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Sample records for range initial boiling

  1. 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

  2. Initiation of a boiling liquid expanding vapour explosion

    Energy Technology Data Exchange (ETDEWEB)

    McDevitt, C.A.; Chan, C.K.; Steward, F.R.; Tennankore, K.N.; Venart, J.E.S.

    1991-06-01

    Boiling liquid expanding vapour explosions (BLEVEs) occur when a pressure liquefied gas tank is burst and the fluid is ejected to the atmosphere. As the liquid is exposed to a lower pressure it boils violently, causing an explosion which destroys the container. One litre tanks were filled with propane or R12, heated, and caused to rupture at specific test conditions to determine the parameters required for a BLEVE. Results showed that the energy stored in the liquid under pressure relative to the liquid at atmospheric conditions was a quantity which can be used to predict a BLEVE for the particular apparatus and conditions used. The possible importance of the development of a shock wave within the container during a BLEVE was noted. A shock tube was used to study the shock waves generated during a BLEVE. Temperature, liquid volume, rupture location, rupture area, and the fluid involved were varied. The pressure was measured vs time for periods immediately after the rupture. Photographs of the formation of pressure waves were obtained using spark Schlieren photography and showed that the BLEVE phenomenon is definitely a shock-related event. Pressure information was also gathered during BLEVEs of one litre cylinders, and this information is compared to that from the shock tube. Shock tube data showed that transverse waves formed from the initial pressure wave could be amplified. 37 refs., 54 figs., 11 tabs.

  3. Initiation of a boiling liquid expanding vapour explosion

    Energy Technology Data Exchange (ETDEWEB)

    McDevitt, C.A.

    1990-01-01

    Boiling liquid expanding vapour explosions (BLEVEs) occur when a pressure liquefied gas tank is burst and the fluid is ejected to the atmosphere. As the liquid is exposed to a lower pressure it boils violently, causing an explosion which destroys the container. One litre tanks were filled with propane or R12, heated, and caused to rupture at specific test conditions to determine the parameters required for a BLEVE. Results showed that the energy stored in the liquid under pressure relative to the liquid at atmospheric conditions was a quantity which can be used to predict a BLEVE for the particular apparatus and conditions used. The possible importance of the development of a shock wave within the container during a BLEVE was noted. A shock tube was used to study the shock waves generated during a BLEVE. Temperature, liquid volume, rupture location, rupture area, and the fluid involved were varied. The pressure was measured vs time for periods immediately after the rupture. Photographs of the formation of pressure waves were obtained using spark Schlieren photography. Similarities to waves measured during detonations in ducts were noted. Pressure information was also gathered during BLEVEs of one litre cylinders, and this data is compared to that from the shock tube. Shock tube data showed that transverse waves formed from the initial pressure wave could be amplified. 37 refs., 54 figs., 11 tabs.

  4. 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

  5. Influence of initial conditions on rod behaviour during boiling crisis phase following a reactivity initiated accident

    International Nuclear Information System (INIS)

    Georgenthum, V.; Sugiyama, T.

    2010-01-01

    In the frame of their research programs on high burn-up fuel safety, the French Institute for Radioprotection and Nuclear Safety (IRSN) and the Japan Atomic Energy Agency (JAEA) performed a large set of tests devoted to the study of PWR fuel rod behavior during Reactivity Initiated Accident (RIA) respectively in the CABRI reactor and in the NSRR reactor. The reactor test conditions are different in terms of coolant nature, temperature and pressure. In the CABRI reactor, tests were performed until now with sodium coolant at 280 Celsius degrees and 3 bar. In the NSRR reactor most of the tests were performed with stagnant water at 20 C. degrees and atmospheric pressure but recently a new high temperature high pressure capsule has been developed which allows to performed tests at up to 280 Celsius degrees and 70 bar. The paper discusses the influence of test conditions on rod behaviour during boiling phase, based on tests results and SCANAIR code calculations. The study shows that when the boiling crisis is reached, the initial inner and outer rod pressure have an essential impact on the clad straining and possible ballooning. The analysis of the different test conditions makes it possible to discriminate the influence of initial conditions on the different phases of the transient and is useful for modelling and code development. (authors)

  6. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling... for public comment draft regulatory guide (DG), DG-1277, ``Initial Test Program of Emergency Core..., entitled, ``Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors,'' is...

  7. 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

  8. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0134] Initial Test Program of Emergency Core Cooling....79.1, ``Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors.'' This... emergency core cooling systems (ECCSs) for boiling- water reactors (BWRs) whose licenses are issued after...

  9. 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

  10. Global reaction mechanism for the auto-ignition of full boiling range gasoline and kerosene fuels

    Science.gov (United States)

    Vandersickel, A.; Wright, Y. M.; Boulouchos, K.

    2013-12-01

    Compact reaction schemes capable of predicting auto-ignition are a prerequisite for the development of strategies to control and optimise homogeneous charge compression ignition (HCCI) engines. In particular for full boiling range fuels exhibiting two stage ignition a tremendous demand exists in the engine development community. The present paper therefore meticulously assesses a previous 7-step reaction scheme developed to predict auto-ignition for four hydrocarbon blends and proposes an important extension of the model constant optimisation procedure, allowing for the model to capture not only ignition delays, but also the evolutions of representative intermediates and heat release rates for a variety of full boiling range fuels. Additionally, an extensive validation of the later evolutions by means of various detailed n-heptane reaction mechanisms from literature has been presented; both for perfectly homogeneous, as well as non-premixed/stratified HCCI conditions. Finally, the models potential to simulate the auto-ignition of various full boiling range fuels is demonstrated by means of experimental shock tube data for six strongly differing fuels, containing e.g. up to 46.7% cyclo-alkanes, 20% napthalenes or complex branched aromatics such as methyl- or ethyl-napthalene. The good predictive capability observed for each of the validation cases as well as the successful parameterisation for each of the six fuels, indicate that the model could, in principle, be applied to any hydrocarbon fuel, providing suitable adjustments to the model parameters are carried out. Combined with the optimisation strategy presented, the model therefore constitutes a major step towards the inclusion of real fuel kinetics into full scale HCCI engine simulations.

  11. 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

  12. 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

  13. 5 CFR 9701.372 - Creating initial pay ranges.

    Science.gov (United States)

    2010-01-01

    ... 5 Administrative Personnel 3 2010-01-01 2010-01-01 false Creating initial pay ranges. 9701.372... HUMAN RESOURCES MANAGEMENT SYSTEM Pay and Pay Administration Transitional Provisions § 9701.372 Creating initial pay ranges. (a) DHS must, after coordination with OPM, set the initial band rate ranges for the...

  14. Pellet-Cladding Mechanical Interaction Failure Threshold for Reactivity Initiated Accidents for Pressurized Water Reactors and Boiling Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Beyer, Carl E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Geelhood, Kenneth J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

    2013-06-01

    Pacific Northwest National Laboratory (PNNL) has been requested by the U.S. Nuclear Regulatory Commission to evaluate the reactivity initiated accident (RIA) tests that have recently been performed in the Nuclear Safety Research Reactor (NSRR) and CABRI (French research reactor) on uranium dioxide (UO2) and mixed uranium and plutonium dioxide (MOX) fuels, and to propose pellet-cladding mechanical interaction (PCMI) failure thresholds for RIA events. This report discusses how PNNL developed PCMI failure thresholds for RIA based on least squares (LSQ) regression fits to the RIA test data from cold-worked stress relief annealed (CWSRA) and recrystallized annealed (RXA) cladding alloys under pressurized water reactor (PWR) hot zero power (HZP) conditions and boiling water reactor (BWR) cold zero power (CZP) conditions.

  15. Damage analysis of ceramic boron absorber materials in boiling water reactors and initial model for an optimum control rod management

    International Nuclear Information System (INIS)

    Schulz, W.

    2000-01-01

    concept - to calculate the control rod's working life both in the control position and the shut-down position - will automatically lead to an optimization of the control rod strategy. Control rod optimisation is demonstrated by accumulating the total amount of control rods required in a medium-sized BWR up to the total reactor holding period. At least 60% of the first core inventory - for this control rod type an existing EMPIRICAL MODEL is already available - may be used up to the total operating period without any safety loss. Looking to the present disposal situation this concept represents a practical way to reduce all high level waste. In addition benefit of utilizing this concept is that it minimizes tritium emission. Control-rods utilized within Boiling Water Reactors (BWR) are designed for the purpose to control and shape the neutron flux profile in the reactor, to adjust the range of regulation referring to the weight rate of the reactor coolant and thirdly by- shutting down the reactor at any time and under any conditions with regard to nuclear aspects, mechanical integrity and control rod history. The designation control- or shut down rod characterize the particular field of activity for a given control rod. The focal point of my work had shown to be a calculation of the nuclear working life of any control rod design as well as an optimisation method with reference to the holding period for a given control rod inventory as a result of measuring data and a theoretical analysis describing the parameters in a general validity form. (author)

  16. Experimental study on transient boiling heat transfer

    International Nuclear Information System (INIS)

    Visentini, R.

    2012-01-01

    Boiling phenomena can be found in the everyday life, thus a lot of studies are devoted to them, especially in steady state conditions. Transient boiling is less known but still interesting as it is involved in the nuclear safety prevention. In this context, the present work was supported by the French Institute of Nuclear Safety (IRSN). In fact, the IRSN wanted to clarify what happens during a Reactivity-initiated Accident (RIA). This accident occurs when the bars that control the nuclear reactions break down and a high power peak is passed from the nuclear fuel bar to the surrounding fluid. The temperature of the nuclear fuel bar wall increases and the fluid vaporises instantaneously. Previous studies on a fuel bar or on a metal tube heated by Joule effect were done in the past in order to understand the rapid boiling phenomena during a RIA. However, the measurements were not really accurate because the measurement techniques were not able to follow rapid phenomena. The main goal of this work was to create an experimental facility able to simulate the RIA boiling conditions but at small scale in order to better understand the boiling characteristics when the heated-wall temperature increases rapidly. Moreover, the experimental set-up was meant to be able to produce less-rapid transients as well, in order to give information on transient boiling in general. The facility was built at the Fluid-Mechanics Institute of Toulouse. The core consists of a metal half-cylinder heated by Joule effect, placed in a half-annulus section. The inner half cylinder is made of a 50 microns thick stainless steel foil. Its diameter is 8 mm, and its length 200 mm. The outer part is a 34 mm internal diameter glass half cylinder. The semi-annular section is filled with a coolant, named HFE7000. The configuration allows to work in similarity conditions. The heated part can be place inside a loop in order to study the flow effect. The fluid temperature influence is taken into account as

  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. 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.

  19. 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.

  20. 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

  1. Boiling Suppression in Convective Flow

    International Nuclear Information System (INIS)

    Aounallah, Y.

    2004-01-01

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

  2. 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.

  3. 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.

  4. Singular system analysis of the Local Power Range Monitor (LPRM) readings of a Boiling Water Reactor (BWR) in an unstable event

    International Nuclear Information System (INIS)

    Ginestar Peiro, D.; Verdu, G.; Miro, R.

    2006-01-01

    Singular system analysis is a successful technique to separate oscillating components from a given signal. A methodology is proposed to apply this technique to the signals obtained from the LPRMs of a boiling water reactor core and extract the contributions of the in-phase oscillation and the out-of-phase oscillations from the LPRM readings during an unstable event. This methodology has been validated with synthetic signals and simulations of in-phase and out-of-phase oscillations of the Leibstadt reactor. Finally, one case of Ringhals I Stability Benchmark has been analysed. (author)

  5. 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)

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

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  7. Pool Boiling of Hydrocarbon Mixtures on Water

    Energy Technology Data Exchange (ETDEWEB)

    Boee, R.

    1996-09-01

    In maritime transport of liquefied natural gas (LNG) there is a risk of spilling cryogenic liquid onto water. The present doctoral thesis discusses transient boiling experiments in which liquid hydrocarbons were poured onto water and left to boil off. Composition changes during boiling are believed to be connected with the initiation of rapid phase transition in LNG spilled on water. 64 experimental runs were carried out, 14 using pure liquid methane, 36 using methane-ethane, and 14 using methane-propane binary mixtures of different composition. The water surface was open to the atmosphere and covered an area of 200 cm{sup 2} at 25 - 40{sup o}C. The heat flux was obtained by monitoring the change of mass vs time. The void fraction in the boiling layer was measured with a gamma densitometer, and a method for adapting this measurement concept to the case of a boiling cryogenic liquid mixture is suggested. Significant differences in the boil-off characteristics between pure methane and binary mixtures revealed by previous studies are confirmed. Pure methane is in film boiling, whereas the mixtures appear to enter the transitional boiling regime with only small amounts of the second component added. The results indicate that the common assumption that LNG will be in film boiling on water because of the high temperature difference, may be questioned. Comparison with previous work shows that at this small scale the results are influenced by the experimental apparatus and procedures. 66 refs., 76 figs., 28 tabs.

  8. 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

  9. 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)

  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. Boiling nucleation

    International Nuclear Information System (INIS)

    Cole, R.

    1974-01-01

    Experimental results of flash evaporation of a pool of water subjected to sudden pressure drop are reported. The experiments were conducted with pure water at equilibrium temperatures between 40 to 80 0 C and with superheat in the range of about 3 to 5 0 C. Two distinct exponential decaying processes were identified for flash evaporation and the flashing time was found to decrease with an increase of equilibrium temperature and with the decrease of superheat. Basic experiments on flash evaporation of distilled water were conducted. However, the results may not be quantitatively applicable to seawater flash evaporators as the presence of salts in the seawater will considerably change the surface tension and in turn affect the nonequilibrium fraction

  12. Molecular dynamics simulation of bubble nucleation in explosive boiling

    International Nuclear Information System (INIS)

    Zou Yu; Chinese Academy of Sciences, Beijing; Huai Xiulan; Liang Shiqiang

    2009-01-01

    Molecular dynamics (MD) simulation is carried out for the bubble nucleation of liquid nitrogen in explosive boiling. The heat is transferred into the simulation system by rescaling the velocity of the molecules. The results indicate that the initial equilibrium temperature of liquid and molecular cluster size affect the energy conversion in the process of bubble nucleation. The potential energy of the system violently varies at the beginning of the bubble nucleation, and then varies around a fixed value. At the end of bubble nucleation, the potential energy of the system slowly increases. In the bubble nucleation of explosive boiling, the lower the initial equilibrium temperature, the larger the size of the molecular cluster, and the more the heat transferred into the system of the simulation cell, causing the increase potential energy in a larger range. (authors)

  13. 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.

  14. 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

  15. 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…

  16. 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.)

  17. 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

  18. 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…

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

    Science.gov (United States)

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

    2014-01-01

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

  20. 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)

  1. 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)

  2. 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.

  3. 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

  4. 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)

  5. An experimental investigation of triggered film boiling destabilisation

    International Nuclear Information System (INIS)

    Naylor, P.

    1985-03-01

    Film boiling was established on a polished brass rod in water, collapse being initiated by either a pressure pulse or a transient bulk water flow. This work is relevant to the triggering stage of a molten fuel-coolant interaction, and a criterion is proposed for triggered film boiling collapse with pressure pulse. (U.K.)

  6. 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)

  7. Some observations on boiling heat transfer with surface oscillation

    International Nuclear Information System (INIS)

    Miyashita, H.

    1992-01-01

    The effects of surface oscillation on pool boiling heat transfer are experimentally studied. Experiments were performed in saturated ethanol and distilled water, covering the range from nucleate to film boiling except in the transition region. Two different geometries were employed as the heating surface with the same wetting area, stainless steel pipe and molybdenum ribbon. The results confirm earlier work on the effect of surface oscillation especially in lower heat flux region of nucleate boiling. Interesting boiling behavior during surface oscillation is observed, which was not referred to in previous work. (2 figures) (Author)

  8. Critical superheats upon boiling of dissociating liquids

    International Nuclear Information System (INIS)

    Kolykhan, L.I.; Solov'ev, V.N.

    1985-01-01

    The experimental data on critical superheats of dissociating liquids, i.e. nitrogen tetroxide and nitrine are presented (nitrine is the solution of nitrogen oxide in nitrogen tetroxide). The experiments with boiling N 2 O 4 have been carried out in the pressure range 0.1-3.0 MPa and with boiling nitrine within the pressure range 0.2-9.0 MPa. The experiments have revealed an anomalous dependence of critical superheats on pressure P, thus at P>=2.5 MPa the critical superheat values exceed the limiting ones, and at P=4.5 MPa this excess amounts to more than 16 K, essentially exceeding the errors of the experiments. The results for N 2 O 4 critical superheats agree with experimental data of other authors. Complex phenomena observed upon boiling of dissociating liquids require further theoretical and experimental studies

  9. 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)

  10. Determination of the boiling-point distribution by simulated distillation from n-pentane through n-tetratetracontane in 70 to 80 seconds.

    Science.gov (United States)

    Lubkowitz, Joaquin A; Meneghini, Roberto I

    2002-01-01

    This work presents the carrying out of boiling-point distributions by simulated distillation with direct-column heating rather than oven-column heating. Column-heating rates of 300 degrees C/min are obtained yielding retention times of 73 s for n-tetratetracontane. The calibration curves of the retention time versus the boiling point, in the range of n-pentane to n-tetratetracontane, are identical to those obtained by slower oven-heating rates. The boiling-point distribution of the reference gas oil is compared with that obtained with column oven heating at rates of 15 to 40 degrees C/min. The results show boiling-point distribution values nearly the same (1-2 degrees F) as those obtained with oven column heating from the initial boiling point to 80% distilled off. Slightly higher differences are obtained (3-4 degrees F) for the 80% distillation to final boiling-point interval. Nonetheless, allowed consensus differences are never exceeded. Precision of the boiling-point distributions (expressed as standard deviations) are 0.1-0.3% for the data obtained in the direct column-heating mode.

