closure laws for mass, momentum and heat transfers between the wall, the vapour film, the vapour-liquid interface, and the liquid core. Such closure laws have been developed and adjusted using IAFB-relevant experimental results, including heat flux, wall temperature and void fraction data. The model is extensively assessed against data from three independent sources. A total of 46 experiments have been analyzed. The overall predictions are good. The IAFB-specific closure laws proposed have also intrinsic value, and may be used in other two-fluid models. They should allow to improve the description of post-dryout, low quality heat transfer by the safety codes.

  2. 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)

  3. Fiscal 2000 project of inviting proposals for international joint research - invitation for international proposal (Energy conservation No.3). Achievement report on development of micro-scale boiling aided high efficiency heat removing device; 2000 nendo kokusai kyodo kenkyu teian kobo jigyo - kokusai teian kobo (shoe No.3). Micro scale boiling ni yoru kokoritsu honetsu device no kaihatsu seika hokokusho

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-03-01

    Studies are conducted about basic matters of heat transfer with boiling, such as critical heat flux intensification, prevention of dry-out, and the development of refrigerants suitable for use for heat transfer with boiling, for the purpose of developing boiling heat conduction type high-efficiency heat removing devices for use in electronics, and then heat removing devices usable as power devices in the future are experimentally designed. Activities are conducted in the three fields of (1) the study of basic micro-boiling technology, (2) development of micro-scale boiling element technology, and (3) international joint studies. Efforts are made to develop the technology of removing heat from ultrahigh heat fluxes using a micro-valve in field (1), to develop the technology of heat transfer by boiling in a micro-channel in field (2); and to develop the technology of critical heat flux intensification in a boiling heat exchanger in an electromagnetic field (3). In an effort to develop the technology of heat removal, a heat transfer plate is installed at the bottom of a path which is narrow, horizontal, and rectangular, and distilled water is allowed to flow and boil. Micro-bubble emission boiling occurs by a subcooling degree of 40K at an average flow rate of 0.5 m/s, and an ultrahigh heat flux of 2-7 times 10{sup 6} W/m{sup 2} is obtained. The value is 2-4 times as high as the current IC chip critical heat flux. (NEDO)

  4. Experiment study of the onset of nucleate boiling in narrow annular channel

    International Nuclear Information System (INIS)

    Wang Jiaqiang; Jia Dounan; Guo Yun

    2004-01-01

    The onset of nucleate boiling (ONB) was investigated for water flowing in the annular duct which clearance is 1.2 mm at the pressure range from 1.0 to 4.5 MPa. The effect on ONB of some thermodynamics parameters was also analyzed. The available data dealing with sub-cooled boiling initial point of water in narrow annular clearance duct are analyzed by using regression method. The new developed correlation was obtained by considering the bilateral heating factor

  5. Investigation of subcooled boiling onset propagation

    International Nuclear Information System (INIS)

    Josipovic, M.; Riznic, J.; Vrhovac, M.; Spasojevic, D.

    1986-01-01

    In paper is presented a method for thermohydrodynamicaly and kinematically nonequilibrium two-phase mixture flow basic process and phenomena investigation, during chosen transient. Comparison and brief discussion of results on experimental facility KVP are included. (author)

  6. Development of high conversion boiling water reactor

    International Nuclear Information System (INIS)

    Yamashita, Jun-ichi; Mochida, Takaaki; Uchikawa, Sadao.

    1988-01-01

    It is expected that the period of LWRs being the main source of electric power supply becomes long, therefore, the development of next generation LWRs placing emphasis on the effective utilization of uranium resources and the improvement of economical efficiency is necessary. In this paper, as the next generation BWRs subsequent to ABWRs, the concept of the core of high conversion type BWRs is reported, in which emphasis is placed on the saving of natural uranium resources by raising the rate of conversion to plutonium. This core is that of realizing the high rate of conversion by utilizing the void in the core, which is the feature of BWRs, and the case of making the change of the core structure relatively small by using cross type control rods and the case of changing the core structure for further heightening the rate of conversion and making control rods into cluster type are described. In order to meet the demand like this, Hitachi Ltd. has engaged in the development of the concept of the core of next generation LWRs. In the high conversion type BWRs, there is not large change in the reactor system and turbine system from the current BWRs. The features and the concept of the core of high conversion type BWRs are described. (Kako, I.)

  7. Development of advanced boiling water reactor for medium capacity

    International Nuclear Information System (INIS)

    Kazuo Hisajima; Yutaka Asanuma

    2005-01-01

    This paper describes a result of development of an Advanced Boiling Water Reactor for medium capacity. 1000 MWe was selected as the reference. The features of the current Advanced Boiling Water Reactors, such as a Reactor Internal Pump, a Fine Motion Control Rod Drive, a Reinforced Concrete Containment Vessel, and three-divisionalized Emergency Core Cooling System are maintained. In addition, optimization for 1000 MWe has been investigated. Reduction in thermal power and application of the latest fuel reduced the number of fuel assemblies, Control Rods and Control Rod Drives, Reactor Internal Pumps, and Safety Relief Valves. The number of Main Steam lines was reduced from four to two. As for the engineered safety features, the Flammability Control System was removed. Special efforts were made to realize a compact Turbine Building, such as application of an in line Moisture Separator, reduction in the number of pumps in the Condensate and Feedwater System, and change from a Turbine-Driven Reactor Feedwater Pump to a Motor-Driven Reactor Feedwater Pump. 31% reduction in the volume of the Turbine Building is expected in comparison with the current Advanced Boiling Water Reactors. (authors)

  8. Assessment of CHF characteristics at subcooled conditions for the CANFLEX bundle

    International Nuclear Information System (INIS)

    Onder, E.N.; Leung, L.K.H.

    2013-01-01

    Boiling-Length-Average (BLA) Critical Heat Flux (CHF) values for the CANFLEX bundle at cross-sectional average subcooled conditions have been evaluated using the ASSERT-PV subchannel code. The predicted BLA CHF values supplement experimental BLA CHF values obtained with full-scale bundle simulators at saturated conditions in developing a BLA CHF correlation applicable over the interested range of cross-sectional average thermodynamic quality in regional overpower protection (ROP) trip and safety analyses. The BLA CHF correlation exhibits similar characteristics to those observed in tubes at subcooled and saturated conditions. Applying this correlation has led to similar prediction accuracy in dryout power to that using the BLA CHF-data-based correlation at saturated conditions. However, it provides improved prediction accuracy in dryout power at dryout conditions near saturation compared to the BLA CHF-data-based correlation (which tends to underpredict the dryout power)

  9. Effects of Parallel Channel Interactions, Steam Flow, Liquid Subcool ...

    African Journals Online (AJOL)

    Tests were performed to examine the effects of parallel channel interactions, steam flow, liquid subcool and channel heat addition on the delivery of liquid from the upper plenum into the channels and lower plenum of Boiling Water Nuclear Power Reactors during reflood transients. Early liquid delivery into the channels, ...

  10. Development of surface wettability characteristics for enhancing pool boiling heat transfer

    International Nuclear Information System (INIS)

    Kim, Moo Hwan; Jo, Hang Jin

    2010-05-01

    For several centuries, many boiling experiments have been conducted. Based on literature survey, the characteristic of heating surface in boiling condition played as an important role which mainly influenced to boiling performance. Among many surface factor, the fact that wettability effect is significant to not only the enhancement of critical heat flux(CHF) but also the nucleate boiling heat transfer is also supported by other kinds of boiling experiments. In this regard, the excellent boiling performance (a high CHF and heat transfer performance) in pool boiling could be achieved through some favorable surface modification which satisfies the optimized wettability condition. To find the optimized boiling condition, we design the special heaters to examine how two materials, which have different wettability (e.g. hydrophilic and hydrophobic), affect the boiling phenomena. The special heaters have hydrophobic dots on hydrophilic surface. The contact angle of hydrophobic surface is 120 .deg. to water at the room temperature. The contact angle of hydrophilic surface is 60 .deg. at same conditions. To conduct the experiment with new surface condition, we developed new fabrication method and design the pool boiling experimental apparatus. Through this facility, we can the higher CHF on pattern surface than that on hydrophobic surface, and the higher boiling heat transfer performance on pattern surface than that on hydrophilic surface. Based on this experimental results, we concluded that we proposed new heating surface condition and surface fabrication method to realize the best boiling condition by modified heating surface condition

  11. Domestic and overseas development of advanced boiling water reactors

    International Nuclear Information System (INIS)

    Hatazawa, Mamoru; Fuchino, Satoshi; Nakada, Kotaro

    2010-01-01

    Since Toshiba delivered the world's first advanced boiling water reactor (ABWR) to The Tokyo Electric Power Company, Inc. in 1996, we have been devoting continuous efforts to the construction and operational support of ABWR systems as major products. We are now promoting the construction of domestic and overseas ABWR systems along with the standardization of ABWRs. We are also engaged in the research and development of core technologies to support further promotion of ABWRs as a concurrent solution to the issues of global warming and energy security for individual countries. (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. 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)

  14. 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)

  15. Boiling Heat Transfer Mechanisms in Earth and Low Gravity: Boundary Condition and Heater Aspect Ratio Effects

    Science.gov (United States)

    Kim, Jungho

    2004-01-01

    Boiling is a complex phenomenon where hydrodynamics, heat transfer, mass transfer, and interfacial phenomena are tightly interwoven. An understanding of boiling and critical heat flux in microgravity environments is of importance to space based hardware and processes such as heat exchange, cryogenic fuel storage and transportation, electronic cooling, and material processing due to the large amounts of heat that can be removed with relatively little increase in temperature. Although research in this area has been performed in the past four decades, the mechanisms by which heat is removed from surfaces in microgravity are still unclear. Recently, time and space resolved heat transfer data were obtained in both earth and low gravity environments using an array of microheaters varying in size between 100 microns to 700 microns. These heaters were operated in both constant temperature as well as constant heat flux mode. Heat transfer under nucleating bubbles in earth gravity were directly measured using a microheater array with 100 m resolution operated in constant temperature mode with low and high subcooled bulk liquid along with images from below and from the side. The individual bubble departure diameter and energy transfer were larger with low subcooling but the departure frequency increased at high subcooling, resulting in higher overall heat transfer. The bubble growth for both subcoolings was primarily due to energy transfer from the superheated liquid layer relatively little was due to wall heat transfer during the bubble growth process. Oscillating bubbles and sliding bubbles were also observed in highly subcooled boiling. Transient conduction and/or microconvection was the dominant heat transfer mechanism in the above cases. A transient conduction model was developed and compared with the experimental data with good agreement. Data was also obtained with the heater array operated in a constant heat flux mode and measuring the temperature distribution across

  16. 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)

  17. Subcooled Liquid Oxygen Cryostat for Magneto-Archimedes Particle Separation by Density

    Science.gov (United States)

    Hilton, D. K.; Celik, D.; Van Sciver, S. W.

    2008-03-01

    An instrument for the separation of particles by density (sorting) is being developed that uses the magneto-archimedes effect in liquid oxygen. With liquid oxygen strongly paramagnetic, the magneto-archimedes effect is an extension of diamagnetic levitation in the sense of increasing the effective buoyancy of a particle. The instrument will be able to separate ensembles of particles from 100 μm to 100 nm in size, and vertically map or mechanically deliver the separated particles. The instrument requires a column of liquid oxygen that is nearly isothermal, free of thermal convection, subcooled to prevent nucleate boiling, and supported against the strong magnetic field used. Thus, the unique cryostat design that meets these requirements is described in the present article. It consists in part of a column of liquid nitrogen below for cooling the liquid oxygen, with the liquid oxygen pressurized by helium gas to prevent nucleate boiling.

  18. Modification and validation of the natural heat convection and subcooled void formation models in the code PARET

    International Nuclear Information System (INIS)

    Hainoun, A.; Alhabit, F.; Ghazi, N.

    2008-01-01

    Two new modifications have been included in the current PARET code that is widely applied in the dynamic and safety analysis of research reactors. A new model was implemented for the simulation of void formation in the subcooled boiling regime, the other modification dealt with the implementation of a new approach to improve the prediction of heat transfer coefficient under natural circulation condition. The modified code was successfully validated using adequate single effect tests covering the physical phenomena of interest for both natural circulation and subcooled void formation at low pressure and low heat flux. The validation results indicate significant improvement of the code compared to the default version. Additionally, to simplify the code application an interactive user interface was developed enabling pre and post-processing of the code predictions. (author)

  19. Evaluation of subcooled CHF correlations using the PU-BTPFL CHF database for vertical upflow of water in a uniformly heated round tube

    International Nuclear Information System (INIS)

    Hall, D.D.; Mudawar, I.

    1996-01-01

    A simple methodology for assessing the predictive ability of critical heat flux (CHF) correlations applicable to subcooled flow boiling in a uniformly heated vertical tube was developed. Popular correlations published in handbooks and review articles as well as the most recent correlations were analyzed with a database compiled by the authors. The PU-BTPFL CHF Database, which contains 29,718 CHF data points, is the largest collection of CHF data ever cited in the world literature. The parametric ranges of the CHF database are diameters from 0.3 to 45 mm, length-to-diameter ratios from 2 to 2484, mass velocities from 0.01 x 10 3 to 138 x 10 3 kg m -2 s -1 , pressures from 1 to 223 bars, inlet subcoolings from 0 to 347 C, inlet qualities from -2.63 to 0.00, outlet subcoolings from 0 to 305 C, outlet qualities from -2.13 to 1.00, and critical heat fluxes from 0.05 x 10 6 to 276 x 10 6 W m -2 . The database contained 4357 data points having a subcooled outlet condition at CHF. The correlation published in Caira et al. (1993) was the most accurate in both low and high mass velocity regions having been developed with a larger database than most correlations. In general, CHF correlations developed from data covering a limited range of flow conditions can not be extended to other flow conditions without much uncertainty. Subcooled flow boiling has great potential for accommodating the high heat fluxes in such diverse applications as fusion and fission reactors, manufacturing and materials processing, advanced space thermal management systems, accelerator targets, avionic cold plates, X-ray anodes, and high-density multi-chip modules in supercomputers and other modular electronics

  20. Characteristic of onset of nucleate boiling in natural circulation

    International Nuclear Information System (INIS)

    Zhou Tao; Yang Ruichang; Liu Ruolei

    2006-01-01

    Two kinds of thermodynamics quality at onset of nucleate boiling with sub-cooled boiling were calculated for force circulation by using Bergles and Rohesenow method or Davis and Anderson method, and natural circulation by using Tsinghua University project group's empirical equations suggested in our natural circulation experiment at same condition. The characteristic of onset of nucleate boiling with subcooled boiling in natural circulation were pointed out. The research result indicates that the thermodynamics quality at onset of nucleate boiling with subcooled boiling in natural circulation is more sensitive for heat and inlet temperature and system pressure. Producing of onset of nucleate boiling with subcooled boiling is early at same condition. The research result also indicates more from microcosmic angle of statistical physics that the phenomena are caused by the effects of characteristic of dissipative structure of natural circulation in self organization, fluctuation force and momentum force of dynamics on thermodynamics equilibrium. these can lay good basis for study and application on sub-cooled boiling in natural circulation in future. (authors)

  1. Development of NUFREQ-N, an analytical model for the stability analysis of nuclear coupled density-wave oscillations in boiling water nuclear reactors

    International Nuclear Information System (INIS)

    Park, G.C.

    1983-01-01

    A state-of-the-art one-dimensional thermal-hydraulic model has been developed to be used for the linear analysis of nuclear-coupled density-wave oscillations in a boiling water nuclear reactor (BWR). The model accounts for phasic slip, distributed spacers, subcooled boiling, space/time-dependent power distributions and distributed heated wall dynamics. In addition to a parallel channel stability analysis, a detailed model was derived for the BWR loop analysis of both the natural and forced circulation modes of operation. In its final form, this model constitutes a multi-input, multi-output (MIMO) linear system, which features a general nodal neutron kinetics model. Kinetics parameters for use in the kinetics model have been obtained by utilizing self-consistent nodal data and power distributions. The stability characteristics of a typical BWR/4 has been investigated with the Nyquist criterion. The computer implementation of this mode, NUFREQ-N, was used for the parametric study of a typical BWR/4 and comparison were made with existing in-core and out-of-core data. Also, NUFREQ-N was used to analyze the expected stability characteristics of a typical BWR/4. The parametric results revealed important factors influencing BWR stability margin. It was found that NUFREQ-N generally agreed well with out-of-core data. This was especially true for the predicted power-to-flow transfer function, which is the most important transfer function in thermal-hydraulic stability analysis

  2. 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)

  3. 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.

  4. Evaluation of subcooled critical heat flux correlations using the PU-BTPFL CHF database for vertical upflow of water in a uniformly heated round tube

    International Nuclear Information System (INIS)

    Hall, D.D.; Mudawar, I.

    1997-01-01

    A simple methodology for assessing the predictive ability of critical heat flux (CHF) correlations applicable to subcooled flow boiling in a uniformly heated vertical tube is developed. Popular correlations published in handbooks and review articles as well as the most recent correlations are analyzed with the PU-BTPFL CHF database, which contains 29,718 CHF data points. This database is the largest collection of CHF data (vertical upflow of water in a uniformly heated round tube) ever cited in the world literature. The parametric ranges of the CHF database are diameters from 0.3 to 45 mm, length-to-diameter ratios from 2 to 2484, mass velocities from 0.01 x 10 3 to 138 x 10 3 kg/m 2 ·s, pressures from 1 to 223 bars, inlet subcoolings from 0 to 347 C, inlet qualities from -2.63 to 0.00, outlet subcoolings from 0 to 305 C, outlet qualities from -2.13 to 1.00, and CHFs from 0.05 x 10 6 to 276 x 10 6 W/m 2 . The database contains 4,357 data points having a subcooled outlet condition at CHF. A correlation published elsewhere is the most accurate in both low- and high-mass velocity regions, having been developed with a larger database than most correlations. In general, CHF correlations developed from data covering a limited range of flow conditions cannot be extended to other flow conditions without much uncertainty

  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. Integrating artificial neural networks and empirical correlations for the prediction of water-subcooled critical heat flux

    International Nuclear Information System (INIS)

    Mazzola, A.

    1997-01-01

    The critical heat flux (CHF) is an important parameter for the design of nuclear reactors, heat exchangers and other boiling heat transfer units. Recently, the CHF in water-subcooled flow boiling at high mass flux and subcooling has been thoroughly studied in relation to the cooling of high-heat-flux components in thermonuclear fusion reactors. Due to the specific thermal-hydraulic situation, very few of the existing correlations, originally developed for operating conditions typical of pressurized water reactors, are able to provide consistent predictions of water-subcooled-flow-boiling CHF at high heat fluxes. Therefore, alternative predicting techniques are being investigated. Among these, artificial neural networks (ANN) have the advantage of not requiring a formal model structure to fit the experimental data; however, their main drawbacks are the loss of model transparency ('black-box' character) and the lack of any indicator for evaluating accuracy and reliability of the ANN answer when 'never-seen' patterns are presented. In the present work, the prediction of CHF is approached by a hybrid system which couples a heuristic correlation with a neural network. The ANN role is to predict a datum-dependent parameter required by the analytical correlation; ; this parameter was instead set to a constant value obtained by usual best-fitting techniques when a pure analytical approach was adopted. Upper and lower boundaries can be possibly assigned to the parameter value, thus avoiding the case of unexpected and unpredictable answer failure. The present approach maintains the advantage of the analytical model analysis, and it partially overcomes the 'black-box' character typical of the straight application of ANNs because the neural network role is limited to the correlation tuning. The proposed methodology allows us to achieve accurate results and it is likely to be suitable for thermal-hydraulic and heat transfer data processing. (author)

  7. Development, implementation and assessment of specific closure laws for inverted-annular film-boiling in a two-fluid model

    International Nuclear Information System (INIS)

    Cachard, F. de

    1994-10-01

    Inverted-annular film-boiling (IAFB) is one of the post-burnout heat transfer modes taking place, in particular, during the reflooding phase of the loss-of-coolant accident, when the liquid at the quench front is subcooled. Under IAFB conditions, a continuous liquid core is separated from the wall by a superheated vapour film. The heat transfer rate in IAFB is influenced by the flooding rate, liquid subcooling, pressure, and the wall geometry and temperature. These influences can be accounted by a two-fluid model with physically sound closure laws for mass, momentum and heat transfer between the wall, the vapour film, the vapour-liquid interface, and the liquid core. The applicability of existing IAFB two-fluid models is limited. This is attributed to shortcomings in the description of heat transfer within the liquid core, to use of certain correlations outside their validity range, and to a limited use of experimental information on IAFB. The usual approach has been to develop models employing generally applicable closure laws including, however, adjustable parameters, and to adjust these using global experimental results. The present approach has been to develop IAFB-specific closure laws in such a form that they could be adjusted separately using detailed, IAFB-relevant, experimental result. Steady-state results, including heat flux, wall temperature and void fraction data have been used for the adjustment. A key issue in IAFB modeling is to predict how the heat flux reaching the vapour-liquid interface is split into a liquid heating term and a vaporization term. In the model proposed, convective liquid heating is related to the liquid velocity relative to the interface, and not to the absolute liquid velocity, as in previous models. This relative velocity is deduced from the interfacial shear stress, using the liquid-interface friction law. With this modification, the prediction of the experimental trends is greatly improved. (author) figs., tabs., refs

  8. 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)

  9. Boiling hysteresis of impinging circular submerged jets with highly wetting liquids

    International Nuclear Information System (INIS)

    Zhou, D.W.; Ma, C.F.; Yu, J.

    2004-01-01

    An experimental study was carried out to characterize the boiling hysteresis of impinging circular submerged jets with highly wetting liquids. The effects of noncondensable gases and surface aging on boiling curves were considered. The present study focused on the effects of jet parameters (jet exit velocity, radial distance from the stagnation point and nozzle diameter) and fluid subcooling on incipient boiling superheat and superheat excursion, as well as the physical mechanism of boiling hysteresis. Results show that the incipient boiling superheat decreases only with fluid subcooling regardless of jet parameters, and that the superheat excursion increases with nozzle diameter and radial distance from the stagnation point and decreasing jet exit velocity and fluid subcooling. Boiling hysteresis occurs due to deactivation of vapor embryos within larger cavities. Three anomalous phenomena at boiling inception are recorded and discussed in terms of irregular activation of vapor embryos

  10. Study on the effect of subcooling on vapor film collapse on high temperature particle surface

    International Nuclear Information System (INIS)

    Abe, Yutaka; Tochio, Daisuke; Yanagida, Hiroshi

    2000-01-01

    Thermal detonation model is proposed to describe vapor explosion. According to this model, vapor film on pre-mixed high temperature droplet surface is needed to be collapsed for the trigger of the vapor explosion. It is pointed out that the vapor film collapse behavior is significantly affected by the subcooling of low temperature liquid. However, the effect of subcooling on micro-mechanism of vapor film collapse behavior is not experimentally well identified. The objective of the present research is to experimentally investigate the effect of subcooling on micro-mechanism of film boiling collapse behavior. As the results, it is experimentally clarified that the vapor film collapse behavior in low subcooling condition is qualitatively different from the vapor film collapse behavior in high subcooling condition. In case of vapor film collapse by pressure pulse, homogeneous vapor generation occurred all over the surface of steel particle in low subcooling condition. On the other hand, heterogeneous vapor generation was observed for higher subcooling condition. In case of vapor film collapse spontaneously, fluctuation of the gas-liquid interface after quenching propagated from bottom to top of the steel particle heterogeneously in low subcooling condition. On the other hand, simultaneous vapor generation occurred for higher subcooling condition. And the time transient of pressure, particle surface temperature, water temperature and visual information were simultaneously measured in the vapor film collapse experiment by external pressure pulse. Film thickness was estimated by visual data processing technique with the pictures taken by the high-speed video camera. Temperature and heat flux at the vapor-liquid interface were estimated by solving the heat condition equation with the measured pressure, liquid temperature and vapor film thickness as boundary conditions. Movement of the vapor-liquid interface were estimated with the PIV technique with the visual observation

  11. The development of fuel elements for boiling water reactors

    International Nuclear Information System (INIS)

    Holzer, R.; Kilian, P.

    1984-01-01

    The longevity of today's standard fuel elements constitutes a sound basis for designing advanced fuel elements for higher discharge burnups. Operating experience as well as postirradiation examinations of discharged fuel elements indicate that the technical limits have not reached by far. However, measures to achieve an economic and reliable fuel cycle are not restricted to the design of fuel elements, but also extend into such fields as fuel management and the mode of reactor operation. Fuel elements can be grouped together in zones in the core as a function of burnup and reactivity. The loading scheme can be aligned to this approach by concentrating on typical control rod positions. Reloads can also be made up of two sublots of fuel elements with different gadolinium contents. Longer cycles, e.g., of eighteen instead of twelve months, are easy to plan reactivitywise by increasing the quantity to be replaced from at present one quarter to one third. In fuel elements designed for higher burnups, the old scheme of reloading one quarter of the fuel inventory can be retained. The measures already introduced or in the planning stage incorporate a major potential for technical and economic optimization of the fuel cycle in boiling water reactors. (orig.) [de

  12. A study on the correlations development for film boiling heat transfer on spheres

    International Nuclear Information System (INIS)

    Jeong, Yong Hoon; Baek, Won Pil; Chang, Soon Heung

    1998-01-01

    Film boiling is the heat transfer mechanism that can occurs when large temperature differences exist between a cold liquid and hot material. In the nuclear reactor safety analysis, film boiling has become an important issue in recent years. During severe accident, hot molten corium fall into relatively cool water, and fragment into spheres or sphere-like particles. If the steam explosion is triggered, the thermal energy of corium is converted into the mechanical energy that can threaten the integrity of reactor vessel or reactor cavity. One of the important concerns in the heat transfer analysis during pre-mixing stage is the film boiling heat transfer between the corium and water/steam two-phase flow. Until now, considerable works on film boiling have been performed. However, there is no available correlation adequate for severe accident analysis. In this study, film boiling heat transfer correlations have been developed, and their applicable ranges have been enlarged and their prediction accuracy has been enhanced

  13. A study on the correlations development for film boiling heat transfer on spheres

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Hoon; Baek, Won Pil; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1999-12-31

    Film boiling is the heat transfer mechanism that can occurs when large temperature differences exist between a cold liquid and hot material. In the nuclear reactor safety analysis, film boiling has become an important issue in recent years. During severe accident, hot molten corium fall into relatively cool water, and fragment into spheres or sphere-like particles. If the steam explosion is triggered, the thermal energy of corlium is converted into the mechanical energy that can threaten the integrity of reactor vessel or reactor cavity. One of the important concerns in the heat transfer analysis during pre-mixing stage is the film boiling heat transfer between the corium and water/steam two-phase flow. Until now, considerable works on film boiling have been performed. However, there is no available correlation adequate for severe accident analysis. In this study, film boiling heat transfer correlations have been developed, and their applicable ranges have been enlarged and their prediction accuracy has been enhanced. 7 refs., 5 figs., 5 tabs. (Author)

  14. A study on the correlations development for film boiling heat transfer on spheres

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Yong Hoon; Baek, Won Pil; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1998-12-31

    Film boiling is the heat transfer mechanism that can occurs when large temperature differences exist between a cold liquid and hot material. In the nuclear reactor safety analysis, film boiling has become an important issue in recent years. During severe accident, hot molten corium fall into relatively cool water, and fragment into spheres or sphere-like particles. If the steam explosion is triggered, the thermal energy of corlium is converted into the mechanical energy that can threaten the integrity of reactor vessel or reactor cavity. One of the important concerns in the heat transfer analysis during pre-mixing stage is the film boiling heat transfer between the corium and water/steam two-phase flow. Until now, considerable works on film boiling have been performed. However, there is no available correlation adequate for severe accident analysis. In this study, film boiling heat transfer correlations have been developed, and their applicable ranges have been enlarged and their prediction accuracy has been enhanced. 7 refs., 5 figs., 5 tabs. (Author)

  15. Nucleate Boiling Heat Transfer Studied Under Reduced-Gravity Conditions

    Science.gov (United States)

    Chao, David F.; Hasan, Mohammad M.