  11. 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)

  12. 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)

  13. Pool film boiling heat transfer, 5

    International Nuclear Information System (INIS)

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

    1981-01-01

    Steady minimum film boiling heat flux and temperature were experimentally studied for a horizontal cylinder test heater in a pool of saturated water under pressures ranging from 0.1 to 2 MPa. Minimum temperature of film boiling may be determined by hydrodynamic Taylor instability for the pressures lower than around 1.0 MPa and by homogeneous nucleation temperature for the higher pressures. However, conventional correlations of minimum heat flux based on the hydrodynamic Taylor instability cannot at all predict the pressure dependency of the experimental data in the lower pressure region. Semi-empirical equation of the minimum heat flux based on the hydrodynamic Taylor instability was given. (author)

  14. Boiling point measurements on liquid UO2

    International Nuclear Information System (INIS)

    Bober, M.; Singer, J.; Trapp, M.

    1986-01-01

    In analogy to the classic boiling point method, a quasi-stationary millisecond laser-heating technique was applied to measure the saturated-vapour pressure curve of liquid UO 2 in the temperature range of 3500 to 4500 K. The result is represented by log p(MPa) 5.049 -23042/T(K) according to an average heat of vaporization of 441 kJ/mol and a normal boiling point of 3808 K. Besides, spectral emissivities of liquid UO 2 were measured at the pyrometer wavelengths of 752 and 1064 nm. (author)

  15. Fragmentation of molten metal drop with instantaneous contact temperature below the boiling point of Na

    Energy Technology Data Exchange (ETDEWEB)

    Inukai, S.; Sugiyama, K. [Hokkaido Univ., Dept. of Nuclear Engineering, Sapporo (Japan); Nishimura, S.; Kinoshita, I. [Central Research Institute of Electric Power Industry, Tokyo (Japan)

    2001-07-01

    The consequence of the core disruptive accidents in metallic-fueled Na-cooled reactors is strongly affected by the feedback reactivity originating in the boiling of Na and the dispersion of molten fuel due to fuel-coolant interactions. The design of the core configuration to promote the dispersion of molten fuel is therefore very important for social acceptance. It has been recognized in this context that metallic fuel has a potentiality to make liquefied fuel with fuel pin tube even in the temperature range below the boiling point of Na. If the liquefied fuel solidified without fuel-coolant interactions in the core region, this event leads the core condition to a pessimistic scenario of re-criticality. As a basic study related to this problem, the present experimental study investigates the possibility of fragmentation of metal drop with instantaneous contact temperature below the boiling point of Na (883 C). The molten Al drop, which has a melting point of 660 C above the operational temperature range of core, was selected as a simulant of liquefied fuel in the present study. Al particles of 5 g or 0.56 g were heated up to the initial temperature ranging from 850 C to 1113 C in a crucible by using an electric heater. The molten Al drop was dropped into a sodium pool adjusted the temperature from 280 C to 499 C. The Al drop at initial temperature sufficiently higher that the boiling point of Na was observed to fragment into pieces under the condition of instantaneous contact temperature below the boiling point of Na. It is confirmed that the fragmentation is caused due to the thermal interactions between the molten Al and the Na entrapped into the drop. (author)

  16. Fragmentation of molten metal drop with instantaneous contact temperature below the boiling point of Na

    International Nuclear Information System (INIS)

    Inukai, S.; Sugiyama, K.; Nishimura, S.; Kinoshita, I.

    2001-01-01

    The consequence of the core disruptive accidents in metallic-fueled Na-cooled reactors is strongly affected by the feedback reactivity originating in the boiling of Na and the dispersion of molten fuel due to fuel-coolant interactions. The design of the core configuration to promote the dispersion of molten fuel is therefore very important for social acceptance. It has been recognized in this context that metallic fuel has a potentiality to make liquefied fuel with fuel pin tube even in the temperature range below the boiling point of Na. If the liquefied fuel solidified without fuel-coolant interactions in the core region, this event leads the core condition to a pessimistic scenario of re-criticality. As a basic study related to this problem, the present experimental study investigates the possibility of fragmentation of metal drop with instantaneous contact temperature below the boiling point of Na (883 C). The molten Al drop, which has a melting point of 660 C above the operational temperature range of core, was selected as a simulant of liquefied fuel in the present study. Al particles of 5 g or 0.56 g were heated up to the initial temperature ranging from 850 C to 1113 C in a crucible by using an electric heater. The molten Al drop was dropped into a sodium pool adjusted the temperature from 280 C to 499 C. The Al drop at initial temperature sufficiently higher that the boiling point of Na was observed to fragment into pieces under the condition of instantaneous contact temperature below the boiling point of Na. It is confirmed that the fragmentation is caused due to the thermal interactions between the molten Al and the Na entrapped into the drop. (author)

  17. 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.

  18. 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

    boiling from given boiling conditions with the pool CHF data measured by Dhir and Liaw and Paul and Abdel-Khalik and the subcooled flow CHF data measured by Del Valle M. and Kenning and with the heat flux data in transition boiling measured by Dhir and Liaw. The predictions show good agreement with the existing data. To use the present phenomenological model as a prediction tool, a study has been performed to predict CHF in pool and subcooled forced convection boiling using existing correlations for active site density, maximum bubble diameter, and heat transfer coefficients in nucleate boiling. Comparison of the model predictions with experimental data for pool boiling of water and upward flow boiling of water in vertical, uniformly-heated round tubes is performed. The data set (2438 data points) for CHF in subcooled forced convection boiling covers wide ranges of operating conditions (0.1≤P≤14.0 MPa, 0.00033≤D≤0.0375 m: 0.002≤L≤2 m: 660 ≤G≤90000 kg/m 2 s: 70≤Δh,≤1456 kJ/kg). Without any tuning factor, 1492 data points out of 2438 (61.2%) are calculated with a r.m.s. error of 41.3% and about 80% of the calculated data points are predicted within ±50%. It is also shown that by a modification of suppression factor in subcooled boiling, the predictive capability of the present model can be improved, i.e., 2421 data points (99.3%) are calculated with a r.m.s. error of 20.5% and 82.3% of the calculated data points are predicted within ±25%. In addition, the parametric trends of CHF in subcooled forced convection boiling have been investigated under local conditions hypothesis

  19. Pool Boiling CHF in Inclined Narrow Annuli

    International Nuclear Information System (INIS)

    Kang, Myeong Gie

    2010-01-01

    by Kang to identify the combined effects of the surface orientation and a confined space on pool boiling heat transfer in annuli. The gap size was 15 mm and the annuli with both open and closed bottoms were considered. At a given heat flux, the heat transfer coefficient was increased with the inclination angle increase. However, no occurrence of the CHF was observed regardless of the flow inlet condition for the given gap size and heat fluxes tested. Summarizing the published results, it can be said that the narrow gap size, restriction of the bottom inlet flow into the confined space, and the inclination angle not only changes nucleate boiling heat transfer but also initiates the CHF. Therefore, the present study is aimed at the investigation of the effects of a narrow gap size (5 mm) on pool boiling heat transfer in inclined annuli to improve Kang's previous results

  20. 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.)

  1. 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.)

  2. 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

  3. 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.

  4. 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)

  5. 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

  6. 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

  7. 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

  8. 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

  9. Initial land reclamation procedures related to possible Pu-cleanup activities at the Tonopah Test Range

    International Nuclear Information System (INIS)

    Wallace, A.; Romney, E.M.

    1976-02-01

    If areas of the Tonopah Test Range (TTR) are to be used for experimental tests of procedures for clean-up of 239 Pu contamination, there are experiences in the Great Basin Desert portions of the Nevada Test Site (NTS) which can serve as guides to reclamation and revegetation of such arid lands. Procedures which will encourage development of the grasses Hilaria jamesii and Oryzopsis hymenoides, as well as the perennial shrubs Eurotia lanata and Atriplex canescens would greatly improve the area as range land

  10. 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)

  11. 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)

  12. A Mitigation Approach to Counter Initial Ranging Based DoS Attacks on IEEE 802.16-2009

    International Nuclear Information System (INIS)

    Saleem, Y.; Asif, K.H.; Ahmad, T.; Bashir, K.

    2013-01-01

    In recent years increase in wireless accessed devices does not prerequisite any evidence. Security is the main concern for the researchers in 802.16e now-a-days. The layer structures defines that the security sub-layer resides over the physical layer and provides security on the link layer. This paper discusses the security threats present and still unsolved at the initial network entry stage. A mitigation approach to counter Initial Ranging Based DoS attacks on IEEE 802.16-2009 are particularized in this paper. Furthermore the existing solutions of initial ranging vulnerability are analyzed and their limitations are discussed. Proposed solution was checked against these limitations to ensure their absence. Moreover the solution was implemented in OMNET++ and results were analyzed to ensure the practicality and efficiency. (author)

  13. 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

  14. Research progresses and future directions on pool boiling heat transfer

    Directory of Open Access Journals (Sweden)

    M. Kumar

    2015-12-01

    Full Text Available This paper reviews the previous work carried on pool boiling heat transfer during heating of various liquids and commodities categorized as refrigerants and dielectric fluids, pure liquids, nanofluids, hydrocarbons and additive mixtures, as well as natural and synthetic colloidal solutions. Nucleate pool boiling is an efficient and effective method of boiling because high heat fluxes are possible with moderate temperature differences. It is characterized by the growth of bubbles on a heated surface. It occurs during boiling of liquids for excess temperature ranging from 5 to 30 °C in various processes related to high vaporization of liquid for specific purposes like sugarcane juice heating for jaggery making, milk heating for khoa making, steam generation, cooling of electronic equipments, refrigeration and etcetera. In this review paper, pool boiling method during heating of liquids for specific purpose is depicted. It is inferred that enhancement in pool boiling heat transfer is a challenging and complex task. Also, recent research and use of various correlations for natural convection pool boiling is reviewed.

  15. 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

  16. 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)

  17. 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.

  18. Onset of a nucleate boiling and incipient point of net vapor generation in narrow channel

    International Nuclear Information System (INIS)

    Hong, G.

    2014-01-01

    An experimental study on onset of nucleate boiling (ONB) and incipient point of net vapor generation (IPNVG) in narrow rectangular channel was presented. Flow direction in the channel was vertical upward. The experimental results indicate that the classical correlations of ONB for conventional channels were not suitable for the present narrow rectangular channel. The wall superheat needed to initiate boiling is found to be higher for the same given values of heat and mass flux. The experimental results of IPNVG indicate that the heat flux, triggering net vapor generation in narrow rectangular channel, is litter lower than that calculated by correlations for conventional channels. The relative prediction error of qIPNVG by Griffith model, Saha model and Sun model ranges from -17.9% to +9.6%. A new correlation was developed to predict the ONB in narrow rectangular channel. The proposed correlation predictions agreed well with the experimental data. (author)

  19. 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

  20. Investigation of Superficial Resistance of Different Purity Copper at Boiling Nitrogen Temperature Depending on Treatment of Current-Conducting Layer

    International Nuclear Information System (INIS)

    Kutovoj, V.A.; Nikolaenko, A.A.; Stoev, P.I.

    2007-01-01

    Results of this scientific work show influence of annealing temperature and deformation degree of initial MOB copper and after electron beam refining on superficial resistance at temperature of boiling nitrogen. It is shown, that 30 % deformation and annealing in 873...923 K temperature range results in appreciable reduction of superficial resistance at the investigated samples of copper. The lowest values of superficial resistance after thermal and mechanical treatment were observed in the samples after electron beam refinement

  1. 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)

  2. Heat transfer phenomena related to the boiling crisis

    International Nuclear Information System (INIS)

    Groenveld, D.C.

    1981-03-01

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

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

    International Nuclear Information System (INIS)

    Kenning, David B.R.

    2004-01-01

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

  4. 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.

  5. Study of film boiling collapse behavior during vapor explosion

    International Nuclear Information System (INIS)

    Yagi, Masahiro; Yamano, Norihiro; Sugimoto, Jun; Abe, Yutaka; Adachi, Hiromichi; Kobayashi, Tomoyoshi.

    1996-06-01

    Possible large scale vapor explosions are safety concern in nuclear power plants during severe accident. In order to identify the occurrence of the vapor explosion and to estimate the magnitude of the induced pressure pulse, it is necessary to investigate the triggering condition for the vapor explosion. As a first step of this study, scooping analysis was conducted with a simulation code based on thermal detonation model. It was found that the pressure at the collapse of film boiling much affects the trigger condition of vapor explosion. Based on this analytical results, basic experiments were conducted to clarify the collapse conditions of film boiling on a high temperature solid ball surface. Film boiling condition was established by flooding water onto a high temperature stainless steel ball heated by a high frequency induction heater. After the film boiling was established, the pressure pulse generated by a shock tube was applied to collapse the steam film on the ball surface. As the experimental boundary conditions, materials and size of the balls, magnitude of pressure pulse and initial temperature of the carbon and stainless steel balls were varied. The transients of pressure and surface temperature were measured. It was found that the surface temperature on the balls sharply decreased when the pressure wave passed through the film on balls. Based on the surface temperature behavior, the film boiling collapse pattern was found to be categorized into several types. Especially, the pattern for stainless steel ball was categorized into three types; no collapse, collapse and reestablishment after collapse. It was thus clarified that the film boiling collapse behavior was identified by initial conditions and that the pressure required to collapse film boiling strongly depended on the initial surface temperature. The present results will provide a useful information for the analysis of vapor explosions based on the thermal detonation model. (J.P.N.)

  6. Flow boiling test of GDP replacement coolants

    International Nuclear Information System (INIS)

    Park, S.H.

    1995-01-01

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

  7. 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

  8. Evaluation of onset of nucleate boiling models

    Energy Technology Data Exchange (ETDEWEB)

    Huang, LiDong [Heat Transfer Research, Inc., College Station, TX (United States)], e-mail: lh@htri.net

    2009-07-01

    This article discusses available models and correlations for predicting the required heat flux or wall superheat for the Onset of Nucleate Boiling (ONB) on plain surfaces. It reviews ONB data in the open literature and discusses the continuing efforts of Heat Transfer Research, Inc. in this area. Our ONB database contains ten individual sources for ten test fluids and a wide range of operating conditions for different geometries, e.g., tube side and shell side flow boiling and falling film evaporation. The article also evaluates literature models and correlations based on the data: no single model in the open literature predicts all data well. The prediction uncertainty is especially higher in vacuum conditions. Surface roughness is another critical criterion in determining which model should be used. However, most models do not directly account for surface roughness, and most investigators do not provide surface roughness information in their published findings. Additional experimental research is needed to improve confidence in predicting the required wall superheats for nucleation boiling for engineering design purposes. (author)

  9. Evaluation of onset of nucleate boiling models

    International Nuclear Information System (INIS)

    Huang, LiDong

    2009-01-01

    This article discusses available models and correlations for predicting the required heat flux or wall superheat for the Onset of Nucleate Boiling (ONB) on plain surfaces. It reviews ONB data in the open literature and discusses the continuing efforts of Heat Transfer Research, Inc. in this area. Our ONB database contains ten individual sources for ten test fluids and a wide range of operating conditions for different geometries, e.g., tube side and shell side flow boiling and falling film evaporation. The article also evaluates literature models and correlations based on the data: no single model in the open literature predicts all data well. The prediction uncertainty is especially higher in vacuum conditions. Surface roughness is another critical criterion in determining which model should be used. However, most models do not directly account for surface roughness, and most investigators do not provide surface roughness information in their published findings. Additional experimental research is needed to improve confidence in predicting the required wall superheats for nucleation boiling for engineering design purposes. (author)

  10. Study on water boiling noises in a large volume

    International Nuclear Information System (INIS)

    Masagutov, R.F.; Krivtsov, V.A.

    1977-01-01

    Presented are the results of measurement of the noise spectra during boiling of water in a large volume at the pressure of 1 at. Boiling of the distilled water has been accomplished with the use of the heaters made of the Kh18N10T steel, 50 mm in length, 2 mm in the outside diameter, with the wall thickness of 0.1 mm. The degree of water under heating changed during the experiments from 0 to 80 deg C, and the magnitude of the specific heat flux varied from o to 0.7 - 0.9 qsup(x), where qsup(x) was the specific heat flux of the tube burn-out. The noise spectrum of the boiling water was analyzed at frequencies of 0.5 to 200 kHz. The submerge-type pressure-electric transmitters were used for measurements. At underheating boiling during the experiment the standing waves have formed which determine the structure of the measured spectra. During saturated boiling of water no standing waves were revealed. At underheating over 15 - 20 deg C the water boiling process is accompanied by the noises within the ultrasonic frequency range. The maximum upper boundary of the noise in the experiments amounts to 90 - 100 kHz

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

    International Nuclear Information System (INIS)

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

    1981-10-01

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

  12. Explosive Boiling of Superheated Cryogenic Liquids

    CERN Document Server

    Baidakov, V G

    2007-01-01

    The monograph is devoted to the description of the kinetics of spontaneous boiling of superheated liquefied gases and their solutions. Experimental results are given on the temperature of accessible superheating, the limits of tensile strength of liquids due to processes of cavitation and the rates of nucleation of classical and quantum liquids. The kinetics of evolution of the gas phase is studied in detail for solutions of cryogenic liquids and gas-saturated fluids. The properties of the critical clusters (bubbles of critical sizes) of the newly evolving gas phase are analyzed for initial st

  13. 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)

  14. The initial magnetic susceptibility of polydisperse ferrofluids: A comparison between experiment and theory over a wide range of concentration

    International Nuclear Information System (INIS)

    Solovyova, Anna Y.; Goldina, Olga A.; Ivanov, Alexey O.; Elfimova, Ekaterina A.; Lebedev, Aleksandr V.

    2016-01-01

    Temperature dependencies of the static initial magnetic susceptibility for ferrofluids at various concentrations are studied using experiment and statistical-mechanical theories. Magnetic susceptibility measurements are carried out for twelve samples of magnetite-based fluids stabilized with oleic acid over a wide range of temperatures (210 K ≲T ≲ 390 K); all samples have the same granulometric composition but different volume ferroparticle concentrations (0.2 ≲ φ ≲ 0.5). Experimental results are analyzed using three theories: the second-order modified mean-field theory (MMF2) [A. O. Ivanov and O. B. Kuznetsova, Phys. Rev. E 64, 41405 (2001)]; its correction for polydisperse ferrofluids arising from Mayer-type cluster expansion and taking into account the first terms of the polydisperse second virial coefficient [A. O. Ivanov and E. A. Elfimova, J. Magn. Magn. Mater 374, 327 (2015)]; and a new theory based on MMF2 combined with the first terms of the polydisperse second and third virial contributions to susceptibility. It turns out that the applicability of each theory depends on the experimental sample density. If twelve ferrofluid samples are split into three groups of strong, moderate, and low concentrated fluids, the temperature dependences of the initial magnetic susceptibility in each group are very precisely described by one of the three theories mentioned above. The determination of a universal formula predicting a ferrofluid susceptibility over a broad range of concentrations and temperatures remains as a challenge.

  15. 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.

  16. 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

  17. A numerical investigation of the effect of surface wettability on the boiling curve.

    Directory of Open Access Journals (Sweden)

    Hua-Yi Hsu

    Full Text Available Surface wettability is recognized as playing an important role in pool boiling and the corresponding heat transfer curve. In this work, a systematic study of pool boiling heat transfer on smooth surfaces of varying wettability (contact angle range of 5° - 180° has been conducted and reported. Based on numerical simulations, boiling curves are calculated and boiling dynamics in each regime are studied using a volume-of-fluid method with contact angle model. The calculated trends in critical heat flux and Leidenfrost point as functions of surface wettability are obtained and compared with prior experimental and theoretical predictions, giving good agreement. For the first time, the effect of contact angle on the complete boiling curve is shown. It is demonstrated that the simulation methodology can be used for studying pool boiling and related dynamics and providing more physical insights.

  18. Nonlinear dynamics of boiling water reactors

    International Nuclear Information System (INIS)

    March-Leuba, J.; Cacuci, D.G.; Perez, R.B.

    1983-01-01

    Recent stability tests in Boiling Water Reactors (BWRs) have indicated that these reactors can exhibit the special nonlinear behavior of following a closed trajectory called limit cycle. The existence of a limit cycle corresponds to an oscillation of fixed amplitude and period. During these tests, such oscillations had their amplitudes limited to about +- 15% of the operating power. Since limit cycles are fairly insensitive to parameter variations, it is possible to operate a BWR under conditions that sustain a limit cycle (of fixed amplitude and period) over a finite range of reactor parameters

  19. Flow boiling in expanding microchannels

    CERN Document Server

    Alam, Tamanna

    2017-01-01

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

  20. 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. 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 density change for sodium boiling. Results of the initial comparison with data (the W-1 SLSF pipe rupture experiment) are very promising and provide an optimistic basis for proceeding with further development

  2. 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)

  3. 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

  4. 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

  5. A stability identification system for boiling water nuclear reactors

    International Nuclear Information System (INIS)

    Belblidia, L.A.; Chevrier, A.

    1994-01-01

    Boiling water reactors are subject to instabilities under low-flow, high-power operating conditions. These instabilities are a safety concern and it is therefore important to determine stability margins. This paper describes a method to estimate a measure of stability margin, called the decay ratio, from autoregressive modelling of time series data. A phenomenological model of a boiling water reactor with known stability characteristics is used to generate time series to validate the program. The program is then applied to signals from local power range monitors from the cycle 7 stability tests at the Leibstadt plant. (author) 7 figs., 2 tabs., 12 refs

  6. Hydrodynamic instability induced liquid--solid contacts in film boiling

    International Nuclear Information System (INIS)

    Yao, S.; Henry, R.E.

    1976-01-01

    The film boiling liquid-solid contacts of saturated ethanol and water to horizontal flat gold plated copper are examined by using electric conductance probe. It is observed that the liquid-solid contacts occur over a wide temperature range, and generally, induced by hydrodynamic instabilities. The area of contact decreases exponentially with interface temperature and is liquid depth dependent. The averaged duration of contacts is strongly influenced by the dominant nucleation process, and thus, depends on the interface temperature and the wettability of the solid during the contact. The frequency of major contacts is about 1.5 times the bubble detaching frequency. It is found that the liquid-solid contacts may account for a large percentage of the film boiling heat transfer near the low temperature end of film boiling and decreases as the interface temperature increases

  7. 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.

  8. 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

  9. 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

  10. 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

  11. 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)

  12. 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.

  13. 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

  14. 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

  15. 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

  16. Boiling water reactor life extension monitoring

    International Nuclear Information System (INIS)

    Stancavage, P.

    1991-01-01

    In 1991 the average age of GE-supplied Boiling Water Reactors (BWRs) reached 15 years. The distribution of BWR ages range from three years to 31 years. Several of these plants have active life extension programmes, the most notable of which is the Monticello plant in Minnesota which is the leading BWR plant for license renewal in the United States. The reactor pressure vessel and its internals form the heart of the boiling water reactor (BWR) power plant. Monitoring the condition of the vessel as it operates provides a continuous report on the structural integrity of the vessel and internals. Monitors for fatigue, stress corrosion and neutron effects can confirm safety margins and predict residual life. Every BWR already incorporates facilities to track the key aging mechanisms of fatigue, stress corrosion and neutron embrittlement. Fatigue is measured by counting the cycles experienced by the pressure vessel. Stress corrosion is gauged by periodic measurements of primary water conductivity and neutron embrittlement is tracked by testing surveillance samples. The drawbacks of these historical procedures are that they are time consuming, they lag the current operation, and they give no overall picture of structural integrity. GE has developed an integrated vessel fitness monitoring system to fill the gaps in the historical, piecemetal monitoring of the BWR vessel and internals and to support plant life extension. (author)

  17. 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.)

  18. Some fundamental aspects of boiling in nuclear reactors

    International Nuclear Information System (INIS)

    Mondin, H.; Lavigne, P.; Semeria, R.

    1964-01-01

    The main results obtained at Grenoble during the last four years in the field of boiling mechanisms and related phenomena in nuclear reactors are reported. 1 - Observation Of Boiling: By the use of photography and ultrafast cinematography (8000 frames per second maximum), boiling in a vessel or a tube was observed up to 140 kg/cm 2 . The populations of bubble-generating seeds (sites) were counted, and a correlation established giving their number per unit of surface area as a function of the thermal flux and the pressure. The diameter of the bubbles breaking of from the wall was studied up to 140 kg/cm 2 : three types of bubble have been shown to exist: - those in equilibrium, their diameter following the formula of Fritz and Ende, - bubbles found by boiling, the diameters of which decrease rapidly with the pressure (1/100 mm to 140 kg/cm 2 ), - the coalescences which appear in saturated liquid above 15 W/cm 2 , their proportion being independent of the pressure. Strioscopic observations were made of the movements of the thermal film associated with the generation of the seeds, at the initiation and condensation of the bubbles, the mechanisms responsible for the highly efficient heat transfer could thus be defined. 2 - Pressure Losses In Two-Phase Flow: A physical model of the continuous variation of the free space content in a boiling channel has been proposed by means of which the pressure losses can be calculated without invoking a break in the coefficient of friction when free boiling begins. Agreement between theory and experiment is satisfactory. The various forms which total pressure loss in a boiling tube may present as a function of flow rate have been studied. Special features are observed at very low and very high speeds. 3 - Burn-Out: Under steady operating conditions, it is shown that in a uniformly heated channel the burn-out flux as a function of output rate is generally independent of the length. When burn-out is a result of output oscillation, the

  19. Out-of-pile experiments with an electrical boiler for Acoustic Boiling Detection

    International Nuclear Information System (INIS)

    Aberle, J.; Bartholomay, R.; Reimann, G.; Rohrbacher, H.A.; Schleisiek, K.

    1978-03-01

    This report contains the experimental results of boiling tests obtained during the first testing phase in spring 1977 with an electrically heated 18-rod boiling generator installed in the sodium tank facility (NABEA) of IRE. The layout and performance of the boiling facility together with its instrumentation and criteria of selection of acoustic sensors for the detection of sodium boiling are described and discussed. The report provides information about the thermodynamics, blockage design and thermal conduction within the range of installation of the electric connecting head. The evaluation of the acoustic signals shows that boiling is indicated promptly and with a sufficiently high signal-to-noise ratio both by solid-born sensors and by high temperature microphones placed in the sodium

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

    International Nuclear Information System (INIS)

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

    1976-06-01

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

  1. 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

    Full text of publication follows: nano-fluids, or conventional liquids, e.g., water, with small concentration of nano-particles uniformly suspended, have attracted attention as a new heat transport medium with enhanced thermo-physical properties. Up to the present, only exploratory experiments on nano-fluids have been reported. Das et al (Int. J. Heat Mass Transfer 43, pp 3701-3707, 2003) conducted boiling experiments with water containing 38 nm Al 2 O 3 nano-particles. They observed deterioration in the nucleate boiling heat transfer due to the deposition of nano-particles. Boiling experiments conducted by Vassallo et al (Int. J. Heat Mass Transfer 47, pp 407-411, 2004) using silica nano-fluid using 0.4 mm diameter NiCr wire showed three times higher critical heat flux (CHF) and the wire traversed the film boiling region before it failed. Another independent experiment performed on 1 cm 2 square plate with a very low concentration of nano-particles ranging from 0.01 to 0.05 g/liter and at under pressure (2.89 psia), nano-fluids resulted in drastic 2∼3 times enhancement of the CHF (You and Kim, Appl. Phys. Lett. 83. No 16, 2003). However in all the aforementioned studies no appropriate explanation of the CHF enhancement has been advanced. The measured 2-3 times higher critical heat flux for very dilute nano-fluids may have high significance if such nano-fluids could be employed in heat transport systems. Recently, we investigated the effect of nano-particles on film boiling, which governs heat transfer during accident conditions in a reactor plant, e.g., in coolability of a degraded core, or a particulate debris bed or a core melt, and in steam explosions. Our previous experiments performed on film boiling in nano-fluids having larger concentrations of 5, 10, and 20 g/liter than those in You's experiments showed that the nano-fluids lower the film boiling temperature, decrease the film boiling heat transfer and provide a much thicker and more stable film than

  2. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-05-15

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

  4. An analytical model for particulate deposition on vertical heat transfer surfaces in a boiling environment

    International Nuclear Information System (INIS)

    Keefer, R.H.; Rider, J.L.; Waldman, L.A.

    1993-01-01

    A frequent problem in heat exchange equipment is the deposition of particulates entrained in the working fluid onto heat transfer surfaces. These deposits increase the overall heat transfer resistance and can significantly degrade the performance of the heat exchanger. Accurate prediction of the deposition rate is necessary to ensure that the design and specified operating conditions of the heat exchanger adequately address the effects of this deposit layer. Although the deposition process has been studied in considerable detail, much of the work has focused on investigating individual aspects of the deposition process. This paper consolidates this previous research into a mechanistically based analytical prediction model for particulate deposition from a boiling liquid onto vertical heat transfer surfaces. Consistent with the well known Kern-Seaton approach, the model postulates net particulate accumulation to depend on the relative contributions of deposition and reentrainment processes. Mechanisms for deposition include boiling, momentum, and diffusion effects. Reentrainment is presumed to occur via an intermittent erosion process, with the energy for particle removal being supplied by turbulent flow instabilities. The contributions of these individual mechanisms are integrated to obtain a single equation for the deposit thickness versus time. The validity of the resulting model is demonstrated by comparison with data published in the open literature. Model estimates show good agreement with data obtained over a range of thermal-hydraulic conditions in both flow and pool boiling environments. The utility of the model in performing parametric studies (e.g. to determine the effect of flow velocity on net deposition) is also demonstrated. The initial success of the model suggests that it could prove useful in establishing a range of heat exchanger. operating conditions to minimize deposition

  5. Heat Transfer in Boiling Dilute Emulsion with Strong Buoyancy

    Science.gov (United States)

    Freeburg, Eric Thomas

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

  6. BWR [boiling-water reactor] and PWR [pressurized-water reactor] off-normal event descriptions

    International Nuclear Information System (INIS)

    1987-11-01

    This document chronicles a total of 87 reactor event descriptions for use by operator licensing examiners in the construction of simulator scenarios. Events are organized into four categories: (1) boiling-water reactor abnormal events; (2) boiling-water reactor emergency events; (3) pressurized-water reactor abnormal events; and (4) pressurized-water reactor emergency events. Each event described includes a cover sheet and a progression of operator actions flow chart. The cover sheet contains the following general information: initial plant state, sequence initiator, important plant parameters, major plant systems affected, tolerance ranges, final plant state, and competencies tested. The progression of operator actions flow chart depicts, in a flow chart manner, the representative sequence(s) of expected immediate and subsequent candidate actions, including communications, that can be observed during the event. These descriptions are intended to provide examiners with a reliable, performance-based source of information from which to design simulator scenarios that will provide a valid test of the candidates' ability to safely and competently perform all licensed duties and responsibilities

  7. Spray structure as generated under homogeneous flash boiling nucleation regime

    International Nuclear Information System (INIS)

    Levy, M.; Levy, Y.; Sher, E.

    2014-01-01

    We show the effect of the initial pressure and temperature on the spatial distribution of droplets size and their velocity profile inside a spray cloud that is generated by a flash boiling mechanism under homogeneous nucleation regime. We used TSI's Phase Doppler Particle Analyzer (PDPA) to characterize the spray. We conclude that the homogeneous nucleation process is strongly affected by the initial liquid temperature while the initial pressure has only a minor effect. The spray shape is not affected by temperature or pressure under homogeneous nucleation regime. We noted that the only visible effect is in the spray opacity. Finally, homogeneous nucleation may be easily achieved by using a simple atomizer construction, and thus is potentially suitable for fuel injection systems in combustors and engines. - Highlights: • We study the characteristics of a spray that is generated by a flash boiling process. • In this study, the flash boiling process occurs under homogeneous nucleation regime. • We used Phase Doppler Particle Analyzer (PDPA) to characterize the spray. • The SMD has been found to be strongly affected by the initial liquid temperature. • Homogeneous nucleation may be easily achieved by using a simple atomizer unit

  8. Boiling on fins with wire screen of variable effective conductivity

    Directory of Open Access Journals (Sweden)

    Orzechowski Tadeusz

    2017-01-01

    Full Text Available The high scale of integration of modern equipment used for medical, military and other purposes puts heavy demands as regards the removal of great heat fluxes. This can be achieved only in exchangers that apply the phase change phenomena. Among many methods to improve boiling heat transfer, the wire mesh covering demonstrates some advantages due to the possibilities of designing the desired microstructure parameters, availability on the market, and low cost. The wire mesh microstucture with specified geometrical parameters produces anisotropy in conductivity. The different arrangement of the mesh layers relative to the direction of the heat flux is a cause of the change of temperature distribution within the layer. The consequence is a respective change in the discharge conditions of the gas phase and liquid feed. The experiments were conducted on fins covered with a single layer of copper mesh with lumen of 38 % and boiling FC-72 at ambient pressure. Compared with the smooth surface, the wire mesh structures yield an increase in the heat transfer rate at boiling. It is also shown that nucleate boiling is initiated at lower wall superheat. Formulas for longitudinal and perpendicular thermal conductivity are given for different mesh structure arrangements.

  9. Boiling Patterns of Iso-asymmetric Nuclear Matter

    International Nuclear Information System (INIS)

    Tõke, Jan

    2013-01-01

    Limits of thermodynamic metastability of self-bound neutron-rich nuclear matter are explored within the framework of microcanonical thermodynamics of interacting Fermi Gas model in Thomas-Fermi approximation. It is found that as the excitation energy per nucleon of the system is increased beyond a certain limiting value, the system loses metastability and becomes unstable with respect to joint fluctuations in excitation energy per nucleon and in isospin per nucleon. As a result, part of the system is forced to boil off in a form of iso-rich non-equilibrated vapors. Left behind in such a process, identifiable with distillation, is a more iso-symmetric metastable residue at a temperature characteristic of its residual isospin content. With a progressing increase in the initial excitation energy per nucleon, more neutron-rich matter is boiled off and a more iso-symmetric residue is left behind with progressively increasing characteristic temperature. Eventually, when all excess neutrons are shed, the system boils uniformly with a further supply of excitation energy, leaving behind a smaller and smaller residue at a characteristic boiling-point temperature of iso-symmetric matter.

  10. 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

  11. Gamma heated subassembly for sodium boiling experiments

    International Nuclear Information System (INIS)

    Artus, S.C.

    1975-01-01

    The design of a system to boil sodium in an LMFBR is examined. This design should be regarded as a first step in a series of boiling experiments. The reactor chosen for the design of the boiling apparatus is the Experimental Breeder Reactor-II (EBR-II), located at the National Reactor Testing Station in Idaho. Criteria broadly classified as design objectives and design requirements are discussed

  12. Gamma heated subassembly for sodium boiling experiments

    Energy Technology Data Exchange (ETDEWEB)

    Artus, S.C.

    1975-01-01

    The design of a system to boil sodium in an LMFBR is examined. This design should be regarded as a first step in a series of boiling experiments. The reactor chosen for the design of the boiling apparatus is the Experimental Breeder Reactor-II (EBR-II), located at the National Reactor Testing Station in Idaho. Criteria broadly classified as design objectives and design requirements are discussed.

  13. Instability in flow boiling in microchannels

    CERN Document Server

    Saha, Sujoy Kumar

    2016-01-01

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

  14. Preliminary results from film boiling destabilisation experiments

    International Nuclear Information System (INIS)

    Naylor, P.

    1984-05-01

    A series of experiments to investigate the triggered destabilisation of film boiling has been undertaken. Film boiling was established on a polished brass rod immersed in water and the effects of various triggers were investigated. Preliminary results are presented and two thresholds have been observed: an impulse threshold below which triggered destabilisation will not occur and a thermal threshold above which film boiling will re-establish following triggered destabilisation. (author)

  15. Film boiling heat transfer in liquid helium

    International Nuclear Information System (INIS)

    Inai, Nobuhiko

    1979-01-01

    The experimental data on the film boiling heat transfer in liquid helium are required for investigating the stability of superconducting wires. On the other hand, liquid helium has the extremely different physical properties as compared with the liquids at normal temperature such as water. In this study, the experiments on pool boiling were carried out, using the horizontal top surface of a 20 mm diameter copper cylinder in liquid helium. For observing individual bubbles, the experiments on film boiling from a horizontal platinum wire were performed separately in liquid nitrogen and liquid helium, and photographs of floating-away bubbles were taken. The author pointed out the considerable upward shift of the boiling curve near the least heat flux point in film boiling from the one given by the Berenson's equation which has been said to agree comparatively well with the data on the film boiling of the liquids at normal temperature, and the reason was investigated. Consequently, a model for film boiling heat transfer was presented. Also one equation expressing the film boiling at low heat flux for low temperature liquids was proposed. It represents well the tendency to shift from Berenson's equation of the experimental data on film boiling at the least heat flux point for liquid helium, liquid nitrogen and water having extremely different physical properties. Some discussions are added at the end of the paper. (Wakatsuki, Y.)