    2000-01-01

    Boiling is known to be a very efficient mode of heat transfer, and as such, it is employed in component cooling and in various energy-conversion systems. In space, boiling heat transfer may be used in thermal management, fluid handling and control, power systems, and on-orbit storage and supply systems for cryogenic propellants and life-support fluids. Recent interest in the exploration of Mars and other planets and in the concept of in situ resource utilization on the Martian and Lunar surfaces highlights the need to understand how gravity levels varying from the Earth's gravity to microgravity (1g = or > g/g(sub e) = or > 10(exp -6)g) affect boiling heat transfer. Because of the complex nature of the boiling process, no generalized prediction or procedure has been developed to describe the boiling heat transfer coefficient, particularly at reduced gravity levels. Recently, Professor Vijay K. Dhir of the University of California at Los Angeles proposed a novel building-block approach to investigate the boiling phenomena in low-gravity to microgravity environments. This approach experimentally investigates the complete process of bubble inception, growth, and departure for single bubbles formed at a well-defined and controllable nucleation site. Principal investigator Professor Vijay K. Dhir, with support from researchers from the NASA Glenn Research Center at Lewis Field, is performing a series of pool boiling experiments in the low-gravity environments of the KC 135 microgravity aircraft s parabolic flight to investigate the inception, growth, departure, and merger of bubbles from single- and multiple-nucleation sites as a function of the wall superheat and the liquid subcooling. Silicon wafers with single and multiple cavities of known characteristics are being used as test surfaces. Water and PF5060 (an inert liquid) were chosen as test liquids so that the role of surface wettability and the magnitude of the effect of interfacial tension on boiling in reduced

  16. Development of a new simulation code for evaluation of criticality transients involving fissile solution boiling

    International Nuclear Information System (INIS)

    Basoglu, Benan; Yamamoto, Toshihiro; Okuno, Hiroshi; Nomura, Yasushi

    1998-03-01

    In this work, we report on the development of a new computer code named TRACE for predicting the excursion characteristics of criticality excursions involving fissile solutions. TRACE employs point neutronics coupled with simple thermal-hydraulics. The temperature, the radiolytic gas effects, and the boiling phenomena are estimated using the transient heat conduction equation, a lumped-parameter energy model, and a simple boiling model, respectively. To evaluate the model, we compared our results with the results of CRAC experiments. The agreement in these comparisons is quite satisfactory. (author)

  17. 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

  18. 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)

  19. Analytical modeling of inverted annular film boiling

    International Nuclear Information System (INIS)

    Analytis, G.T.; Yadigaroglu, G.

    1987-01-01

    By employing a two-fluid formulation similar to the one used in the most recent LWR accident analysis codes, a model for the Inverted Annular Film Boiling region is developed. The conservation equations, together with appropriate closure relations are solved numerically. Successful comparisons are made between model predictions and heat transfer coefficient distributions measured in a series of single-tube reflooding experiments. Generally, the model predicts correctly the dependence of the heat transfer coefficient on liquid subcooling and flow rate; for some cases, however, heat transfer is still under-predicted, and an enhancement of the heat exchange from the liquid-vapour interface to the bulk of the liquid is required. The importance of the initial conditions at the quench front is also discussed. (orig.)

  20. Analytical modeling of inverted annular film boiling

    International Nuclear Information System (INIS)

    Analytis, G.T.; Yadigaroglu, G.

    1985-01-01

    By employing a two-fluid formulation similar to the one used in the most recent LWR accident analysis codes, a model for the Inverted Annular Film Boiling region is developed. The conservation equations, together with appropriate constitutive relations are solved numerically and successful comparisons are made between model predictions and heat transfer coefficient distributions measured in a series of single-tube reflooding experiments. The model predicts generally correctly the dependence of the heat transfer coefficient on liquid subcooling and flow rate, through, for some cases, heat transfer is still under-predicted, and an enhancement of the heat exchange from the liquid-vapour interface to the bulk of the liquid is required

  1. 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

  2. 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)

  3. Pressure drops in low pressure local boiling

    International Nuclear Information System (INIS)

    Courtaud, Michel; Schleisiek, Karl

    1969-01-01

    For prediction of flow reduction in nuclear research reactors, it was necessary to establish a correlation giving the pressure drop in subcooled boiling for rectangular channels. Measurements of pressure drop on rectangular channel 60 and 90 cm long and with a coolant gap of 1,8 and 3,6 mm were performed in the following range of parameters. -) 3 < pressure at the outlet < 11 bars abs; -) 25 < inlet temperature < 70 deg. C; -) 200 < heat flux < 700 W/cm 2 . It appeared that the usual parameter, relative length in subcooled boiling, was not sufficient to correlate experimental pressure losses on the subcooled boiling length and that there was a supplementary influence of pressure, heat flux and subcooling. With an a dimensional parameter including these terms a correlation was established with an error band of ±10%. With a computer code it was possible to derive the relation giving the overall pressure drop along the channel and to determine the local gradients of pressure drop. These local gradients were then correlated with the above parameter calculated in local conditions. 95 % of the experimental points were computed with an accuracy of ±10% with this correlation of gradients which can be used for non-uniform heated channels. (authors) [fr

  4. Theoretical and experimental studies on critical heat flux in subcooled boiling and vertical flow geometry; Badania teoretyczne i eksperymentalne kryzysu wrzenia w warunkach wrzenia przechlodzonego w przeplywie w kanale pionowym

    Energy Technology Data Exchange (ETDEWEB)

    Staron, E [Institute of Atomic Energy, Otwock-Swierk (Poland)

    1997-12-31

    Critical Heat Flux is a very important subject of interest due to design, operation and safety analysis of nuclear power plants. Every new design of the core must be thoroughly checked. Experimental studies have been performed using freon as a working fluid. The possibility of transferring of results into water equivalents has been proved. The experimental study covers vertical flow, annular geometry over a wide range of pressure, mass flow and temperature at inlet of test section. Theoretical models of Critical Heat Flux have been presented but only those which cover DNB. Computer programs allowing for numerical calculations using theoretical models have been developed. A validation of the theoretical models has been performed in accordance with experimental results. (author). 83 refs, 32 figs, 4 tabs.

  5. Theoretical and experimental studies on critical heat flux in subcooled boiling and vertical flow geometry; Badania teoretyczne i eksperymentalne kryzysu wrzenia w warunkach wrzenia przechlodzonego w przeplywie w kanale pionowym

    Energy Technology Data Exchange (ETDEWEB)

    Staron, E. [Institute of Atomic Energy, Otwock-Swierk (Poland)

    1996-12-31

    Critical Heat Flux is a very important subject of interest due to design, operation and safety analysis of nuclear power plants. Every new design of the core must be thoroughly checked. Experimental studies have been performed using freon as a working fluid. The possibility of transferring of results into water equivalents has been proved. The experimental study covers vertical flow, annular geometry over a wide range of pressure, mass flow and temperature at inlet of test section. Theoretical models of Critical Heat Flux have been presented but only those which cover DNB. Computer programs allowing for numerical calculations using theoretical models have been developed. A validation of the theoretical models has been performed in accordance with experimental results. (author). 83 refs, 32 figs, 4 tabs.

  6. Letter Report: Progress in developing EQ3/6 for modeling boiling processes

    Energy Technology Data Exchange (ETDEWEB)

    Wolery, T. J., LLNL

    1995-08-28

    EQ3/6 is a software package for geochemical modeling of aqueous systems, such as water/rock or waste/water rock. It is being developed for a variety of applications in geochemical studies for the Yucca Mountain Site Characterization Project. The present focus is on development of capabilities to be used in studies of geochemical processes which will take place in the near-field environment and the altered zone of the potential repository. We have completed the first year of a planned two-year effort to develop capabilities for modeling boiling processes. These capabilities will interface with other existing and future modeling capabilities to provide a means of integrating the effects of various kinds of geochemical processes in complex systems. This year, the software has been modified to allow the formation of a generalized gas phase in a closed system for which the temperature and pressure are known (but not necessarily constant). The gas phase forms when its formation is thermodynamically favored; that is, when the system pressure is equal to the sum of the partial pressures of the gas species as computed from their equilibrium fugacities. It disappears when this sum falls below that pressure. `Boiling` is the special case in which the gas phase which forms consists mostly of water vapor. The reverse process is then `condensation.` To support calculations of boiling and condensation, we have added a capability to calculate the fugacity coefficients of gas species in the system H{sub 2}O-CO{sub 2}-CH{sub 4}-H{sub 2},-Awe{sub 2}-N{sub 2},-H{sub 2}S-NH3. This capability at present is accurate only at relatively low pressures, but is adequate for all likely repository boiling conditions. We have also modified the software to calculate changes in enthalpy (heat) and volume functions. Next year we will be extending the boiling capability to calculate the pressure or the temperature at known enthalpy. We will also add an option for open system boiling.

  7. Multi-physical developments for safety related investigations of low moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Schlenker, Markus Thomas

    2014-12-19

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  8. Multi-physical Developments for Safety Related Investigations of Low Moderated Boiling Water Reactors

    OpenAIRE

    Schlenker, Markus Thomas

    2014-01-01

    The main objective of this dissertation is the development and optimization of a low moderated boiling water reactor (BWR) core with improved fuel utilization to be incorporated in a Gen II BWR nuclear power plant. The assessment of the new core design is done by comparing it with a full MOX BWR core design regarding neutron physical and thermal-hydraulic design and safety criteria (e.g. inherent reactivity coefficients) and different sustainability parameters (e.g. conversion ratio).

  9. 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

  10. 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)

  11. Development of alarm handling methods for boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yukiharu, Ohga; Hiroshi, Seki; Setsuo, Arita [Power and Industrial Systems R and D Div., Hitachi, Ltd., Hitachi, Ibaraki (Japan)

    1997-09-01

    A method was developed to select important alarms in two steps: first, selection is based on the physical relationship between the alarms, and second, selection is according to the initial event. An approach combining a neural network and knowledge processing was proposed to identify the event rapidly. A prototype system was evaluated in the Kashiwazaki/Kariwa-4 Nuclear Power Plant during the startup test. The evaluation test confirmed that about 30% of the alarms are selected from among the many activated alarms. The second method, dealing with presentation, supports operators in their selection and confirmation of the required information for plant operation. The method selects and offers plant information in response to plant status changes and operators` demands. The selection procedure is based on the knowledge and data as structured by the plant functional structure; i.e. a means-ends abstraction hierarchy model. A prototype system was evaluated using a BWR simulator. The results showed that appropriate information items are automatically selected according to plant status changes and information on generated alarms is presented to operators together with the related trend graph and system diagram. Answers are generated in reply to the operators` demands and operators can confirm the generated alarms on each plant function, such as systems and components. 8 refs, 10 figs, 2 tabs.

  12. Development of alarm handling methods for boiling water reactors

    International Nuclear Information System (INIS)

    Ohga Yukiharu; Seki Hiroshi; Arita Setsuo

    1997-01-01

    A method was developed to select important alarms in two steps: first, selection is based on the physical relationship between the alarms, and second, selection is according to the initial event. An approach combining a neural network and knowledge processing was proposed to identify the event rapidly. A prototype system was evaluated in the Kashiwazaki/Kariwa-4 Nuclear Power Plant during the startup test. The evaluation test confirmed that about 30% of the alarms are selected from among the many activated alarms. The second method, dealing with presentation, supports operators in their selection and confirmation of the required information for plant operation. The method selects and offers plant information in response to plant status changes and operators' demands. The selection procedure is based on the knowledge and data as structured by the plant functional structure; i.e. a means-ends abstraction hierarchy model. A prototype system was evaluated using a BWR simulator. The results showed that appropriate information items are automatically selected according to plant status changes and information on generated alarms is presented to operators together with the related trend graph and system diagram. Answers are generated in reply to the operators' demands and operators can confirm the generated alarms on each plant function, such as systems and components. 8 refs, 10 figs, 2 tabs

  13. State of development progress of advanced boiling water reactors

    International Nuclear Information System (INIS)

    Tomono, Katsuya

    1982-01-01

    Advanced BWRs being developed at present are those aiming at the improvement of reliability and safety, the reduction of radiation exposure, the improvement of operation performance and capacity ratio of plants, and the heightening of economical efficiency by concentrating the experience and excellent technology of BWR manufacturers in the world. Now in Japan, the independence with Japanese technology is possible in almost all fields of nuclear power generation, and the improvement and standardization project is in progress to obtain the steady results. However, in order to pursue the most desirable BWRs conceivable at present, five BWR manufacturers in the world organized the Advanced Engineering Team in July, 1978, and performed the feasibility study of advanced BWRs for more than one year. Tokyo Electric Power Co., Inc., evaluated the report on the results, and judged that it is desirable to advance into the next stage aiming at the practical use of advanced BWRs. For the purpose, the electric power common research on advanced BWRs has been in progress, and the A-BWR project is to be examined in the third improvement and standardization project of MITI. The main technical features such as the coolant recirculation system of internal pump type, reinforced concrete containment vessels, fine motion control rod drive, improved core and fuel and others are explained. (Kako, I.)

  14. 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.

  15. Numerical simulation on the explosive boiling phenomena on the surface of molten metal

    International Nuclear Information System (INIS)

    Chen Deqi; Peng Cheng; Wang Qinghua; Pan Liangming

    2014-01-01

    In this paper, numerical simulation was carried out to investigate the explosive boiling phenomenon on high temperature surface also the influence of vapor growth rate during explosive boiling, vapor condensation in sub-cooled water and the subsequent effect on flowing and heat transfer. The simulation result indicates that the steam on the molten metal surface grows with very high speed, and it pushes away the sub-cooled water around and causes severe flowing. The steam clusters which block the sub-cooled water to rewet the molten metal surface are appearing at the same time. During the growth, lifting off as well as condensation of the steam clusters, the sub-cooled water around is strongly disturbed, and obvious vortexes appear. Conversely, the vortex will influence the steam cluster detachment and cub-cooled water rewetting the metal surface. This simulation visually displays the complex explosive boiling phenomena on the molten metal surface with high temperature. (authors)

  16. 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.

  17. 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

  18. Development and testing of high-performance fuel pin simulators for boiling experiments in liquid metal flow

    International Nuclear Information System (INIS)

    Casal, V.

    1976-01-01

    There are unknown phenomena, about local and integral boiling events in the core of sodium cooled fast breeder reactors. Therefore at GfK depend out-of-pile boiling experiments have been performed using electrically heated dummies of fuel element bundles. The success of these tests and the amount of information derived from them depend exclusively on the successful simulation of the fuel pins by electrically heated rods as regards the essential physical properties. The report deals with the development and testing of heater rods for sodium boiling experiments in bundles including up to 91 heated pins

  19. Vertical downward subcooled bubbly flow modelling with RELAP5/MOD3.2.2 gamma

    International Nuclear Information System (INIS)

    Ristevski, R.; Parzer, I.; Markov, Z.

    2000-01-01

    The presented paper will consider the correlation for void fraction distribution in the subcooled boiling flow regime of downward liquid flow at low velocities. More specifically, it will focus on the choice of the most appropriate heat and mass transfer correlation. The experimental findings and theoretical consideration of these processes and phenomena will be compared with RELAP5/MOD3.2.2 Gamma predictions. (author)

  20. 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

  1. 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

  2. Disappearance of a detached vapor mass in subcooled water

    International Nuclear Information System (INIS)

    Inada, Shigeaki; Miyasaka, Yoshiki; Izumi, Ryotaro.

    1986-01-01

    Experiments on pool transition boiling of water under atmospheric pressure on a heated surface 10 mm in diameter were conducted for subcooling 15 - 50 K. The mass flux of condensation of a detached coalescent vapor bubble was experimentally estimated by a mathematical model based on the mass transfer mechanism of condensation. As a result, it is clarified that the mass flux of condensation of the detached bubble was influenced by the initial growing velocity of a vapor bubble immediately following the detached bubble. The disappearance velocity of the detached bubble defined as a ratio of the bubble diameter at the departure to the time required until the disappearance, is in the range 0.2 to 2.0 m/sec. The disappearance velocity is proportional to the initial growing velocity of the bubble, to the square of the heat flux of the heated surface and to the cube of the wall superheat, separately. (author)

  3. Mechanistic model of the inverted annular film boiling

    International Nuclear Information System (INIS)

    Seok, Ho; Chang, Soon Heung

    1989-01-01

    An analytical model is developed to predict the heat transfer coefficient and the friction factor in the inverted annular film boiling. The developed model is based on two-fluid mass, momentum and energy balance equations and a theoretical velocity profile. The predictions of the proposed model are compared with the experimental data and the well-established correlations. For the heat transfer coefficient, they agree with the experimental data and are more promising than those of Bromely and Berenson correlations. The present model also accounts the effects of the mass flux and subcooling on the heat transfer. The friction factor predictions agree qualitatively with the experimental measurements, while some cases show a similar behavior with those of the post-CHF dispersed flow obtained from Beattie's correlation

  4. Film boiling heat transfer and vapour film collapse for various geometries

    International Nuclear Information System (INIS)

    Jouhara, H.I.; Axcell, B.P.

    2005-01-01

    increased. The minimum vapour film thickness was found to be nearly independent of water subcooling for most specimens. The sphere diameter was found to influence the heat transfer data at collapse; in contrast the length of cylinders and plates did not seem to affect results. Three types of vapour film collapse were observed on all of specimens, namely: 'explosive', 'explosive-progressive' and 'progressive'. The mode depended largely on water subcooling and the second mode has not been reported before the current investigation. The steady state film boiling investigation was conducted on a plate similar to the short plate used in the transient experiments. The heat transfer results and the types of vapour/liquid interface during the vapour film period were similar to those in the transient experiments although the heat transfer coefficient was found to be somewhat lower than in the transient experiments. The vapour film on the plate was also noted to be more stable than that in the transient experiments. A model based on Nusselt's falling film condensation theory was developed for predicting subcooled film boiling heat transfer from a flat plate and this gave heat transfer coefficients within 10% of the experimental values for water subcooling greater than 10 K and within 30% in all cases. (authors)

  5. Development of boiling transition analysis code TCAPE-INS/B based on mechanistic methods for BWR fuel bundles. Models and validations with boiling transition experimental data

    International Nuclear Information System (INIS)

    Ishida, Naoyuki; Utsuno, Hideaki; Kasahara, Fumio

    2003-01-01

    The Boiling Transition (BT) analysis code TCAPE-INS/B based on the mechanistic methods coupled with subchannel analysis has been developed for the evaluation of the integrity of Boiling Water Reactor (BWR) fuel rod bundles under abnormal operations. Objective of the development is the evaluation of the BT without using empirical BT and rewetting correlations needed for different bundle designs in the current analysis methods. TCAPE-INS/B consisted mainly of the drift-flux model, the film flow model, the cross-flow model, the thermal conductivity model and the heat transfer correlations. These models were validated systematically with the experimental data. The accuracy of the prediction for the steady-state Critical Heat Flux (CHF) and the transient temperature of the fuel rod surface after the occurrence of BT were evaluated on the validations. The calculations for the experiments with the single tube and bundles were carried out for the validations of the models incorporated in the code. The results showed that the steady-state CHF was predicted within about 6% average error. In the transient calculations, BT timing and temperature of the fuel rod surface gradient agreed well with experimental results, but rewetting was predicted lately. So, modeling of heat transfer phenomena during post-BT is under modification. (author)

  6. 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

  7. Model for boiling and dryout in particle beds

    International Nuclear Information System (INIS)

    Lipinski, R.J.

    1982-06-01

    Over the last ten years experiments and modeling of dryout in particle beds have produced over fifty papers. Considering only volume-heated beds, over 250 dryout measurements have been made, and are listed in this work. In addition, fifteen models to predict dryout have been produced and are discussed. A model is developed in this report for one-dimensional boiling and dryout in a porous medium. It is based on conservation laws for mass, momentum, and energy. The initial coupled differential equations are reduced to a single first-order differential equation with an algebraic equation for the upper boundary condition. The model includes the effects of both laminar and turbulent flow, two-phase friction, and capillary force. The boundary condition at the bed bottom includes the possibility of inflowing liquid and either an adiabatic or a bottom-cooled support structure. The top of the bed may be either channeled or subcooled. In the first case the channel length and the saturation at the base of the channels are predicted. In the latter case, a criterion for penetration of the subcooled zone by channels is obtained

  8. Development Potentials for LH2 Storage System with Advanced Boil-off Management

    International Nuclear Information System (INIS)

    Takashi Maemura; Takanobu Kamiya; Shuichi Kawasaki; Ryo Nakamura; Kenji Nakamichi

    2006-01-01

    This paper describes our R and D until 2004 for liquid hydrogen components and system, and current development status summary from 2005 for the LH2 storing, transporting, and refuelling system with the advanced boil-off management using 'slush hydrogen', sponsored by NEDO (domestic projects). The objectives of our study from 2005 are to prove the reduction of the evaporation loss (BOG loss) by utilizing the slush hydrogen, which is the mixture of solids and triple point liquid hydrogen. Use of slush hydrogen rather than atmospheric pressure liquid hydrogen provides the advantage in density and cooling capacity. Assuming a vehicle storage tank size such as 100 to 200 litter ones, the BOG rate can be reduced to 30 percent less than the atmospheric pressure liquid hydrogen is. Present execution plan is to develop, built, and test experimental equipments composed of a slush hydrogen generator, a transfer line, and a storage tank during three years from 2005 to 2007. (authors)

  9. Current status of the post boiling transition research in Japan. Integrity evaluation of nuclear fuel assemblies after boiling transition and development of rewetting correlations

    International Nuclear Information System (INIS)

    Hara, Takashi; Mizokami, Shinya; Kudo, Yoshiro; Komura, Seiichi; Nagata, Yoshifumi; Morooka, Shinichi

    2003-01-01

    Development of rewetting correlation formula was the key to predict fuel-cladding temperature after Boiling Transition (BT). Japanese BWR utilities and vendors performed some tests of rewetting and made two rewetting correlation formulas. The effect on fuel integrity after BT depends on temperature of fuel rod and time of dryout. Main cause of losing fuel integrity during BWR's Anticipated Operational Occurrences (AOO) after BT is embrittlement of the claddings due to oxidation. Ballooning of fuel rod is excepted because its pressure boundary isn't broken. In Japan, the Standards Committee of Atomic Energy Society of Japan (AESJ) is making post BT standard. This standard provides guidelines based on the latest knowledge to judge fuel integrity in case of BT and the validity of reusing the fuel assembly that experienced BT in BWRs. (author)

  10. 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)

  11. 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)

  12. IAEA ICSP on HWR moderator subcooling requirements to demonstrate backup heat sink

    International Nuclear Information System (INIS)

    Choi, J.; Nitheanandan, T.

    2013-01-01

    The IAEA launched a new International Collaborative Standard Problem (ICSP) on 'HWR Moderator Subcooling Requirements to Demonstrate Backup Heat Sink Capabilities of Moderator during Accidents'. The purpose of the ICSP is to benchmark analysis computer codes in simulating contact boiling experimental data to assess the subcooling requirements for an overheated pressure tube, plastically deforming into contact with the calandria tube during a postulated large break loss of coolant accident. The experimental data obtained for the ICSP blind simulation can be used to assess safety analysis computer codes simulating thermal radiation heat transfer to the pressure tube, pressure tube deformation or failure, pressure tube to calandria tube heat transfer, calandria tube to moderator heat transfer, and calandria tube deformation or failure. (author)

  13. Acoustic phenomena during boiling

    International Nuclear Information System (INIS)

    Dorofeev, B.M.

    1985-01-01

    Applied and theoretical significance of investigation into acoustic phenomena on boiling is discussed. Effect of spatial and time conditions on pressure vapour bubble has been elucidated. Collective effects were considered: acoustic interaction of bubbles, noise formation ion developed boiling, resonance and hydrodynamic autooscillations. Different methods for predicting heat transfer crisis using changes of accompanying noise characteristics were analysed. Principle peculiarities of generation mechanism of thermoacoustic autooscillations were analysed as well: formation of standing waves; change of two-phase medium contraction in a channel; relation of alternating pressure with boiling process as well as with instantaneous and local temperatures of heat transfer surface and liquid in a boundary layer

  14. Development of a novel infrared-based visualization technique to detect liquid-gas phase dynamics on boiling surfaces

    International Nuclear Information System (INIS)

    Kim, Hyung Dae

    2011-01-01

    modeling of all these two-phase heat transfer phenomena, there is clearly a need to detect the liquid-vapor-solid triple contact line and measure its physical characteristics (extension, speed, temperature). In this paper we demonstrate the application to boiling heat transfer of a recently-developed experimental technique, named DEPIcT

  15. Development of a MELCOR self-initialization algorithm for boiling water reactors

    International Nuclear Information System (INIS)

    Chien, C.S.; Wang, S.J.; Cheng, S.K.

    1996-01-01

    The MELCOR code, developed by Sandia National Laboratories, is suitable for calculating source terms and simulating severe accident phenomena of nuclear power plants. Prior to simulating a severe accident transient with MELCOR, the initial steady-state conditions must be generated in advance. The current MELCOR users' manuals do not provide a self-initialization procedure; this is the reason users have to adjust the initial conditions by themselves through a trial-and-error approach. A MELCOR self-initialization algorithm for boiling water reactor plants has been developed, which eliminates the tedious trial-and-error procedures and improves the simulation accuracy. This algorithm adjusts the important plant variable such as the dome pressure, downcomer level, and core flow rate to the desired conditions automatically. It is implemented through input with control functions provided in MELCOR. The reactor power and feedwater temperature are fed as input data. The initialization work of full-power conditions of the Kuosheng nuclear power station is cited as an example. These initial conditions are generated successfully with the developed algorithm. The generated initial conditions can be stored in a restart file and used for transient analysis. The methodology in this study improves the accuracy and consistency of transient calculations. Meanwhile, the algorithm provides all MELCOR users an easy and correct method for establishing the initial conditions

  16. Design-development and operation of the Experimental Boiling-Water Reactor (EBWR) facility, 1955--1967

    International Nuclear Information System (INIS)

    Boing, L.E.; Wimunc, E.A.; Whittington, G.A.

    1990-11-01

    The Experimental Boiling-Water Reactor (EBWR) was designed, built, and operated to provide experience and engineering data that would demonstrate the feasibility of the direct-cycle, boiling-water reactor and be applicable to improved, larger nuclear power stations; and was based on information obtained in the first test boiling-water reactors, the BORAX series. EBWR initially produced 20 MW(t), 5 MW(e); later modified and upgraded, as described and illustrated, it was operated at up to 100 MW(t). The facility fulfilled its primary mission -- demonstrating the practicality of the direct-boiling concept -- and, in fact, was the prototype of some of the first commercial plants and of reactor programs in some other countries. After successful completion of the Water-Cooled Reactor Program, EBWR was utilized in the joint Argonne-Hanford Plutonium Recycle Program to develop data for the utilization of plutonium as a fuel in light- water thermal systems. Final shutdown of the EBWR facility followed the termination of the latter program. 13 refs., 12 figs

  17. Development of a real-time stability measurement system for boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.; King, W.T.

    1987-01-01

    This paper describes the development of a portable, real time system for boiling water reactor (BWR) stability measurements. The system provides a means for the operator to monitor the reactor stability using existing plant instrumentation and commercially available hardware. The noise component (i.e., perturbations around steady state) of the neutron signal in BWRs has been shown to contain information about reactor stability, and several algorithms have been developed to extract that information. For the present work, the authors have used an algorithm that has been implemented on a portable personal computer. This algorithm uses the autocorrelation function of naturally occurring neutron noise (measured without special plant perturbations) and an autoregressive modeling technique to produce the asymptotic DR. For this real-time implementation, neutron noise data is preconditioned (i.e., filtered and amplified) and sampled at a 5-Hz sampling rate using a commercial data-acquisition system. Approximately every 1.5 min, the current (snapshot) autocorrelation function is computed directly from the data, and the average autocorrelation is updated. The current and average DR estimates are evaluated with the same periodicity and are displayed on the screen along with the autocorrelations and average power spectrum of the neutron noise

  18. 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

  19. Development of a prediction model of severe reaction in boiled egg challenges.

    Science.gov (United States)

    Sugiura, Shiro; Matsui, Teruaki; Nakagawa, Tomoko; Sasaki, Kemal; Nakata, Joon; Kando, Naoyuki; Ito, Komei

    2016-07-01

    We have proposed a new scoring system (Anaphylaxis SCoring Aichi: ASCA) for a quantitative evaluation of the anaphylactic reaction that is observed in an oral food challenge (OFC). Furthermore, the TS/Pro (Total Score of ASCA/cumulative protein dose) can be a marker to represent the overall severity of a food allergy. We aimed to develop a prediction model for a severe allergic reaction that is provoked in a boiled egg white challenge. We used two separate datasets to develop and validate the prediction model, respectively. The development dataset included 198 OFCs, that tested positive. The validation dataset prospectively included 140 consecutive OFCs, irrespective of the result. A 'severe reaction' was defined as a TS/Pro higher than 31 (the median score of the development dataset). A multivariate logistic regression analysis was performed to identify the factors associated with a severe reaction and develop the prediction model. The following four factors were independently associated with a severe reaction: ovomucoid specific IgE class (OM-sIgE: 0-6), aged 5 years or over, a complete avoidance of egg, and a total IgE prediction model. The model showed good discrimination in a receiver operating characteristic analysis; area under the curve (AUC) = 0.84 in development dataset, AUC = 0.85 in validation dataset. The prediction model significantly improved the AUC in both datasets compared to OM-sIgE alone. This simple scoring prediction model was useful for avoiding risky OFC. Copyright © 2016 Japanese Society of Allergology. Production and hosting by Elsevier B.V. All rights reserved.

  20. A two-step method for developing a control rod program for boiling water reactors

    International Nuclear Information System (INIS)

    Taner, M.S.; Levine, S.H.; Hsiao, M.Y.

    1992-01-01

    This paper reports on a two-step method that is established for the generation of a long-term control rod program for boiling water reactors (BWRs). The new method assumes a time-variant target power distribution in core depletion. In the new method, the BWR control rod programming is divided into two steps. In step 1, a sequence of optimal, exposure-dependent Haling power distribution profiles is generated, utilizing the spectral shift concept. In step 2, a set of exposure-dependent control rod patterns is developed by using the Haling profiles generated at step 1 as a target. The new method is implemented in a computer program named OCTOPUS. The optimization procedure of OCTOPUS is based on the method of approximation programming, in which the SIMULATE-E code is used to determine the nucleonics characteristics of the reactor core state. In a test in cycle length over a time-invariant, target Haling power distribution case because of a moderate application of spectral shift. No thermal limits of the core were violated. The gain in cycle length could be increased further by broadening the extent of the spetral shift

  1. Latent heat transport and microlayer evaporation in nucleate boiling

    International Nuclear Information System (INIS)

    Jawurek, H.H.

    1977-08-01

    Part 1 of this work provides a broad overview and, where possible, a quantitative assessment of the complex physical processes which together constitute the mechanism of nucleate boiling heat transfer. It is shown that under a wide range of conditions the primary surface-to-liquid heat flows within an area of bubble influence are so redistributed as to manifest themselves predominantly as latent heat transport, that is, as vaporisation into attached bubbles. Part 2 deals in greater detail with one of the component processes of latent heat transport, namely microlayer evaporation. A literature review reveals the need for synchronised records of microlayer geometry versus time and of normal bubble growth and departure. An apparatus developed to provide such records is described. High-speed cine interference photography from beneath and through a transparent heating surface provided details of microlayer geometry and an image reflection system synchronised these records with the bubble profile views. Results are given for methanol and ethanol boiling at sub-atmospheric pressures and at various heat fluxes and bulk subcoolings. In all cases it is found that microlayers were of sub-micron thickness, that microlayer thinning was restricted to the inner layer edge (with the thickness elsewhere remaining constant or increasing with time) and that the contribution of this visible evaporation to the total vapour flow into bubbles was negligible. The observation of thickening towards the outer microlayer edge, however, demonstrates that a liquid replenishment flow occurred simultaneously with the evaporation process

  2. 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)

  3. 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)

  4. An Experimental investigation of critical flow rates of subcooled water through short pipes with small diameters

    International Nuclear Information System (INIS)

    Park, Choon Kyung

    1997-02-01

    The primary objective of this study is to improve our understanding on critical flow phenomena in a small size leak and to develop a model which can be used to estimate the critical mass flow rates through reactor vessel or primary coolant pipe wall. For this purpose, critical two-phase flow phenomena of subcooled water through short pipes (100 ≤ L ≤ 400 mm) with small diameters (3.4 ≤ D ≤ 7.15 mm) have been experimentally investigated for wide ranges of subcooling (0∼199 .deg. C) and pressure (0.5∼2.0MPa). To examine the effects of various parameters (i.e., the location of flashing inception, the degree of subcooling, the stagnation temperature and pressure, and the pipe size) on the critical two-phase flow rates of subcooled water, a total of 135 runs were made for various combinations of test parameters using four different L/D test sections. Experimental results that show effects of various parameters on subcooled critical two-phase flow rates are presented. The measured static pressure profiles along the discharge pipe show that the critical flow rate can be strongly influenced by the flashing location. The locations of saturation pressure for different values of the stagnation subcooling have been consistently determined from the pressure profiles. Based upon the test results, two important parameters have been identified. These are cold state discharge coefficient and dimensionless subcooling, which are found to efficiently take into account the test section geometry and the stagnation conditions, respectively. A semi-empirical model has been developed to predict subcooled two-phase flow rates through small size openings. This model provides a simple and direct calculation of the critical mass flow rates with information on the initial condition and on the test section geometry. Comparisons between the mass fluxes calculated by present model and a total of 755 selected experimental data from 9 different investigators show that the agreement is

  5. Assessment of CHF characteristics at subcooled conditions for the CANDU CANFLEX bundle

    International Nuclear Information System (INIS)

    Onder, E.N.; Leung, L.K.H.

    2011-01-01

    An analysis has been performed to assess the Critical Heat Flux (CHF) characteristics for the CANFLEX bundle at subcooled conditions. CHF characteristics for CANDU bundles have been established from experiments using full-scale bundle simulators. These experiments covered flow conditions of interest to normal operation and postulated loss-of-flow and small break loss-of-coolant accidents. Experimental CHF values obtained from these experiments were applied to develop correlations for analyses of regional overpower protection and safety trips. These correlations are applicable to the saturated region in the reference uncrept channel and the slightly subcooled region in postulated high-creep channels. Expanding the CHF data to subcooled conditions facilitates the evaluation of the margin to dryout at upstream bundle locations, even though dryout occurrences are not anticipated there. In view of the lack of experimental data, the ASSERT-PV subchannel code has been applied to establish CHF values at low qualities and high subcoolings (thermodynamic qualities corresponding to -25%). These CHF values have been applied to extend the CHF correlation to the highly subcooled conditions. (author)

  6. THE DEVELOPMENT OF THE CALCULATION MODEL FOR THE ESTIMATION OF THE BOILING POINT OF THE ­POLYMER-SOLVENT MIXTURES

    Directory of Open Access Journals (Sweden)

    Matseevich Andrey Vyacheslavovich

    2018-03-01

    Full Text Available Subject of the study: one of the most promising areas in the field of polymer physics is the development of the calculation models allowing to quantify the properties of polymers. This work provides the calculation model for the quantitative assessment of the boiling point of solutions of polymer in the organic solvent. The model is based on the chemical structure of polymer and solvent. For the components the Hildebrand solubility parameter, the latent heat of vaporization and the boiling point of the solvent are calculated. Objectives: to generate the equation connecting the boiling point of polymer solution in the chosen solvent with the boiling point of the pure solvent, the molecular weights of the repeating unit of polymer and the molecule of solvent, the weight fraction of polymer in solution, the Hildebrand solubility parameter and the molar volume of the repeating unit of polymer. Materials and methods: the Hildebrand solubility parameter of solutions and polymers and also the van der Waals volume were calculated using the method of A.A. Askadsky; the enthalpy of vaporization of the solvent at the boiling point was expressed through the Hildebrand solubility parameter. The dependence of the enthalpy of vaporization from the temperature was taken into consideration. The computerization of the method was implemented, according to which all calculations are performed automatically after entering the information on the chemical structure of polymer and solvent into the computer. Results: the equation connecting the ebulliometric constant of the low concentration polymer solution with the boiling point of the solvent, the molar volume of the solvent and the Hildebrand parameter was generated. The results of the analysis were checked with regard to the system of polystyrene/toluene; the possibility of practical application of the offered method was shown. Conclusions: the method presented in this article allows to predict the ebulliometric

  7. Pool boiling from rotating and stationary spheres in liquid nitrogen

    Science.gov (United States)

    Cuan, Winston M.; Schwartz, Sidney H.

    1988-01-01

    Results are presented for a preliminary experiment involving saturated pool boiling at 1 atm from rotating 2 and 3 in. diameter spheres which were immersed in liquid nitrogen (LN2). Additional results are presented for a stationary, 2 inch diameter sphere, quenched in LN2, which were obtained utilizing a more versatile and complete experimental apparatus that will eventually be used for additional rotating sphere experiments. The speed for the rotational tests was varied from 0 to 10,000 rpm. The stationary experiments parametrically varied pressure and subcooling levels from 0 to 600 psig and from 0 to 50 F, respectively. During the rotational tests, a high speed photographic analysis was undertaken to measure the thickness of the vapor film surrounding the sphere. The average Nusselt number over the cooling period was plotted against the rotational Reynolds number. Stationary sphere results included local boiling heat transfer coefficients at different latitudinal locations, for various pressure and subcooling levels.

  8. Development of septum-free injector for gas chromatography and its application to the samples with a high boiling point.

    Science.gov (United States)

    Ito, Hiroshi; Hayakawa, Kazuichi; Yamamoto, Atsushi; Murase, Atsushi; Hayakawa, Kazumi; Kuno, Minoru; Inoue, Yoshinori

    2006-11-03

    A novel apparatus with a simple structure has been developed for introducing samples into the vaporizing chamber of a gas chromatograph. It requires no septum due to the gas sealing structure over the carrier gas supply line. The septum-free injector made it possible to use injection port temperatures as high as 450 degrees C. Repetitive injection of samples with boiling points below 300 degrees C resulted in peak areas with relative standard deviations between 1.25 and 3.28% (n=5) and good linearity (r(2)>0.9942) for the calibration curve. In the analysis of polycyclic aromatic hydrocarbons and a base oil, the peak areas of components with high boiling points increased as the injection port temperature was increased to 450 degrees C.

  9. Critical discharge of initially subcooled water through slits

    International Nuclear Information System (INIS)

    Amos, C.N.; Schrock, V.E.

    1983-09-01

    This report describes an experimental investigation into the critical flow of initially subcooled water through rectangular slits. The study of such flows is relevant to the prediction of leak flow rates from cracks in piping, or pressure vessels, which contain sufficient enthalpy that vaporization will occur if they are allowed to expand to the ambient pressure. Two new analytical models, which allow for the generation of a metastable liquid phase, are developed. Experimental results are compared with the predictions of both these new models and with a Fanno Homogeneous Equilibrium Model

  10. Critical discharge of initially subcooled water through slits. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Amos, C N; Schrock, V E

    1983-09-01

    This report describes an experimental investigation into the critical flow of initially subcooled water through rectangular slits. The study of such flows is relevant to the prediction of leak flow rates from cracks in piping, or pressure vessels, which contain sufficient enthalpy that vaporization will occur if they are allowed to expand to the ambient pressure. Two new analytical models, which allow for the generation of a metastable liquid phase, are developed. Experimental results are compared with the predictions of both these new models and with a Fanno Homogeneous Equilibrium Model.

  11. Development of an in-core fuel management tool for boiling water reactors

    International Nuclear Information System (INIS)

    Gilli, Luca; Wakker, Pieter H.; Elder, Brian R.

    2017-01-01

    The in-core fuel management of a nuclear reactor is a challenging task due to the virtually infinite number of loading patterns one could theoretically adopt. The ROSA (Reloading Optimization by Simulated Annealing) code is an optimization tool that has been successfully used in the last two decades to facilitate the core design of several Pressurized Water Reactors (PWRs). It is designed to perform a stochastic search for an optimal Loading Pattern (LP) using a simulated annealing algorithm. This corresponds to performing a depletion calculation for each one of the hundreds of thousands of unique LPs generated during the stochastic search. Therefore, speed is one of the most important requirements that the solvers used by the depletion tool must fulfill. ROSA's depletion analysis tool makes use of a particularly fast nodal method (known as the kernel method) for the evaluation of the power distribution associated with a particular LP. One of the strongest assumptions behind the kernel method is that the neutron migration length does not change considerably between the point where a neutron is generated and the point where the same neutron is absorbed. Although strong, this assumption is quite compatible with the neutronic characteristics of PWRs cores. In this paper we give an overview of the work done in order to develop a version of ROSA capable of performing the core design of Boiling Water Reactors (BWRs). We focus the discussion on the development of the depletion analysis tool by outlining the modifications of the kernel methods implemented in order to make the solver accurate for BWR cores. An improvement of the definition of the transport kernel is necessary to take the strong anisotropies characterizing the neutronic problem into account. These anisotropies arise due to the presence of strong changes in the moderator density and due to the presence of control blades. Furthermore, we are going to discuss how the boundary conditions are adopted by the

  12. Investigation of film boiling thermal hydraulics under FCI conditions. Results of a numerical study

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Dinh, A.T.; Nourgaliev, R.R.; Sehgal, B.R. [Div. of Nuclear Power Safety Royal Inst. of Tech. (RIT), Brinellvaegen 60, 10044 Stockholm (Sweden)

    1998-01-01

    Film boiling on the surface of a high-temperature melt jet or of a melt particle is one of key phenomena governing the physics of fuel-coolant interactions (FCIs) which may occur during the course of a severe accident in a light water reactor (LWR). A number of experimental and analytical studies have been performed, in the past, to address film boiling heat transfer and the accompanying hydrodynamic aspects. Most of the experiments have, however, been performed for temperature and heat flux conditions, which are significantly lower than the prototypic conditions. For ex-vessel FCIs, high liquid subcooling can significantly affect the FCI thermal hydraulics. Presently, there are large uncertainties in predicting natural-convection film boiling of subcooled liquids on high-temperature surfaces. In this paper, research conducted at the Division of Nuclear Power Safety, Royal Institute of Technology (RIT/NPS), Stockholm, concerning film-boiling thermal hydraulics under FCI condition is presented. Notably, the focus is placed on the effects of (1) water subcooling, (2) high-temperature steam properties, (3) the radiation heat transfer and (4) mixing zone boiling dynamics, on the vapor film characteristics. Numerical investigations are performed using a novel CFD modeling concept named as the local-homogeneous-slip model (LHSM). Results of the analytical and numerical studies are discussed with respect to boiling dynamics under FCI conditions. (author)

  13. Development of Empirical Correlation to Calculate Pool Boiling Heat Transfer of Tandem Tubes

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Myeong-Gie [Andong National University, Andong (Korea, Republic of)

    2015-10-15

    The heat exchanging tubes are in vertical alignment. For the cases, the upper tube is affected by the lower tube. Since heat transfer is closely related to the conditions of tube surface, bundle geometry, and liquid, lots of studies have been carried out for the several decades to investigate the combined effects of those factors on pool boiling heat transfer. One of the most important parameters in the analysis of a tube array is the pitch ( P ) between tubes. Many researchers have been investigated its effect on heat transfer enhancement for the tube bundles and the tandem tubes. The effect of a tube array on heat transfer enhancement was also studied for application to the flooded evaporators. Cornwell and Schuller studied the sliding bubbles by high speed photography to account the enhancement of heat transfer observed at the upper tubes of a bundle. The study by Memory et al. shows the effects of the enhanced surface and oil adds to the heat transfer of tube bundles. They identified that, for the structured and porous bundles, oil addition leads to a steady decrease in performance. The flow boiling of n-pentane across a horizontal tube bundle was investigated experimentally by Roser et al. They identified that convective evaporation played a significant part of the total heat transfer. The fouling of the tube bundle under pool boiling was also studied by Malayeri et al. They identified that the mechanisms of fouling on the middle and top heater substantially differ from those at the bottom heater.

  14. 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.)

  15. Axial propagation of free surface boiling into superheated liquids in vertical tubes

    International Nuclear Information System (INIS)

    Grolmes, M.A.; Fauske, H.K.

    1974-01-01

    A unique free surface boiling phenomenon has been observed as a result of rapid depressurization of an initially saturated or slightly subcooled stagnant liquid column in the absence of wall and bulk nucleation sites. Closeup high-speed photographs of water, refrigerant-11, and methyl alcohol in tubes from 0.2 to 15 in. dia reveal that the initiation of violent free surface flashing (vapor plus entrained liquid) follows from the development of Marangoni-type surface waves. The rate of propagation of the flashing surface shows evidence of choked flow limitations and proceeds at a rate which is several orders of magnitude greater than surface evaporation (vapor only) alone. The onset of free surface flashing was found to be dependent upon both the degree of initial liquid superheat and the tube diameter. (U.S.)

  16. Boiling performance and material robustness of modified surfaces with multi scale structures for fuel cladding development

    Energy Technology Data Exchange (ETDEWEB)

    Jo, HangJin; Kim, Jin Man [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Yeom, Hwasung [Department of Nuclear Engineering and Engineering physics, UW-Madison, Madison, WI 53706, Unities States (United States); Lee, Gi Cheol [Department of Mechanical Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Kiyofumi, Moriyama; Kim, Moo Hwan [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784, Gyungbuk (Korea, Republic of); Sridharan, Kumar; Corradini, Michael [Department of Nuclear Engineering and Engineering physics, UW-Madison, Madison, WI 53706, Unities States (United States)

    2015-09-15

    Highlights: • We improved boiling performance and material robustness using surface modification. • We combined micro/millimeter post structures and nanoparticles with heat treatments. • Compactly-arranged micrometer posts had improved boiling performance. • CHF increased significantly due to capillary pumping by the deposited NP layers. • Sintering procedure increased mechanical strength of the NP coating surface. - Abstract: By regulating the geometrical characteristics of multi-scale structures and by adopting heat treatment for protective layer of nanoparticles (NPs), we improved critical heat flux (CHF), boiling heat transfer (BHT), and mechanical robustness of the modified surface. We fabricated 1-mm and 100-μm post structures and deposited NPs on the structured surface as a nano-scale structured layer and protective layer at the same time, then evaluated the CHF and BHT and material robustness of the modified surfaces. On the structured surfaces without NPs, the surface with compactly-arranged micrometer posts had improved CHF (118%) and BHT (41%). On the surface with structures on which NPs had been deposited, CHF increased significantly (172%) due to capillary pumping by the deposited NP layers. The heat treatment improved robustness of coating layer in comparison to the one of before heat treatment. In particular, low-temperature sintering increased the hardness of the modified surface by 140%. The increased mechanical strength of the NP coating is attributed to reduction in coating porosity during sintering. The combination of micrometer posts structures and sintered NP coating can increase the safety, efficiency and reliability of advanced nuclear fuel cladding.

  17. Boiling performance and material robustness of modified surfaces with multi scale structures for fuel cladding development

    International Nuclear Information System (INIS)

    Jo, HangJin; Kim, Jin Man; Yeom, Hwasung; Lee, Gi Cheol; Park, Hyun Sun; Kiyofumi, Moriyama; Kim, Moo Hwan; Sridharan, Kumar; Corradini, Michael

    2015-01-01

    Highlights: • We improved boiling performance and material robustness using surface modification. • We combined micro/millimeter post structures and nanoparticles with heat treatments. • Compactly-arranged micrometer posts had improved boiling performance. • CHF increased significantly due to capillary pumping by the deposited NP layers. • Sintering procedure increased mechanical strength of the NP coating surface. - Abstract: By regulating the geometrical characteristics of multi-scale structures and by adopting heat treatment for protective layer of nanoparticles (NPs), we improved critical heat flux (CHF), boiling heat transfer (BHT), and mechanical robustness of the modified surface. We fabricated 1-mm and 100-μm post structures and deposited NPs on the structured surface as a nano-scale structured layer and protective layer at the same time, then evaluated the CHF and BHT and material robustness of the modified surfaces. On the structured surfaces without NPs, the surface with compactly-arranged micrometer posts had improved CHF (118%) and BHT (41%). On the surface with structures on which NPs had been deposited, CHF increased significantly (172%) due to capillary pumping by the deposited NP layers. The heat treatment improved robustness of coating layer in comparison to the one of before heat treatment. In particular, low-temperature sintering increased the hardness of the modified surface by 140%. The increased mechanical strength of the NP coating is attributed to reduction in coating porosity during sintering. The combination of micrometer posts structures and sintered NP coating can increase the safety, efficiency and reliability of advanced nuclear fuel cladding

  18. Development of conversion efficient processes for +525[degree]C pitch to low boiling distillates

    Energy Technology Data Exchange (ETDEWEB)

    Pandey, R.N.

    1992-10-22

    Catalytic hydroprocessing of Esso pitch (+525[degree]C fraction) was investigated using a continuous reactor system at various operating conditions. The catalysts studied included ZnCl[sub 2], SnCl[sub 4], SiO[sub 2], Ni/W, Co/Mo, Zn/Cr, and H[sub 3]PO[sub 4]/SiO[sub 2]. The catalysts were characterized by surface area and acidity measurements. The gas and liquid products were collected and analyzed, and results are presented. The work demonstrates the effectiveness of a continuous flash hydropyrolysis process for the conversion of petroleum residuals to low boiling distillates. It has been found that in the presence of ZnCl[sub 2] catalyst, conversions up to 92% can be obtained under relatively moderate conditions of temperature and hydrogen pressure. The formation of gaseous products is less than 5 wt %. The favorable conditions of operation, a reduction in sulfur content, increase in H/C ratio, and a significant yield of low boiling distillates is obtained. The process variables, such as temperature, hydrogen pressure, and liquid residence times influence product yield and quality. Coke formation was almost negligible even under conditions of high pitch conversion. A review of various techniques of pyrolysis and hydropyrolysis of materials such as coal, bitumen, and heavy oil is included. 198 refs., 46 figs., 40 tabs.

  19. Modeling of subcooling and solidification of phase change materials

    Science.gov (United States)

    Günther, Eva; Mehling, Harald; Hiebler, Stefan

    2007-12-01

    Phase change materials (PCM) are able to store thermal energy in small temperature intervals very efficiently due to their high latent heat. Particularly high storage capacity is found in salt hydrates. Salt hydrates however often show subcooling, thus inhibiting the release of the stored heat. In the state of the art simulations of PCM, the effect of subcooling is almost always neglected. This is a practicable approach for small subcooling, but it is problematic for subcooling in the order of the driving temperature gradient on unloading the storage. In this paper, we first present a new algorithm to simulate subcooling in a physically proper way. Then, we present a parametric study to demonstrate the main features of the algorithm and a comparison of computed and experimentally obtained data. The new algorithm should be particularly useful in simulating applications with low cooling rates, for example building applications.

  20. Development of generalized boiling transition model applicable for wide variety of fuel bundle geometries. Basic strategy and numerical approaches

    International Nuclear Information System (INIS)

    Ninokata, Hisashi; Sadatomi, Michio; Okawa, Tomio

    2003-01-01

    In order to establish a key technology to realize advanced BWR fuel designs, a three-year project of the advanced subchannel analysis code development had been started since 2002. The five dominant factors involved in the boiling transitional process in the fuel bundles were focused. They are, (1) inter-subchannel exchanges, (2) influences of obstacles (3) dryout of liquid film, (4) transition of two-phase flow regimes and (5) deposition of droplets. It has been recognized that present physical models or constitutive equations in subchannel formulations need to be improved so that they include geometrical effects in the fuel bundle design more mechanistically and universally. Through reviewing literatures and existent experimental results, underlying elementary processes and geometrical factors that are indispensable for improving subchannel codes were identified. The basic strategy that combines numerical and experimental approaches was proposed aiming at establishment of mechanistic models for the five dominant factors. In this paper, the present status of methodologies for detailed two-phase flow studies has been summarized. According to spatial scales of focused elementary processes, proper numerical approaches were selected. For some promising numerical approaches, preliminary calcitonins were performed for assessing their applicability to investigation of elementary processes involved in the boiling transition. (author)

  1. Simulation of heat and mass transfer in boiling water with the Melodif code

    International Nuclear Information System (INIS)

    Freydier, P.; Chen, O.; Olive, J.; Simonin, O.

    1991-04-01

    The Melodif code is developed at Electricite de France, Research and Development Division. It is an eulerian two dimensional code for the simulation of turbulent two phase flows (a three dimensional code derived from Melodif, ASTRID, is currently being prepared). Melodif is based on the two fluid model, solving the equations of conservation for mass, momentum and energy, for both phases. In such a two fluid model, the description of interfacial transfers between phases is a crucial issue. The model used applies to a dominant continuous phase, and a dispersed phase. A good description of interfacial momentum transfer exists in the standard MELODIF code: the drag force, the apparent mass force... are taken into account. An important factor for interfacial transfers is the interfacial area per volume unit. With the assumption of spherical gas bubbles, an equation has been written for this variable. In the present wok, a model has been tested for interfacial heat and mass transfer in the case of boiling water: it is assumed that mass transfer is controlled by heat transfer through the latent massic energy taken in the phase that vaporizes (or condenses). This heat and mass transfer model has been tested in various configurations: - a cylinder with water flowing inside, is being heated. Boiling takes place near the wall, while bubbles migrating to the core of the flow recondense. This roughly simulates a sub-cooled boiling phenomenon. - a box containing liquid water is depressurized. Boiling takes place in the whole volume of the fluid. The Melodif code can simulate this configuration due to the implicitation of the relation between interphase mass transfer and the pressure variable

  2. Diagnostics of the boiling state of coolant based on neutron fluctuation at the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Por, G.; Gloeckler, O.; Izsak, E.; Valko, J.

    1985-09-01

    A short summary of theory and early experiments on the effect of propagating perturbation on neutron fluctuations in nuclear reactors is given. Boiling noise was examined in the Rheisenberg reactor of 70 MWe. Comparing the results of measurements with those carried out in the Paks nuclear power plant it seems possible that a small subcooled boiling took place during the 2nd fuel cycle. (author)

  3. Performance evaluation of an ejector subcooled vapor-compression refrigeration cycle

    International Nuclear Information System (INIS)

    Xing, Meibo; Yan, Gang; Yu, Jianlin

    2015-01-01

    Highlights: • An ejector subcooled vapor-compression refrigeration cycle is proposed. • The performance of the cycle with ejector subcooling is evaluated theoretically. • Increase in refrigeration capacity can be achieved by the ejector subcooled cycle. • The new cycle exhibits higher COP compared to the basic single-stage cycle. • Performance of the new cycle depends on the operation pressures of the ejector. - Abstract: In this study, a novel vapor-compression refrigeration cycle with mechanical subcooling using an ejector is proposed to improve the performance of a conventional single-stage vapor-compression refrigeration cycle. In the theoretical study, a mathematical model is developed to predict the performance of the cycle by using R404A and R290, and then compared with that of the conventional refrigeration cycle. The simulation results show that the performance of the ejector subcooled cycle is better than that of the conventional cycle. When the evaporator temperature ranges from −40 to −10 °C and the condenser temperature is 45 °C, the novel cycle displays volumetric refrigeration capacity improvements of 11.7% with R404A and 7.2% with R290. And the novel cycle achieves COP improvements of 9.5% with R404A and 7.0% with R290. In addition, the improvement of the COP and cooling capacity of this novel cycle largely depends on the operation pressures of the ejector. The potential practical advantages offered by the cycle may be worth further attention in future studies

  4. 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

  5. Performance enhancement of a subcooled cold storage air conditioning system

    International Nuclear Information System (INIS)

    Hsiao, M.-J.; Cheng, C.-H.; Huang, M.-C.; Chen, S.-L.

    2009-01-01

    This article experimentally investigates the enhancement of thermal performance for an air conditioning system utilizing a cold storage unit as a subcooler. The cold storage unit is composed of an energy storage tank, liquid-side heat exchanger, suction-side heat exchanger and energy storage material (ESM), water. When the cooling load is lower than the nominal cooling capacity of the system, the cold storage unit can store extra cold energy of the system to subcool the condenser outlet refrigerant. Hence, both the cooling capacity and coefficient of performance (COP) of the system will be increased. This experiment tests the two operation modes: subcooled mode with energy storage and non-subcooled mode without energy storage. The results show that for fixed cooling loads at 3.05 kW, 3.5 kW and 3.95 kW, the COP of the subcooled mode are 16.0%, 15.6% and 14.1% higher than those of the non-subcooled mode, respectively. In the varied cooling load experiments, the COP of the subcooled cold storage air conditioning system is 15.3% higher than the conventional system.

  6. 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

  7. Evaluation of subcooled critical heat flux correlations for tubes with and without internal twisted tapes

    International Nuclear Information System (INIS)

    Inasaka, F.; Nariai, H.

    1996-01-01

    Eleven correlations and models for critical heat flux (CHF) of subcooled flow boiling in water were evaluated. Both a direct substitution method (DSM) and a heat balance condition method (HBM) were compared in the evaluations. The HBM was recommended as a better prediction method in the present study. For straight tubes under uniform heating conditions, the correlations of the Gunther, Knoebel, modified Tong, W-2, and Tong-75, and also the Celata and Weisman-Pei models were confirmed to give reasonably good predictions. Among them, the Celata model was the best with respect to accuracy. For swirl flow under uniform heating conditions, Tong-75-I (involving modification of the water velocity parameter) and Nariai-Inasaka correlations were confirmed to give reasonably good predictions, even though their predictions were too low for the CHF under non-uniform heating conditions. (orig.)