  16. Boiling of the Interface between Two Immiscible Liquids below the Bulk Boiling Temperatures of Both Components

    OpenAIRE

    Pimenova, Anastasiya V.; Goldobin, Denis S.

    2014-01-01

    We consider the problem of boiling of the direct contact of two immiscible liquids. An intense vapour formation at such a direct contact is possible below the bulk boiling points of both components, meaning an effective decrease of the boiling temperature of the system. Although the phenomenon is known in science and widely employed in technology, the direct contact boiling process was thoroughly studied (both experimentally and theoretically) only for the case where one of liquids is becomin...

  17. Boils

    Science.gov (United States)

    ... of Giving Governance By-Laws Committees Committee Service Conflict of Interest Policy Meeting Minutes Archive History Mission ... the normal, harmless bacteria we all carry. The source may be a family member, a pet or ...

  18. 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

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

    Kim, M.C.

    1981-01-01

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

  1. Signal processing for boiling noise detection

    International Nuclear Information System (INIS)

    Ledwidge, T.J.; Black, J.L.

    1989-01-01

    The present paper deals with investigations of acoustic signals from a boiling experiment performed on the KNS I loop at KfK Karlsruhe. Signals have been analysed in frequency as well as in time domain. Signal characteristics successfully used to detect the boiling process have been found in time domain. (author). 6 refs, figs

  2. Feedback stabilization of transition boiling states

    NARCIS (Netherlands)

    Gils, van R.W.; Speetjens, M.F.M.; Nijmeijer, H.

    2010-01-01

    A nonlinear one-dimensional heat-transfer model for pool boiling systems is considered. The model involves only the temperature distribution within the heater and models the heat exchange with the boiling medium via a nonlinear boundary condition imposed at the fluid-heater interface. This compact

  3. Contribution to the boiling curve of sodium

    International Nuclear Information System (INIS)

    Schins, H.E.J.

    1975-01-01

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

  4. 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)

  5. Enhanced mobility of poly(3-hexylthiophene) transistors by spin-coating from high-boiling-point solvents

    DEFF Research Database (Denmark)

    Chang, J.F.; Sun, B.Q.; Breiby, Dag Werner

    2004-01-01

    chloroform are typically on the order of 0.01 cm(2)/(V s). Here we investigate a range of solvents with higher boiling points. We find that 1,2,4-trichlorobenzene with good solubility and a high boiling point significantly improves the field-effect mobilities up to 0.12 cm(2)/(V s) with on:off ratios of 10...

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

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  7. Is goniometry suitable for measuring ankle range of motion in female ballet dancers? An initial comparison with radiographic measurement.

    Science.gov (United States)

    Russell, Jeffrey A; Shave, Ruth M; Kruse, David W; Nevill, Alan M; Koutedakis, Yiannis; Wyon, Matthew A

    2011-06-01

    Female ballet dancers require extreme ankle motion to attain the demi-plié (weight-bearing full dorsiflexion [DF]) and en pointe (weight-bearing full plantar flexion [PF]) positions of ballet. However, techniques for assessing this amount of motion have not yet received sufficient scientific scrutiny. Therefore, the purpose of this study was to examine possible differences between weight-bearing goniometric and radiographic ankle range of motion measurements in female ballet dancers. Ankle range of motion in 8 experienced female ballet dancers was assessed by goniometry and 2 radiographic measurement methods. The latter were performed on 3 mediolateral x-rays, in demi-plié, neutral, and en pointe positions; one of them used the same landmarks as goniometry. DF values were not significantly different among the methods, but PF values were (P ballet dancers and suggest that goniometry may not be ideal for assessing ankle range of motion in these individuals. Therefore, further research is needed to standardize how DF and PF are measured in ballet dancers. Diagnostic, Level I.

  8. Investigation of contact line dynamics under a vapor bubble at boiling on the transparent heater

    Science.gov (United States)

    Surtaev, A. S.; Serdyukov, V. S.

    2018-01-01

    The paper presents the results of an experimental study of dynamics of vapor bubble growth and departure at pool boiling, obtained with the use of high-speed video recording and IR thermography. The study was carried out at saturated water boiling under the atmospheric pressure in the range of heat fluxes of 30-150 kW/m2. To visualize the process and determine the growth rates of the outer bubble diameter, microlayer region and dry spot area, transpa-rent thin film heater with the thickness of 1 μm deposited on sapphire substrate was used in the experiments, and video recording was performed from the bottom side of the heating surface. To study integral heat transfer as well as local non-stationary thermal characteristics, high-speed infrared thermography with a frequency of up to 1000 FPS was used. High-speed video recording showed that after formation of vapor bubble and microlayer region, dry spot appears in a short time (up to 1 ms) under the vapor bubble. Various stages of contact line boundary propagation were ob-served. It was shown that at the initial stage before the development of small-scale perturbations, the dry spot propaga-tion rate is constant. It was also showed that the bubble departure stage begins after complete evaporation of liquid in the microlayer region.

  9. Study of vapour phase dynamics with nitrogen boiling in the field of centrifugal forces

    International Nuclear Information System (INIS)

    Levchenko, N.M.; Kolod'ko, I.M.

    1987-01-01

    The vapour phase dynamics during film boiling of liquid nitrogen on horizontal wire in the field of centrifugal forces has been studied experimentally in a wide range of overloads(1 ≤ η ≤ 375) and heat fluxes (q kp2 ≤ q ≤ 4q kpi ). The available data confirmed and the theoretical relationships suggested make it possible to calculate the hydrodynamic film boiling parameters (wave length, bubble departure diameter and frequency) for other liquids

  10. Boiling water reactor fuel bundle

    International Nuclear Information System (INIS)

    Weitzberg, A.

    1986-01-01

    A method is described of compensating, without the use of control rods or burnable poisons for power shaping, for reduced moderation of neutrons in an uppermost section of the active core of a boiling water nuclear reactor containing a plurality of elongated fuel rods vertically oriented therein, the fuel rods having nuclear fuel therein, the fuel rods being cooled by water pressurized such that boiling thereof occurs. The method consists of: replacing all of the nuclear fuel in a portion of only the upper half of first predetermined ones of the fuel rods with a solid moderator material of zirconium hydride so that the fuel and the moderator material are axially distributed in the predetermined ones of the fuel rods in an asymmetrical manner relative to a plane through the axial midpoint of each rod and perpendicular to the axis of the rod; placing the moderator material in the first predetermined ones of the fuel rods in respective sealed internal cladding tubes, which are separate from respective external cladding tubes of the first predetermined ones of the fuel rods, to prevent interaction between the moderator material and the external cladding tube of each of the first predetermined ones of the fuel rods; and wherein the number of the first predetermined ones of the fuel rods is at least thirty, and further comprising the steps of: replacing with the moderator material all of the fuel in the upper quarter of each of the at least thirty rods; and also replacing with the moderator material all of the fuel in the adjacent lower quarter of at least sixteen of the at least thirty rods

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

    International Nuclear Information System (INIS)

    Katto, Yoshiro; Ishii, Kazunori.

    1978-01-01

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

  12. A re-evaluation of the initial yield of the hydrated electron in the picosecond time range

    International Nuclear Information System (INIS)

    Muroya, Yusa; Lin Mingzhang; Wu, Guozhong; Iijima, Hokuto; Yoshii, Koji; Ueda, Toru; Kudo, Hisaaki; Katsumura, Yosuke

    2005-01-01

    The yield of the hydrated electron in the picosecond time range has been re-evaluated with an ultrafast pulse radiolysis system using a laser photocathode RF-gun in combination with a conventional one, and a value of 4.1±0.2 per 100 eV of absorbed energy at 20 ps was derived. This is consistent with recent experimental results using a time correlation method [Bartels et al., J. Phys. Chem. A 104, 1686-1691 (2000)] and with Monte-Carlo calculations [Muroya et al., Can. J. Chem. 80 1367-1374 (2002)

  13. Void fraction calculation in a channel containing boiling coolant

    International Nuclear Information System (INIS)

    Norelli, F.

    1978-01-01

    The problem of void fraction calculation was studied for a channel containing boiling coolant, when a slip ratio correlation is used. Use of fitting (e.g. polinomial or rational algebraic) for slip ratio correlation and the characteristic method are proposed in this work. In this way we are reduced to some elementary quadrature problem. Another problem discussed in the present work concerns what we must consider as ''initial condition'' in any initial value problem, in order to take into account different error distributions in steady state and in successive time-dependent calculations

  14. Effect of boiling regime on melt stream breakup in water

    International Nuclear Information System (INIS)

    Spencer, B.W.; Gabor, J.D.; Cassulo, J.C.

    1986-01-01

    A study has been performed examining the breakup and mixing behavior of an initially coherent stream of high-density melt as it flows downward through water. This work has application to the quenching of molten core materials as they drain downward during a postulated severe reactor accident. The study has included examination of various models of breakup distances based upon interfacial instabilities dominated either by liquid-liquid contact or by liquid-vapor contact. A series of experiments was performed to provide a data base for assessment of the various modeling approaches. The experiments involved Wood's metal (T/sub m/ = 73 0 C, rho = 9.2 g/cm 3 , d/sub j/ = 20 mm) poured into a deep pool of water. The temperature of the water and wood's metal were varied to span the range from single-phase, liquid-liquid contact to the film boiling regime. Experiment results showed that breakup occurred largely as a result of the spreading and entrainment from the leading edge of the jet. However, for streams of sufficient lengths a breakup length could be discerned at which there was no longer a coherent central core of the jet to feed the leading edge region. The erosion of the vertical trailing column is by Kelvin-Helmoltz instabilities and related disengagement of droplets from the jet into the surrounding fluid. For conditions of liquid-liquid contact, the breakup length has been found to be about 20 jet diameters; when substantial vapor is produced at the interface due to heat transfer from the jet to the water, the breakup distance was found to range to as high as 50 jet diameters. The former values are close to the analytical prediction of Taylor, whereas the latter values are better predicted by the model of Epstein and Fauske

  15. Elucidating the role of vegetation in the initiation of rainfall-induced shallow landslides: Insights from an extreme rainfall event in the Colorado Front Range

    Science.gov (United States)

    McGuire, Luke; Rengers, Francis K.; Kean, Jason W.; Coe, Jeffrey A.; Mirus, Benjamin B.; Baum, Rex L.; Godt, Jonathan W.

    2016-01-01

    More than 1100 debris flows were mobilized from shallow landslides during a rainstorm from 9 to 13 September 2013 in the Colorado Front Range, with the vast majority initiating on sparsely vegetated, south facing terrain. To investigate the physical processes responsible for the observed aspect control, we made measurements of soil properties on a densely forested north facing hillslope and a grassland-dominated south facing hillslope in the Colorado Front Range and performed numerical modeling of transient changes in soil pore water pressure throughout the rainstorm. Using the numerical model, we quantitatively assessed interactions among vegetation, rainfall interception, subsurface hydrology, and slope stability. Results suggest that apparent cohesion supplied by roots was responsible for the observed connection between debris flow initiation and slope aspect. Results suggest that future climate-driven modifications to forest structure could substantially influence landslide hazards throughout the Front Range and similar water-limited environments where vegetation communities may be more susceptible to small variations in climate.

  16. Heat transfer enhancement on nucleate boiling

    International Nuclear Information System (INIS)

    Zhuang, M.; Guibai, L.

    1990-01-01

    This paper reports on enhancement of nucleate boiling heat transfer with additives that was investigated experimentally. More than fifteen kinds of additives were chosen and tested. Eight kinds of effective additives which can enhance nucleate boiling heat transfer were selected. Experimental results showed that boiling heat transfer coefficient of water was increased by 1 to 5 times and that of R-113 was increased by 1 to 4 times when trace amount additives were put in the two boiling liquids. There exist optimum concentrations for the additives, respectively, which can enhance nucleate boiling heat transfer rate best. In order to analyze the mechanism of the enhancement of boiling heat transfer with additives, the surface tension and the bubble departure diameter were measured. The nucleation sites were investigated by use of high-speed photograph. Experimental results showed that nucleation sites increase with additive amount increasing and get maximum. Increasing nucleation sites is one of the most important reason why nucleate boiling heat transfer can be enhanced with additives

  17. Boiling Heat Transfer to Halogenated Hydrocarbon Refrigerants

    Science.gov (United States)

    Yoshida, Suguru; Fujita, Yasunobu

    The current state of knowledge on heat transfer to boiling refrigerants (halogenated hydrocarbons) in a pool and flowing inside a horizontal tube is reviewed with an emphasis on information relevant to the design of refrigerant evaporators, and some recommendations are made for future research. The review covers two-phase flow pattern, heat transfer characteristics, correlation of heat transfer coefficient, influence of oil, heat transfer augmentation, boiling from tube-bundle, influence of return bend, burnout heat flux, film boiling, dryout and post-dryout heat transfer.

  18. Retention of Underrepresented Minority Faculty: Strategic Initiatives for Institutional Value Proposition Based on Perspectives from a Range of Academic Institutions.

    Science.gov (United States)

    Whittaker, Joseph A; Montgomery, Beronda L; Martinez Acosta, Veronica G

    2015-01-01

    The student and faculty make-up of academic institutions does not represent national demographics. Racial and ethnic minorities are disproportionately underrepresented nationally, and particularly at predominantly white institutions (PWIs). Although significant efforts and funding have been committed to increasing points of access or recruitment of under-represented minority (URM) students and faculty at PWIs, these individuals have not been recruited and retained at rates that reflect their national proportions. Underrepresentation of URMs is particularly prevalent in Science, Technology, Engineering, and Mathematics (STEM) disciplines. This reality represents a national crisis given a predicted shortage of workers in STEM disciplines based on current rates of training of all individuals, majority and URM, and the intersection of this limitation with persistent challenges in the recruitment, training, retention and advancement of URMs who will soon represent the largest pool of future trainees. An additional compounding factor is the increasingly disproportionate underrepresentation of minorities at higher professorial and administrative ranks, thus limiting the pool of potential mentors who are correlated with successful shepherding of URM students through STEM training and development. We address issues related to improving recruitment and retention of URM faculty that are applicable across a range of academic institutions. We describe challenges with recruitment and retention of URM faculty and their advancement through promotion in the faculty ranks and into leadership positions. We offer specific recommendations, including identifying environmental barriers to diversity and implementing strategies for their amelioration, promoting effective and innovative mentoring, and addressing leadership issues related to constructive change for promoting diversity.

  19. 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)

  20. Sodium boiling noise topics in the German Democratic Republic

    International Nuclear Information System (INIS)

    Ziegenbein, D.

    1982-01-01

    surveillance of the core is of basic importance because of the fact that local positive reactivity can be the consequence of the sodium void effect. Sodium boiling experiments were carried out on BOR-60 reactor. These studies have shown that valuable information can be obtained from the signal properties in the frequency range as well as in the time domain. In a contribution to this conference some conclusions are drawn from these boiling experiments using different methods of signal analysis

  1. 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

  2. Recovering low-boiling hydrocarbons, etc

    Energy Technology Data Exchange (ETDEWEB)

    Pier, M

    1934-10-03

    A process is described for the recovery of low-boiling hydrocarbons of the nature of benzine through treatment of liquid carbonaceous materials with hydrogen under pressure at raised temperature, suitably in the presence of catalysts. Middle oils (practically saturated with hydrogen) or higher boiling oils at a temperature above 500/sup 0/ (with or without the addition of hydrogen) containing cyclic hydrocarbons not saturated with hydrogen are changed into low boiling hydrocarbons of the nature of benzine. The cracking takes place under strongly hydrogenating conditions (with the use of a strongly active hydrogenating catalyst or high pressure) at temperatures below 500/sup 0/. If necessary, the constituents boiling below 200/sup 0/ can be reconverted into cyclic hydrocarbons partially saturated with hydrogen. (BLM)

  3. 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

  4. Dynamic model for a boiling water reactor

    International Nuclear Information System (INIS)

    Muscettola, M.

    1963-07-01

    A theoretical formulation is derived for the dynamics of a boiling water reactor of the pressure tube and forced circulation type. Attention is concentrated on neutron kinetics, fuel element heat transfer dynamics, and the primary circuit - that is the boiling channel, riser, steam drum, downcomer and recirculating pump of a conventional La Mont loop. Models for the steam and feedwater plant are not derived. (author)

  5. Numerical simulation of single bubble boiling behavior

    Directory of Open Access Journals (Sweden)

    Junjie Liu

    2017-06-01

    Full Text Available The phenomena of a single bubble boiling process are studied with numerical modeling. The mass, momentum, energy and level set equations are solved using COMSOL multi-physics software. The bubble boiling dynamics, the transient pressure field, velocity field and temperature field in time are analyzed, and reasonable results are obtained. The numeral model is validated by the empirical equation of Fritz and could be used for various applications.

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

    International Nuclear Information System (INIS)

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

    1982-01-01

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

  7. An efficient reliable method to estimate the vaporization enthalpy of pure substances according to the normal boiling temperature and critical properties

    Directory of Open Access Journals (Sweden)

    Babak Mehmandoust

    2014-03-01

    Full Text Available The heat of vaporization of a pure substance at its normal boiling temperature is a very important property in many chemical processes. In this work, a new empirical method was developed to predict vaporization enthalpy of pure substances. This equation is a function of normal boiling temperature, critical temperature, and critical pressure. The presented model is simple to use and provides an improvement over the existing equations for 452 pure substances in wide boiling range. The results showed that the proposed correlation is more accurate than the literature methods for pure substances in a wide boiling range (20.3–722 K.

  8. An efficient reliable method to estimate the vaporization enthalpy of pure substances according to the normal boiling temperature and critical properties.

    Science.gov (United States)

    Mehmandoust, Babak; Sanjari, Ehsan; Vatani, Mostafa

    2014-03-01

    The heat of vaporization of a pure substance at its normal boiling temperature is a very important property in many chemical processes. In this work, a new empirical method was developed to predict vaporization enthalpy of pure substances. This equation is a function of normal boiling temperature, critical temperature, and critical pressure. The presented model is simple to use and provides an improvement over the existing equations for 452 pure substances in wide boiling range. The results showed that the proposed correlation is more accurate than the literature methods for pure substances in a wide boiling range (20.3-722 K).