  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. Stability analysis on natural circulation boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Metz, Peter

    1999-05-01

    The purpose of the study is a stability analysis of the simplified boiling water reactor concept. A fluid dynamics code, DYNOS, was developed and successfully validated against FRIGG and DESIRE data and a stability benchmark on the Ringhals 1 forced circulation BWR. Three simplified desings were considered in the analysis: The SWRIOOO by Siemens and the SBWR and ESBWR from the General Electric Co. For all three design operational characteristics, i.e. power versus flow rate maps, were calculated. The effects which different geometric and operational parameters, such as the riser height, inlet subcooling etc., have on the characteristics have been investigated. Dynamic simulations on the three simplified design revealed the geysering and the natural circulation oscillations modes only. They were, however, only encountered at pressure below 0.6 MPa. Stability maps for all tree simplified BWRs were calculated and plotted. The study concluded that a fast pressurisation of the reactor vessel is necessary to eliminate the possibility of geysering or natural circulation oscillations mode instability. (au) 26 tabs., 88 ills.

  10. Stability analysis on natural circulation boiling water reactors

    International Nuclear Information System (INIS)

    Metz, Peter

    1999-05-01

    The purpose of the study is a stability analysis of the simplified boiling water reactor concept. A fluid dynamics code, DYNOS, was developed and successfully validated against FRIGG and DESIRE data and a stability benchmark on the Ringhals 1 forced circulation BWR. Three simplified desings were considered in the analysis: The SWRIOOO by Siemens and the SBWR and ESBWR from the General Electric Co. For all three design operational characteristics, i.e. power versus flow rate maps, were calculated. The effects which different geometric and operational parameters, such as the riser height, inlet subcooling etc., have on the characteristics have been investigated. Dynamic simulations on the three simplified design revealed the geysering and the natural circulation oscillations modes only. They were, however, only encountered at pressure below 0.6 MPa. Stability maps for all tree simplified BWRs were calculated and plotted. The study concluded that a fast pressurisation of the reactor vessel is necessary to eliminate the possibility of geysering or natural circulation oscillations mode instability. (au)

  11. Thermoeconomic optimization of subcooled and superheated vapor compression refrigeration cycle

    International Nuclear Information System (INIS)

    Selbas, Resat; Kizilkan, Onder; Sencan, Arzu

    2006-01-01

    An exergy-based thermoeconomic optimization application is applied to a subcooled and superheated vapor compression refrigeration system. The advantage of using the exergy method of thermoeconomic optimization is that various elements of the system-i.e., condenser, evaporator, subcooling and superheating heat exchangers-can be optimized on their own. The application consists of determining the optimum heat exchanger areas with the corresponding optimum subcooling and superheating temperatures. A cost function is specified for the optimum conditions. All calculations are made for three refrigerants: R22, R134a, and R407c. Thermodynamic properties of refrigerants are formulated using the Artificial Neural Network methodology

  12. Analytical study of flow instability behaviour in a boiling two-phase natural circulation loop under low quality conditions

    International Nuclear Information System (INIS)

    Nayak, A.K.; Kumar, N.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2002-01-01

    Analytical investigations have been carried out to study the flow instability behaviour in a boiling two-phase natural circulation loop under low quality conditions. For this purpose, the computer code TINFLO-S has been developed. The code solves the conservation equations of mass, momentum and energy and equation of state for homogeneous equilibrium twophase flow using linear analytical technique. The results of the code have been validated with the experimental data of the loop for both the steady state and stability. The study reveals that the stability behaviour of low quality flow oscillations is different from that of the high quality flow oscillations. The instability reduces with increase in power and throttling at the inlet of the heater. The instability first increases and then reduces with increase in pressure at any subcooling. The effects of diameter of riser pipe, heater and the height of the riser on this instability are also investigated. (orig.) [de

  13. 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

  14. Subcooler assembly for SSC single magnet test program

    International Nuclear Information System (INIS)

    Wu, K.C.; Brown, D.P.; Sondericker, J.H.; Farah, Y.; Zantopp, D.; Nicoletti, A.

    1991-01-01

    A subcooler assembly has been designed, constructed and installed in the MAGCOOL magnet test area at Brookhaven National Laboratory. Since July 1989, it has been used for testing SSC magnets. This subcooler assembly and cryogenic system are the first of its kind ever built. Today, with more than 5000 hours of operating time, the subcooler has proved to be a reliable unit with individual components meeting design expectations. The lowest temperatures achieved with one SSC dipole are 3.0 K at the suction of the cold vacuum pump and 3.2 K at the return of the magnet. The system performs well in both steady state operation and during magnet quench, subcooling, cooldown and warmup. 4 refs., 7 figs

  15. Velocity field measurement in micro-bubble emission boiling

    International Nuclear Information System (INIS)

    Ito, Daisuke; Saito, Yasushi; Natazuka, Jun

    2017-01-01

    Liquid inlet behavior to a heat surface in micro-bubble emission boiling (MEB) was investigated by flow measurement using particle image velocimetry (PIV). Subcooled pool boiling experiments under atmospheric pressure were carried out using a heat surface with a diameter of 10 mm. An upper end of a heater block made of copper was used as the heat surface. Working fluid was the deionized water and the subcooling was varied from 40 K to 70 K. Three K-type thermocouples were installed in the copper block to measure the temperature gradient, and the heat flux and wall superheat were estimated from these temperature data to make a boiling curve. The flow visualization around the heat surface was carried out using a high-speed video camera and a light sheet. The microbubbles generated in the MEB were used as tracer particles and the velocity field was obtained by PIV analysis of the acquired image sequence. As a result, the higher heat fluxes than the critical heat flux could be obtained in the MEB region. In addition, the distribution characteristics of the velocity in MEB region were studied using the PIV results and the location of the stagnation point in the velocity fields was discussed. (author)

  16. 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

  17. Thermal-hydraulic oscillations in a low pressure two-phase natural circulation loop at low powers and high inlet subcoolings

    International Nuclear Information System (INIS)

    Wang, S.B.; Wu, J.Y.; Chin Pan; Lin, W.K.

    2004-01-01

    The stability of a natural circulation boiling loop is of great importance and interests for both academic researches and many industrial applications, such as next generation boiling water reactors. The present study investigated the thermal-hydraulic oscillation behavior in a low pressure two-phase natural circulation loop at low powers and high inlet subcoolings. The experiments were conducted at atmospheric pressure with heating power ranging from 4 to 8 kW and inlet subcooling ranging from 27 to 75 deg. C. Significant oscillations in loop mass flow rate, pressure drop in each section, and heated wall and fluid temperatures are present for all the cases studied here. The oscillation is typically quasi-periodic and with flow reversal with magnitudes smaller than forward flows. The magnitude of wall temperature oscillation could be as high as 60 deg. C, which will be of serious concern for practical applications. It is found that the first fundamental oscillation (large magnitude oscillation) frequency increases with increase in heated power and with decrease in inlet subcooling. (author)

  18. 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

  19. Vapour phase motion in cryogenic systems containing superheated and subcooled liquids

    Science.gov (United States)

    Kirichenko, Yu. A.; Chernyakov, P. S.; Seregin, V. E.

    The development of vent pipelines, and venting storage tanks for cryogenic liquids requires the knowledge of the law of motion as well as regularities of vapour content variation in the liquid and heat dissipation by the vapour phase. This is a theoretical study of the effect of superheating (subcooling) of the liquid, relative acceleration and reduced pressure upon the size and velocity of noninteracting vapour bubbles, moving in the liquid, and upon their resistance and heat transfer coefficients.

  20. LOFT system structural response during subcooled blowdown

    International Nuclear Information System (INIS)

    Martinell, J.S.

    1978-01-01

    The Loss-of-Fluid Test (LOFT) facility is a highly instrumented, pressurized water reactor test system designed to be representative of large pressurized water reactors (LPWRs) for the simulation of loss-of-coolant accidents (LOCAs). Detailed structural analysis and appropriate instrumentation (accelerometers and strain gages) on the LOFT system provided information for evaluation of the structural response of the LOFT facility for loss-of-coolant experiment (LOCE) induced loads. In general, the response of the system during subcooled blowdown was small with typical structural accelerations below 2.0 G's and dynamic strains less than 150 x 10 - 6 m/m. The accelerations measured at the steam generator and simulated steam generator flange exceeded LOCE design values; however, integration of the accelerometer data at these locations yielded displacements which were less than one half of the design values associated with a safe shutdown earthquake (SSE), which assures structural integrity for LOCE loads. The existing measurement system was adequate for evaluation of the LOFT system response during the LOCEs. The conditions affecting blowdown loads during nuclear LOCEs will be nearly the same as those experienced during the nonnuclear LOCEs, and the characteristics of the structural response data in both types of experiments are expected to be the same. The LOFT system is concluded to be adequately designed and further analysis of the LOFT system with structural codes is not required for future LOCE experiments

  1. Interaction of the nucleation phenomena at adjacent sites in nucleate boiling

    International Nuclear Information System (INIS)

    Sultan, M.; Judd, R.L.

    1983-01-01

    The present investigation is an original study in nucleate pool boiling heat transfer combining theory and experiment in which water boiling at atmospheric pressure on a single copper surface at two different levels of heat and different levels of subcooling was studied. Cross spectral analysis of the signals generated by the emission of bubbles at adjacent nucleation sites was used to determine the relationship of the time elapsed between the start of bubble growth at the two neighbouring active sites with the distance separating them. The experimental results obtained indicated that for the lower level of heat flux at three different levels of subcooling, the elapsed time and distance were directly related. Theoretical predictions of a temperature disturbance propagating through the heating surface in the radial direction gave good agreement with the experimental findings, suggesting that this is the mechanism responsible for the activation of the surrounding nucleation sites

  2. Analysis of Void Fraction Distribution and Departure from Nucleate Boiling in Single Subchannel and Bundle Geometries Using Subchannel, System, and Computational Fluid Dynamics Codes

    Directory of Open Access Journals (Sweden)

    Taewan Kim

    2012-01-01

    Full Text Available In order to assess the accuracy and validity of subchannel, system, and computational fluid dynamics codes, the Paul Scherrer Institut has participated in the OECD/NRC PSBT benchmark with the thermal-hydraulic system code TRACE5.0 developed by US NRC, the subchannel code FLICA4 developed by CEA, and the computational fluid dynamic code STAR-CD developed by CD-adapco. The PSBT benchmark consists of a series of void distribution exercises and departure from nucleate boiling exercises. The results reveal that the prediction by the subchannel code FLICA4 agrees with the experimental data reasonably well in both steady-state and transient conditions. The analyses of single-subchannel experiments by means of the computational fluid dynamic code STAR-CD with the CD-adapco boiling model indicate that the prediction of the void fraction has no significant discrepancy from the experiments. The analyses with TRACE point out the necessity to perform additional assessment of the subcooled boiling model and bulk condensation model of TRACE.

  3. Lung development and postnatal survival for rats exposed in utero to a high-boiling coal liquid

    Energy Technology Data Exchange (ETDEWEB)

    Springer, D.L.; Hackett, P.L.; Miller, R.A.; Buschbom, R.L.

    1986-01-01

    The study reported determines postnatal viability and development of survivors following in utero exposure to Harmarville process solvent (HPS), a wide-boiling-range (150 to > 455/sup 0/C) coal liquid. For this study, 0.74 g kg/sup -1/ of the coal liquid was administered (by intragastric intubation) to rats from 12 to 14 dg. Offspring were evaluated for postnatal survival, growth and lung and thymus weights. Fifty-four percent of the exposed pups and 9% of the control pups died between birth and 3 days postpartum. Of the treated pups that died, 10% (6/5; pups/litters) had cleft palate, 27% (17/9) had small lungs and 33% (21/8) had both cleft palate and small lungs. No gross malformations were observed in the remaining 30% of the dead pups. Microscopic examination of lungs from HPS-treated pups revealed no evident histological abnormalities. Body, lung and thymus weights for treated animals that died were significantly less than those of controls. Surviving exposed pups weighed significantly less than control pups from 0.25 to 21 days postpartum and their thymus weights were also depressed through 21 days postpartum. These data suggest that retarded lung growth during prenatal life as a result of in utero exposure to the coal liquid contributes to a significant portion of the observed neonatal mortality. Furthermore, lung weights of survivors, although significantly lower than control values through 7 days postpartum, appeared to have recovered by 21 days postpartum.

  4. 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)

  5. Parametric investigation on transient boiling heat transfer of metal rod cooled rapidly in water pool

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chi Young [Department of Fire Protection Engineering, Pukyong National University, 45, Yongso-ro, Nam-gu, Busan 48513 (Korea, Republic of); Kim, Sunwoo, E-mail: swkim@alaska.edu [Mechanical Engineering Department, University of Alaska Fairbanks, P. O. Box 755905, Fairbanks, AK 99775-5905 (United States)

    2017-03-15

    Highlights: • Effects of liquid subcooling, surface coating, material property, and surface oxidation are examined. • Liquid subcooling affects remarkably the quenching phenomena. • Cr-coated surfaces for ATF might extend the quenching duration. • Solids with low heat capacity shorten the quenching duration. • Surface oxidation can affect strongly the film boiling heat transfer and MFB point. - Abstract: In this work, the effects of liquid subcooling, surface coating, material property, and surface oxidation on transient pool boiling heat transfer were investigated experimentally using the vertical metal rod and quenching method. The change in rod temperature was measured with time during quenching, and the visualization of boiling around the test specimen was performed using the high-speed video camera. As the test materials, the zircaloy (Zry), stainless steel (SS), niobium (Nb), and copper (Cu) were tested. In addition, the chromium-coated niobium (Cr-Nb) and chromium-coated stainless steel (Cr-SS) were prepared for accident tolerant fuel (ATF) application. Low liquid subcooling and Cr-coating shifted the quenching curve to the right, which indicates a prolongation of quenching duration. On the other hand, the material with small heat capacity and surface oxidation caused the quenching curve to move to the left. To examine the influence of the material property and surface oxidation on the film boiling heat transfer performance and minimum film boiling (MFB) point in more detail, the wall temperature and heat flux were calculated from the present transient temperature profile using the inverse heat transfer analysis, and then the curves of wall temperature and heat flux in the film boiling regime were obtained. In the present experimental conditions, the effect of material property on the film boiling heat transfer performance and MFB point seemed to be minor. On the other hand, based on the experimental results of the Cu test specimen, the surface

  6. An electrical simulator of a nuclear fuel rod cooled by nucleate boiling

    International Nuclear Information System (INIS)

    Costa, Antonio Carlos Lopes da; Machado, Luiz; Koury, Ricardo Nicolau Nassar; Passos, Julio Cesar

    2009-01-01

    This study investigates an electrical heated test section designed to simulate a nuclear fuel rod. This simulator comprises a stainless steel vertical tube, with length and outside diameter of 600 mm and 10 mm, respectively, inside which there is a high power electrical resistor. The heat generated is removed by means of enhanced confined subcooled nucleate boiling of water in an annular space containing 153 small metal inclined discs. The tests were performed under electrical power and pressure up to 48 kW and 40 bar, respectively. The results show that the experimental boiling heat transfer coefficients are in good agreement with those calculated using the Jens-Lottes correlation. (author)

  7. An electrical simulator of a nuclear fuel rod cooled by nucleate boiling

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Antonio Carlos Lopes da [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)], e-mail: aclc@cdtn.br; Machado, Luiz; Koury, Ricardo Nicolau Nassar [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Mecanica], e-mail: luizm@demec.ufmg.br; Bonjour, Jocelyn [CETHIL, UMR5008, CNRS, INSA-Lyon (France)], e-mail: jocelyn.bonjour@insa-lyon.fr; Passos, Julio Cesar [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil). Dept. de Engenharia Mecanica. LEPTEN/Boiling], e-mail: jpassos@emc.ufsc.br

    2009-07-01

    This study investigates an electrical heated test section designed to simulate a nuclear fuel rod. This simulator comprises a stainless steel vertical tube, with length and outside diameter of 600 mm and 10 mm, respectively, inside which there is a high power electrical resistor. The heat generated is removed by means of enhanced confined subcooled nucleate boiling of water in an annular space containing 153 small metal inclined discs. The tests were performed under electrical power and pressure up to 48 kW and 40 bar, respectively. The results show that the experimental boiling heat transfer coefficients are in good agreement with those calculated using the Jens-Lottes correlation. (author)

  8. Investigation on the minimum film boiling temperature on metallic and ceramic heaters

    International Nuclear Information System (INIS)

    Ladisch, R.

    1980-06-01

    The minimum film boiling temperature on ceramic and metallic heaters has been experimentally studied. The knowledge of this temperature boundary is important in safety considerations on all liquid cooled nuclear reactors. The experiments have been carried out by quenching a hot metal cylinder with and without ceramic coating of aluminium in water. Results show that the minimum film boiling temperature Tsub(min) increases with water subcooling and is dependend upon the thermophysical properties of the heating surface. The roughness of the heater does not affect Tsub(min). At low subcoolings the vapour film is more stable and seems to break down when the specific heatflux upon liquid solid contact is lower than a threshold value above which film boiling can be reestablished. At higher subcoolings instead the vapour film is thinner and more stable. In this case the surface temperature decreases beyond the value by which the specific heatflux upon liquid solid contact would be lower than the threshold value. As soon as the vapour film becomes unstable, it collapses. (orig.) [de

  9. Enabling Highly Effective Boiling from Superhydrophobic Surfaces

    Science.gov (United States)

    Allred, Taylor P.; Weibel, Justin A.; Garimella, Suresh V.

    2018-04-01

    A variety of industrial applications such as power generation, water distillation, and high-density cooling rely on heat transfer processes involving boiling. Enhancements to the boiling process can improve the energy efficiency and performance across multiple industries. Highly wetting textured surfaces have shown promise in boiling applications since capillary wicking increases the maximum heat flux that can be dissipated. Conversely, highly nonwetting textured (superhydrophobic) surfaces have been largely dismissed for these applications as they have been shown to promote formation of an insulating vapor film that greatly diminishes heat transfer efficiency. The current Letter shows that boiling from a superhydrophobic surface in an initial Wenzel state, in which the surface texture is infiltrated with liquid, results in remarkably low surface superheat with nucleate boiling sustained up to a critical heat flux typical of hydrophilic wetting surfaces, and thus upends this conventional wisdom. Two distinct boiling behaviors are demonstrated on both micro- and nanostructured superhydrophobic surfaces based on the initial wetting state. For an initial surface condition in which vapor occupies the interstices of the surface texture (Cassie-Baxter state), premature film boiling occurs, as has been commonly observed in the literature. However, if the surface texture is infiltrated with liquid (Wenzel state) prior to boiling, drastically improved thermal performance is observed; in this wetting state, the three-phase contact line is pinned during vapor bubble growth, which prevents the development of a vapor film over the surface and maintains efficient nucleate boiling behavior.

  10. Modeling of the acoustic boiling noise of sodium during an assembly blockage in sodium-cooled reactors

    International Nuclear Information System (INIS)

    Vanderhaegen, M.

    2013-01-01

    In the framework of the fourth generation of nuclear reactors safety requirements, the acoustic boiling detection is studied to detect subassembly blockages. Boiling, that might occur during subassembly blockages and that can lead to clad failure, generates hydrodynamic noise that can be related to the two-phase flow. A bubble dynamics study shows that the sound source during subassembly boiling is condensation. This particular phenomenon generates most noise as a high subcooling is present in the subassembly and because of the high thermal diffusivity of sodium. This result leads to an estimate of the form of the acoustic spectrum that will be filtered and amplified during propagation inside the liquid. And even though it is unlikely that bubbles will be present inside the subassembly, due to the very gradual temperature profile at the wall and due to the geometry that leads to a strong confinement of the vapor, the historical bubble dynamics approach gives some insight in previous measurements. Additionally, some hypotheses can be disproved. These theoretical ideas are validated with a small water experiment, yet it also shows that a simple experience in sodium doesn't lead to a better knowledge of the acoustic source. A theoretical analysis also revealed that a realistic experiment with a simulant fluid, such as water or mercury, isn't representative. A similar conclusion is obtained when studying cavitation as a simulant acoustic source. As such, the acoustic detection of boiling, in comparison with other detection systems, isn't sufficiently developed yet to be applied as a reactor protective system. (author) [fr

  11. Large-scale boiling experiments of the flooded cavity concept for in-vessel core retention

    International Nuclear Information System (INIS)

    Chu, T.Y.; Slezak, S.E.; Bentz, J.H.; Pasedag, W.F.

    1994-01-01

    This paper presents results of ex-vessel boiling experiments performed in the CYBL (CYlindrical BoiLing) facility. CYBL is a reactor-scale facility for confirmatory research of the flooded cavity concept for accident management. CYBL has a tank-within-a-tank design; the inner tank simulates the reactor vessel and the outer tank simulates the reactor cavity. Experiments with uniform and edge-peaked heat flux distributions up to 20 W/cm 2 across the vessel bottom were performed. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling is mainly due to the gravity head which results from flooding the sides of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid/solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion (ejection). The results suggest that under prototypic heat load and heat flux distributions, the flooded cavity in a passive pressurized water reactor like the AP-600 should be capable of cooling the reactor pressure vessel in the central region of the lower head that is addressed by these tests

  12. 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

  13. Development of natural circulation small and medium sized boiling water reactor: HSBWR-600

    International Nuclear Information System (INIS)

    Miki, Minoru; Horiuchi, Tetsuo; Yoshimoto, Yuichiro; Sumida, Isao; Murase, Michio; Akita, Minoru; Niino, Tsuyoshi

    1988-01-01

    In nuclear power generation, the development of large reactors has been promoted as the main energy source in Japan. However, world economy entered low growth age, and the growth of electric power demand slowed down. Accordingly, attention has been paid to the medium and small reactors that can cope with whatever needs by serializing their types in addition to the nuclear power plants of medium output matching to electric power demand. In order to cope with these new needs, the economical efficiency of medium and small reactors must be as close as possible to that of large reactors, and as the countermeasures to the demerits due to small size, those must be made into the plants having simplified systems and the safety easily acceptable to public. Hitachi Ltd. plans to develop the natural circulation type medium and small BWRs of 600 NWe output class, HSBWR-600, on the basis of the nuclear power plant technology based on the rich results of design and operation of BWRs obtained so far, and to rank them as one of the BWR series. The target of their development design, the circumstance of their development, the core design and the thermo-hydraulic characteristics, the reactor pressure vessel and in-core structures, the safety design, system design, building layout and the evaluation are reported. (Kako, I.)

  14. Critical heat flux of subcooled flow boiling in narrow rectangular channels

    International Nuclear Information System (INIS)

    Kureta, Masatoshi; Akimoto, Hajime

    1999-01-01

    In relation to the high-heat-load devices such as a solid-target cooling channel of a high-intensity neutron source, burnout experiments were performed to obtain critical heat flux (CHF) data systematically for vertical upward flow in one-side heated rectangular channels. One of the objectives of this study was to study an extensibility of existing CHF correlations and models, which were proposed for a round tube, to rectangular channels for design calculation. Existing correlations and models were reviewed and compared with obtained data. Sudo's thin liquid layer dryout model, Griffel correlation and Bernath correlation were in good agreement with the experimental data for short-heated-length and low inlet water temperature conditions. (author)

  15. Critical heat flux for downward-facing pool boiling on CANDU calandria tube surface

    Energy Technology Data Exchange (ETDEWEB)

    Behdadi, Azin, E-mail: behdada@mcmaster.ca; Talebi, Farshad; Luxat, John

    2017-04-15

    Highlights: • Pressure tube-calandria tube contact may challenge fuel channel integrity in CANDU. • Critical heat flux variation is predicted on the outer surface of CANDU calandria tube. • A two-phase boundary layer flow driven by buoyancy is modeled on the surface. • Different slip ratios and flow regimes are considered inside the boundary layer. • Subcooling effects are added to the model using wall heat flux partitioning. - Abstract: One accident scenario in CANDU reactors that can challenge the integrity of the primary pressure boundary is a loss of coolant accident, referred to as critical break LOCA, in which the pressure tube (PT) can undergo thermal creep strain deformation and contact its calandria tube (CT). In such case, rapid redistribution of stored heat from PT to CT, leads to a large spike in heat flux to the moderator which can cause bubble accumulation and dryout on the CT surface. A challenge to fuel channel integrity is posed if critical heat flux occurs on the surface of the CT and results in sustained film boiling. If the post-dryout temperature becomes sufficiently high then continued creep strain of the PT and CT may lead to fuel channel failure. In this study, a mechanistic model is developed to predict the critical heat flux variations along the downward facing outer surface of CT. The hydrodynamic model considers a liquid macrolayer beneath an elongated vapor slug on the surface. Local dryout is postulated to occur whenever the fresh liquid supply to the macrolayer is not sufficient to compensate for the liquid depletion. A boundary layer analysis is performed, treating the two phase motion as an external buoyancy driven flow. The model shows good agreement with the available experimental data and has been modified to take into account the effect of subcooling.

  16. influence of sub-cooling on the energy performance of two eco

    African Journals Online (AJOL)

    PUBLICATIONS1

    frigerants, consistently exhibited better performance than R22 in sub-cooling heat ... 2014 Kwame Nkrumah University of Science and Technology (KNUST) ... sales volume among all refrigerants. .... The sub-cooling heat exchanger affects the.

  17. Development of a model to predict flow oscillations in low-flow sodium boiling

    International Nuclear Information System (INIS)

    Levin, A.E.; Griffith, P.

    1980-04-01

    Tests performed in a small scale water loop showed that voiding oscillations, similar to those observed in sodium, were present in water, as well. An analytical model, appropriate for either sodium or water, was developed and used to describe the water flow behavior. The experimental results indicate that water can be successfully employed as a sodium simulant, and further, that the condensation heat transfer coefficient varies significantly during the growth and collapse of vapor slugs during oscillations. It is this variation, combined with the temperature profile of the unheated zone above the heat source, which determines the oscillatory behavior of the system. The analytical program has produced a model which qualitatively does a good job in predicting the flow behavior in the wake experiment. The amplitude discrepancies are attributable to experimental uncertainties and model inadequacies. Several parameters (heat transfer coefficient, unheated zone temperature profile, mixing between hot and cold fluids during oscillations) are set by the user. Criteria for the comparison of water and sodium experiments have been developed

  18. Development of a water boil-off spent-fuel calorimeter system

    International Nuclear Information System (INIS)

    Creer, J.M.; Shupe, J.W. Jr.

    1981-05-01

    A calorimeter system was developed to measure decay heat generation rates of unmodified spent fuel assemblies from commercial nuclear reactors. The system was designed, fabricated, and successfully tested using the following specifications: capacity of one BWR or PWR spent fuel assembly; decay heat generation range 0.1 to 2.5 kW; measurement time of < 12 h; and an accuracy of +-10% or better. The system was acceptance tested using a dc reference heater to simulate spent fuel assembly heat generation rates. Results of these tests indicated that the system could be used to measure heat generation rates between 0.5 and 2.5 kW within +- 5%. Measurements of heat generation rates of approx. 0.1 kW were obtained within +- 15%. The calorimeter system has the potential to permit measurements of heat generation rates of spent fuel assemblies and other devices in the 12- to 14-kW range. Results of calorimetry of a Turkey Point spent fuel assembly indicated that the assembly was generating approx. 1.55 kW

  19. Current status and development of heat exchangers for boiling water reactor nuclear power plant

    International Nuclear Information System (INIS)

    Uchiyama, Yoshio; Nishioka, Shuji; Ito, Shizuo

    1975-01-01

    More efficient and reliable operation is required for BWR heat exchangers because of nuclear plant safety and other reasons. Heat exchangers are classified into two categories of systems, the system for normal operation and the system for emergency operation. The present state and future improvement of heat exchangers are described in view of heat transfer performance, material selection, structural design, vibration, and so on. When noncondensing gas exists in vapour, heat transfer performance deteriorates, so that the heat transfer characteristics should be corrected by the adaption of venting the non condensing gas from the system. Heat transfer tubes should have high corrosion resistance to working fluid as well as high thermal conductivity, strength and economy. From that point of view, 30% cupro-nickel tubes will be replaced with 10% cupro-nickel tubes or titanium tubes though some technical development is necessary. These heat exchangers are now designed according to the MITI criteria for the technology concerning nuclear and thermal power generation, ASME Boiler and Pressure Vessel Code Sec. III and some other criteria. Most of heat transfer tube failures are caused from the tube vibration induced by working fluid flow, so that the vibration test and analysis were performed on U-tube elements. Some correction was obtained for design and fabrication techniques. (Iwase, T.)

  20. LMFBR safety and sodium boiling

    Energy Technology Data Exchange (ETDEWEB)

    Hinkle, W.D.; Tschamper, P.M.; Fontana, M.H.; Henry, R.E.; Padilla, A. Jr.

    1978-01-01

    Within the U.S. Fast Breeder Reactor Safety R and D Work Breakdown Structure for Line of Assurance 2, Limit Core Damage, the influence of sodium boiling upon the progression and termination of accidents is being studied in loss of flow, transient overpower, loss of piping integrity, loss of shutdown heat removal system and local fault situations. The pertinent analytical and experimental results of this research to date are surveyed and compared with the requirements for demonstrating the effectiveness of this line of assurance. A discussion of specific technical issues concerned with sodium boiling and the need for future development work is also presented.

  1. Study of the hovering period and bubble size in fully developed pool nucleate boiling of saturated liquid with a time-dependent heat source

    International Nuclear Information System (INIS)

    Pasamehmetoglu, K.O.; Nelson, R.A.

    1987-01-01

    In this paper, the bubble behavior in saturated pool boiling with a time-dependent heat source is analyzed. The study is restricted to the period from fully developed nucleate boiling until critical heat flux occurs. The hovering period and the departure volume of the bubble are selected as the characteristic parameters for bubble behavior. These parameters are quantified by solving the equation of motion for an idealized bubble. This equation is solved for cases in which the surface heat flux changes linearly and exponentially as a function of time. After nondimensionalization, the results are compared directly with the results of the steady-state problem. The comparison shows that the transient heat input has practically no effect on the hovering period. However, the transient heat flux causes a decreased volume at bubble departure. The volume decrease is dependent on the severity of the transient. These results are in qualitative agreement with the experimental observation quoted in the literature

  2. New methods of subcooled water recognition in dew point hygrometers

    Science.gov (United States)

    Weremczuk, Jerzy; Jachowicz, Ryszard

    2001-08-01

    Two new methods of sub-cooled water recognition in dew point hygrometers are presented in this paper. The first one- impedance method use a new semiconductor mirror in which the dew point detector, the thermometer and the heaters were integrated all together. The second one an optical method based on a multi-section optical detector is discussed in the report. Experimental results of both methods are shown. New types of dew pont hydrometers of ability to recognized sub-cooled water were proposed.

  3. Boiling and fragmentation behaviour during fuel-sodium interactions

    International Nuclear Information System (INIS)

    Schins, H.; Gunnerson, F.S.

    1986-01-01

    A selection of the results and subsequent analysis of molten fuel-sodium interaction experiments conducted within the JRC BETULLA I and II facilities are reported. The fuels were copper and stainless steel, at initial temperatures far above their melting points; or urania and alumina, initially at their melting points. For each test, the molten fuel masses were in lower kilogram range and the subcooled pool mass was either 160 or 4 kg. The sodium pool was instrumented continually monitor the system temperature and pressure. Post-test examination results of the fragmented fuel debris sizes, shape and crystalline structure are given. The results of this study suggest the following: Transition boiling is the dominant boiling mode for the tested fuels in subcooled sodium. Two fragmentation mechanisms, vapour bubble formation/collapse and thermal stress shrinkage cracking prevailed for the oxide fuels. This was evidenced by the presence of both smooth and fractured particulate. In contrast, all metal fuel debris was smooth, suggesting fragmentation by the vapour bubble formation/collapse mechanism only during the molten state and for each test, there was no evidence of an energetic fuel-coolant interaction. (orig.)

  4. Acoustic emissions of a boiling liquid - an experimental survey in water and extrapolation to SFRs

    International Nuclear Information System (INIS)

    Vanderhaegen, M.; Paumel, K.; Tourin, A.

    2013-06-01

    The acoustic detection of sodium boiling is seen as a promising and innovative surveillance technique for sodium-cooled fast reactors (SFRs). It could be especially useful to detect in-core boiling that are the consequence of initiating accidents or whilst the mean subassembly temperature is very close to the nominal value. This latter is a consequence of the fuel assembly design of SFRs. Furthermore, it is a technique that has been proven to be successful in the past to follow the boiling behavior during SFR experiments that were aimed at simulating accidental conditions. However its effectiveness as in-core instrumentation still has to be demonstrated. In that aim, the acoustic emissions of sodium boiling in subassemblies are studied. Experimental studies are however limited to the boiling of common coolants due to the complications that arise when boiling liquid metals. As such, simple water experiments are performed. And although the results of these experiments are not completely representative for sodium boiling due to the incomplete thermo-hydraulic similarities between sodium and water, they can provide an interesting knowledge of the many influences that control the acoustic pressure field. In this article we'll specifically show how the condensation of vapor in subcooled liquid, the principal contribution to the acoustic emissions during boiling and hence the acoustic spectrum, is influenced by a pin-bundle geometry. We study this influence by comparing pool boiling experimental acoustic recordings with those of a simple pin-bundle geometry boiling experiment. The qualitative link, between this relatively simple pin-bundle experiment and the condensation phenomena that take place during sodium boiling inside SFR subassemblies, is used as a basis for this analysis. This simple experimental evidence, together with other theoretical arguments based on a thorough analysis of the sodium material properties, enables us to deduce that simple sodium

  5. Converting high boiling hydrocarbons

    Energy Technology Data Exchange (ETDEWEB)

    Terrisse, H; DuFour, L

    1929-02-12

    A process is given for converting high boiling hydrocarbons into low boiling hydrocarbons, characterized in that the high boiling hydrocarbons are heated to 200 to 500/sup 0/C in the presence of ferrous chloride and of such gases as hydrogen, water gas, and the like gases under a pressure of from 5 to 40 kilograms per square centimeter. The desulfurization of the hydrocarbons occurs simultaneously.

  6. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    Nishimura, N.; Nakai, H.; Ross, M.A.

    1999-01-01

    In the Boiling Water Reactor (BWR) system, steam generated within the nuclear boiler is sent directly to the main turbine. This direct cycle steam delivery system enables the BWR to have a compact power generation building design. Another feature of the BWR is the inherent safety that results from the negative reactivity coefficient of the steam void in the core. Based on the significant construction and operation experience accumulated on the BWR throughout the world, the ABWR was developed to further improve the BWR characteristics and to achieve higher performance goals. The ABWR adopted 'First of a Kind' type technologies to achieve the desired performance improvements. The Reactor Internal Pump (RIP), Fine Motion Control Rod Drive (FMCRD), Reinforced Concrete Containment Vessel (RCCV), three full divisions of Emergency Core Cooling System (ECCS), integrated digital Instrumentation and Control (I and C), and a high thermal efficiency main steam turbine system were developed and introduced into the ABWR. (author)

  7. On subcooler design for integrated two-temperature supermarket refrigeration system

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Liang [Institute of Refrigeration and Cryogenics, Shanghai Jiaotong University, Shanghai 200240 (China); Zhang, Chun-Lu [College of Mechanical Engineering, Tongji University, No. 4800, Cao An Highway, Shanghai 201804 (China)

    2011-01-15

    The energy saving opportunity of supermarket refrigeration systems using subcooler between the medium-temperature (MT) refrigeration system and the low-temperature (LT) refrigeration system has been identified in the previous work. This paper presents a model-based comprehensive analysis on the subcooler design. The optimal subcooling control is discussed as well. With optimal subcooler size and subcooling control, the maximum energy savings of integrated two-temperature supermarket refrigeration system using R404A or R134a as working fluid can achieve 27% or 20%, respectively. The load ratio of MT to LT system and the operating conditions have considerable impact on the energy savings. (author)

  8. 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

  9. Evaporation, Boiling and Bubbles

    Science.gov (United States)

    Goodwin, Alan

    2012-01-01

    Evaporation and boiling are both terms applied to the change of a liquid to the vapour/gaseous state. This article argues that it is the formation of bubbles of vapour within the liquid that most clearly differentiates boiling from evaporation although only a minority of chemistry textbooks seems to mention bubble formation in this context. The…

  10. 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.)

  11. Studies on boiling heat transfer on a hemispherical downward heating surface supposing IVR-AM

    International Nuclear Information System (INIS)

    Yoshida, Kenji; Matsumoto, Hiroyuki; Matsumoto, Tadayoshi; Kataoka, Isao

    2006-01-01

    The scale-down experiments supposing the IVR-AM were made on the pool boiling heat transfer from hemispherical downward facing heating surface. The boiling phenomena were realized by flooding the heated hemispherical vessel into the sub-cooled water or saturated water under the atmospheric pressure. The hemispherical vessel supposing the scale-down pressure vessel was made of SUS304 stainless steel. Molten lead, which was preheated up to about 500 degrees Celsius, was put into the vessel and used as the heat source. The vessel was cooled down by flooding into the water to realize the quenching process. The direct observation by using the digital video camera was performed and made clear the special characteristics of boiling phenomena such as the film boiling, the transition boiling and the nucleate boiling taking place in order during the cooling process. The measurement for the wall superheat and heat flux by using thermocouples was also carried out to make clear the boiling heat transfer characteristics during the cooling process. Fifteen thermocouples are inserted in the wall of the hemispherical bowl to measure the temperature distributions and heat flux in the hemispherical bowl. (author)

  12. Nuclear fuel performance in boiling water reactors

    International Nuclear Information System (INIS)

    Elkins, R.B.; Baily, W.E.; Proebstle, R.A.; Armijo, J.S.; Klepfer, H.H.

    1981-01-01

    A major development program is described to improve the performance of Boiling Water Reactor fuel. This sustained program is described in four parts: 1) performance monitoring, 2) fuel design changes, 3) plant operating recommendations, and 4) advanced fuel programs

  13. Flow behaviour in a CANDU horizontal fuel channel from stagnant subcooled initial conditions

    International Nuclear Information System (INIS)

    Caplan, M.Z.; Gulshani, P.; Holmes, R.W.; Wright, A.C.D.

    1984-01-01

    The flow behaviour in a CANDU primary system with horizontal fuel channels is described following a small inlet header break. With the primary pumps running, emergency coolant injection is in the forward direction so that the channel outlet feeders remain warmer than the inlet thereby promoting forward natural circulation. However, the break force opposes the forward driving force. Should the primary pumps run down after the circuit has refilled, there is a break size for which the natural circulation force is balanced by the break force and channels could, theoretically, stagnate. Result of visualization and of full-size channel tests on channel flow behaviour from an initially stagnant channel condition are discussed. After a channel stagnation, the decay power heats the coolant to saturation. Steam is then formed and the coolant stratifies. The steam expands into the subcooled water in the end fitting in a chugging type of flow regime due to steam condensation. After the end fitting reaches the saturation temperature, steam is able to penetrate into the vertical feeder thereby initiating a large buoyancy induced flow which refills the channel. The duration of stagnation is shown to be sensitive to small asymmetries in the initial conditions. A small initial flow can significantly shorten the occurrence and/or duration of boiling as has been confirmed by reactor experience. (author)

  14. 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

  15. Investigation of the minimum film boiling temperature of water during rewetting under forced convective conditions

    International Nuclear Information System (INIS)

    Huang, X.C.; Bartsch, G.; Wang, B.X.

    1992-01-01

    The minimum film boiling temperature of water has been measured on a copper hollow cylinder of 50 mm length with the mass flux rate ranging from 25 to 500 kg/m 2 s and the pressure from 0.1 to 1.0 MPa at subcoolings of 5 to 50 K. Film boiling is established with help of a temperature-controlled system. Rewetting can be initiated by cutting off or very gradually reducing the power supply to the test section. A numerical method for solving the two-dimensional nonlinear inverse heat conduction problem is utilized in the data reduction, taking into account the axial heat conduction. The results are compared with the steady-state maximum transition boiling temperatures measured on the same test section and with the true quench temperatures available in the literature so far. (4 figures, 1 table) (Author)

  16. Burnout in boiling heat transfer. part I: pool boiling systems

    International Nuclear Information System (INIS)

    Bergles, A.E.

    1977-01-01

    Recent experimental and analytical developments in pool-boiling burnout are reviewed, and results are summarized that clarify the dependence of critical heat flux on heater geometry and fluid properties. New analytical interpretations of burnout are discussed, and the effects of surface condition, aging, acceleration, and transient heating (or cooling) are described. The relation of sound to burnout and new techniques for stabilizing electric heaters at burnout are also considered

  17. 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

  18. Effect of degree of subcooling on vapor explosion

    International Nuclear Information System (INIS)

    Xu Zhihong; Yang Yanhua; Li Tianshu

    2010-01-01

    In order to investigate the mechanism of the vapor explosion, an observable experiment equipment for low-temperature molten materials to be dropped into water was designed. In the experiment, molten material jet was injected into water to experimentally obtain the visualized information. This experiment results show that the degree of subcooling restrains the explosion. In order to validate the result by other aspects, the breakup experiment was conducted. Results show that the degree of water subcooling is important to melt breakup. High temperature of water is easy to increase the vapor generation during molten material falling, which decrease the drag and accelerated the molten material falling. At the same time, more vapor appear around the molten metal decrease the heat transfer amount between water and molten materials. The two experimental results coincide. (authors)

  19. Downward transfer of a sub-cooled cryoliquid

    CERN Document Server

    Wertelaers, P

    2016-01-01

    An alternative is proposed to the traditional transfer of a cryo fluid in gaseous -- and warm -- form, a method of low productivity and high energy cost. In order to prevent the much-feared geysering, focus is on sub-cooling of the liquid, and the safe maintaining of such state all along the journey. A cryogenic transfer line of simplest construction is proposed, and the difficulties with such line extending over a transfer depth of the order of the kilometre, are discussed.

  20. Odd-Boiled Eggs

    Science.gov (United States)

    Kaminsky, Kenneth; Scheman, Naomi

    2010-01-01

    At a Shabbat lunch in Madrid not long ago, the conversation turned to the question of boiling eggs. One of the guests mentioned that a Dutch rabbi he knew had heard that in order to make it more likely that boiled eggs be kosher, you should add an egg to the pot if the number you began with was even. According to the laws of Kashruth, Jews may not…

  1. Critical flashing flows in nozzles with subcooled inlet conditions

    International Nuclear Information System (INIS)

    Abuaf, N.; Jones, O.C. Jr.; Wu, B.J.C.

    1983-01-01

    Examination of a large number of experiments dealing with flashing flows in converging and converging-diverging nozzles reveals that knowledge of the flashing inception point is the key to the prediction of critical flow rates. An extension of the static flashing inception correlation of Jones [16] and Alamgir and Lienhard [17] to flowing systems has allowed the determination of the location of flashing inception in nozzle flows with subcooled inlet conditions. It is shown that in all the experiments examined with subcooled inlet regardless of the degree of inlet subcooling, flashing inception invariably occurred very close to the throat. A correlation is given to predict flashing inception in both pipes and nozzles which matches all data available, but is lacking verification in intermediate nozzle geometries where turbulence may be important. A consequence of this behavior is that the critical mass flux may be correlated to the pressure difference between the nozzle inlet and flashing inception, through a single phase liquid discharge coefficient and an accurate prediction of the flashing inception pressure at the throat. Comparison with the available experiments indicate that the predicted mass fluxes are within 5 percent of the measurements

  2. Extending cavitation models to subcooled and superheated nozzle flow

    International Nuclear Information System (INIS)

    Schmidt, D.P.; Corradini, M.L.

    1997-01-01

    Existing models for cavitating flow are extended to apply to discharge of hot liquid through nozzles. Two types of models are considered: an analytical model and a two-dimensional numerical model. The analytical model of cavitating nozzle flow is reviewed and shown to apply to critical nozzle flow where the liquid is subcooled with respect to the downstream conditions. In this model the liquid and vapor are assumed to be in thermodynamic equilibrium. The success of this analytical model suggests that hydrodynamic effects dominate the subcooled nozzle flow. For more detailed predictions an existing multi-dimensional cavitation model based on hydrodynamic non-equilibrium is modified to apply to discharge of hot liquid. Non-equilibrium rate data from experimental measurements are used to close the equations. The governing equations are solved numerically in time and in two spatial dimensions on a boundary fitted grid. Results are shown for flow through sharp nozzles, and the coefficient of discharge is found to agree with experimental measurements for both subcooled and flashing fluid. (author)

  3. 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.)

  4. Nucleate boiling heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Saiz Jabardo, J.M. [Universidade da Coruna (Spain). Escola Politecnica Superior], e-mail: mjabardo@cdf.udc.es

    2009-07-01

    Nucleate boiling heat transfer has been intensely studied during the last 70 years. However boiling remains a science to be understood and equated. In other words, using the definition given by Boulding, it is an 'insecure science'. It would be pretentious of the part of the author to explore all the nuances that the title of the paper suggests in a single conference paper. Instead the paper will focus on one interesting aspect such as the effect of the surface microstructure on nucleate boiling heat transfer. A summary of a chronological literature survey is done followed by an analysis of the results of an experimental investigation of boiling on tubes of different materials and surface roughness. The effect of the surface roughness is performed through data from the boiling of refrigerants R-134a and R-123, medium and low pressure refrigerants, respectively. In order to investigate the extent to which the surface roughness affects boiling heat transfer, very rough surfaces (4.6 {mu}m and 10.5 {mu}m ) have been tested. Though most of the data confirm previous literature trends, the very rough surfaces present a peculiar behaviour with respect to that of the smoother surfaces (Ra<3.0 {mu}m). (author)

  5. Nucleate boiling heat transfer

    International Nuclear Information System (INIS)

    Saiz Jabardo, J.M.

    2009-01-01

    Nucleate boiling heat transfer has been intensely studied during the last 70 years. However boiling remains a science to be understood and equated. In other words, using the definition given by Boulding, it is an 'insecure science'. It would be pretentious of the part of the author to explore all the nuances that the title of the paper suggests in a single conference paper. Instead the paper will focus on one interesting aspect such as the effect of the surface microstructure on nucleate boiling heat transfer. A summary of a chronological literature survey is done followed by an analysis of the results of an experimental investigation of boiling on tubes of different materials and surface roughness. The effect of the surface roughness is performed through data from the boiling of refrigerants R-134a and R-123, medium and low pressure refrigerants, respectively. In order to investigate the extent to which the surface roughness affects boiling heat transfer, very rough surfaces (4.6 μm and 10.5 μm ) have been tested. Though most of the data confirm previous literature trends, the very rough surfaces present a peculiar behaviour with respect to that of the smoother surfaces (Ra<3.0 μm). (author)

  6. Thermal-hydraulic instabilities in pressure tube graphite - moderated boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling channels in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement.

  7. Thermal-hydraulic instabilities in pressure tube graphite-moderated boiling water reactors

    International Nuclear Information System (INIS)

    Tsiklauri, G.; Schmitt, B.

    1995-09-01

    Thermally induced two-phase instabilities in non-uniformly heated boiling charmers in RBMK-1000 reactor have been analyzed using RELAP5/MOD3 code. The RELAP5 model of a RBMK-1000 reactor was developed to investigate low flow in a distribution group header (DGH) supplying 44 fuel pressure tubes. The model was evaluated against experimental data. The results of the calculations indicate that the period of oscillation for the high power tube varied from 3.1s to 2.6s, over the power range of 2.0 MW to 3.0 MW, respectively. The amplitude of the flow oscillation for the high powered tube varied from +100% to -150% of the tube average flow. Reverse flow did not occur in the lower power tubes. The amplitude of oscillation in the subcooled region at the inlet to the fuel region is higher than in the saturated region at the outlet. In the upper fuel region and outlet connectors the flow oscillations are dissipated. The threshold of flow instability for the high powered tubes of a RBMK reactor is compared to Japanese data and appears to be in good agreement

  8. Conservatism in methodologies for moderator subcooling sufficiency for fuel channel integrity upon pressure tube and calandria tube contact

    Energy Technology Data Exchange (ETDEWEB)

    Sun, L., E-mail: LSun@nbpower.com [Point Lepreau Generating Station, Lepreau, NB, (Canada)

    2015-07-01

    During a postulated large LOCA event in CANDU reactors, the pressure tube may balloon to contact with its surrounding calandria tube to transfer heat to the moderator. To confirm the integrity of the fuel channel in this case, many experiments have been performed in the last three decades. Based on the extant database of the pressure tube/calandria tube (PT/CT) contact, an analytical methodology was developed by Canadian Nuclear Industry to determine the sufficiency of moderator subcooling for fuel channel integrity. At the same time a semi-empirical methodology with an idea of Equivalent Moderator Subcooling (EMS) was also developed to judge the sufficiency of the moderator. In this work, some discussions were made over the two methodologies on their conservatism and it is demonstrated that the analytical approach is over conservative comparing with the EMS methodology. By using the EMS methodology, it is demonstrated that applying glass-peened calandria tubes, the requirement to moderator subcooling can be reduced by 10{sup o}C from that for smooth calandria tubes. (author)

  9. Relationship between high quality CHF and boiling length in annulus geometry with uniformly heated rod

    International Nuclear Information System (INIS)

    Chun, S. Y.; Mun, S. K.; Park, J. K.; Yang, S. K.; Jung, M. K.

    1999-01-01

    The relationship between the boiling length and the CHF in annulus geometry with uniformly heated rod has been studied. In this study the CHF data under pressure of 0.57∼15.01 MPa, flow rate of 200∼650 kg/m 2 s, inlet subcooling of 85∼353 kJ/kg and exit quality of 0.106∼0.536 have been applied. As a result of examining the flow pattern over the heated section, all of the CHF data were the dryout type CHF in annular flow and the locations of the churn to annular flow transition moved down stream of the heated section with increasing the pressure. The effect of pressure on the boiling length under the CHF conditions showed the trends similar to the effect of pressure on the CHF. The relationship between the non-dimensional CHF, q CHF and mass flux taking into account of the boiling length, G ( L h / L B ) indicated the linear relationship without scatter and regardless of pressure and inlet subcooling. The CHF calculated by using the relationship between the non-dimensionless CHF, q CHF and mass flux, G ( L h / L B ) predicted very well the experimental CHF data with the pressure dependence

  10. Visualization of nucleate pool boiling of freon 113

    International Nuclear Information System (INIS)

    Afify, M.A.; Fruman, D.H.

    1987-01-01

    The purpose of this investigation is to give a fine description of the behaviour of vapour bubbles in nucleate pool boiling at sites of known sizes using high speed photography. The shapes and growth history of isolated bubbles were determined for a variety of experimental conditions. Coalescence effects between two adjacent or consecutive bubbles were also visualized and the occurrence of vapour patches and continuous vapour columns was demonstrated. Quantitative analysis of the films allows to determine the history and nucleation characteristics of bubbles as a function of various parameters such as heat flux, liquid subcooling and size and nature of nucleation sites. These results are in good agreement with those found in the literature

  11. 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)

  12. CARR-CNS with crescent-shape moderator cell and sub-cooling helium jacket around cell

    International Nuclear Information System (INIS)

    Yu, Qingfeng; Feng, Quanke; Kawai, Takeshi; Cheng, Liang; Shen, Feng; Yuan, Luzheng

    2005-01-01

    The new type of the moderator cell was developed for the Cold Neutron Source (CNS) of the China Advanced Research Reactor (CARR) which is now constructing at the China Institute of Atomic Energy in Beijing. A crescent-shape moderator cell covered by the sub-cooling helium jacket is adopted. A crescent-shape would help to increase the volume of the moderator cell for corresponding it to the 4 cold neutron guide tubes, even if liquid hydrogen not liquid deuterium were used as a cold moderator. The sub-cooling helium jacket covering the moderator cell removes the nuclear heating of the outer shell wall of the cell. It contributes to reduce the void fraction of liquid hydrogen in the inner shell. Such a type of a moderator cell is suitable for the CNS with higher nuclear heating. The cold helium gas flows down firstly into the sub-cooling helium jacket and then flows up to the condenser. Therefore, the theory of the self-regulation for the thermo-siphon type of the CNS is also applicable

  13. Strength analysis of CARR-CNS with crescent-shape moderator cell and helium sub-cooling jacket covering cell

    International Nuclear Information System (INIS)

    Yu Qingfeng; Feng Quanke; Kawai Takeshi; Shen Feng; Yuan Luzheng; Cheng Liang

    2005-01-01

    The new type of the moderator cell was developed for the cold neutron source (CNS) of the China Advanced Research Reactor (CARR) which is now being constructed at the China Institute of Atomic Energy in Beijing. A crescent-shape moderator cell covered by the helium sub-cooling jacket is adopted. The structure of the moderator cell is optimized by the stress FEM analysis. A crescent-shape would help to increase the volume of the moderator cell for fitting it to the four cold neutron guide tubes, even if liquid hydrogen, not liquid deuterium, was used as a cold moderator. The helium sub-cooling jacket covering the moderator cell removes the nuclear heating of the outer shell wall of the cell. It contributes to reduce the void fraction of liquid hydrogen in the outer shell of the moderator cell. Such a type of a moderator cell is suitable for the CNS with higher nuclear heating. The cold helium gas flows down first into the helium sub-cooling jacket and then flows up to the condenser. The theory of the self-regulation suitable to the thermo-siphon type of the CNS is also applicable and validated

  14. CARR-CNS with crescent-shape moderator cell and sub-cooling helium jacket surrounding cell

    International Nuclear Information System (INIS)

    Yu, Qingfeng; Feng, Quanke; Kawai, Takeshi; Shen, Feng; Yuan, Luzheng

    2005-01-01

    The new type of the moderator cell was developed for the Cold Neutron Source (CNS) of the China Advanced Research Reactor (CARR) which is now constructing at the China Institute of Atomic Energy in Beijing. A crescent-shape moderator cell covered by the sub-cooling helium jacket is adopted. A crescent-shape would help to increase the volume of the moderator cell for corresponding it to the 4 cold neutron guide tubes, even if liquid hydrogen not liquid deuterium were used as a cold moderator. The sub-cooling helium jacket covering the moderator cell removes the nuclear heating of the outer shell wall of the cell. It contributes to reduce the void fraction of liquid hydrogen in the inner shell. Such a type of a moderator cell is suitable for the CNS with higher nuclear heating. The cold helium gas flows down firstly into the sub-cooling helium jacket and then flows up to the condenser. Therefore, the theory of the self-regulation for the thermo-siphon type of the CNS is also applicable

  15. Development and validation of models for simulation of supercritical carbon dioxide Brayton cycles and application to self-propelling heat removal systems in boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Venker, Jeanne

    2015-03-31

    The objective of the current work was to develop a model that is able to describe the transient behavior of supercritical carbon dioxide (sCO{sub 2}) Brayton cycles, to be applied to self-propelling residual heat removal systems in boiling water reactors. The developed model has been implemented into the thermohydraulic system code ATHLET. By means of this improved ATHLET version, novel residual heat removal systems, which are based on closed sCO{sub 2} Brayton cycles, can be assessed as a retrofit measure for present light water reactors. Transient simulations are hereby of great importance. The heat removal system has to be modeled explicitly to account for the interaction between the system and the behavior of the plant during different accident conditions. As a first step, transport and thermodynamic fluid properties of supercritical carbon dioxide have been implemented in ATHLET to allow for the simulation of the new working fluid. Additionally, a heat transfer correlation has been selected to represent the specific heat transfer of supercritical carbon dioxide. For the calculation of pressure losses due to wall friction, an approach for turbulent single phase flow has been adopted that is already implemented in ATHLET. In a second step, a component model for radial compressors has been implemented in the system code. Furthermore, the available model for axial turbines has been adapted to simulate the transient behavior of radial turbines. All extensions have been validated against experimental data. In order to simulate the interaction between the self-propelling heat removal system and a generic boiling water reactor, the components of the sCO{sub 2} Brayton cycle have been dimensioned with first principles. An available input deck of a generic BWR has then been extended by the residual heat removal system. The modeled application has shown that the extended version of ATHLET is suitable to simulate sCO{sub 2} Brayton cycles and to evaluate the introduced

  16. Development and validation of models for simulation of supercritical carbon dioxide Brayton cycles and application to self-propelling heat removal systems in boiling water reactors

    International Nuclear Information System (INIS)

    Venker, Jeanne

    2015-01-01

    The objective of the current work was to develop a model that is able to describe the transient behavior of supercritical carbon dioxide (sCO 2 ) Brayton cycles, to be applied to self-propelling residual heat removal systems in boiling water reactors. The developed model has been implemented into the thermohydraulic system code ATHLET. By means of this improved ATHLET version, novel residual heat removal systems, which are based on closed sCO 2 Brayton cycles, can be assessed as a retrofit measure for present light water reactors. Transient simulations are hereby of great importance. The heat removal system has to be modeled explicitly to account for the interaction between the system and the behavior of the plant during different accident conditions. As a first step, transport and thermodynamic fluid properties of supercritical carbon dioxide have been implemented in ATHLET to allow for the simulation of the new working fluid. Additionally, a heat transfer correlation has been selected to represent the specific heat transfer of supercritical carbon dioxide. For the calculation of pressure losses due to wall friction, an approach for turbulent single phase flow has been adopted that is already implemented in ATHLET. In a second step, a component model for radial compressors has been implemented in the system code. Furthermore, the available model for axial turbines has been adapted to simulate the transient behavior of radial turbines. All extensions have been validated against experimental data. In order to simulate the interaction between the self-propelling heat removal system and a generic boiling water reactor, the components of the sCO 2 Brayton cycle have been dimensioned with first principles. An available input deck of a generic BWR has then been extended by the residual heat removal system. The modeled application has shown that the extended version of ATHLET is suitable to simulate sCO 2 Brayton cycles and to evaluate the introduced heat removal system

  17. 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

  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. Ex-vessel boiling experiments: laboratory- and reactor-scale testing of the flooded cavity concept for in-vessel core retention. Pt. II. Reactor-scale boiling experiments of the flooded cavity concept for in-vessel core retention

    International Nuclear Information System (INIS)

    Chu, T.Y.; Bentz, J.H.; Slezak, S.E.; Pasedag, W.F.