  9. 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

  10. Boiling experiments in DFR and PFR

    International Nuclear Information System (INIS)

    Judd, A.M.

    1994-01-01

    At the end of its life, in 1975-1977, a series of Special Experiments was conducted in the Dounreay Fast Reactor. Fuel pins were deliberately subjected to overheating, up to the coolant boiling point, for periods of several hours at a time. The boiling was monitored by acoustic sensors and thermocouples, and after the tests the fuel pins were examined to determine the extent of damage. The results of these experiments have been widely reported. The present paper summarises the results as a reminder of their significance. The outstanding conclusion was that coolant boiling had no severe consequences. In some, but not all, cases the pins failed, but little fuel was released, no local blockages were formed, and there was no fuel melting. At around the same time PFR was being commissioned, and for a time the primary coolant circuit was operated with a dummy core, containing no nuclear fuel. An electrically-heated boiling rig was deployed in the dummy core, and observed by acoustic monitors. The data gathered enabled the noise of boiling to be compared with the background noise, and provided valuable support for the design of acoustic boiling noise detection systems. (author)

  11. Boiling-up of a liquid in a large volume at fast pressure drop

    International Nuclear Information System (INIS)

    Isaev, O.A.; Pavlov, P.A.

    1980-01-01

    Experiment results on sharp pressure drop in overheated water and carbon dioxide are presented. Pressure fields are investigated upon seal failure of the tube for various initial temperatures varying in the 0.57-0.97 interval on critical temperature. The depth of the liOuid inlet into the metastable region can be compared with maximum permissible superheating of a pure liquid. The applicability of fluctuation embrion formation for pressure calculation in the initial phase of explosive boiling-up at seal failure of the system is considered. The nature of boiling centers origin is discussed

  12. Enhancement of pool boiling heat transfer coefficients using carbon nanotubes

    International Nuclear Information System (INIS)

    Park, Ki Jung; Jung, Dong Soo

    2007-01-01

    In this study, the effect of carbon nanotubes (CNTs) on nucleate boiling heat transfer is investigated. Three refrigerants of R22, R123, R134a, and water were used as working fluids and 1.0 vol.% of CNTs was added to the working fluids to examine the effect of CNTs. Experimental apparatus was composed of a stainless steel vessel and a plain horizontal tube heated by a cartridge heater. All data were obtained at the pool temperature of 7 .deg. C for all refrigerants and 100 .deg. C for water in the heat flux range of 10∼80 kW/m 2 . Test results showed that CNTs increase nucleate boiling heat transfer coefficients for all fluids. Especially, large enhancement was observed at low heat fluxes of less than 30 kW/m 2 . With increasing heat flux, however, the enhancement was suppressed due to vigorous bubble generation. Fouling on the heat transfer surface was not observed during the course of this study. Optimum quantity and type of CNTs and their dispersion should be examined for their commercial application to enhance nucleate boiling heat transfer in many applications

  13. Modelling of Control Bars in Calculations of Boiling Water Reactors

    International Nuclear Information System (INIS)

    Khlaifi, A.; Buiron, L.

    2004-01-01

    The core of a nuclear reactor is generally composed of a neat assemblies of fissile material from where neutrons were descended. In general, the energy of fission is extracted by a fluid serving to cool clusters. A reflector is arranged around the assemblies to reduce escaping of neutrons. This is made outside the reactor core. Different mechanisms of reactivity are generally necessary to control the chain reaction. Manoeuvring of Boiling Water Reactor takes place by controlling insertion of absorbent rods to various places of the core. If no blocked assembly calculations are known and mastered, blocked assembly neutronic calculation are delicate and often treated by case to case in present studies [1]. Answering the question how to model crossbar for the control of a boiling water reactor ? requires the choice of a representation level for every chain of variables, the physical model, and its representing equations, etc. The aim of this study is to select the best applicable parameter serving to calculate blocked assembly of a Boiling Water Reactor. This will be made through a range of representative configurations of these reactors and used absorbing environment, in order to illustrate strategies of modelling in the case of an industrial calculation. (authors)

  14. 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.

  15. Determination of Sodium, Potassium, Magnesium, and Calcium Minerals Level in Fresh and Boiled Broccoli and Cauliflower by Atomic Absorption Spectrometry

    Science.gov (United States)

    Nerdy

    2018-01-01

    Vegetables from the cabbage family vegetables consumed by many people, which is known healthful, by eaten raw, boiled, or cooked (stir fry or soup). Vegetables like broccoli and cauliflower contain vitamins, minerals, and fiber. This study aims to determine the decrease percentage of sodium, potassium, magnesium, and calcium minerals level caused by boiled broccoli and cauliflower by atomic absorption spectrometry. Boiled broccoli and cauliflower prepared by given boiled treatment in boiling water for 3 minutes. Fresh and boiled broccoli and cauliflower carried out dry destruction, followed by quantitative analysis of sodium, potassium, magnesium, and calcium minerals respectively at a wavelength of 589.0 nm; 766.5 nm; 285.2 nm; and 422.7 nm, using atomic absorption spectrometry methods. After the determination of the sodium, potassium, magnesium, and calcium minerals level followed by validation of analytical methods with accuracy, precision, linearity, range, limit of detection (LOD), and limit of quantitation (LOQ) parameters. Research results show a decrease in the sodium, potassium, magnesium, and calcium minerals level in boiled broccoli and cauliflower compared with fresh broccoli and cauliflower. Validation of analytical methods gives results that spectrometry methods used for determining sodium, potassium, magnesium, and calcium minerals level are valid. It concluded that the boiled gives the effect of decreasing the minerals level significantly in broccoli and cauliflower.

  16. Nutrition content of brisket point end of part Simental Ongole Crossbred meat in boiled various temperature

    Science.gov (United States)

    Riyanto, J.; Sudibya; Cahyadi, M.; Aji, A. P.

    2018-01-01

    This aim of this study was to determine the quality of nutritional contents of beef brisket point end of Simental Ongole Crossbred meat in various boiling temperatures. Simental Ongole Crossbred had been fattened for 9 months. Furthermore, they were slaughtered at slaughterhouse and brisket point end part of meat had been prepared to analyse its nutritional contents using Food Scan. These samples were then boiled at 100°C for 0 (TR), 15 (R15), and 30 (R30) minutes, respectively. The data was analysed using Randomized Complete Design (CRD) and Duncan’s multiple range test (DMRT) had been conducted to differentiate among three treatments. The results showed that boiling temperatures significantly affected moisture, and cholesterol contents of beef (P<0.05) while fat content was not significantly affected by boiling temperatures. The boiling temperature decreased beef water contents from 72.77 to 70.84%, on the other hand, the treatment increased beef protein and cholesterol contents from 20.77 to 25.14% and 47.55 to 50.45 mg/100g samples, respectively. The conclusion of this study was boiling of beef at 100°C for 15 minutes and 30 minutes decreasing water content and increasing protein and cholesterol contents of brisket point end of Simental Ongole Crossbred beef.

  17. Specific features of hydrogen boiling heat transfer on the AMg-6 alloy massive heater

    International Nuclear Information System (INIS)

    Kirichenko, Yu.A.; Kozlov, S.M.; Rusanov, K.V.; Tyurina, E.G.

    1989-01-01

    Heat transfer and nucleate burns-out saturated with hydrogen at a plate heater (thickness-13 mm, diameter of heat-transferring surface - 30 mm) made of an aluminium alloy with the low value of a heat assimilation coefficient in the pressure range from 7.2x10 3 to 6x10 5 Pa is experimentally investigated. Value of start of boiling characteristics and heat transfer coefficients during nucleate burn-out, as well as the first critical densities of a heat flux and temperature heads are obtained. Existence of certain differrences of heat exchange during boiling is shown using a massive heater made of low-heat-conductive material in comparison with other cases of hydrogen boiling. Hypothesis concerning the existence of so-called mixed boiling on the heat transfer surface, which has been detected earlier only in helium boiling, as well as concerning possible reasons of stability of film boiling ficii in preburn-out region of heat duty is discussed

  18. An efficient reliable method to estimate the vaporization enthalpy of pure substances according to the normal boiling temperature and critical properties

    OpenAIRE

    Mehmandoust, Babak; Sanjari, Ehsan; Vatani, Mostafa

    2014-01-01

    The heat of vaporization of a pure substance at its normal boiling temperature is a very important property in many chemical processes. In this work, a new empirical method was developed to predict vaporization enthalpy of pure substances. This equation is a function of normal boiling temperature, critical temperature, and critical pressure. The presented model is simple to use and provides an improvement over the existing equations for 452 pure substances in wide boiling range. The results s...

  19. 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.

  20. 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

  1. Specialists' meeting on sodium boiling noise detection. Summary report

    International Nuclear Information System (INIS)

    1982-01-01

    The purpose of the meeting was to review and discuss methods available for detection of the initial stage of accidents in fast reactors, with most attention on reliable detection by acoustic techniques, which could provide a valuable addition to the safety protection. Results obtained from reactor experiments were also discussed and recommendations made for future developments. The meeting was divided into five technical sessions as follows: Signals from sodium boiling; Transmission of acoustic waves and background noise; Detection techniques; Reactor experiments; and Future requirements

  2. Specialists' meeting on sodium boiling noise detection. Summary report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1982-01-01

    The purpose of the meeting was to review and discuss methods available for detection of the initial stage of accidents in fast reactors, with most attention on reliable detection by acoustic techniques, which could provide a valuable addition to the safety protection. Results obtained from reactor experiments were also discussed and recommendations made for future developments. The meeting was divided into five technical sessions as follows: Signals from sodium boiling; Transmission of acoustic waves and background noise; Detection techniques; Reactor experiments; and Future requirements.

  3. Incorporating Water Boiling in the Numerical Modelling of Thermal Remediation by Electrical Resistance Heating

    Science.gov (United States)

    Molnar, I. L.; Krol, M.; Mumford, K. G.

    2017-12-01

    Developing numerical models for subsurface thermal remediation techniques - such as Electrical Resistive Heating (ERH) - that include multiphase processes such as in-situ water boiling, gas production and recovery has remained a significant challenge. These subsurface gas generation and recovery processes are driven by physical phenomena such as discrete and unstable gas (bubble) flow as well as water-gas phase mass transfer rates during bubble flow. Traditional approaches to multiphase flow modeling soil remain unable to accurately describe these phenomena. However, it has been demonstrated that Macroscopic Invasion Percolation (MIP) can successfully simulate discrete and unstable gas transport1. This has lead to the development of a coupled Electro Thermal-MIP Model2 (ET-MIP) capable of simulating multiple key processes in the thermal remediation and gas recovery process including: electrical heating of soil and groundwater, water flow, geological heterogeneity, heating-induced buoyant flow, water boiling, gas bubble generation and mobilization, contaminant mass transport and removal, and additional mechanisms such as bubble collapse in cooler regions. This study presents the first rigorous validation of a coupled ET-MIP model against two-dimensional water boiling and water/NAPL co-boiling experiments3. Once validated, the model was used to explore the impact of water and co-boiling events and subsequent gas generation and mobilization on ERH's ability to 1) generate, expand and mobilize gas at boiling and NAPL co-boiling temperatures, 2) efficiently strip contaminants from soil during both boiling and co-boiling. In addition, a quantification of the energy losses arising from steam generation during subsurface water boiling was examined with respect to its impact on the efficacy of thermal remediation. While this study specifically targets ERH, the study's focus on examining the fundamental mechanisms driving thermal remediation (e.g., water boiling) renders

  4. Investigation on the applicability of Piety's on-line PSD-pattern recognition algorithm to boiling detection by neutron-noise at a swimming-pool reactor

    International Nuclear Information System (INIS)

    Behringer, K.; Spiekerman, G.; Yadigaroglu, G.

    1984-11-01

    The neutron noise signal of an initiation-of-boiling experiment performed at the SAPHIR reactor has been analyzed by the PSD-pattern recognition algorithm of Piety (1977); the results indicate that the onset of boiling can be detected by this method. Improved confidence statements for the statistical decision discriminants are given. (Auth.)

  5. 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)

  6. A Review of Wettability Effect on Boiling Heat Transfer Enhancement

    International Nuclear Information System (INIS)

    Seo, Gwang Hyeok; Jeun, Gyoo Dong; Kim, Sung Joong

    2012-01-01

    Critical heat flux (CHF) and nucleate boiling heat transfer coefficient (NBHTC) are the key parameters characterizing pool boiling heat transfer. These variables are complicatedly related to thermal-hydraulic parameters of surface wettability, nucleation site density, bubble departure diameter and frequency, to mention a few. In essence, wettability effect on pool boiling heat transfer has been a major fuel to enhance the CHF. Often, however, the improved wettability effect hinders the nucleate boiling. Thus a comprehensive review of such wettability effect may enlighten a further study in this boiling heat transfer area. Phan et al. described surface wettability effects on boiling heat transfer

  7. 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.)

  8. SU-C-207A-06: On-Line Beam Range Verification with Multiple Scanning Particle Beams: Initial Feasibility Study with Simulations

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, Y; Sun, X; Lu, W; Jia, X; Wang, J; Shao, Y [The University of Texas Southwestern Medical Ctr., Dallas, TX (United States)

    2016-06-15

    Purpose: To investigate the feasibility and requirement for intra-fraction on-line multiple scanning particle beam range verifications (BRVs) with in-situ PET imaging, which is beyond the current single-beam BRV with extra factors that will affect the BR measurement accuracy, such as beam diameter, separation between beams, and different image counts at different BRV positions. Methods: We simulated a 110-MeV proton beam with 5-mm diameter irradiating a uniform PMMA phantom by GATE simulation, which generated nuclear interaction-induced positrons. In this preliminary study, we simply duplicated these positrons and placed them next to the initial protons to approximately mimic the two spatially separated positron distributions produced by two beams parallel to each other but with different beam ranges. These positrons were then imaged by a PET (∼2-mm resolution, 10% sensitivity, 320×320×128 mm^3 FOV) with different acquisition times. We calculated the positron activity ranges (ARs) from reconstructed PET images and compared them with the corresponding ARs of original positron distributions. Results: Without further image data processing and correction, the preliminary study show the errors between the measured and original ARs varied from 0.2 mm to 2.3 mm as center-to-center separations and range differences were in the range of 8–12 mm and 2–8 mm respectively, indicating the accuracy of AR measurement strongly depends on the beam separations and range differences. In addition, it is feasible to achieve ≤ 1.0-mm accuracy for both beams with 1-min PET acquisition and 12 mm beam separation. Conclusion: This study shows that the overlap between the positron distributions from multiple scanning beams can significantly impact the accuracy of BRVs of distributed particle beams and need to be further addressed beyond the established method of single-beam BRV, but it also indicates the feasibility to achieve accurate on-line multi-beam BRV with further improved

  9. Nucleate pool-boiling heat transfer - I. Review of parametric effects of boiling surface

    International Nuclear Information System (INIS)

    Pioro, I.L.; Rohsenow, W.; Doerffer, S.S.

    2004-01-01

    The objective of this paper is to assess the state-of-the-art of heat transfer in nucleate pool-boiling. Therefore, the paper consists of two parts: part I reviews and examines the effects of major boiling surface parameters affecting nucleate-boiling heat transfer, and part II reviews and examines the existing prediction methods to calculate the nucleate pool-boiling heat transfer coefficient (HTC). A literature review of the parametric trends points out that the major parameters affecting the HTC under nucleate pool-boiling conditions are heat flux, saturation pressure, and thermophysical properties of a working fluid. Therefore, these effects on the HTC under nucleate pool-boiling conditions have been the most investigated and are quite well established. On the other hand, the effects of surface characteristics such as thermophysical properties of the material, dimensions, thickness, surface finish, microstructure, etc., still cannot be quantified, and further investigations are needed. Particular attention has to be paid to the characteristics of boiling surfaces. (author)

  10. Angular distribution of thick-target bremsstrahlung produced by electrons with initial energies ranging from 10 to 20 keV incident on Ag

    Energy Technology Data Exchange (ETDEWEB)

    Gonzales, D.; Cavness, B.; Williams, S. [Department of Physics, Angelo State University, San Angelo, Texas 76909 (United States)

    2011-11-15

    Experimental results are presented comparing the intensities of the bremsstrahlung produced by electrons with initial energies ranging from 10 to 20 keV incident on a thick Ag target, measured at forward angles in the range of 0 degree sign to 55 degree sign . When the data are corrected for attenuation due to photon absorption within the target, the results indicate that the detected radiation is distributed anisotropically only at photon energies k that are approximately equal to the initial energy of the incident electrons E{sub 0}. The results of our experiments suggest that, as k/E{sub 0}{yields} 0, the detected radiation essentially becomes isotropic due primarily to the scattering of electrons within the target. A comparison to the theory of Kissel et al.[At. Data Nucl. Data Tables 28, 381 (1983)] suggests that the angular distribution of bremsstrahlung emitted by electrons incident on thick targets is similar to the angular distribution of bremsstrahlung emitted by electrons incident on free-atom targets only when k/E{sub 0}{approx_equal} 1. The experimental data also are in approximate agreement with the angular distribution predictions of the Monte Carlo program penelope.

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

    Science.gov (United States)

    Xu, Bin; Shi, Yumei; Chen, Dongsheng

    2014-03-01

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

  12. CONTINUOUS ANALYZER UTILIZING BOILING POINT DETERMINATION

    Science.gov (United States)

    Pappas, W.S.

    1963-03-19

    A device is designed for continuously determining the boiling point of a mixture of liquids. The device comprises a distillation chamber for boiling a liquid; outlet conduit means for maintaining the liquid contents of said chamber at a constant level; a reflux condenser mounted above said distillation chamber; means for continuously introducing an incoming liquid sample into said reflux condenser and into intimate contact with vapors refluxing within said condenser; and means for measuring the temperature of the liquid flowing through said distillation chamber. (AEC)

  13. The boiling point of stratospheric aerosols.

    Science.gov (United States)

    Rosen, J. M.

    1971-01-01

    A photoelectric particle counter was used for the measurement of aerosol boiling points. The operational principle involves raising the temperature of the aerosol by vigorously heating a portion of the intake tube. At or above the boiling point, the particles disintegrate rather quickly, and a noticeable effect on the size distribution and concentration is observed. Stratospheric aerosols appear to have the same volatility as a solution of 75% sulfuric acid. Chemical analysis of the aerosols indicates that there are other substances present, but that the sulfate radical is apparently the major constituent.

  14. Boiling crisis during liquid motion at high mass rate and underheating in channels

    International Nuclear Information System (INIS)

    Solov'ev, D.S.; Solov'ev, S.L.

    2007-01-01

    One describes a physical model of liquid boiling crisis in smooth wall channels for high values of underheating and of mass rate of flow. One worded a condition ensuring crisis initiation. Paper presents the calculated ratio for the critical density of thermal flow in channels enabling to analyze the effect of channel diameter, of underheating and of mass rate on it [ru

  15. 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)

  16. How Hydroclimate Influences the Effectiveness of Particle Filter Data Assimilation of Streamflow in Initializing Short- to Medium-range Streamflow Forecasts

    Science.gov (United States)

    Clark, E.; Wood, A.; Nijssen, B.; Clark, M. P.

    2017-12-01

    Short- to medium-range (1- to 7-day) streamflow forecasts are important for flood control operations and in issuing potentially life-save flood warnings. In the U.S., the National Weather Service River Forecast Centers (RFCs) issue such forecasts in real time, depending heavily on a manual data assimilation (DA) approach. Forecasters adjust model inputs, states, parameters and outputs based on experience and consideration of a range of supporting real-time information. Achieving high-quality forecasts from new automated, centralized forecast systems will depend critically on the adequacy of automated DA approaches to make analogous corrections to the forecasting system. Such approaches would further enable systematic evaluation of real-time flood forecasting methods and strategies. Toward this goal, we have implemented a real-time Sequential Importance Resampling particle filter (SIR-PF) approach to assimilate observed streamflow into simulated initial hydrologic conditions (states) for initializing ensemble flood forecasts. Assimilating streamflow alone in SIR-PF improves simulated streamflow and soil moisture during the model spin up period prior to a forecast, with consequent benefits for forecasts. Nevertheless, it only consistently limits error in simulated snow water equivalent during the snowmelt season and in basins where precipitation falls primarily as snow. We examine how the simulated initial conditions with and without SIR-PF propagate into 1- to 7-day ensemble streamflow forecasts. Forecasts are evaluated in terms of reliability and skill over a 10-year period from 2005-2015. The focus of this analysis is on how interactions between hydroclimate and SIR-PF performance impact forecast skill. To this end, we examine forecasts for 5 hydroclimatically diverse basins in the western U.S. Some of these basins receive most of their precipitation as snow, others as rain. Some freeze throughout the mid-winter while others experience significant mid-winter melt

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

    International Nuclear Information System (INIS)

    Klein, G.A.

    1979-01-01

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

  18. Relationships between boiling regimes and chemical concentration processes in tube support plate crevices

    International Nuclear Information System (INIS)

    Baum, A.

    2002-01-01

    The results of a test inter-relating crevice boiling regimes and chemical concentration in tube support plate crevices are presented. Testing of highly soluble, non-volatile autoclave chemistries produced characteristic crevice pH and impedance distributions during nucleate boiling, initiation of dryout, steady-state operation, and following shutdown. However, the patterns changed as a function of the solubility and volatility of the autoclave chemistry, the solute concentration, and the presence of residual solutes from previous testing. The changes were related to variations in the rates of concentrated solution formation, transport, volatilization, and precipitation. (authors)

  19. 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)

  20. Radionuclide buildup in BWR [boiling water reactor] reactor coolant recirculation piping

    International Nuclear Information System (INIS)

    Duce, S.W.; Marley, A.W.; Freeman, A.L.

    1989-12-01

    Since the spring of 1985, thermoluminescent dosimeter, dose rate, and gamma spectral data have been acquired on the contamination of boiling water reactor primary coolant recirculation systems as part of a Nuclear Regulatory Commission funded study. Data have been gathered for twelve facilities by taking direct measurements and/or obtaining plant and vendor data. The project titled, ''Effectiveness and Safety Aspects of Selected Decontamination Processes'' (October 1983) initially reviewed the application of chemical decontamination processes on primary coolant recirculation system piping. Recontamination of the system following pipe replacement or chemical decontamination was studied as a second thrust of this program. During the course of this study, recontamination measurements were made at eight different commercial boiling water reactors. At four of the reactors the primary coolant recirculation system piping was chemically decontaminated. At the other four the piping was replaced. Vendor data were obtained from two boiling water reactors that had replaced the primary coolant recirculation system piping. Contamination measurements were made at two newly operating boiling water reactors. This report discusses the results of these measurements as they apply to contamination and recontamination of boiling water reactor recirculation piping. 16 refs., 29 figs., 9 tabs

  1. Lattice Boltzmann Methods to Address Fundamental Boiling and Two-Phase Problems

    Energy Technology Data Exchange (ETDEWEB)

    Uddin, Rizwan

    2012-01-01

    This report presents the progress made during the fourth (no cost extension) year of this three-year grant aimed at the development of a consistent Lattice Boltzmann formulation for boiling and two-phase flows. During the first year, a consistent LBM formulation for the simulation of a two-phase water-steam system was developed. Results of initial model validation in a range of thermo-dynamic conditions typical for Boiling Water Reactors (BWRs) were shown. Progress was made on several fronts during the second year. Most important of these included the simulation of the coalescence of two bubbles including the surface tension effects. Work during the third year focused on the development of a new lattice Boltzmann model, called the artificial interface lattice Boltzmann model (AILB model) for the 3 simulation of two-phase dynamics. The model is based on the principle of free energy minimization and invokes the Gibbs-Duhem equation in the formulation of non-ideal forcing function. This was reported in detail in the last progress report. Part of the efforts during the last (no-cost extension) year were focused on developing a parallel capability for the 2D as well as for the 3D codes developed in this project. This will be reported in the final report. Here we report the work carried out on testing the AILB model for conditions including the thermal effects. A simplified thermal LB model, based on the thermal energy distribution approach, was developed. The simplifications are made after neglecting the viscous heat dissipation and the work done by pressure in the original thermal energy distribution model. Details of the model are presented here, followed by a discussion of the boundary conditions, and then results for some two-phase thermal problems.

  2. Tube temperature rise limits: Boiling considerations

    Energy Technology Data Exchange (ETDEWEB)

    Vanderwater, R.G.

    1952-03-26

    A revision of tube power limits based on boiling considerations was presented earlier. The limits were given on a basis of tube power versus header pressure. However, for convenience of operation, the limits have been converted from tube power to permissible water temperature rise. The permissible {triangle}t`s water are given in this document.

  3. Boiling Heat Transfer in Battery Electric vehicles

    NARCIS (Netherlands)

    Gils, van R.W.; Speetjens, M.F.M.; Nijmeijer, H.

    2011-01-01

    In this paper the feedback stabilisation of a boiling-based cooling scheme is discussed. Application of such cooling schemes in practical setups is greatly limited by the formation of a thermally insulating vapour film on the to-be-cooled device, called burn-out. In this study a first step is made,

  4. 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

  5. The film boiling look-up table: an improvement in predicting post-chf temperatures

    International Nuclear Information System (INIS)

    Groeneveld, D.C.; Leung, L.K.H.; Vasic, A.Z.; Guo, Y.J.; El Nakla, M.; Cheng, S.C.

    2002-01-01

    During the past 50 years more than 60 film boiling prediction methods have been proposed (Groeneveld and Leung, 2000). These prediction methods generally are applicable over limited ranges of flow conditions and do not provide reasonable predictions when extrapolated well outside the range of their respective database. Leung et al. (1996, 1997) and Kirillov et al. (1996) have proposed the use of a film-boiling look-up table as an alternative to the many models, equations and correlations for the inverted annular film boiling (IAFB) and the dispersed flow film-boiling (DFFB) regime. The film-boiling look-up table is a logical follow-up to the development of the successful CHF look-up table (Groeneveld et al., 1996). It is basically a normalized data bank of heat-transfer coefficients for discrete values of pressure, mass flux, quality and heat flux or surface-temperature. The look-up table proposed by Leung et al. (1996, 1997), and referred to as PDO-LW-96, was based on 14,687 data and predicted the surface temperature with an average error of 1.2% and an rms error of 6.73%. The heat-transfer coefficient was predicted with an average error of -4.93% and an rms error of 16.87%. Leung et al. clearly showed that the look-up table approach, as a general predictive tool for film-boiling heat transfer, was superior to the correlation or model approach. Error statistics were not provided for the look-up table proposed by Kirillov et al. (1996). This paper reviews the look-up table approach and describes improvements to the derivation of the film-boiling look-up table. These improvements include: (i) a larger data base, (ii) a wider range of thermodynamic qualities, (iii) use of the wall temperature instead of the heat flux as an independent parameter, (iv) employment of fully-developed film-boiling data only for the derivation of the look-up table, (v) a finer subdivision and thus more table entries, (vi) smoother table, and (vii) use of the best of five prediction methods

  6. An experimental investigation of untriggered film boiling collapse

    International Nuclear Information System (INIS)

    Naylor, P.

    1985-03-01

    Film boiling has been investigated in a stagnant pool, using polished brass or anodised aluminium alloy rods in water. Experimental boiling curves were obtained, and pronounced ripples on the vapour/liquid interface were photographed. A criterion for untriggered film boiling collapse is proposed, consistent with experimental results. Application of the results to molten fuel coolant interaction studies is discussed. (U.K.)

  7. Estimation of boiling point of radon by radiogas chromatography

    International Nuclear Information System (INIS)

    Takahashi, N.; Otozai, K.

    1986-01-01

    The retention volume of radon was measured by means of radiogas chromatography. The boiling point of radon was estimated from the retention volume by the use of the semi-empirical formula relating the boiling point to the retention volume. The obtained boiling point (198+-2)K was lower by 13 K than that measured by Gray and Ramsay. (author)

  8. Fluid inclusion from drill hole DW-5, Hohi geothermal area, Japan: Evidence of boiling and procedure for estimating CO2 content

    Science.gov (United States)

    Sasada, M.; Roedder, E.; Belkin, H.E.

    1986-01-01

    Fluid inclusion studies have been used to derive a model for fluid evolution in the Hohi geothermal area, Japan. Six types of fluid inclusions are found in quartz obtained from the drill core of DW-5 hole. They are: (I) primary liquid-rich with evidence of boiling; (II) primary liquid-rich without evidence of boiling; (III) primary vapor-rich (assumed to have been formed by boiling); (IV) secondary liquid-rich with evidence of boiling; (V) secondary liquid-rich without evidence of boiling; (VI) secondary vapor-rich (assumed to have been formed by boiling). Homogenization temperatures (Th) range between 196 and 347??C and the final melting point of ice (Tm) between -0.2 and -4.3??C. The CO2 content was estimated semiquantitatively to be between 0 and 0.39 wt. % based on the bubble behavior on crushing. NaCl equivalent solid solute salinity of fluid inclusions was determined as being between 0 and 6.8 wt. % after minor correction for CO2 content. Fluid inclusions in quartz provide a record of geothermal activity of early boiling and later cooling. The CO2 contents and homogenization temperatures of fluid inclusions with evidence of boiling generally increase with depth; these changes, and NaCl equivalent solid solute salinity of the fluid can be explained by an adiabatic boiling model for a CO2-bearing low-salinity fluid. Some high-salinity inclusions without CO2 are presumed to have formed by a local boiling process due to a temperature increase or a pressure decrease. The liquid-rich primary and secondary inclusions without evidence of boiling formed during the cooling process. The salinity and CO2 content of these inclusions are lower than those in the boiling fluid at the early stage, probably as a result of admixture with groundwater. ?? 1986.

  9. Impurity concentration behaviors in a boiling tubesheet crevice Part II. Packed crevice

    International Nuclear Information System (INIS)

    Bahn, Chi Bum; Oh, Si Hyoung; Park, Byung Gi; Hwang, Il Soon; Rhee, In Hyoung; Kim, Uh Chul; Na, Jung Won

    2003-01-01

    The impurity concentration behavior of a boiling crevice packed with magnetite particles was investigated with thermocouples and electrodes for the measurement of temperature and electrochemical corrosion potential (ECP), respectively, in order to understand chemical change in a pressurized water reactor (PWR) steam generator (SG) crevice. A secondary solution composed of 50 ppm Na and 200 ppb hydrogen was supplied at a flow rate of about 4 l/h. Sodium hydroxide (NaOH) concentration process in the crevice and the resultant boiling point elevation behavior were characterized with temperature and ECP data. The temperature in the packed crevice was about 2-3 deg. C higher than that for the open crevice. In the same conditions, the magnetite-packed crevice showed a greater amount of boiling point elevation with a longer time to reach a steady state compared with the case of an open crevice. It was found that the bottom region of the crevice was initially filled with steam, and then the concentrated liquid region initially located at the middle of crevice expanded to both the crevice bottom and the upper region. To analytically estimate the wetted length, a closed form model was introduced. The model results estimated the initial wetted length shorter as compared with the measurement results. Measured ECP results of packed crevice showed similar behaviors as compared with calculated results by using Nernst equation. ECP results reasonably coincided with the boiling point elevation estimated from the temperature data except one unusual case

  10. The effect of vapor polarity and boiling point on breakthrough for binary mixtures on respirator carbon.

    Science.gov (United States)

    Robbins, C A; Breysse, P N

    1996-08-01

    This research evaluated the effect of the polarity of a second vapor on the adsorption of a polar and a nonpolar vapor using the Wheeler model. To examine the effect of polarity, it was also necessary to observe the effect of component boiling point. The 1% breakthrough time (1% tb), kinetic adsorption capacity (W(e)), and rate constant (kv) of the Wheeler model were determined for vapor challenges on carbon beds for both p-xylene and pyrrole (referred to as test vapors) individually, and in equimolar binary mixtures with the polar and nonpolar vapors toluene, p-fluorotoluene, o-dichlorobenzene, and p-dichlorobenzene (referred to as probe vapors). Probe vapor polarity (0 to 2.5 Debye) did not systematically alter the 1% tb, W(e), or kv of the test vapors. The 1% tb and W(e) for test vapors in binary mixtures can be estimated reasonably well, using the Wheeler model, from single-vapor data (1% tb +/- 30%, W(e) +/- 20%). The test vapor 1% tb depended mainly on total vapor concentration in both single and binary systems. W(e) was proportional to test vapor fractional molar concentration (mole fraction) in mixtures. The kv for p-xylene was significantly different (p boiling point; however, these differences were apparently of limited importance in estimating 1% tb for the range of boiling points tested (111 to 180 degrees C). Although the polarity and boiling point of chemicals in the range tested are not practically important in predicting 1% tb with the Wheeler model, an effect due to probe boiling point is suggested, and tests with chemicals of more widely ranging boiling point are warranted. Since the 1% tb, and thus, respirator service life, depends mainly on total vapor concentration, these data underscore the importance of taking into account the presence of other vapors when estimating respirator service life for a vapor in a mixture.

  11. 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

  12. Mineralogical, micromorphological and geochemical transformations in the initial steps of the weathering process of charnockite from the Caparaó Range, southeastern Brazil

    Science.gov (United States)

    Soares, Caroline Cibele Vieira; Varajão, Angélica Fortes Drummond Chicarino; Varajão, César Augusto Chicarino; Boulangé, Bruno

    2014-12-01

    X-ray diffraction (XRD), X-ray Fluorescence (XRF), optical microscopy, Scanning Electron Microscopy coupled with Energy Dispersive Spectrometry (SEM-EDS) and Electron Probe micro-analyser (EPMA) and Wavelength-Dispersive Spectroscopy (WDS) were conducted on charnockite from the Caparaó Suite and its alteration cortex to determine the mineralogical, micromorphological and geochemical transformations resulting from the weathering process. The hydrolysis of the charnockite occurred in different stages, in accordance with the order of stability of the minerals with respect to weathering: andesine/orthopyroxene, pargasite and alkali feldspar. The rock modifications had begun with the formation of a layer of incipient alteration due to the percolation of weathering solutions first in the pressure relief fractures and then in cleavage and mineral edges. The iron exuded from ferromagnesian minerals precipitated in the intermineral and intramineral discontinuities. The layer of incipient alteration evolves into an inner cortex where the plagioclase changes into gibbsite by direct alitisation, the ferromagnesian minerals initiate the formation of goethitic boxworks with kaolinitic cores, and the alkali feldspar initiates indirect transformation into gibbsite, forming an intermediate phase of illite and kaolinite. In the outer cortex, mostly traces of alkali feldspar remain, and they are surrounded by goethite and gibbsite as alteromorphics, characterising the formation of the isalteritic horizon that occurs along the slope and explains the bauxitization process at the Caparaó Range, SE Brazil.

  13. Pool boiling performance of lithium bromide solution on enhanced tubes

    International Nuclear Information System (INIS)

    Sim, Yong Sub; Kim, Nae Hyun

    2015-01-01

    In the present study, pool boiling heat transfer coefficients of LiBr solution were obtained for smooth, corrugated, floral and low fin tubes. Test range covered saturation pressure from 7.38 kPa to 101.3 kPa, LiBr concentration from 0% to 50%. Low fin tube yielded the highest heat transfer coefficient followed by floral, corrugated and smooth tube. The sequence was the same independent of saturation pressure or LiBr concentration. For the low fin tube, the heat transfer enhancement ratio ranged from 1.34 to 2.37. Those of floral tube are 1.16 to 1.64 and those of corrugated tube are 0.90 to 1.43. The enhancement ratio decreased as LiBr concentration increased. No general trend is observed for the effect of saturation pressure. Exceptionally high heat transfer coefficient was obtained for the low fin tube in pure water. The reason may be attributed to many small bubbles at high frequency due to the existence of low fins. Boiling heat transfer correlations were developed which are applicable for saturation pressure from 7.38 kPa to 101.3 kPa and LiBr concentration from 0% to 50%.

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

    Science.gov (United States)

    Merte, Herman, Jr.