    1997-01-01

    For pt.I see ibid., p.77-88 (1997). This paper summarizes the results of a reactor-scale ex-vessel boiling experiment for assessing the flooded cavity design of the heavy water new production reactor. The simulated reactor vessel has a cylindrical diameter of 3.7 m and a torispherical bottom head. Boiling outside the reactor vessel was found to be subcooled nucleate boiling. The subcooling mainly results from the gravity head, which in turn results from flooding the side of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid-solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion. The results show that, under prototypic heat load and heat flux distributions, the flooded cavity will be effective for in-vessel core retention in the heavy water new production reactor. The results also demonstrate that the heat dissipation requirement for in-vessel core retention, for the central region of the lower head of an AP-600 advanced light water reactor, can be met with the flooded cavity design. (orig.)

  20. Density wave oscillations of a boiling natural circulation loop induced by flashing

    Energy Technology Data Exchange (ETDEWEB)

    Furuya, Masahiro; Inada, Fumio; Yasuo, Akira [Central Research Institute of Electric Power Industry, Tokyo (Japan)

    1995-09-01

    Experiments are conducted to investigate two-phase flow instabilities in a boiling natural circulation loop with a chimney due to flashing in the chimney at lower pressure. The test facility used in this experiment is designed to have non-dimensional values which are nearly equal to those of natural circulation BWR. Stability maps in reference to the heat flux, the inlet subcooling, the system pressure are presented. This instability is suggested to be density wave oscillations due to flashing in the chimney, and the differences from other phenomena such as flow pattern oscillations and geysering phenomena are discussed by investigating the dynamic characteristics, the oscillation period, and the transient flow pattern.

  1. Effect of liquid subcooling on acoustic characteristics during the condensation process of vapor bubbles in a subcooled pool

    International Nuclear Information System (INIS)

    Tang, Jiguo; Yan, Changqi; Sun, Licheng; Li, Ya; Wang, Kaiyuan

    2015-01-01

    Highlights: • Deviations of signals increase first and then decrease with increase in subcooling. • Two typical waveforms are observed and correspond to bubble split-up and collapse. • Dominant frequency in low frequency region is found for all condensation regimes. • Peaks in high frequency region were only found in capillary wave regime. • Bubble collapse frequency is close to frequency of first peak in amplitude spectra. - Abstract: Sound characteristics of direct contact condensation of vapor bubbles in a subcooled pool were investigated experimentally with a hydrophone and a high-speed video camera. Three different condensation modes were observed, which were referred to as shape oscillation regime, transition regime and capillary wave regime in the paper. Time domain analysis indicated that the acoustic signals were boosted in their maximum amplitude with increase in subcooling, while their standard and average absolute deviations shifted to decrease after reaching a peak value. In addition, two different waveforms were found, possible sources of which were split-up and collapse of bubbles, respectively. From the amplitude spectra obtained by FFT, the first dominant frequency was found at frequency of 150–300 Hz for all condensation regimes, whereas some peaks in high frequency region were observed only for the capillary wave regime. The first dominant frequency was the result of the periodic variation in the vapor bubble volume, and the peaks in high frequency region were due to the high-frequency oscillation of water in pressure caused by sudden bubble collapse. The frequency of first peak was considered to be resulted from the periodic bubble collapse or split-up and thus was close to the bubble collapse frequency obtained from snapshots of bubble condensation. Moreover, according to results of short-time Fourier transform (STFT), the time intervals in which a certain process of bubble condensing occurred could be well known.

  2. Dry patch formed boiling and burnout in potassium pool boiling

    International Nuclear Information System (INIS)

    Michiyoshi, I.; Takenaka, N.; Takahashi, O.

    1986-01-01

    Experimental results are presented on dry patch formed boiling and burnout in saturated potassium pool boiling on a horizontal plane heater for system pressures from 30 to 760 torr and liquid levels from 5 to 50 mm. The dry patch formation occurs in the intermittent boiling which is often encountered when liquid alkali metals are used under relatively low pressure conditions. Burnout is caused from both continuous nucleate and dry patch formed boiling. The burnout heat flux together with nucleate boiling heat transfer coefficients are empirically correlated with system pressures. A model is also proposed to predict the minimum heat flux to form the dry patch. (author)

  3. 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)

  4. 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)

  5. Research on axial total pressure distributions of sonic steam jet in subcooled water

    International Nuclear Information System (INIS)

    Wu Xinzhuang; Li Wenjun; Yan Junjie

    2012-01-01

    The axial total pressure distributions of sonic steam jet in subcooled water were experimentally investigated for three different nozzle diameters (6.0 mm, 8.0 mm and 10.0 mm). The inlet steam pressure, and pool subcooling subcooled water temperature were in the range of 0.2-0.6 MPa and 420-860 ℃, respectively. The effect of steam pressure, subcooling water temperature and nozzle size on the axial pressure distributions were obtained, and also the characteristics of the maximum pressure and its position were studied. The results indicated that the characteristics of the maximum pressure were influenced by the nozzle size for low steam pressure, but the influence could be ignored for high steam pressure. Moreover, a correlation was given to correlate the position of the maximum pressure based on steam pressure and subcooling water temperature, and the discrepancies of predictions and experiments are within ±15%. (authors)

  6. 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.

  7. Boiling phenomenon and heat transfer in bead-packed porous structure

    International Nuclear Information System (INIS)

    Zhang Xiaojie; ZHu Yanlei; Bai Bofeng; Yan Xiao; Xiao Zejun

    2009-01-01

    A visual study on pool boiling behavior and phase distribution was conducted on the porous structures made of staggered glass beads at atmospheric pressure. The bead-packed structure was heated on the bottom. The investigations were carried out respectively at different glass bead diameters which were 4 mm, 6 mm and 8 mm. The results show that during subcooled boiling, small isolated bubbles are formed on the heated surface and combine into main-bubbles, the dispersion frequency of the main-bubbles is low and the small bubbles scatter in the bead-packed porous structures. At the initial stage of saturated boiling, the bubble growth rate, the volume of main-bubbles and the range of continuous vapor phase increase. The dispersion frequency of main-bubbles increases with the increasing of heat flux. During film boiling, the heated surface is absolutely covered with vapor film and the porous structure is full of liquid. The larger the diameter of beads is, the higher heat flux is needed for the same phenomenon, and the higher maximum value of heat transfer coefficient will be. During the whole saturated boiling, and the heat transfer enhanced firstly and then weakened. Being opposite to that of the diameters of 4 mm and 8 mm, the heat transfer coefficient in the 6 mm-bead-packed porous structure decreases with the increasing of the heat flux. (authors)

  8. 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.

  9. Measurements of Burnout Conditions for Flow of Boiling Water in Vertical Round Ducts (Part 2)

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Kurt M; Mathisen, R P; Eklind, O; Norman, B

    1964-01-15

    The hydrodynamic stability and the burnout conditions for flow of boiling water have been studied in a natural circulation loop in the pressure range from 10 to 70 atg. The test section was a round, duct of 20 mm inner diameter and 4890 mm heated length. The experimental results showed that within the ranges tested the stability of the flow increases with increasing pressure, increasing throttling before the test section, but decreases with increasing inlet sub-cooling and increasing throttling after the test section. The measured thresholds of instability compared well with the analytical results by Jahnberg. For an inlet sub-cooling temperature of about 2 deg C the measured burnout steam qualities were low by a factor of about 1.3 compared to forced circulation data obtained with the same test section. At higher sub-cooling temperatures the discrepancy between forced and natural circulation data increased, so that at {delta}t{sub sub} = 16 deg C, the natural circulation data were low by a factor of about 2.5. However, by applying inlet throttling of the flow the burnout values approached and finally coincided with the forced circulation data.

  10. Mechanistic modeling of pool film-boiling and quench on a Candu calandria tube following a critical break LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, J.T.; Luxat, J.C. [McMaster University, A315 JHE Building, 1280 Main St.W. Hamilton, ON, L8S 4L7 (Canada)

    2008-07-01

    Following a postulated critical LBLOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a Candu CT (approximately 13 cm). A second order, non-linear and non-homogeneous ODE for vapour film thickness has been derived. The variation of steady state vapour film thickness prior to quench as a function of subcooling temperature, wall superheat, and incident heat flux is examined. The CT outer surface heatup rate and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (authors)

  11. Mechanistic modeling of pool film-boiling and quench on a Candu calandria tube following a critical break LOCA

    International Nuclear Information System (INIS)

    Jiang, J.T.; Luxat, J.C.

    2008-01-01

    Following a postulated critical LBLOCA a pressure tube (PT) can experience creep deformation and balloon uniformly into contact with the calandria tube (CT). The resultant heat flux to CT is high as stored heat is transferred out of the hot PT. This heat flux can cause dryout on the outer surface of the CT and establish film boiling. This paper presents a model of buoyancy-driven natural convection film boiling on the outside of a horizontal tube with diameter relevant to a Candu CT (approximately 13 cm). A second order, non-linear and non-homogeneous ODE for vapour film thickness has been derived. The variation of steady state vapour film thickness prior to quench as a function of subcooling temperature, wall superheat, and incident heat flux is examined. The CT outer surface heatup rate and effective film boiling heat transfer coefficient from the model are in good agreement with available experimental data. (authors)

  12. Geysering in boiling channels

    Energy Technology Data Exchange (ETDEWEB)

    Aritomi, Masanori; Takemoto, Takatoshi [Tokyo Institute of Technology, Tokyo (Japan); Chiang, Jing-Hsien [Japan NUS Corp. Ltd., Toyko (Japan)] [and others

    1995-09-01

    A concept of natural circulation BWRs such as the SBWR has been proposed and seems to be promising in that the primary cooling system can be simplified. The authors have been investigating thermo-hydraulic instabilities which may appear during the start-up in natural circulation BWRs. In our previous works, geysering was investigated in parallel boiling channels for both natural and forced circulations, and its driving mechanism and the effect of system pressure on geysering occurrence were made clear. In this paper, geysering is investigated in a vertical column and a U-shaped vertical column heated in the lower parts. It is clarified from the results that the occurrence mechanism of geysering and the dependence of system pressure on geysering occurrence coincide between parallel boiling channels in circulation systems and vertical columns in non-circulation systems.

  13. Applications of Artificial Neural Network for the Prediction of Pool Boiling Curves

    International Nuclear Information System (INIS)

    Su, Guanghui; Fukuda, K.; Morita, K.

    2002-01-01

    Artificial neural network (ANN) has the advantage that the best-fit correlations of experimental data will no longer be necessary for predicting unknowns from the known parameters. The ANN was applied to predict the pool boiling curves in this paper. The database of experimental data presented by Berenson, Dhuga et al., and Bui and Dhir etc. were used in the analysis. The database is subdivided in two subsets. The first subset is used to train the network and the second one is used to test the network after the training process. The input parameters of the ANN are: wall superheat ΔT w , surface roughness, steady/transient heating/transient cooling, subcooling, Surface inclination and pressure. The output parameter is heat flux q. The proposed methodology allows us to achieve the accuracy that satisfies the user's convergence criterion and it is suitable for pool boiling curve data processing. (authors)

  14. 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)

  15. High boiling point hydrocarbons

    Energy Technology Data Exchange (ETDEWEB)

    Pier, M

    1929-04-29

    A process is given for the production of hydrocarbons of high boiling point, such as lubricating oils, from bituminous substances, such as varieties of coal, shale, or other solid distillable carbonaceous materials. The process consists of treating the initial materials with organic solvents and then subjecting the products extracted from the initial materials, preferably directly, to a reducing treatment in respect to temperature, pressure, and time. The reduction treatment is performed by means of hydrogen under pressure.

  16. 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

  17. 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.

  18. An analytic model of pool boiling critical heat flux on an immerged downward facing curved surface

    International Nuclear Information System (INIS)

    He, Hui; Pan, Liang-ming; Wu, Yao; Chen, De-qi

    2015-01-01

    Highlights: • Thin liquid film and supplement of liquid contribute to the CHF. • CHF increases from the bottom to the upper of the lowerhead. • Evaporation of thin liquid film is dominant nearby bottom region. • The subcooling has significant effects on the CHF. - Abstract: In this paper, an analytical model of the critical heat flux (CHF) on the downward facing curved surface for pool boiling has been proposed, which hypothesizes that the CHF on the downward facing curved is composed of two parts, i.e. the evaporation of the thin liquid film underneath the elongated bubble adhering to the lower head outer surface and the depletion of supplement of liquid due to the relative motion of vapor bubbles along with the downward facing curved. The former adopts the Kelvin–Helmholtz instability analysis of vapor–liquid interface of the vapor jets which penetrating in the thin liquid film. When the heat flux closing to the CHF point, the vapor–liquid interface becomes highly distorted, which block liquid to feed the thin liquid film and the thin liquid film will dry out gradually. While the latter considers that the vapor bubbles move along with the downward facing curved surface, and the liquid in two-phase boundary layer enter the liquid film that will be exhausted when the CHF occurs. Based on the aforementioned mechanism and the energy balance between the thin liquid film evaporation and water feeding, and taking the subcooling of the bulk water into account, the mathematic model about the downward facing curved surface CHF has been proposed. The CHF of the downward facing curved surface for pool boiling increases along with the downward facing orientation except in the vicinity of bottom center region, because in this region the vapor bubble almost stagnates and the evaporation of the thin liquid film is dominant. In addition, the subcooling has significant effect on the CHF. Comparing the result of this model with the published experimental results show

  19. 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)

  20. Revision of nucleated boiling mechanisms

    International Nuclear Information System (INIS)

    Converti, J.; Balino, J.L.

    1987-01-01

    The boiling occurrence plays an important role in the power reactors energy transfer. But still, there is not a final theory on the boiling mechanisms. This paper presents a critical analysis of the most important nucleated boiling models that appear in literature. The conflicting points are identified and experiments are proposed to clear them up. Some of these experiments have been performed at the Thermohydraulics laboratory (Bariloche Atomic Center). (Author)

  1. Technology, safety and costs of decommissioning a reference boiling water reactor power station: Comparison of two decommissioning cost estimates developed for the same commercial nuclear reactor power station

    International Nuclear Information System (INIS)

    Konzek, G.J.; Smith, R.I.

    1990-12-01

    This study presents the results of a comparison of a previous decommissioning cost study by Pacific Northwest Laboratory (PNL) and a recent decommissioning cost study of TLG Engineering, Inc., for the same commercial nuclear power reactor station. The purpose of this comparative analysis on the same plant is to determine the reasons why subsequent estimates for similar plants by others were significantly higher in cost and external occupational radiation exposure (ORE) than the PNL study. The primary purpose of the original study by PNL (NUREG/CR-0672) was to provide information on the available technology, the safety considerations, and the probable costs and ORE for the decommissioning of a large boiling water reactor (BWR) power station at the end of its operating life. This information was intended for use as background data and bases in the modification of existing regulations and in the development of new regulations pertaining to decommissioning activities. It was also intended for use by utilities in planning for the decommissioning of their nuclear power stations. The TLG study, initiated in 1987 and completed in 1989, was for the same plant, Washington Public Supply System's Unit 2 (WNP-2), that PNL used as its reference plant in its 1980 decommissioning study. Areas of agreement and disagreement are identified, and reasons for the areas of disagreement are discussed. 31 refs., 3 figs., 22 tabs

  2. Analysis of two-phase flow instability in vertical boiling channels I: development of a linear model for the inlet velocity perturbation

    International Nuclear Information System (INIS)

    Hwang, D.H.; Yoo, Y.J.; Kim, K.K.

    1998-08-01

    A linear model, named ALFS, is developed for the analysis of two-phase flow instabilities caused by density wave oscillation and flow excursion in a vertical boiling channel with constant pressure drop conditions. The ALFS code can take into account the effect of the phase velocity difference and the thermally non-equilibrium phenomena, and the neutral boundary of the two-phase flow instability was analyzed by D-partition method. Three representative two-phase flow models ( i.e. HEM, DEM, and DNEM) were examined to investigate the effects on the stability analysis. As the results, it reveals that HEM shows the most conservative prediction of heat flux at the onset of flow instability. three linear models, Ishiis DEM, Sahas DNEM, and ALFS model, were applied to Sahas experimental data of density wave oscillation, and as the result, the mean and standard deviation of the predicted-to-measured heat flux at the onset of instability were calculated as 0.93/0.162, 0.79/0.112, and 0.95/0.143, respectively. For the long test section, however, ALFS model tends to predict the heat fluxes about 30 % lower than the measured values. (author). 14 refs

  3. 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

  4. Development of BWR [boiling water reactor] and PWR [pressurized water reactor] event descriptions for nuclear facility simulator training

    International Nuclear Information System (INIS)

    Carter, R.J.; Bovell, C.R.

    1987-01-01

    A number of tools that can aid nuclear facility training developers in designing realistic simulator scenarios have been developed. This paper describes each of the tools, i.e., event lists, events-by-competencies matrices, and event descriptions, and illustrates how the tools can be used to construct scenarios

  5. Heat transfer correlation development and assessment: a summary and assessment of return to nucleate boiling phenomena during blowdown tests conducted at the Idaho National Engineering Laboratory (INEL)

    International Nuclear Information System (INIS)

    Eaton, A.M.; Tolman, E.L.

    1979-04-01

    The data are presented which were obtained in Loss-of-Coolant Experiments (LOCE) at Idaho National Engineering Laboratory (INEL) which demonstrate the presence of cladding rewetting after the critical heat flux has been exceeded as a viable cooling mechanism during the blowdown phase of a LOCE. A brief review of the mechanisms associated with the boiling crisis and rewetting is also provided. The relevance of INEL LOCE rewetting data to nuclear reactor licensing Evaluation Model Requirements is considered, and the conclusion is made that the elimination of rewetting and return to nucleate boiling (RNB) in Evaluation Models represents a definite conservatism

  6. The CEA program on boiling noise detection

    International Nuclear Information System (INIS)

    Le Guillou, G.; Brunet, M.; Girard, J.P.; Flory, D.

    1982-01-01

    The research program on the application of noise analysis on boiling detection in a fast subassembly began 10 years ago at the CEA, mainly in the Nuclear Center of Cadarache. Referring exclusively to the aspects of premature detection of the boiling phenomenon it can be said that this program is organized around the following three detection techniques: acoustic noise analysis; neutron noise analysis; temperature noise analysis. Its development is in conjunction with in-pile experiments in Phenix or Rapsodie as well as 'ex-pile' (boiling experiments through electric heating). Three detection techniques were developed independent of each other, but that they were regrouped during the execution of the most important experiments and with the 'Super Phenix' project. The noise analysis system ANABEL with which Superphenix will be equipped with shows the industrial interest in detection methods based on noises. One of the results of the CEA program today is the possibility to evaluate the potential capacity for boiling detection in the subassembly. But in order to obtain the necessary funds from the commercial nuclear plant operators it is mandatory to have successful demonstrations which will be the objective of the future program

  7. Size-exclusion chromatography for the determination of the boiling point distribution of high-boiling petroleum fractions.

    Science.gov (United States)

    Boczkaj, Grzegorz; Przyjazny, Andrzej; Kamiński, Marian

    2015-03-01

    The paper describes a new procedure for the determination of boiling point distribution of high-boiling petroleum fractions using size-exclusion chromatography with refractive index detection. Thus far, the determination of boiling range distribution by chromatography has been accomplished using simulated distillation with gas chromatography with flame ionization detection. This study revealed that in spite of substantial differences in the separation mechanism and the detection mode, the size-exclusion chromatography technique yields similar results for the determination of boiling point distribution compared with simulated distillation and novel empty column gas chromatography. The developed procedure using size-exclusion chromatography has a substantial applicability, especially for the determination of exact final boiling point values for high-boiling mixtures, for which a standard high-temperature simulated distillation would have to be used. In this case, the precision of final boiling point determination is low due to the high final temperatures of the gas chromatograph oven and an insufficient thermal stability of both the gas chromatography stationary phase and the sample. Additionally, the use of high-performance liquid chromatography detectors more sensitive than refractive index detection allows a lower detection limit for high-molar-mass aromatic compounds, and thus increases the sensitivity of final boiling point determination. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  8. Development of a model to predict flow oscillations in low-flow sodium boiling. [Loss-of-Piping Integrity accidents

    Energy Technology Data Exchange (ETDEWEB)

    Levin, A.E.; Griffith, P.

    1980-04-01

    Tests performed in a small scale water loop showed that voiding oscillations, similar to those observed in sodium, were present in water, as well. An analytical model, appropriate for either sodium or water, was developed and used to describe the water flow behavior. The experimental results indicate that water can be successfully employed as a sodium simulant, and further, that the condensation heat transfer coefficient varies significantly during the growth and collapse of vapor slugs during oscillations. It is this variation, combined with the temperature profile of the unheated zone above the heat source, which determines the oscillatory behavior of the system. The analytical program has produced a model which qualitatively does a good job in predicting the flow behavior in the wake experiment. The amplitude discrepancies are attributable to experimental uncertainties and model inadequacies. Several parameters (heat transfer coefficient, unheated zone temperature profile, mixing between hot and cold fluids during oscillations) are set by the user. Criteria for the comparison of water and sodium experiments have been developed.

  9. Customer-operator partnership. A boiling water reactor developed jointly by AREVA NP and E.ON Kernkraft

    International Nuclear Information System (INIS)

    Pasler, Doris; Gauthier, Jean Claude; Diercks, Frank; Fuchs, Michael

    2009-01-01

    Many countries spread all over the world have publicly expressed their intention to pursue the construction of new nuclear power plants with improved safety, economy and more straight forwarded operation and maintenance. Reasons for the intention are: The world wide increasing demand for energy and hence the general necessity to build new power plants. The concerns for increased emissions of green house gases leading to a change in the climate have brought into question the primary reliance on plants utilizing fossil fuels. A new reactor type matching the previously stated issues is AREVA NP's further development of proven BWR design. Combining AREVA's and E.ON's expertise, a project was launched to customize the final basic design for this advanced nuclear power plant having a net power output of about 1,250 MW, a net efficiency of about 37% and a design service life of 60 years. Within this joint venture the overall plant design was simplified and additionally all active safety systems have passive safety related backup systems utilizing basic laws of physics, such as gravity, enabling them to function without electrical power supplies or activation by powered instrumentation and control systems. The development takes into account the technical and accumulated operating experience of the project partners. Based on the operating experience of the project partners a simplification of the overall system engineering was performed, flexible fuel cycle length (12 to 24 months) are possible as well as a reduction of process waste was achieved. These improvements regarding the operation and economics result on the one hand in lower investment cost and on the other hand in a high availability of the plant, hence in low maintenance costs. Generally, the electrical generation costs are accomplished, which are competitive to larger-capacity nuclear power plants and fossil-fired plants. (author)

  10. Breakup Behavior of Molten Wood's Metal Jet in Subcooled Water

    Energy Technology Data Exchange (ETDEWEB)

    Heo, Hyo; Park, Seong Dae; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook [Chung-Ang Univ., Seoul (Korea, Republic of)

    2014-10-15

    There are safety characteristics of the metal fueled sodium fast-cooled reactor (SFR), by identifying the possibility of early termination of severe accidents. If the molten fuel is ejected from the cladding, the ejected molten fuel can interact with the coolant in the reactor vessel. This phenomenon is called as fuel-coolant interaction (FCI). The FCI occurs at the initial phase leading to severe accidents like core disruptive accident (CDA) in the SFR. A part of the corium energy is intensively transferred to the coolant in a very short time during the FCI. The coolant vaporizes at high pressure and expands so results in steam explosion that can threat to the integrity of nuclear reactor. The intensity of steam explosion is determined by jet breakup and the fragmentation behavior. Therefore, it is necessary to understand the jet breakup between the molten fuel jet and the coolant in order to evaluate whether the steam explosion occurs or not. The liquid jet breakup has been studied in various areas, such as aerosols, spray and combustion. In early studies, small diameter jets of low density liquids were studied. The jet breakup for large density liquids has been studied in nuclear reactor field with respect to safety. The existence of vapor film layer between the melt and liquid fluid is only in case of large density breakup. This paper deals with the jet breakup experiment in non-boiling conditions in order to analyze hydraulic effect on the jet behavior. In the present study, the wood's metal was used as the jet material. It has similar properties to the metal fuel. The physical properties of molten materials and coolants are listed in Table I, respectively. It is easy to conduct the experiment due to low melting point of the wood's metal. In order to clarify the dominant factors determining jet breakup and size distribution of the debris, the experiment that the molten wood's metal was injected into the subcooled condition was conducted. The

  11. Local pool boiling heat transfer on a 3 Degree inclined tube surface

    International Nuclear Information System (INIS)

    Kang, Myeong Gie

    2012-01-01

    Mechanisms of pool boiling heat transfer have been studied for a long time. Recently, it has been widely investigated in nuclear power plants for the purpose of acquiring inherent safety functions in case of no power supply. To design more efficient heat exchangers, effects of several parameters on heat transfer must be studied in detail. One of the major issues is variation in local heat transfer coefficients on a tube. Lance and Myers reported that the type of boiling liquid can change the trend of local heat transfer coefficients along the tube periphery. Lance and Myers said that as the liquid is methanol the maximum local heat transfer coefficient was observed at the tube bottom while the maximum was at the tube sides as the boiling liquid was n hexane. Corn well and Einarsson reported that the maximum local heat transfer coefficient was observed at the tube bottom, as the boiling liquid was R113. Corn well and Houston explained the reason of the difference in local heat transfer coefficients along the tube circumference with introducing effects of sliding bubbles on heat transfer. According to Gu pta et al., the maximum and the minimum local heat transfer coefficients were observed at the bottom and top regions of the tube circumference, respectively, using a tube bundle and water. Kang also reported the similar results using a single horizontal tube and water. However, the maximum heat transfer coefficient was observed at the angle of 45 deg. Sateesh et al. investigated variations in local heat transfer coefficients along a tube periphery as the inclination angle was changed. Summarizing the published results, some parts are still remaining to be investigated in detail. Although pool boiling analysis on a nearly horizontal tube is necessary for the design of the advanced power reactor plus, no previous results are published yet. Therefore, the present study is aimed to study variations in local pool boiling heat transfer coefficients for a 3 degree inclined

  12. Boil-off experiments with the EIR-NEPTUN Facility: Analysis and code assessment overview report

    International Nuclear Information System (INIS)

    Aksan, S.N.; Stierli, F.; Analytis, G.T.

    1992-03-01

    The NEPTUN data discussed in this report are from core uncovery (boil-off) experiments designed to investigate the mixture level decrease and the heat up of the fuel rod simulators above the mixture level for conditions simulating core boil-off for a nuclear reactor under small break loss-of-coolant accident conditions. The first series of experiments performed in the NEPTUN test facility consisted of ten boil-off (uncovery) and one adiabatic heat-up tests. In these tests three parameters were varied: rod power, system pressure and initial coolant subcooling. The NEPTUN experiments showed that the external surface thermocouples do not cause a significant cooling influence in the rods to which they are attached under boil-off conditions. The reflooding tests performed later on indicated that the external surface thermocouples have some effect during reflooding for NEPTUN electrically heated rod bundle. Peak cladding temperatures are reduced by about 30--40C and quench times occur 20--70 seconds earlier than rods with embedded thermocouples. Additionally, the external surface-thermocouples give readings up to 20 K lower than those obtained with internal surface thermocouples (in the absence of external thermocouples) in the peak cladding temperature zone. Some of the boil-off data obtained from the NEPTUN test facility are used for the assessment of the thermal-hydraulic transient computer codes. These calculations were performed extensively using the frozen version of TRAC-BD1/MOD1 (version 22). A limited number of assessment calculations were done with RELAP5/MOD2 (version 36.02). In this report the main results and conclusions of these calculations are presented with the identification of problem areas in relation to models relevant to boil-off phenomena. On the basis of further analysis and calculations done, changing some of the models such as the bubbly/slug flow interfacial friction correlation which eliminate some of the problems are recommended

  13. Development of a water boil-off spent-fuel calorimeter system. [To measure decay heat generation rate

    Energy Technology Data Exchange (ETDEWEB)

    Creer, J.M.; Shupe, J.W. Jr.

    1981-05-01

    A calorimeter system was developed to measure decay heat generation rates of unmodified spent fuel assemblies from commercial nuclear reactors. The system was designed, fabricated, and successfully tested using the following specifications: capacity of one BWR or PWR spent fuel assembly; decay heat generation range 0.1 to 2.5 kW; measurement time of < 12 h; and an accuracy of +-10% or better. The system was acceptance tested using a dc reference heater to simulate spent fuel assembly heat generation rates. Results of these tests indicated that the system could be used to measure heat generation rates between 0.5 and 2.5 kW within +- 5%. Measurements of heat generation rates of approx. 0.1 kW were obtained within +- 15%. The calorimeter system has the potential to permit measurements of heat generation rates of spent fuel assemblies and other devices in the 12- to 14-kW range. Results of calorimetry of a Turkey Point spent fuel assembly indicated that the assembly was generating approx. 1.55 kW.