    1988-01-01

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

  15. 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

  16. 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)

  17. 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)

  18. 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

  19. 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)

  20. Study of vapour phase dynamics with nitrogen boiling in the field of centrifugal forces

    Energy Technology Data Exchange (ETDEWEB)

    Levchenko, N M; Kolod' ko, I M

    1987-07-01

    The vapour phase dynamics during film boiling of liquid nitrogen on horizontal wire in the field of centrifugal forces has been studied experimentally in a wide range of overloads(1 less than or equal to eta less than or equal to 375) and heat fluxes (q/sub kp2/ less than or equal to q less than or equal to 4q/sub kpi/). The available data confirmed and the theoretical relationships suggested make it possible to calculate the hydrodynamic film boiling parameters (wave length, bubble departure diameter and frequency) for other liquids.

  1. Vapor pressure determination of liquid UO/sub 2/ using a boiling point technique

    International Nuclear Information System (INIS)

    Bober, M.; Singer, J.

    1987-01-01

    By analogy with the classic boiling point method, a quasi-stationary millisecond laser-heating technique was applied to measure the saturated vapor pressure curve of liquid UO/sub 2/ in the temperature range of 3500 to 4500 K. The results are represented by log rho (MPa)=5.049 - 23 042/T (K), which gives an average heat of vaporization of 441 kJ/mol and a normal boiling point of 3808 K. In addition, spectral emissivities of liquid UO/sub 2/ were determined as a function of the temperature at the pyrometer wavelengths of 752 and 1064 nm

  2. 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)

  3. Startup transient simulation for natural circulation boiling water reactors in PUMA facility

    International Nuclear Information System (INIS)

    Kuran, S.; Xu, Y.; Sun, X.; Cheng, L.; Yoon, H.J.; Revankar, S.T.; Ishii, M.; Wang, W.

    2006-01-01

    In view of the importance of instabilities that may occur at low-pressure and -flow conditions during the startup of natural circulation boiling water reactors, startup simulation experiments were performed in the Purdue University Multi-Dimensional Integral Test Assembly (PUMA) facility. The simulations used pressure scaling and followed the startup procedure of a typical natural circulation boiling water reactor. Two simulation experiments were performed for the reactor dome pressures ranging from 55 kPa to 1 MPa, where the instabilities may occur. The experimental results show the signature of condensation-induced oscillations during the single-phase-to-two-phase natural circulation transition. The results also suggest that a rational startup procedure is needed to overcome the startup instabilities in natural circulation boiling water reactor designs

  4. Boiling enriches the linear polysulfides and the hydrogen sulfide-releasing activity of garlic

    NARCIS (Netherlands)

    Tocmo, Restituto; Wu, Yuchen; Liang, Dong; Fogliano, Vincenzo; Huang, Dejian

    2017-01-01

    Garlic is rich in polysulfides, and some of them can be H2S donors. This study was conducted to explore the effect of cooking on garlic's organopolysulfides and H2S-releasing activity. Garlic bulbs were crushed and boiled for a period ranging from 3 to 30 min and the solvent extracts were analyzed

  5. Pool boiling performance of NovecTM 649 engineered fluid

    International Nuclear Information System (INIS)

    Forrest, Eric; Buongiorno, Jacopo; McKrell, Thomas; Hu, Lin-Wen

    2009-01-01

    A new fluorinated ketone, C 2 F 5 C(O)CF(CF 3 ) 2 , is currently being considered as an environmentally friendly alternative for power electronics cooling applications due to its high dielectric strength and low global warming potential (GWP). Sold commercially by the 3M Company as Novec TM 649 Engineered Fluid, C 2 F 5 C(O)CF(CF 3 ) 2 exhibits very low acute toxicity while maintaining long-term stability. To assess the general two-phase heat transfer performance of Novec TM 649, pool boiling tests were conducted by resistively heating a 0.01 in. diameter nickel wire at the fluid's atmospheric saturation temperature of 49 deg C. The nucleate boiling heat transfer coefficient and critical heat flux (CHF) obtained for the fluorinated ketone compare favorably with results obtained for FC-72, a fluorocarbon widely used for the direct cooling of electronic devices. Initial results indicate that Novec TM 649 may prove to be a viable alternative to FC-72 and other halo alkanes for the cooling of high power density electronic devices. (author)

  6. Water inventory management in condenser pool of boiling water reactor

    International Nuclear Information System (INIS)

    Gluntz, D.M.

    1996-01-01

    An improved system for managing the water inventory in the condenser pool of a boiling water reactor has means for raising the level of the upper surface of the condenser pool water without adding water to the isolation pool. A tank filled with water is installed in a chamber of the condenser pool. The water-filled tank contains one or more holes or openings at its lowermost periphery and is connected via piping and a passive-type valve (e.g., squib valve) to a high-pressure gas-charged pneumatic tank of appropriate volume. The valve is normally closed, but can be opened at an appropriate time following a loss-of-coolant accident. When the valve opens, high-pressure gas inside the pneumatic tank is released to flow passively through the piping to pressurize the interior of the water-filled tank. In so doing, the initial water contents of the tank are expelled through the openings, causing the water level in the condenser pool to rise. This increases the volume of water available to be boiled off by heat conducted from the passive containment cooling heat exchangers. 4 figs

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

    CERN Document Server

    Alam, Tamanna; Jin, Li-Wen

    2013-01-01

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

  8. Suppression of saturated nucleate boiling by forced convective flow

    International Nuclear Information System (INIS)

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

    1980-01-01

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

  9. Boiling point of volatile liquids at various pressures

    Directory of Open Access Journals (Sweden)

    Luisa Maria Valencia

    2017-07-01

    Full Text Available Water, under normal conditions, tends to boil at a “normal boiling temperature” at which the atmospheric pressure fixes the average amount of kinetic energy needed to reach its boiling point. Yet, the normal boiling temperature of different substances varies depending on their nature, for which substances like alcohols, known as volatile, boil faster than water under same conditions. In response to this phenomenon, an investigation on the coexistence of both gas and liquid phases of a volatile substance in a closed system was made, establishing vapor pressure as the determining tendency of a substance to vaporize, which increases exponentially with temperature until a critical point is reached. Since atmospheric pressure is fixed, the internal pressure of the system was varied to determine its relationship with vapor pressure and thus with the boiling point of the substance, concluding that the internal pressure and boiling point of a volatile liquid in a closed system are negatively proportional.

  10. The mechanism of heat transfer in transition boiling

    International Nuclear Information System (INIS)

    Chin Pan; Hwang, J.Y.; Lin, T.L.

    1989-01-01

    Liquid-solid contact in transition boiling is modelled by involving transient conduction, boiling incipience, macrolayer evaporation and vapour film boiling. The prediction of liquid contact duration and time fraction agrees reasonably well with experimental data, and the model is able to predict both of the boiling curve transitions - the critical and minimum heat fluxes. The study concludes that the liquid turbulence due to buoyancy forces and bubble agitation is an important parameter for transition boiling. It is found that surface coating (oxidation or deposition) tends to improve the transition boiling heat transfer and elevate the wall superheats at both the critical heat flux and the minimum film boiling points, which agrees with the experimental observations. (author)

  11. 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

  12. 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)

  13. 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)

  14. 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

  15. 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

  16. Investigation on Minimum Film Boiling Point of Highly Heated Vertical Metal Rod in Aqueous Surfactant Solution

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chi Young; Kim, Jae Han [Pukyong Nat’l Univ., Busan (Korea, Republic of)

    2017-09-15

    In this study, experiments were conducted on the MFB(minimum film boiling) point of highly heated vertical metal rod quenched in aqueous surfactant solution at various temperature conditions. The aqueous Triton X-100 solution(100 wppm) and pure water were used as the liquid pool. Their temperatures ranged from 77 °C to 100 °C. A stainless steel vertical rod of initial center temperature of 500 °C was used as a test specimen. In both liquid pools, as the liquid temperature decreased, the time to reach the MFB point decreased with a parallel increase in the temperature and heat flux of the MFB point. However, over the whole present temperature range, in the aqueous Triton X-100 solution, the time to reach the MFB point was longer, while the temperature and heat flux of the MFB point were reduced when compared with pure water. Based on the present experimental data, this study proposed the empirical correlations to predict the MFB temperature of a high temperature vertical metal rod in pure water and in aqueous Triton X-100 solution.

  17. Compatibility of refractory materials with boiling sodium

    International Nuclear Information System (INIS)

    Meacham, S.A.

    1976-01-01

    The program employed to determine the compatibility of commercially available refractories with boiling sodium is described. The effects of impurities contained within the refractory material, and their relations with the refractory's physical stability are discussed. Also, since consideration of refractories for use as an insulating material within Liquid Metal Fast Breeder Reactor Plants (LMFBR's) is currently under investigation; recommendations, based upon this program, are presented

  18. The fluid similarity of the boiling crisis

    International Nuclear Information System (INIS)

    Katsaounis, A.

    1986-01-01

    Most of the measurements related to the boiling crisis have, until now, been undertaken for a wide parameter variation in the water, and were mainly related to the water-cooled reactor. This article investigates, whether or how the measuring results can be transferred to other fluids. Derived dimensionless similarity figures and those taken from literature are verified by measurements from complex geometries in water and freon 12. (orig.) [de

  19. The fluid similarity of the boiling crisis

    International Nuclear Information System (INIS)

    Katsaounis, A.

    1987-01-01

    Most of the measurements related to the boiling crisis have, until now, been undertaken for a wide parameter variation in the water, and were mainly related to the water-cooled reactor. This article investigates, whether or how the measuring results can be transferred to other fluids. Derived dimensionless similarity figures and those taken from literature are verified by measurements from complex geometries in water and freon 12. (orig./GL) [de

  20. 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)

  1. 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

  2. Full evaporation headspace gas chromatography for sensitive determination of high boiling point volatile organic compounds in low boiling matrices.

    Science.gov (United States)

    Mana Kialengila, Didi; Wolfs, Kris; Bugalama, John; Van Schepdael, Ann; Adams, Erwin

    2013-11-08

    Determination of volatile organic components (VOC's) is often done by static headspace gas chromatography as this technique is very robust and combines easy sample preparation with good selectivity and low detection limits. This technique is used nowadays in different applications which have in common that they have a dirty matrix which would be problematic in direct injection approaches. Headspace by nature favors the most volatile compounds, avoiding the less volatile to reach the injector and column. As a consequence, determination of a high boiling solvent in a lower boiling matrix becomes challenging. Determination of VOCs like: xylenes, cumene, N,N-dimethylformamide (DMF), dimethyl sulfoxide (DMSO), N,N-dimethylacetamide (DMA), N-methyl-2-pyrrolidone (NMP), 1,3-dimethyl-2-imidazolidinone (DMI), benzyl alcohol (BA) and anisole in water or water soluble products are an interesting example of the arising problems. In this work, a headspace variant called full evaporation technique is worked out and validated for the mentioned solvents. Detection limits below 0.1 μg/vial are reached with RSD values below 10%. Mean recovery values ranged from 92.5 to 110%. The optimized method was applied to determine residual DMSO in a water based cell culture and DMSO and DMA in tetracycline hydrochloride (a water soluble sample). Copyright © 2013 Elsevier B.V. All rights reserved.

  3. Boiling anomaly detection by various signal characterization methods

    International Nuclear Information System (INIS)

    Sakuma, M.; Kozma, R.; Kitamura, M.; Schoonewelle, H.; Hoogenboom, J.E.

    1996-01-01

    In order to detect anomalies in the early stage for complex dynamical systems like nuclear power plants, it is important to characterize various statistical features of the data acquired in normal operating condition. In this paper, concept of hierarchical anomaly monitoring method is outlined, which is based on the diversification principle. In addition to usual time and frequency domain analysis (FFT, APDF, MAR-SPRT), other analysis (wavelet, fractal, etc.) are performed. As soon as any inconsistency arises in the results of the analysis on the upper level, a thorough analysis is initiated. A comparison among these methods is performed and the efficiency of the diversification approach has been demonstrated through simulated boiling anomalies in nuclear reactors. (authors)

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  5. 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)

  6. Heterogeneous associations between smoking and a wide range of initial presentations of cardiovascular disease in 1937360 people in England: lifetime risks and implications for risk prediction.

    Science.gov (United States)

    Pujades-Rodriguez, Mar; George, Julie; Shah, Anoop Dinesh; Rapsomaniki, Eleni; Denaxas, Spiros; West, Robert; Smeeth, Liam; Timmis, Adam; Hemingway, Harry

    2015-02-01

    It is not known how smoking affects the initial presentation of a wide range of chronic and acute cardiovascular diseases (CVDs), nor the extent to which associations are heterogeneous. We estimated the lifetime cumulative incidence of 12 CVD presentations, and examined associations with smoking and smoking cessation. Cohort study of 1.93 million people aged ≥30years, with no history of CVD, in 1997-2010. Individuals were drawn from linked electronic health records in England, covering primary care, hospitalizations, myocardial infarction (MI) registry and cause-specific mortality (the CALIBER programme). During 11.6 million person-years of follow-up, 114859 people had an initial non-fatal or fatal CVD presentation. By age 90 years, current vs never smokers' lifetime risks varied from 0.4% vs 0.2% for subarachnoid haemorrhage (SAH), to 8.9% vs 2.6% for peripheral arterial disease (PAD). Current smoking showed no association with cardiac arrest or sudden cardiac death [hazard ratio (HR)=1.04, 95% confidence interval (CI) 0.91-1.19).The strength of association differed markedly according to disease type: stable angina (HR=1.08, 95% CI 1.01-1.15),transient ischaemic attack (HR=1.41, 95% CI 1.28-1.55), unstable angina (HR=1.54, 95% CI 1.38-1.72), intracerebral haemorrhage (HR=1.61, 95% CI 1.37-1.89), heart failure (HR=1.62, 95% CI 1.47-1.79), ischaemic stroke (HR=1.90, 95% CI 1.72-2.10), MI (HR=2.32, 95% CI 2.20-2.45), SAH (HR= 2.70, 95% CI 2.27-3.21), PAD (HR=5.16, 95% CI 4.80-5.54) and abdominal aortic aneurysm (AAA) (HR=5.18, 95% CI 4.61-5.82). Population-attributable fractions were lower for women than men for unheralded coronary death, ischaemic stroke, PAD and AAA. Ten years after quitting smoking, the risks of PAD, AAA (in men) and unheralded coronary death remained increased (HR=1.36, 1.47 and 2.74, respectively). The heterogeneous associations of smoking with different CVD presentations suggests different underlying mechanisms and have important

  7. 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

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  9. The effect of nozzle collar on signle phase and boiling heat transfer by planar impinging jet

    International Nuclear Information System (INIS)

    Shin, Chang Hwan; Yim, Seong Hwan; Cho, Hyung Hee; Wu, Seong Je

    2005-01-01

    The water jet impingement cooling is one of the techniques to remove the heat from high heat flux equipment. Local heat transfer of the confined water impinging jet and the effect of nozzle collar to enhance the heat transfer are investigated in the free surface jet and submerged jet. Boiling is initiated from the farthest downstream and increase of the wall temperature is reduced with developing boiling, forming the flat temperature distributions. The reduction in the nozzle-to-surface distance for H/W≤1 causes significant increases and distribution changes of heat transfer. Developed boiling reduces the differences of heat transfer for various conditions. The nozzle collar is employed at the nozzle exit. The distances from heated surface to nozzle collar, H c are 0.25W, 0.5W and 1.0W. The liquid film thickness is reduced and the velocity of wall jet increases as decreased spacing of collar to heated surface. Heat transfer is enhanced for region from the stagnation to x/W∼8 in the free surface jet and to x/W∼5 in the submerged jet. For nucleate boiling region of further downstream, the heat transfer by the nozzle collar is decreased in submerged jet comparing with higher velocity condition. It is because the increased velocity by collar is de-accelerated downstream

  10. Interactions between bubble formation and heating surface in nucleate boiling

    International Nuclear Information System (INIS)

    Luke, Andrea

    2009-01-01

    The heat transfer and bubble formation is investigated in pool boiling of propane. Size distributions of active nucleation sites on single horizontal copper and steel tubes with different diameter and surface finishes have been calculated from heat transfer measurements over wide ranges of heat flux and selected pressure. The model assumptions of Luke and Gorenflo for the heat transfer near growing and departing bubbles, which were applied in the calculations, have been slightly modified and the calculated results have been compared to experimental investigations by high speed video techniques. The calculated number of active sites shows a good coincidence for the tube with smaller diameter, while the results for the tube with larger diameter describe the same relative increase of the active sites. The comparison of the cumulative size distribution of the active and potential nucleation sites demonstrates the same slope of the curve and that the critical radius of a stable bubble nuclei is smaller than the average cavity size. (author)

  11. 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.

  12. Interactions between bubble formation and heating surface in nucleate boiling

    Energy Technology Data Exchange (ETDEWEB)

    Luke, Andrea [Leibniz University, Hannover (Denmark). Inst. of Thermodynamics], e-mail: ift@ift.uni-hannover.de

    2009-07-01

    The heat transfer and bubble formation is investigated in pool boiling of propane. Size distributions of active nucleation sites on single horizontal copper and steel tubes with different diameter and surface finishes have been calculated from heat transfer measurements over wide ranges of heat flux and selected pressure. The model assumptions of Luke and Gorenflo for the heat transfer near growing and departing bubbles, which were applied in the calculations, have been slightly modified and the calculated results have been compared to experimental investigations by high speed video techniques. The calculated number of active sites shows a good coincidence for the tube with smaller diameter, while the results for the tube with larger diameter describe the same relative increase of the active sites. The comparison of the cumulative size distribution of the active and potential nucleation sites demonstrates the same slope of the curve and that the critical radius of a stable bubble nuclei is smaller than the average cavity size. (author)

  13. Research progresses and future directions on pool boiling heat transfer

    OpenAIRE

    M. Kumar; V. Bhutani; P. Khatak

    2015-01-01

    This paper reviews the previous work carried on pool boiling heat transfer during heating of various liquids and commodities categorized as refrigerants and dielectric fluids, pure liquids, nanofluids, hydrocarbons and additive mixtures, as well as natural and synthetic colloidal solutions. Nucleate pool boiling is an efficient and effective method of boiling because high heat fluxes are possible with moderate temperature differences. It is characterized by the growth of bubbles on a heated s...