  14. 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

  15. Development of design technology on thermal-hydraulic performance in tight-lattice rod bundles. II-rod bowing effect on boiling transition

    International Nuclear Information System (INIS)

    Liu, Wei; Tamai, Hidesada; Kureta, Masatoshi; Ohnuki, Akira; Takase, Kazuyuki; Akimoto, Hajime

    2007-01-01

    A thermal-hydraulic feasibility project for an Innovative Water Reactor for Flexible fuel cycle (FLWR) has been performed since 2002. In this R and D project, large-scale thermal-hydraulic tests, several model experiments and development of advanced numerical analysis codes have been carried out. In this paper, we will describe the critical power characteristics in a 37-rod tight-lattice bundle with rod-bowing under both steady and transient states. It is observed that no matter it is run under a steady or a transient state, boiling transition (BT) always occurs axially at exit elevation of upper high-heat-flux region and transversely in the central area of the bundle. Steady critical power increases monotonically with the increase of mass velocity, with the decrease of inlet water temperature and with the decrease of exit pressure. These trends are same as those in the base case test without rod-bowing. The steady critical power with rod-bowing is about 10% lower than that without rod-bowing. For the postulated power increase and flow decrease cases that may be possibly met in a normal operation of the FLWR, it is confirmed that no BT occurs when Initial Critical Power Ratio (ICPR) is 1.3. Moreover, when the transitions are run under severer ICPR that causes BT, the transient critical powers are generally same as the steady ones. The experiments are analyzed with TRAC-BF1 code. The TRAC-BF1 code shows good prediction for the occurrence or the non occurrence of the BT and predicts the BT starting time within the accuracy of critical power correlation. Traditional quasi - steady state prediction of the transient BT is confirmed being applicable for the postulated abnormal transient processes in the tight lattice bundle with rod - bowing. (author)

  16. When water does not boil at the boiling point.

    Science.gov (United States)

    Chang, Hasok

    2007-03-01

    Every schoolchild learns that, under standard pressure, pure water always boils at 100 degrees C. Except that it does not. By the late 18th century, pioneering scientists had already discovered great variations in the boiling temperature of water under fixed pressure. So, why have most of us been taught that the boiling point of water is constant? And, if it is not constant, how can it be used as a 'fixed point' for the calibration of thermometers? History of science has the answers.

  17. A modified Bernoulli equation for the calculation of the fluid dynamics of the region of subcooled water

    International Nuclear Information System (INIS)

    Pana, P.

    1975-12-01

    A mathematical model is derived to describe the fluid-dynamics for the region of subcooled water. The qualitative changes, when crossing the saturation line, from the wet-steam region to the subcooled region, are discussed with respect to thermodynamical considerations, and the change of state in the subcooled region is treated detailled, showing the limitation of validity for the theoretical model. (orig./TK) [de

  18. Signal processing techniques for sodium boiling noise detection

    International Nuclear Information System (INIS)

    1989-05-01

    At the Specialists' Meeting on Sodium Boiling Detection organized by the International Working Group on Fast Reactors (IWGFR) of the International Atomic Energy Agency at Chester in the United Kingdom in 1981 various methods of detecting sodium boiling were reported. But, it was not possible to make a comparative assessment of these methods because the signal condition in each experiment was different from others. That is why participants of this meeting recommended that a benchmark test should be carried out in order to evaluate and compare signal processing methods for boiling detection. Organization of the Co-ordinated Research Programme (CRP) on signal processing techniques for sodium boiling noise detection was also recommended at the 16th meeting of the IWGFR. The CRP on Signal Processing Techniques for Sodium Boiling Noise Detection was set up in 1984. Eight laboratories from six countries have agreed to participate in this CRP. The overall objective of the programme was the development of reliable on-line signal processing techniques which could be used for the detection of sodium boiling in an LMFBR core. During the first stage of the programme a number of existing processing techniques used by different countries have been compared and evaluated. In the course of further work, an algorithm for implementation of this sodium boiling detection system in the nuclear reactor will be developed. It was also considered that the acoustic signal processing techniques developed for boiling detection could well make a useful contribution to other acoustic applications in the reactor. This publication consists of two parts. Part I is the final report of the co-ordinated research programme on signal processing techniques for sodium boiling noise detection. Part II contains two introductory papers and 20 papers presented at four research co-ordination meetings since 1985. A separate abstract was prepared for each of these 22 papers. Refs, figs and tabs

  19. Zero Boil Off System for Cryogen Storage, Phase I

    Data.gov (United States)

    National Aeronautics and Space Administration — This work proposes to develop a zero boil off (ZBO) dewar using a two-stage pulse-tube cooler together with two innovative, continuous-flow cooling loops and an...

  20. Measurement of partial discharge inception characteristics in sub-cooled liquid nitrogen

    International Nuclear Information System (INIS)

    Koo, J.Y.; Lee, S.H.; Shin, W.J.; Khan, Umer A.; Oh, S.H.; Seong, J.K.; Lee, B.W.

    2011-01-01

    We measured partial discharge and partial discharge initiation voltage of subcooled liquid nitrogen. Various kinds of test samples have been prepared. Sub-cooled temperature in liquid nitrogen were changed. The number of PD pluses were decreased when 68 K liquid nitrogen was used. Sub-cooled liquid nitrogen has positive effects to suppress PD activities. Partial discharge (PD) measurement is one of the effective diagnostic techniques to predict abnormal high voltage dielectric insulation conditions of the electric equipments. PD diagnostic techniques were also could be utilized to evaluate the conditions of cryogenic dielectric insulation media of high temperature superconducting electric equipment in liquid nitrogen. Generally, liquid nitrogen at 77 K is used as cryogenic and dielectric media for high temperature superconducting devices for high voltage electric power systems. But due to generation of bubbles during quench conditions which cause harmful effect on the properties of liquid nitrogen insulation, sub-cooled nitrogen under 77 K was also employed to suppress bubble formation. In this work, investigation of PD characteristics of sub-cooled liquid nitrogen was conducted in order to clarify the relation between PD inception and the temperature of liquid nitrogen. It was observed that measured PDIV (PD inception voltage) shows little differences according to the sub-cooled temperature of liquid nitrogen, but the magnitude and total numbers of PD has been slightly decreased according the decrease of cooled temperature of liquid nitrogen. From experimental results, it was deduced that the sub-cooled liquid nitrogen from 68 K to 77 K, could be applicable without any considerations of the variation of PDIV.

  1. 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

  2. Numerical and Experimental Study of Mechanisms Involved in Boiling Histotripsy.

    Science.gov (United States)

    Pahk, Ki Joo; Gélat, Pierre; Sinden, David; Dhar, Dipok Kumar; Saffari, Nader

    2017-12-01

    The aim of boiling histotripsy is to mechanically fractionate tissue as an alternative to thermal ablation for therapeutic applications. In general, the shape of a lesion produced by boiling histotripsy is tadpole like, consisting of a head and a tail. Although many studies have demonstrated the efficacy of boiling histotripsy for fractionating solid tumors, the exact mechanisms underpinning this phenomenon are not yet well understood, particularly the interaction of a boiling vapor bubble with incoming incident shockwaves. To investigate the mechanisms involved in boiling histotripsy, a high-speed camera with a passive cavitation detection system was used to observe the dynamics of bubbles produced in optically transparent tissue-mimicking gel phantoms exposed to the field of a 2.0-MHz high-intensity focused ultrasound (HIFU) transducer. We observed that boiling bubbles were generated in a localized heated region and cavitation clouds were subsequently induced ahead of the expanding bubble. This process was repeated with HIFU pulses and eventually resulted in a tadpole-shaped lesion. A simplified numerical model describing the scattering of the incident ultrasound wave by a vapor bubble was developed to help interpret the experimental observations. Together with the numerical results, these observations suggest that the overall size of a lesion induced by boiling histotripsy is dependent on the sizes of (i) the heated region at the HIFU focus and (ii) the backscattered acoustic field by the original vapor bubble. Copyright © 2017 World Federation for Ultrasound in Medicine and Biology. Published by Elsevier Inc. All rights reserved.

  3. 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)

  4. 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)

  5. 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

  6. Micro transport phenomena during boiling

    CERN Document Server

    Peng, Xiaofeng

    2011-01-01

    "Micro Transport Phenomena During Boiling" reviews the new achievements and contributions in recent investigations at microscale. It presents some original research results and discusses topics at the frontier of thermal and fluid sciences.

  7. 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

  8. 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

  9. 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

  10. 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

  11. Sodium boiling studies at the CEA state of the art

    International Nuclear Information System (INIS)

    Girard, C.; Grand, D.; Papin, J.; Seiler, J.M.

    1979-08-01

    A description of the general approach used by the CEA to solve sodium boiling problems provides an understanding of our philosophy for code development. From the review of the main results obtained in the out-of-pile experiments, CFNa and CESAR, we deduce the main hypothesis of our basic model of sodium boiling. Our best estimate and simplified codes are briefly described and their results are compared with the experiments

  12. Conjugate heat transfer effects on wall bubble nucleation in subcooled flashing flows

    International Nuclear Information System (INIS)

    Peterson, P.F.; Hijikata, K.

    1990-01-01

    A variety of models have been proposed to explain observations that large liquid superheat is required to initiate nucleation in flashing flows of subcooled liquids in nozzles, cracks and pipes. In such flows an abrupt change in the fluid temperature occurs downstream of the nucleating cavities. This paper examines the subcooling of the nucleating cavities due to conjugate heat transfer to the cold downstream fluid. This examination suggests a mechanism limiting the maximum active cavity size. Simple analysis shows that, of the total superheat required to initiate flashing, a substantial portion results from conjugate wall subcooling, which decreases the cavity vapor pressure. The specific case of flashing critical nozzle flow is examined in detail. Here boundary-layer laminarization due to the strong favorable pressure gradient aids the analysis of conjugate heat transfer

  13. Cryogenic system with the sub-cooled liquid nitrogen for cooling HTS power cable

    Energy Technology Data Exchange (ETDEWEB)

    Fan, Y.F. [Chinese Academy of Sciences, Beijing (China). Technical Institute of Physics and Chemistry; Graduate School of Chinese Academy of Sciences, Beijing (China); Gong, L.H.; Xu, X.D.; Li, L.F.; Zhang, L. [Chinese Academy of Sciences, Beijing (China). Technical Institute of Physics and Chemistry; Xiao, L.Y. [Chinese Academy of Sciences, Beijing (China). Institute of Electrical Engineering

    2005-04-01

    A 10 m long, three-phase AC high-temperature superconducting (HTS) power cable had been fabricated and tested in China August 2003. The sub-cooled liquid nitrogen (LN{sub 2}) was used to cool the HTS cable. The sub-cooled LN{sub 2} circulation was built by means of a centrifugal pump through a heat exchanger in the sub-cooler, the three-phase HTS cable cryostats and a LN{sub 2} gas-liquid separator. The LN{sub 2} was cooled down to 65 K by means of decompressing, and the maximum cooling capacity was about 3.3 kW and the amount of consumed LN{sub 2} was about 72 L/h at 1500 A. Cryogenic system design, test and some experimental results would be presented in this paper. (author)

  14. 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

  15. 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)

  16. 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

  17. 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].

  18. Method for estimating boiling temperatures of crude oils

    International Nuclear Information System (INIS)

    Jones, R.K.

    1996-01-01

    Evaporation is often the dominant mechanism for mass loss during the first few days following an oil spill. The initial boiling point of the oil and the rate at which the boiling point changes as the oil evaporates are needed to initialize some computer models used in spill response. The lack of available boiling point data often limits the usefulness of these models in actual emergency situations. A new computational method was developed to estimate the temperature at which a crude oil boils as a function of the fraction evaporated using only standard distillation data, which are commonly available. This method employs established thermodynamic rules and approximations, and was designed to be used with automated spill-response models. Comparisons with measurements show a strong correlation between results obtained with this method and measured values

  19. Chaotic oscillations in a low pressure two-phase natural circulation loop under low power and high inlet subcooling conditions

    International Nuclear Information System (INIS)

    Wu, C.Y.; Wang, S.B.; Pan, C.

    1996-01-01

    The oscillation characteristics of a low pressure two-phase natural circulation loop have been investigated experimentally in this study. Experimental results indicate that the characteristics of the thermal hydraulic oscillations can be periodic, with 2-5 fundamental frequencies, or chaotic, depending on the heating power and inlet subcooling. The number of fundamental frequencies of oscillation increases if the inlet subcooling is increased at a given heating power or the heating power is decreased at a given inlet subcooling; chaotic oscillations appear if the inlet subcooling is further increased and/or the heating power is further decreased. A map of the oscillation characteristics is thus established. The change in oscillation characteristics is evident from the time evolution and power spectrum of a thermal hydraulic parameter and the phase portraits of two thermal hydraulic parameters. These results reveal that a strange attractor exists in a low pressure two-phase natural circulation loop with low power and very high inlet subcooling. (orig.)

  20. Procedures and instrumentation for sodium boiling experiments in EBR-II

    International Nuclear Information System (INIS)

    Crowe, R.D.

    1976-01-01

    The development of instrumentation capable of detecting localized coolant boiling in a liquid metal cooled breeder reactor (LMFBR) has a high priority in fast reactor safety. The detection must be rapid enough to allow corrective action to be taken before significant damage occurs to the core. To develop and test a method of boiling detection, it is desirable to produce boiling in a reactor and thereby introduce a condition in the reactor the original design concepts were chosen to preclude. The proposed boiling experiments are designed to safely produce boiling in the subassembly of a fast reactor and provide the information to develop boiling detection instrumentation without core damage or safety compromise. The experiment consists of the operation of two separate subassemblies, first, a gamma heated boiling subassembly which produces non-typical but highly conservative boiling and then a fission heated subassembly which simulates a prototypical boiling event. The two boiling subassemblies are designed to operate in the instrumentation subassembly test facility (INSAT) of Experiment Breeder Reactor II

  1. Characteristics of phenomenon and sound in microbubble emission boiling

    International Nuclear Information System (INIS)

    Zhu Guangyu; Sun Licheng; Tang Jiguo

    2014-01-01

    Background: Nowadays, the efficient heat transfer technology is required in nuclear energy. Therefore, micro-bubble emission boiling (MEB) is getting more attentions from many researchers due to its extremely high heat-transfer dissipation capability. Purpose: An experimental setup was built up to study the correspondences between the characteristics on the amplitude spectrum of boiling sound in different boiling modes. Methods: The heat element was a copper block heated by four Si-C heaters. The upper of the copper block was a cylinder with the diameter of 10 mm and height of 10 mm. Temperature data were measured by three T-type sheathed thermocouples fitted on the upper of the copper block and recorded by NI acquisition system. The temperature of the heating surface was estimated by extrapolating the temperature distribution. Boiling sound data were acquired by hydrophone and processed by Fourier transform. Bubble behaviors were captured by high-speed video camera with light system. Results: In nucleate boiling region, the boiling was not intensive and as a result, the spectra didn't present any peak. While the MEB fully developed on the heating surface, an obvious peak came into being around the frequency of 300 Hz. This could be explained by analyzing the video data. The periodic expansion and collapse into many extremely small bubbles of the vapor film lead to MEB presenting an obvious characteristic peak in its amplitude spectrum. Conclusion: The boiling mode can be distinguished by its amplitude spectrum. When the MEB fully developed, it presented a characteristic peak in its amplitude spectrum around the frequency between 300-400 Hz. This proved that boiling sound of MEB has a close relation with the behavior of vapor film. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Becker, Kurt M

    1962-07-01

    The present report deals with the results of the first phase of an experimental investigation of burnout conditions for flow of boiling water in vertical round ducts. Data were obtained in the following ranges of variables. Pressure 2.4subcooling 18<{delta}t{sub sub}< 192 deg C; Steam quality 0.37subcooling and channel length are negligible.

  3. Transient behavior of natural circulation for boiling two-phase flow, 2

    International Nuclear Information System (INIS)

    Aritomi, Masanori; Chiang, Jing-Hsien; Mori, Michitugu.

    1991-01-01

    In this set of experiments, natural circulation in boiling two-phase flow has been investigated for power transients, simulating the start-up process in a natural circulation BWR. This was done in order to understand the underlying mechanism of thermo-hydraulic instability which may appear during a start-up. In this paper, geysering is dealt with especially and the driving mechanism is clarified by investigating the stability related to effects of inlet velocity, subcooling, temperature in an outlet plenum and non-heated length between heated section and the outlet plenum. Furthermore, by considering these results and the operational experience in the Dodewaard reactor, recommendations on how the thermo-hydraulic instabilities can be prevented from occurring are proposed concerning a reactor configuration and start-up procedure for natural circulation BWRs. (author)

  4. Experimental investigation of pool boiling heat transfer and critical heat flux on a downward facing surface

    International Nuclear Information System (INIS)

    Gocmanac, M.; Luxat, J.C.

    2012-01-01

    A separate effects experimental study of heat transfer and Critical Heat Flux (CHF) on a downward facing plate in subcooled water pool boiling is described. Two geometries of downwards facing surfaces are studied. The first is termed the 'confined' study in which bubble motion is restricted to the heated surface. The second is termed the 'unconfined' study where individual bubbles are free to move along the heated surface and vent in any direction. The method used in the confined study is novel and involves the placement of a lip surrounding the heated surface. The CHF as a function of angle of inclination of the surface is presented and is in good agreement with other experimental data from somewhat different test geometries. (author)

  5. Measurements of Burnout Conditions for Flow of Boiling Water in Vertical Round Ducts (Part 2)

    International Nuclear Information System (INIS)

    Becker, Kurt M.; Persson, P.; Nilsson, L.; Eriksson, O.

    1963-06-01

    The present report deals with the results of the second phase of an experimental investigation of burnout conditions for flow of boiling water in vertical round ducts. The following ranges of variables were studied and 809 burnout measurements were obtained. Pressure 5. 3 2 ; Inlet subcooling 56 sub BO 2 ; Mass velocity 100 2 s; Heated length 600 BO , were plotted against the pressure with the surface heat flux as parameter. The data have been correlated by curves, and the scatter around the curves is less than ± 5 per cent. In the ranges investigated, the observed steam quality at burnout, X BO generally decreases with increasing heat flux and mass velocity but increases with increasing pressure. The data have been compared with the empirical correlation by Tong, and excellent agreement was found for pressures higher than 10 kg/cm 2

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

    International Nuclear Information System (INIS)

    Becker, Kurt M.

    1962-01-01

    The present report deals with the results of the first phase of an experimental investigation of burnout conditions for flow of boiling water in vertical round ducts. Data were obtained in the following ranges of variables. Pressure 2.4 sub 2 ; Mass velocity 144 2 /s; Heated length 1040 BO , were plotted against the pressure with the surface heat flux as parameter. The data have been correlated by curves. The scatter of the data around the curves is less than ± 5 per cent. In the ranges investigated the observed steam quality at burnout, x BO generally decreases with increasing heat flux; increases with increasing pressure and decreases with increasing mass velocity. The mass velocity effect has been explained on the basis of climbing film flow theory. Finally we have found that for engineering purposes the effects of inlet subcooling and channel length are negligible

  7. Fuel-coolant interaction in a shock tube with initially-established film boiling

    International Nuclear Information System (INIS)

    Sharon, A.; Bankoff, S.G.

    1979-01-01

    A new mode of thermal interaction has been employed, in which liquid metal is melted in a crucible within a shock tube; the coolant level is raised to overflow the crucible and establish subcooled film boiling with known bulk metal temperature; and a pressure shock is then initiated. With water and lead-tin alloy an initial splash of metal may be obtained after the vapor film has collapsed, due primarily to thermal interaction, followed by a successive cycle of bubble growth and collapse. To obtain large interactions, the interfacial contact temperature must exceed the spontaneous nucleation temperature of the coolant. Other cutoff behavior is observed with respect to the initial system pressure and temperatures and with the shock pressure and rise time. Experiments with butanol and lead-tin alloy show only relatively mild interactions. Qualitative explanations are proposed for the different behaviors of the two liquids

  8. A comprehensive review on pool boiling of nanofluids

    International Nuclear Information System (INIS)

    Ciloglu, Dogan; Bolukbasi, Abdurrahim

    2015-01-01

    Nanofluids are nanoparticle suspensions of small particle size and low concentration dispersed in base fluids such as water, oil and ethylene glycol. These fluids have been considered by researchers as a unique heat transfer carrier because of their thermophysical properties and a great number of potential benefits in traditional thermal engineering applications, including power generation, transportation, air conditioning, electronics devices and cooling systems. Many attempts have been made in the literature on nanofluid boiling; however, data on the boiling heat transfer coefficient (HTC) and the critical heat flux (CHF) have been inconsistent. This paper presents a review of recent researches on the pool boiling heat transfer behaviour of nanofluid. First, the development of nanofluids and their potential applications are briefly given. Then, the effects of various parameters on nanofluids pool boiling are discussed in detail. - Highlights: • A review on the pool boiling heat transfer of nanofluid is presented and discussed. • Nanoparticle deposition considerably affects the boiling heat transfer. • The HTC decreases due to the low contact angle and the high adhesion energy. • The HTC increases due to the formation of the new cavities and liquid suction. • The CHF increases due to the increase in roughness, wettability and capillarity

  9. Outline of advanced boiling water reactor

    International Nuclear Information System (INIS)

    Yoshio Matsuo

    1987-01-01

    The ABWR (Advanced Boiling Water Reactor) is based on construction and operational experience in Japan, USA and Europe. It was developed jointly by the BWR supplieres, General Electric, Hitachi, and Toshiba, as the next generation BWR for Japan. The Tokyo Electric Power Co. provided leadership and guidance in developing the ABWR, and in combination with five other Japanese electric power companies. The major objectives in developing the ABWR are: 1. Enhanced plant operability, maneuverability and daily load-following capability; 2. Increased plant safety and operating margins; 3. Improved plant availability and capacity factor; 4. Reduced occupational radiation exposure; 5. Reduced radwaste volume, and 6. Reduced plant capital and operating costs. (Liu)

  10. Post-CHF low-void heat transfer of water: measurements in the complete transition boiling region at atmospheric pressure

    International Nuclear Information System (INIS)

    Johannsen, K.; Meinen, W.

    1984-01-01

    An experimental investigation of low-void heat transfer of water has been performed in the range of CHF and the minimum stable film boiling temperature. The heat transfer system used consists of a vertically mounted copper tube of 1 cm I.D. and 5 cm length with surface-temperature controlled, indirect Joule heating. Results are presented for upflowing water at inverted annular flow conditions in the inlet subcooling range of 2.5 - 40 0 C and mass flux range of 137-600 kg/m 2 s in terms of boiling curves and heat transfer coefficients versus wall temperature. Heat transfer in the stationary rewetting front, which occurs within the test section during operation in the transition boiling mode, is also dealt with. At high mass flux, occurrence of an inverse rewetting front has been observed. It is also noted that, at fixed location, minimum heat flux observed is usually not associated with the minimum stable film boiling temperature

  11. 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

  12. 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

  13. 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.

  14. Analysis on energy saving potential of integrated supermarket HVAC and refrigeration systems using multiple subcoolers

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Liang [Institute of Refrigeration and Cryogenics, Shanghai Jiaotong University, Shanghai 200240 (China); China R and D Center, Carrier Corporation, No. 3239 Shen Jiang Road, Shanghai 201206 (China); Zhang, Chun-Lu [China R and D Center, Carrier Corporation, No. 3239 Shen Jiang Road, Shanghai 201206 (China)

    2010-02-15

    The paper presents a model-based analysis on the energy saving potential of supermarket HVAC (heating, ventilating, and air-conditioning) and refrigeration systems using multiple subcoolers among the high-temperature HVAC system, the medium-temperature refrigeration system, and the low-temperature refrigeration system. The principle of energy reduction is to have the higher COP (coefficient of performance) system generate more cooling capacity to increase the cooling capacity or reduce the power consumption of the lower COP system. The subcooler could be placed between the medium-temperature and low-temperature systems, between the high-temperature and medium-temperature systems, and between the high-temperature and low-temperature systems. All integration scenarios of adding one, two and three subcoolers have been investigated. The energy saving potential varies with the load ratio between high-, medium- and low-temperature systems, COP of three systems, and the ''on-off'' duty time of HVAC system. The optimal sequence of adding subcoolers is also proposed. (author)

  15. Measurement of local flow pattern in boiling R12 simulating PWR conditions with multiple optical probes

    International Nuclear Information System (INIS)

    Garnier, J.

    1998-01-01

    For a comprehensive approach of boiling crisis phenomenon in order to get more reliable predictions of critical heat flux in PWR core, a flow pattern study is under progress at CEA GRENOBLE (in a joint program with Electricite de France: EdF). The first aim is to get experimental results on flow structure in the range of thermal hydraulic parameters involved in the core of a PWR (pressure up to 16 MPa, heat flux about 1 MW/m 2 , mass velocity up to 5000 kg/s/m 2 . As critical heat flux is a local phenomenon and is the result of the flow development, the data has to be measured from the beginning of boiling until boiling crisis, and from the bulk flow until the boundary layer close to the heating walls. Therefore, these results will be useful in modeling not only boiling crisis phenomenon but also condensation in subcooled boiling, coalescence, splitting up, mass and energy transfers at interfaces, and so on. In a first step, the test section is a vertical tube 19.2 mm internal diameter with an axial uniform heat flux over a 3.5m length. The study is performed on the DEBORA loop with Freon 12 as coolant fluid. We assume that basic boiling phenomena (and the knowledge we get about them) only depend on the fluid properties by means of dimensionless parameters but not on the fluid itself. In a first part, we briefly recall that interfacial detection is the most important parameter of a flow pattern study. Therefore, the use of probes able to measure the Phase Indicator Function (P.I.F.) is necessary. A first study of flow conditions shows that the flow pattern is essentially a bubbly one with vapor particles of low diameter (about 300 clm) and high velocity (up to 7 m/s). These criteria induce that a multiple optical probe is the most appropriate tool provided we improve the technology. We detail the way to obtain probes able to detect small particles at high velocity. Each fiber is stretched to get a tip of 10 Clm with the cladding kept on 50 μm length which defines

  16. 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

  17. Surface boiling of superheated liquid

    Energy Technology Data Exchange (ETDEWEB)

    Reinke, P. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1997-01-01

    A basic vaporization mechanism that possibly affects the qualitative and quantitative prediction of the consequences of accidental releases of hazardous superheated liquids was experimentally and analytically investigated. The studies are of relevance for the instantaneous failure of a containment vessel filled with liquefied gas. Even though catastrophical vessel failure is a rare event, it is considered to be a major technological hazard. Modeling the initial phase of depressurisation and vaporization of the contents is an essential step for the subsequent analysis of the spread and dispersion of the materials liberated. There is only limited understanding of this inertial expansion stage of the superheated liquid, before gravity and atmospheric turbulence begin to dominate the expansion. This work aims at a better understanding of the vaporization process and to supply more precise source-term data. It is also intended to provide knowledge for the prediction of the behavior of large-scale releases by the investigation of boiling on a small scale. Release experiments with butane, propane, R-134a and water were conducted. The vaporization of liquids that became superheated by sudden depressurisation was studied in nucleation-site-free glass receptacles. Several novel techniques for preventing undesired nucleation and for opening the test-section were developed. Releases from pipes and from a cylindrical geometry allowed both linear one-dimensional, and radial-front two-dimensional propagation to be investigated. Releases were made to atmospheric pressure over a range of superheats. It was found that, above a certain superheat temperature, the free surface of the metastable liquid rapidly broke up and ejected a high-velocity vapor/liquid stream. The zone of intense vaporization and liquid fragmentation proceeded as a front that advanced into the test fluids. No nucleation of bubbles in the bulk of the superheated liquid was observed. (author) figs., tabs., refs.

  18. Surface boiling of superheated liquid

    International Nuclear Information System (INIS)

    Reinke, P.

    1997-01-01

    A basic vaporization mechanism that possibly affects the qualitative and quantitative prediction of the consequences of accidental releases of hazardous superheated liquids was experimentally and analytically investigated. The studies are of relevance for the instantaneous failure of a containment vessel filled with liquefied gas. Even though catastrophical vessel failure is a rare event, it is considered to be a major technological hazard. Modeling the initial phase of depressurisation and vaporization of the contents is an essential step for the subsequent analysis of the spread and dispersion of the materials liberated. There is only limited understanding of this inertial expansion stage of the superheated liquid, before gravity and atmospheric turbulence begin to dominate the expansion. This work aims at a better understanding of the vaporization process and to supply more precise source-term data. It is also intended to provide knowledge for the prediction of the behavior of large-scale releases by the investigation of boiling on a small scale. Release experiments with butane, propane, R-134a and water were conducted. The vaporization of liquids that became superheated by sudden depressurisation was studied in nucleation-site-free glass receptacles. Several novel techniques for preventing undesired nucleation and for opening the test-section were developed. Releases from pipes and from a cylindrical geometry allowed both linear one-dimensional, and radial-front two-dimensional propagation to be investigated. Releases were made to atmospheric pressure over a range of superheats. It was found that, above a certain superheat temperature, the free surface of the metastable liquid rapidly broke up and ejected a high-velocity vapor/liquid stream. The zone of intense vaporization and liquid fragmentation proceeded as a front that advanced into the test fluids. No nucleation of bubbles in the bulk of the superheated liquid was observed. (author) figs., tabs., refs

  19. 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

  20. CFD investigation of nucleate boiling in non-circular geometries at high pressure

    Energy Technology Data Exchange (ETDEWEB)

    Thakrar, R., E-mail: rkt08@imperial.ac.uk; Murallidharan, J.; Walker, S.P.

    2017-02-15

    Highlights: • Blind CFD benchmark of high-pressure boiling test case in rectangular geometry. • Influence of turbulence, wall boiling, interfacial area transport and lift force modelling examined. • Good agreement of the area-averaged void with the most mechanistic approaches. • Transition from wall to core void peaking due to regime transition captured only in part. - Abstract: Boiling flows are commonplace in the nuclear industry. Computational Fluid Dynamics (CFD) is slowly beginning to be used to deliver the relevant two-phase thermal hydraulic analyses required for nuclear applications. This paper presents a blind assessment of the capabilities of the commercial CFD code STAR-CCM+ against measurements for a vertically upward mildly subcooled boiling flow approaching saturation in a rectangular channel at a pressure of 41 bar. The available measurements comprised transverse distributions and cross-sectional area averages of void fraction at numerous axial positions along the channel. The predictive ability of several combinations of turbulence, wall heat flux partitioning, interfacial area transport and lift force models was tested. In general, good agreement was obtained for the area-averaged void, with the most mechanistic modelling combination reproducing the measurements accurately. Reasonable agreement was also observed for the distributions of transverse void, however this agreement could not be maintained beyond the channel entrance. The transition from near-wall to core void peaking exhibited in the experiments, attributable presumably to a bubbly to churn-turbulent flow regime transition, could not be reproduced accurately with any of the modelling combinations used, and the basic qualitative trend was captured only in part. Suggestions for future investigation are outlined subsequently.

  1. 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)

  2. MTD-MFC: unified framework for investigation of diversity of boiling heat transfer curves

    Energy Technology Data Exchange (ETDEWEB)

    Shekriladze, I.G. [Georgian Technical University, Tbilisi (Georgia)], e-mail: shekri@geo.net.ge

    2009-07-01

    A keynote paper presents just the next attempt to promote a discussion of modern state of art in the field of boiling heat transfer research. It is shown how longstanding disregard of internal contradictions of applicable approaches has resulted theoretical deadlock. Alternatively, it also is shown how resolution of these contradictions opens the ways to breakthrough in boiling heat transfer theory. Basic experimental facts, physical models and correlations are reconsidered. Principal contradictions between experimental knowledge and traditional model of 'the theatre of actors' (MTA) are discussed. Crucial role of pumping effect of growing bubble (PEGB) in boiling heat transfer and hydrodynamics is shown. Basic role of control of HTC by thermodynamic conditions on nucleation sites is demonstrated and consequent model of 'the theatre of director' (MTD) is discussed. Universal MTD-based correlation of boiling HTC of all types of liquids is considered. Unified consistent research framework for developed boiling heat transfer and diverse specific boiling heat transfer regimes is outlined through supplementing MTD by so-called multifactoring concept (MFC). The latter links transition from developed boiling mode to diverse boiling curves to a phenomenon of multiplication of factors influencing HTC. The ways of further research of the boiling problem are discussed. (author)

  3. MTD-MFC: unified framework for investigation of diversity of boiling heat transfer curves

    International Nuclear Information System (INIS)

    Shekriladze, I.G.

    2009-01-01

    A keynote paper presents just the next attempt to promote a discussion of modern state of art in the field of boiling heat transfer research. It is shown how longstanding disregard of internal contradictions of applicable approaches has resulted theoretical deadlock. Alternatively, it also is shown how resolution of these contradictions opens the ways to breakthrough in boiling heat transfer theory. Basic experimental facts, physical models and correlations are reconsidered. Principal contradictions between experimental knowledge and traditional model of 'the theatre of actors' (MTA) are discussed. Crucial role of pumping effect of growing bubble (PEGB) in boiling heat transfer and hydrodynamics is shown. Basic role of control of HTC by thermodynamic conditions on nucleation sites is demonstrated and consequent model of 'the theatre of director' (MTD) is discussed. Universal MTD-based correlation of boiling HTC of all types of liquids is considered. Unified consistent research framework for developed boiling heat transfer and diverse specific boiling heat transfer regimes is outlined through supplementing MTD by so-called multifactoring concept (MFC). The latter links transition from developed boiling mode to diverse boiling curves to a phenomenon of multiplication of factors influencing HTC. The ways of further research of the boiling problem are discussed. (author)

  4. The myth of the boiling point.

    Science.gov (United States)

    Chang, Hasok

    2008-01-01

    Around 1800, many reputable scientists reported significant variations in the temperature of pure water boiling under normal atmospheric pressure. The reported variations included a difference of over 1 degree C between boiling in metallic and glass vessels (Gay-Lussac), and "superheating" up to 112 degrees C on extracting dissolved air out of water (De Luc). I have confirmed most of these observations in my own experiments, many of which are described in this paper. Water boils at the "boiling point" only under very particular circumstances. Our common-sense intuition about the fixedness of the boiling point is only sustained by our limited experience.

  5. Gravity and Heater Size Effects on Pool Boiling Heat Transfer

    Science.gov (United States)

    Kim, Jungho; Raj, Rishi

    2014-01-01

    The current work is based on observations of boiling heat transfer over a continuous range of gravity levels between 0g to 1.8g and varying heater sizes with a fluorinert as the test liquid (FC-72/n-perfluorohexane). Variable gravity pool boiling heat transfer measurements over a wide range of gravity levels were made during parabolic flight campaigns as well as onboard the International Space Station. For large heaters and-or higher gravity conditions, buoyancy dominated boiling and heat transfer results were heater size independent. The power law coefficient for gravity in the heat transfer equation was found to be a function of wall temperature under these conditions. Under low gravity conditions and-or for smaller heaters, surface tension forces dominated and heat transfer results were heater size dependent. A pool boiling regime map differentiating buoyancy and surface tension dominated regimes was developed along with a unified framework that allowed for scaling of pool boiling over a wide range of gravity levels and heater sizes. The scaling laws developed in this study are expected to allow performance quantification of phase change based technologies under variable gravity environments eventually leading to their implementation in space based applications.

  6. Improvement of boiling heat transfer by radiation induced boiling enhancement

    International Nuclear Information System (INIS)

    Imai, Yasuyuki; Okamoto, Koji; Madarame, Haruki; Takamasa, Tomoji

    2003-01-01

    For nuclear reactor systems, the critical heat flux (CHF) data is very important because it limits reactor efficiency. Improvement of CHF requires that the cooling liquid can contact the heating surface, or a high-wettability, highly hydrophilic heating surface, even if a vapor bubble layer is generated on the surface. In our previous study, we confirmed that the surface wettability changed significantly or that highly hydrophilic conditions were achieved, after irradiation of 60 Co gamma ray, by the Radiation Induced Surface Activation (RISA) phenomenon. To delineate the effect of RISA on boiling phenomena, surface wettability in a high-temperature environment and critical heat flux (CHF) of metal oxides irradiated by gamma rays were investigated. A CHF experiment in the pool boiling condition was carried out under atmospheric pressure. The heating test section made of titanium was 0.2 mm in thickness, 3 mm in height, and 60 mm in length. Oxidation of the surface was carried out by plasma jetting for 40 seconds. The test section was irradiated by 60 Co gamma ray with predetermined radiation intensity and period. The CHF of oxidized titanium was improved up to 100 percent after 800 kGy 60 Co gamma ray irradiation. We call this effect Radiation Induced Boiling Enhancement (RIBE). Before we conducted the CHF experiment, contact angles of the test pieces were measured to show the relationship between wettability and CHF. The CHF in the present experiment increases will surface wettability in the same manner as shown by Liaw and Dhir's results. (author)

  7. Experimental Study on Boiling Crisis in Pool Boiling

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Satbyoul; Kim, Hyungdae [Kyung Hee University, Yongin (Korea, Republic of)

    2016-10-15

    They postulated that failure in re-wetting of a dry patch by a cooling liquid is governed by microhydrodynamics near the wall. Chu et al. commonly observed that active coalescence of newly generated bubbles with preexisting bubbles results in a residual dry patch and prevents the complete rewetting of the dry patch, leading to CHF. In this work, to reveal the key physical mechanism of CHF during the rewetting process of a dry patch, dynamics of dry patches and thermal pattern of a boiling surface are simultaneously observed using TR and IR thermometry techniques. Local dynamics of dry patch and thermal pattern on a boiling surface in synchronized manner for both space and time using TR and IR thermometry were measured during pool boiling of water. Observation and quantitative examination of CHF was performed. - The hydrodynamic and thermal behaviors of irreversible dry patch were observed. The dry patches coalesce into a large dry patch and it locally dried out. Due to the failure of liquid rewetting, the dry patch is not completely rewetted, resulting in the burn out at which temperature is -140°C. - When temperature of a dry patch rises beyond the instantaneous nucleation temperature, several bubbles nucleate at the head of the advancing liquid meniscus and prevents the liquid front, and eventually the overheated dry patch remains alive after the departure of the massive bubble.

  8. Improvement of boiling heat transfer by radiation induced boiling enhancement

    International Nuclear Information System (INIS)

    Imai, Y.; Okamoto, K.; Madarame, H.; Takamasa, T.

    2003-01-01

    For nuclear reactor systems, the Critical Heat Flux (CHF) data is very important because it limits reactor efficiency. Improvement of CHF requires that the cooling liquid can contact the heating surface, or a high-wettability, highly hydrophilic heating surface, even if a vapor bubble layer is generated on the surface. In our previous study, we confirmed that the surface wettability changed significantly or that highly hydrophilic conditions were achieved, after irradiation of 60Co gamma ray, by the Radiation Induced Surface Activation (RISA) phenomenon. To delineate the effect of RISA on boiling phenomena, surface wettability in a high-temperature environment and Critical Heat Flux (CHF) of metal oxides irradiated by gamma rays were investigated. A CHF experiment in the pool boiling condition was carried out under atmospheric pressure. The heating test section made of titanium was 0.2mm in thickness, 3mm in height, and 60mm in length. Oxidation of the surfaces was carried out by plasma jetting for 40 seconds. The test section was irradiated by 60Co gamma ray with predetermined radiation intensity and period. The CHF of oxidized titanium was improved up to 100 percent after 800kGy 60Co gamma ray irradiation. We call this effect Radiation Induced Boiling Enhancement (RIBE). Before we conducted the CHF experiment, contact angles of the test pieces were measured to show the relationship between wettability and CHF. The CHF in the present experiment increases with surface wettability in the same manner as shown by Liaw and Dhir's results

  9. GreenChill Store Certification Protocol for Sub-Cooling Contained on Racks Separate from Refrigeration Equipment

    Science.gov (United States)

    Document describes the protocol used to determine the total load and refrigerant charge of stores that have placed all sub-cooling on a rack separate from all other commercial refrigeration equipment.

  10. Boiling process modelling peculiarities analysis of the vacuum boiler

    Science.gov (United States)

    Slobodina, E. N.; Mikhailov, A. G.

    2017-06-01

    The analysis of the low and medium powered boiler equipment development was carried out, boiler units possible development directions with the purpose of energy efficiency improvement were identified. Engineering studies for the vacuum boilers applying are represented. Vacuum boiler heat-exchange processes where boiling water is the working body are considered. Heat-exchange intensification method under boiling at the maximum heat- transfer coefficient is examined. As a result of the conducted calculation studies, heat-transfer coefficients variation curves depending on the pressure, calculated through the analytical and numerical methodologies were obtained. The conclusion about the possibility of numerical computing method application through RPI ANSYS CFX for the boiling process description in boiler vacuum volume was given.

  11. Drying of upper peat of a low degree of separation in a vibro-boiling layer

    Energy Technology Data Exchange (ETDEWEB)

    Naumovich, V M; Goryachev, V I; Shafeyev, R S

    1981-01-01

    The boiling layer created by a gas flow in its motion upward through friable material finds wide application for the drying of different dispersed materials. In spite of the significant superiority of a boiling layer to other methods of drying, it is not universal and possesses a series of important drawbacks. For an intensification of the boiling layer and an upgrading of its structure, different mechanical systems are employed: agitators, impulses, ultrasound, and other actions on the material. In particular with the development of the vibration technique, vibrated low-frequency oscillations are not only a means of structurally upgrading the boiling layer, but serve as the basic factor in creating an intensively blended layer of friable material. According to the character of the motion of the particles, such a layer resembles a boiling liquid. Therefore, in the given case, it is called vibro-boiling. An appartus working on the principle of a vibro-boiling layer, in many cases, is eliminated of faults, a characteristic worthy of considering by installations with a boiling layer. A considered created department of machines and processes of refining peat, the KPI, is a vibration-drying installatin of the VSU type, designed for the drying of peat with a low degree of separation in a vibro-boiling layer.

  12. A formal approach for the prediction of the critical heat flux in subcooled water

    Energy Technology Data Exchange (ETDEWEB)

    Lombardi, C. [Polytechnic of Milan (Italy)

    1995-09-01

    The critical heat flux (CHF) in subcooled water at high mass fluxes are not yet satisfactory correlated. For this scope a formal approach is here followed, which is based on an extension of the parameters and the correlation used for the dryout prediction for medium high quality mixtures. The obtained correlation, in spite of its simplicity and its explicit form, yields satisfactory predictions, also when applied to more conventional CHF data at low-medium mass fluxes and high pressures. Further improvements are possible, if a more complete data bank will be available. The main and general open item is the definition of a criterion, depending only on independent parameters, such as mass flux, pressure, inlet subcooling and geometry, to predict whether the heat transfer crisis will result as a DNB or a dryout phenomenon.

  13. Determination of heat transfer coefficient for an interaction of sub-cooled gas and metal

    International Nuclear Information System (INIS)

    Sidek, Mohd Zaidi; Kamarudin, Muhammad Syahidan

    2016-01-01

    Heat transfer coefficient (HTC) for a hot metal surface and their surrounding is one of the need be defined parameter in hot forming process. This study has been conducted to determine the HTC for an interaction between sub-cooled gas sprayed on a hot metal surface. Both experiments and finite element have been adopted in this work. Initially, the designated experiment was conducted to obtain temperature history of spray cooling process. Then, an inverse method was adopted to calculate the HTC value before we validate in a finite element simulation model. The result shows that the heat transfer coefficient for interaction of subcooled gas and hot metal surface is 1000 W/m 2 K. (paper)

  14. A formal approach for the prediction of the critical heat flux in subcooled water

    International Nuclear Information System (INIS)

    Lombardi, C.

    1995-01-01

    The critical heat flux (CHF) in subcooled water at high mass fluxes are not yet satisfactory correlated. For this scope a formal approach is here followed, which is based on an extension of the parameters and the correlation used for the dryout prediction for medium high quality mixtures. The obtained correlation, in spite of its simplicity and its explicit form, yields satisfactory predictions, also when applied to more conventional CHF data at low-medium mass fluxes and high pressures. Further improvements are possible, if a more complete data bank will be available. The main and general open item is the definition of a criterion, depending only on independent parameters, such as mass flux, pressure, inlet subcooling and geometry, to predict whether the heat transfer crisis will result as a DNB or a dryout phenomenon

  15. Sensitivity Analysis of RCW Temperature on the Moderator Subcooling Margin for the LBLOCA of Wolsong NPP Unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Si Won; Kim, Jong Hyun; Choi, Sung Soo [Atomic Creative Technology Co., Daejeon (Korea, Republic of); Kim, Sung Min [Central Research Institute, Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

    2016-05-15

    Moderator subcooling margin has been analyzed using the MODTURC{sub C}LAS code in the Large LOCA FSAR PARTs C and F. Performance of moderator heat exchangers depends on RCW (Raw reCirculated Water) temperature. And also the temperature is affected by sea water temperature. Unfortunately, sea water temperature is gradually increasing by global warming. So it will cause increase of RCW temperature inevitably. There is no assessment result of moderator subcooling with increasing RCW temperature even if it is important problem. Therefore, sensitivity analysis is performed to give information about the relation between RCW temperature and moderator subcooling in the present study. The moderator subcooling margin has to be ensured to establish the moderator heat removal when Large LOCA with LOECI and Loss of Class IV Power occurs. However, sea water temperature is increasing gradually due to global warming. So it is necessary that sensitivity analysis of RCW temperature on the moderator subcooling margin to estimate the availability of the moderator heat removal. In the present paper, the moderator subcooling analysis is performed using the same methodology and assumptions except for RCW temperature used in FSAR Large LOCA PART F.

  16. Sensitivity Analysis of RCW Temperature on the Moderator Subcooling Margin for the LBLOCA of Wolsong NPP Unit 1

    International Nuclear Information System (INIS)

    Seo, Si Won; Kim, Jong Hyun; Choi, Sung Soo; Kim, Sung Min

    2016-01-01

    Moderator subcooling margin has been analyzed using the MODTURC_CLAS code in the Large LOCA FSAR PARTs C and F. Performance of moderator heat exchangers depends on RCW (Raw reCirculated Water) temperature. And also the temperature is affected by sea water temperature. Unfortunately, sea water temperature is gradually increasing by global warming. So it will cause increase of RCW temperature inevitably. There is no assessment result of moderator subcooling with increasing RCW temperature even if it is important problem. Therefore, sensitivity analysis is performed to give information about the relation between RCW temperature and moderator subcooling in the present study. The moderator subcooling margin has to be ensured to establish the moderator heat removal when Large LOCA with LOECI and Loss of Class IV Power occurs. However, sea water temperature is increasing gradually due to global warming. So it is necessary that sensitivity analysis of RCW temperature on the moderator subcooling margin to estimate the availability of the moderator heat removal. In the present paper, the moderator subcooling analysis is performed using the same methodology and assumptions except for RCW temperature used in FSAR Large LOCA PART F.

  17. Boiling characteristics of dilute polymer solutions and implications for the suppression of vapor explosions

    Energy Technology Data Exchange (ETDEWEB)

    Bang, K.H.; Kim, M.H. [Univ. of Science and Technology, Pohang (Korea, Republic of)

    1995-09-01

    Quenching experiments of hot solid spheres in dilute aqueous solutions of polyethylene oxide polymer have been conducted for the purpose of investigating the physical mechanisms of the suppression of vapor explosions in this polymer solutions. Two spheres of 22.2mm and 9.5mm-diameter were tested in the polymer solutions of various concentrations at 30{degrees}C. Minimum film boiling temperature ({Delta}T{sub MFB}) in this highly-subcooled liquid rapidly decreased from over 700{degrees}c for pure water to about 150{degrees}C as the polymer concentration was increased up to 300ppm for 22.2mm sphere, and it decreased to 350{degrees}C for 9.5mm sphere. This rapid reduction of minimum film boiling temperature in the PEO aqueous solutions can explain its ability of the suppression of spontaneous vapor explosions. The ability of suppression of vapor explosions by dilute polyethylene oxide solutions against an external trigger pressure was tested by dropping molten tin into the polymer solutions at 25{degrees}C. It was observed that in 50ppm solutions more mass fragmented than in pure water, but produced weaker explosion pressures. The explosion was completely suppressed in 300ppm solutions with the external trigger. The debris size distributions of fine fragments smaller than 0.7mm were shown almost identical regardless of the polymer concentrations.

  18. New Zealand's boiling kettles

    Energy Technology Data Exchange (ETDEWEB)

    Minkowski, H

    1955-04-01

    The Wairakei Valley is covered with mud pots, geysers, and steam vents. Aggressive exploitation of the geothermal resources in this area is underway. The geothermal development experience gained in Italy, where the soil is cold and compact, cannot be applied in New Zealand, as the soil is hot and the rock is porous. Drilling in New Zealand must be accompanied by a constant input of cooling emulsion. Special techniques have been developed for cementing the high-temperature boreholes. Test drillings have been made as deep as 1000 m and pressures on the order of 30 kg/cm/sup 2/ have been encountered. As even higher pressures are expected, valves have been designed to support up to 70 kg/cm/sup 2/. A 70 MW pilot-plant, supplied by 27 wells, was expected to be on-line in 1955.

  19. 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.

  20. On the influence of water subcooling and melt jet parameters on debris formation

    Energy Technology Data Exchange (ETDEWEB)

    Manickam, Louis, E-mail: louis@safety.sci.kth.se; Kudinov, Pavel; Ma, Weimin; Bechta, Sevostian; Grishchenko, Dmitry

    2016-12-01

    Highlights: • Melt and water configuration effects on debris formation is studied experimentally. • Melt superheat and water subcooling are most influential compared to jet size. • Melt-water configuration and material properties influence particle fracture rate. • Results are compared with large scale experiments to study effect of spatial scales. - Abstract: Breakup of melt jet and formation of a porous debris bed at the base-mat of a flooded reactor cavity is expected during the late stages of a severe accident in light water reactors. Debris bed coolability is determined by the bed properties including particle size, morphology, bed height and shape as well as decay heat. Therefore understanding of the debris formation phenomena is important for assessment of debris bed coolability. A series of experiments was conducted in MISTEE-Jet facility by discharging binary-oxide mixtures of WO{sub 3}–Bi{sub 2}O{sub 3} and WO{sub 3}–ZrO{sub 2} into water in order to investigate properties of resulting debris. The effect of water subcooling, nozzle diameter and melt superheat was addressed in the tests. Experimental results reveal significant influence of water subcooling and melt superheat on debris size and morphology. Significant differences in size and morphology of the debris at different melt release conditions is attributed to the competition between hydrodynamic fragmentation of liquid melt and thermal fracture of the solidifying melt droplets. The particle fracture rate increases with increased subcooling. Further the results are compared with the data from larger scale experiments to discern the effects of spatial scales. The present work provides data that can be useful for validation of the codes used for the prediction of debris formation phenomena.

  1. Thermodynamic performance evaluation of transcritical carbon dioxide refrigeration cycle integrated with thermoelectric subcooler and expander

    International Nuclear Information System (INIS)

    Dai, Baomin; Liu, Shengchun; Zhu, Kai; Sun, Zhili; Ma, Yitai

    2017-01-01

    New configurations of transcritical CO_2 refrigeration cycle combined with a thermoelectric (TE) subcooler and an expander (TES+EXP_H_M and TES+EXP_M_L) are proposed. The expander can operate between the high-pressure to the vessel pressure, or from vessel pressure to evaporation pressure. A power system is utilized to balance and supply power to thermoelectric subcooler and compressor. Thermodynamic performance optimizations and analyses are presented. Comparisons are carried out with the BASE, EXP_H_M, EXP_M_L, and TES cycles. The results show that the coefficient of performance (COP) improvement is more notable when the expander is installed between the liquid receiver and the evaporator. Maximum COP is obtained for the new cycles with a simultaneous optimization of discharge pressure and subcooling temperature. The new proposed TES+EXP_M_L cycle shows an excellent and steady performance than other cycles. It operates not only with the highest COP, but also the lowest discharge pressure. Under the working conditions of high gas cooler outlet temperature or low evaporation temperature, the merits of COP improvement and discharge pressure reduction are more prominent. The new cycle is more suitable for the hot regions where the CO_2 can not be sufficiently subcooled or the refrigerated space operates at low evaporation temperature. - Highlights: • New configurations of transcritical CO_2 refrigeration cycle are proposed. • New cycles are optimized and compared with other cycles. • The position of expander has an evident influence on the performance of CO_2 cycle. • TES+EXP_M_L cycle shows the highest COP and lowest discharge pressure. • The range of application for the TES+EXP_M_L cycle is recommended.

  2. 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

  3. Thermodynamic study of the injection of steam bubbles in a subcooled liquid

    International Nuclear Information System (INIS)

    Besset, Jacqueline.

    1980-10-01

    The behaviour of steam bubbles injected in a subcooled liquid has been studied. Water was the fluid chosen for this experiment for the steam and the liquid. The experiment was carried out at atmospheric pressure and the variable parameters were the ΔT subcooling (difference between the saturation temperature at the pressure under consideration and that of the liquid around the bubbles) and the steam output. We first studied the formation of the bubbles in a wide subcooling range (7 0 C 0 C). In this study a straight correlation was obtained giving the volume of the bubbles formed at the injector outlet, which is valid for a wide range of variables. The implosion of free bubbles in the liquid after they separated from the injector was then studied. In these experiments the significant implosion parameters are the Jakob Ja number, that measures the possibility of the liquid to absorb the heat given off by the condensation, and the Peclet Pe(RO) number, that defines the relative participation of conduction and convection in the heat exchanges. These numbers vary in the areas: 35 [fr

  4. Models and Stability Analysis of Boiling Water Reactors

    International Nuclear Information System (INIS)

    Dorning, John

    2002-01-01

    We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model that includes: space-time modal neutron kinetics based on spatial w-modes; single- and two-phase flow in parallel boiling channels; fuel rod heat conduction dynamics; and a simple model of the recirculation loop. The BR model is represented by a set of time-dependent nonlinear ordinary differential equations, and is studied as a dynamical system using the modern bifurcation theory and nonlinear dynamical systems analysis. We first determine the stability boundary (SB) - or Hopf bifurcation set- in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value; then we transform the SB to the practical power-flow map used by BWR operating engineers and regulatory agencies. Using this SB, we show that the normal operating point at 100% power is very stable, that stability of points on the 100% rod line decreases as the flow rate is reduced, and that operating points in the low-flow/high-power region are least stable. We also determine the SB that results when the modal kinetics is replaced by simple point reactor kinetics, and we thereby show that the first harmonic mode does not have a significant effect on the SB. However, we later show that it nevertheless has a significant effect on stability because it affects the basin of attraction of stable operating points. Using numerical simulations we show that, in the important low-flow/high-power region, the Hopf bifurcation that occurs as the SB is crossed is subcritical; hence, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line at points in the low-flow/high-power region. Numerical simulations are also performed to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is

  5. Models and Stability Analysis of Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    John Dorning

    2002-04-15

    We have studied the nuclear-coupled thermal-hydraulic stability of boiling water reactors (BWRs) using a model that includes: space-time modal neutron kinetics based on spatial w-modes; single- and two-phase flow in parallel boiling channels; fuel rod heat conduction dynamics; and a simple model of the recirculation loop. The BR model is represented by a set of time-dependent nonlinear ordinary differential equations, and is studied as a dynamical system using the modern bifurcation theory and nonlinear dynamical systems analysis. We first determine the stability boundary (SB) - or Hopf bifurcation set- in the most relevant parameter plane, the inlet-subcooling-number/external-pressure-drop plane, for a fixed control rod induced external reactivity equal to the 100% rod line value; then we transform the SB to the practical power-flow map used by BWR operating engineers and regulatory agencies. Using this SB, we show that the normal operating point at 100% power is very stable, that stability of points on the 100% rod line decreases as the flow rate is reduced, and that operating points in the low-flow/high-power region are least stable. We also determine the SB that results when the modal kinetics is replaced by simple point reactor kinetics, and we thereby show that the first harmonic mode does not have a significant effect on the SB. However, we later show that it nevertheless has a significant effect on stability because it affects the basin of attraction of stable operating points. Using numerical simulations we show that, in the important low-flow/high-power region, the Hopf bifurcation that occurs as the SB is crossed is subcritical; hence, growing oscillations can result following small finite perturbations of stable steady-states on the 100% rod line at points in the low-flow/high-power region. Numerical simulations are also performed to calculate the decay ratios (DRs) and frequencies of oscillations for various points on the 100% rod line. It is

  6. CAPRICORN subchannel code for sodium boiling in LMFBR fuel bundles

    International Nuclear Information System (INIS)

    Padilla, A. Jr.; Smith, D.E.; O'Dell, L.D.

    1983-01-01

    The CAPRICORN computer code analyzes steady-state and transient, single-phase and boiling problems in LMFBR fuel bundles. CAPRICORN uses the same type of subchannel geometry as the COBRA family of codes and solves a similar system of conservation equations for mass, momentum, and energy. However, CAPRICORN uses a different numerical solution method which allows it to handle the full liquid-to-vapor densit