  14. Downflow film boiling in a rod bundle at low pressure

    International Nuclear Information System (INIS)

    Hochreiter, L.E.; Rosal, E.R.; Fayfich, R.R.

    1978-01-01

    A series of low pressure downflow film boiling heat transfer experiments were conducted in a 14-foot (4.27 m) long electrically heater rod bundle containing 336 heater rods. The resulting data was compared with the Dougall-Rohsenow dispersed flow film boiling correlation. The data was found to lie below this correlation as the quality was increased. It is believed that buoyancy effects decreased the heat transfer in downflow film boiling. (author)

  15. 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

  16. Water Boiling inside Carbon Nanotubes: Towards Efficient Drug Release

    OpenAIRE

    Chaban, Vitaly V.; Prezhdo, Oleg V.

    2012-01-01

    We show using molecular dynamics simulation that spatial confinement of water inside carbon nanotubes (CNT) substantially increases its boiling temperature and that a small temperature growth above the boiling point dramatically raises the inside pressure. Capillary theory successfully predicts the boiling point elevation down to 2 nm, below which large deviations between the theory and atomistic simulation take place. Water behaves qualitatively different inside narrow CNTs, exhibiting trans...

  17. Dynamic behaviour of bubbles of water vapour at a temperature lower than the boiling temperature

    International Nuclear Information System (INIS)

    Jansen, Franz

    1966-01-01

    This research thesis reports the study of the theoretical movement of the wall of vapour water bubbles in a sub-saturated boiling regime, i.e. with an average water temperature lower than the boiling temperature. While assuming that bubbles have an initial translational speed at the beginning of their condensation, the author shows that their shrinkage should result in an accelerated displacement in a direction normal to the wall and inward the liquid. Layers of hot water initially close to the wall would therefore be quickly transported towards cold water areas. Experiments allowed, in some cases, the acceleration of bubbles during their condensation to be noticed: for low sub-saturations in still water and for high sub-saturations in water in forced convection, even though, in this last case, the determination of accelerations is more delicate [fr

  18. Flow dynamics of volume-heated boiling pools

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  19. 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)

  20. 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

  1. Heat transfer coefficient for boiling carbon dioxide

    DEFF Research Database (Denmark)

    Knudsen, Hans Jørgen Høgaard; Jensen, Per Henrik

    1998-01-01

    Heat transfer coefficient and pressure drop for boiling carbon dioxide (R744) flowing in a horizontal pipe has been measured. The calculated heat transfer coeeficient has been compared with the Chart correlation of Shah. The Chart Correlation predits too low heat transfer coefficient but the ratio...... between the measured and the calculated heat transfer coefficient is nearly constant and equal 1.9. With this factor the correlation predicts the measured data within 14% (RMS). The pressure drop is of the same order as the measuring uncertainty and the pressure drop has not been compared with correlation's....

  2. Boiling, condensation, and gas-liquid flow

    International Nuclear Information System (INIS)

    Whalley, P.B.

    1987-01-01

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

  3. Sampling system for a boiling reactor NPP

    International Nuclear Information System (INIS)

    Zabelin, A.I.; Yakovleva, E.D.; Solov'ev, Yu.A.

    1976-01-01

    Investigations and pilot running of the nuclear power plant with a VK-50 boiling reactor reveal the necessity of normalizing the design system of water sampling and of mandatory replacement of the needle-type throttle device by a helical one. A method for designing a helical throttle device has been worked out. The quantitative characteristics of depositions of corrosion products along the line of reactor water sampling are presented. Recommendations are given on the organizaton of the sampling system of a nuclear power plant with BWR type reactors

  4. Pool boiling visualization on open microchannel surfaces

    Directory of Open Access Journals (Sweden)

    Kaniowski Robert

    2017-01-01

    Full Text Available The paper presents visualization investigations into pool boiling heat transfer for open minichannel surfaces. The experiments were carried out wih saturated water at atmospheric pressure. Parallel microchannels fabricated by machining were about 0.3 mm wide and 0.2 to 0.4 mm deep. High-speed videos were used as an aid to understanding the heat transfer mechanism. The visualization study aimed at identifying nucleation sites of the departing bubbles and determining their diameters and frequency at various superheats.

  5. 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)

  6. One component, volume heated, boiling pool thermohydraulics

    International Nuclear Information System (INIS)

    Bede, M.; Perret, C.; Pretrel, H.; Seiler, J.M.

    1993-01-01

    Prior work on boiling pools provided heat exchange correlations valid for bubbly flow with laminar or turbulent boundary layers. New experiments performed with water (SEBULON) and UO 2 (SCARABEE BF2) in a churn-turbulent flow configuration show unexpected heat flux distributions for which the maximum heat flux may be situated well below the pool surface. The origin of this behaviour is attributed to condensation effects, very unstable boundary layer flow and surface oscillation. A calculation model is discussed which permits to approach the experimental heat flux distribution with reasonable accuracy. (authors). 7 figs., 2 appendix., 14 refs

  7. Experimental study and technique for calculation of critical heat fluxes in helium boiling in tubes

    International Nuclear Information System (INIS)

    Arkhipov, V.V.; Kvasnyuk, S.V.; Deev, V.I.; Andreev, V.K.

    1979-01-01

    Studied is the effect of regime parameters on critical heat loads in helium boiling in a vertical tube in the range of mass rates of 80 2 xc) and pressures of 100<=p<=200 kPa for the vapor content range corresponding to the heat exchange crisis of the first kind. The method for calculating critical heat fluxes describing experimental data with the error less than +-15% is proposed. The critical heat loads in helium boiling in tubes reduce with the growth of pressure and vapor content in the regime parameter ranges under investigation. Both positive and negative effects of the mass rate on the critical heat flux are observed. The calculation method proposed satisfactorily describes the experimental data

  8. Outline of the advanced boiling water reactor (ABWR)

    International Nuclear Information System (INIS)

    Hucik, S.A.; Imaoka, T.; Minematsu, A.; Takashima, Y.

    1986-01-01

    The fundamental design of the Advanced Boiling Water Reactor (ABWR) was completed in December 1985. This design represents the next generation of Boiling Water Reactors (BWR) to be introduced into commercial operation in the 1990s. The ABWR is the result of the continuing evolution of the BWR, incorporating state-of-the-art technologies and many new improvements based on an extensive accumulation of world-wide experience through design, construction and operation of BWRs. The ABWR development program was initiated in 1978, with subsequent design and test and development programs started in 1981. Most of the development and verification tests of the new features have been completed. The ABWR development objective focused on an optimized selection of advanced technologies and proven BWR technologies. The ABWR objectives were specific improvements such as operating and safety margins, enhanced availability and capacity factor, and reduced occupational exposure while at the same time achieving significant cost reduction in both capital and operating costs. The ABWR is characterized by an improved NSSS including ten internal recirculation pumps, fine motion electric-hydraulic control rod drives, optimized safety and auxiliary systems, advanced control and instrumentation systems, improved turbine-generator with moisture/separator reheater with plant output increased to 1350 MWe, and an integrated reinforced concrete containment vessel and compact Reactor and Turbine Building design. The turbine system also included improvements in the Turbine-Generator, feedwater/heater system, and condensate treatment systems. The radwaste system was also optimized taking advantage of the plant design improvements and advances in radwaste technology. The ABWR is a truly optimal design which utilizes advanced technologies, capabilities, performance improvements, and yet provides an economic advantage. (author)

  9. 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

  10. 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

  11. Influence of oxide inclusions in marten bath on the quality of metal during boiling

    Energy Technology Data Exchange (ETDEWEB)

    Elanskii, G N; Kudrin, V A; Ageev, E E; Babiskiirh, V K [Moskovskii Vechernij Metallurgicheskij Inst. (USSR)

    1975-01-01

    The effect of oxide inclusions floating in a boiling open-hearth furnace on the quality of steel produced has been determined. During the boiling the metal contains some nonmetallic inclusions. They cannot be completely separated during the boiling period and therefore adversely affect the purity of steel and its mechanical properties. The maximum effect on the mechanical properties (especially on plasticity and impact strenght) is due to relatively large oxide inclusions (greater than 3-3.5 on the scale). The degree of contamination by oxides has been determined for the relative ratio (in %) of polished sections containing oxide inclusions (o.i.). One of the factors determining (%)sub(o.i.) is the rate of carbon oxidation. In the case where the smelting process is carried out under optimal conditions for carbon oxidation, the produced metal is least contaminated with the undesirable large oxide inclusions. Deviation from the optimal range of the oxidation rate during boiling of the steel results in a significant increase in the proportion of smeltings contaminated with the oxides.

  12. 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.

  13. Fouling of Structured Surfaces during Pool Boiling of Aqueous Solutions

    International Nuclear Information System (INIS)

    Esawy, M.

    2011-01-01

    Bubble characteristics in terms of density, size, frequency and motion are key factors that contribute to the superiority of nucleate pool boiling over the other modes of heat transfer. Nevertheless, if heat transfer occurs in an environment which is prone to fouling, the very same parameters may lead to accelerated deposit formation due to concentration effects beneath the growing bubbles. This has led heat exchanger designers frequently to maintain the surface temperature below the boiling point if fouling occurs, e.g. in thermal seawater desalination plants. The present study investigates the crystallization fouling of various structured surfaces during nucleate pool boiling of CaSO 4 solutions to shed light into their fouling behaviour compared with that of plain surfaces for the same operating conditions. As for the experimental part, a comprehensive set of clean and fouling experiments was performed rigorously. The structured tubes included low finned tubes of different fin densities, heights and materials and re-entrant cavity Turbo-B tube types.The fouling experiments were carried out at atmospheric pressure for different heat fluxes ranging from 100 to 300 k W/m 2 and CaSO 4 concentrations of 1.2 and 1.6 g/L. For the sake of comparison, similar runs were performed on plain stainless steel and copper tubes.Overall for the finned tubes, the experimental results showed a significant reduction of fouling resistances of up to 95% compared to those of the stainless steel and copper plain tubes. In addition, the scale formation that occurred on finned tubes was primarily a scattered and thin crystalline layer which differs significantly from those of plain tubes which suffered from a thick and homogenous layer of deposit with strong adhesion. Higher fin densities and lower fin heights always led to better antifouling performance for all investigated finned tubes. It was also shown that the surface material strongly affects the scale formation of finned tubes i

  14. Interaction of centres on nucleate boiling

    International Nuclear Information System (INIS)

    Chekanov, V.V.

    1977-01-01

    An experimental set-up is described which enables to analyse the growth of a bubble on a heater, the frequency of succession of the bubbles, etc., by the change in the electric signal from a photomultiplier, onto whose photocathode the bubble is projected. The change in the first moments of the statistical distribution of the corresponding parameters (growth time, frequency of succession, and so on) is adopted as the measure of the external effect on the vapour formation centre. It is shown that for single-bubble boiling the greatest effect is exerted by the acoustic waves produced by the bubble in the growth period; during developed boiling one observes mutual suppression of centres spaced at a distance of the order of the detachment diameter. As the heat flux increases, the correlation of the motion of the interface over the heater surface increases as well. When the correlation radius becomes equal to the centre-to-centre distance, the first crisis sets in. It is suggested that heaters with a variable coefficient of temperature conductance along the heat-releasing surface must withstand high subcritical heat fluxes

  15. 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.

  16. Boiling heat transfer on horizontal tube bundles

    International Nuclear Information System (INIS)

    Anon.

    1987-01-01

    Nucleate boiling heat transfer characteristics for a tube in a bundle differ from that for a single tube in a pool and this difference is known as 'tube bundle effect.' There exist two bundle effects, positive and negative. The positive bundle effect enhances heat transfer due to convective flow induced by rising bubbles generated from the lower tubes, while the negative bundle effect deteriorates heat transfer due to vapor blanketing caused by accumulation of bubbles. Staggered tube bundles tested and found that the upper tubes in bundles have higher heat transfer coefficients than the lower tubes. The effects of various parameters such as pressure, tube geometry and oil contamination on heat transfer have been examined. Some workers attempted to clarify the mechanism of occurrence of 'bundle effect' by testing tube arrangements of small scale. All reported only enhancement in heat transfer but results showed the symptom of heat transfer deterioration at higher heat fluxes. As mentioned above, it has not been clarified so far even whether the 'tube bundle effect' should serve as enhancement or deterioration of heat transfer in nucleate boiling. In this study, experiments are performed in detail by using bundles of small scale, and effects of heat flux distribution, pressure and tube location are clarified. Furthermore, some consideration on the mechanisms of occurrence of 'tube bundle effect' is made and a method for prediction of heat transfer rate is proposed

  17. Identification of dynamic basins in boiling fluxes

    International Nuclear Information System (INIS)

    Juanico, L.E.

    1997-01-01

    A theoretical and experimental study of the dynamic behavior of a boiling channel is presented. In particular, the existence of different basins of attraction during instabilities was established. A fully analytical treatment of boiling channel dynamics were performed using a algebraic delay model. Subcritical and supercritical Hopf bifurcations could be identified and analyzed using perturbation methods. The derivation of a fully analytical criterion for Hopf bifurcation transcription was applied to determine the amplitude of the limit cycles and the maximum allowed perturbations necessary to break the system stability. A lumped parameters model which allows the representation of flow reversal is presented. The dynamic of very large amplitude oscillations, out of the Hopf bifurcation domain, was studied. The analysis revealed the existence of new dynamical basins of attraction, where the system may evolve to and return from with hysteresis. Finally, an experimental study was conducted, in a water loop at atmospheric pressure, designed to reproduce the operating conditions analyzed in the theory. Different dynamic phase previously predicted in the theory were found and their nonlinear characteristics were studied. In particular, subcritical and supercritical Hopf bifurcations and very large amplitude oscillations with flow reversal were identified. (author). 53 refs., figs

  18. 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)

  19. 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

  20. The mineral products of boiling in the Golden Cross epithermal deposit

    International Nuclear Information System (INIS)

    Simmons, S.F.; Mauk, J.L.; Simpson, M.P.

    2000-01-01

    The Golden Cross low sulfidation epithermal deposit shows a number of features that are directly or indirectly related to boiling hydroythermal fluids. Occurrences of lattice calcite and their quartz pseudomorph equivalents in veins, and occurrences of adularia in veins and in the surrounding altered rocks in the vicinity of ore, are direct evidence of deposition in the presence of boiling hydrothermal fluids. Loss of carbon dioxide causes calcite deposition (platy variety) near the level of first boiling, while adularia deposits due to the attendant pH increase and cooling. Indirect evidence of boiling includes crustiform-colloform quartz banding, late massive calcite veins, and clay-carbonate alteration in the shallow and peripheral parts of the ore zone. The colloform quartz banding strongly resembles the banding in amorophous silica deposits found in geothermal pipes. This implies that fluids ascending the Empire vein structure were saturated in amorphous silica. If so, then they must have undergone phase separation, which initiated at considerable depth (e.g. > or =1000 m) and very hot temperatures (e.g. > or =300 degrees C). On the basis of stable isotope data, late massive veins appear to have deposited from CO 2 -rich steam-heated waters. Calcite deposited along heating paths as these waters descended into the upflow zone during late stage collapse of the hydrothermal plume. In active systems, such steam heated waters form by deep boiling. The high CO 2 contents of these waters promote hydrolytic alteration and the formation of clay-carbonate alteration. (author). 37 refs., 4 figs

  1. Experimental Investigation of Pool Boiling Heat Transfer Enhancement in Microgravity in the Presence of Electric Fields

    Science.gov (United States)

    Herman, Cila

    1996-01-01

    Boiling is an effective mode of heat transfer since high heat flux levels are possible driven by relatively small temperature differences. The high heat transfer coefficients associated with boiling have made the use of these processes increasingly attractive to aerospace engineering. Applications of this type include compact evaporators in the thermal control of aircraft avionics and spacecraft environments, heat pipes, and use of boiling to cool electronic equipment. In spite of its efficiency, cooling based on liquid-vapor phase change processes has not yet found wide application in aerospace engineering due to specific problems associated with the low gravity environment. After a heated surface has reached the superheat required for the initiation of nucleate boiling, bubbles will start forming at nucleation sites along the solid interface by evaporation of the liquid. Bubbles in contact with the wall will continue growing by this mechanism until they detach. In terrestrial conditions, bubble detachment is determined by the competition between body forces (e.g. buoyancy) and surface tension forces that act to anchor the bubble along the three phase contact line. For a given body force potential and a balance of tensions along the three phase contact line, bubbles must reach a critical size before the body force can cause them to detach from the wall. In a low gravity environment the critical bubble size for detachment is much larger than under terrestrial conditions, since buoyancy is a less effective means of bubble removal. Active techniques of heat transfer enhancement in single phase and phase change processes by utilizing electric fields have been the subject of intensive research during recent years. The field of electrohydrodynamics (EHD) deals with the interactions between electric fields, flow fields and temperature fields. Previous studies indicate that in terrestrial applications nucleate boiling heat transfer can be increased by a factor of 50 as

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-12-15

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

  3. A New Computational Tool for Simulation of 3-D Flow and Heat Transfer in Boiling Water Reactors

    International Nuclear Information System (INIS)

    Chen, Hudong

    2002-01-01

    This Phase I work has developed a novel hybrid Lattice Boltzmann Model for the simulation of nonideal fluid thermal dynamics and demonstrated that this model can be used to simulate fundamental two-phase flow processes including boiling initiation, bubble formation and coalescency, and flow-regime formation

  4. 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.

  5. Effect of boiling on the physicochemical properties of Roselle seeds ...

    African Journals Online (AJOL)

    Effect of boiling on the physicochemical composition of Roselle seeds (Hibiscus sabdariffa) grown in Mali was shown. Proximate analysis indicated that boiled whole Roselle seeds (BWRS) are potential high protein source. Moreover, the results of lipid analysis indicated that the seeds were good source of unsaturated fatty ...

  6. Results of KNS-1 acoustic boiling noise benchmark test

    International Nuclear Information System (INIS)

    Foerster, K.; Arnaoutis, N.; Voss, J.

    1989-01-01

    The present paper deals with investigations of acoustic signals from a boiling experiment performed on the KNS I loop at KfK Karlsruhe. Signals have been analysed in frequency as well as in time domain. Signal characteristics successfully used to detect the boiling process have been found in time domain. (author). 13 figs, 1 tab

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

    DEFF Research Database (Denmark)

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

    1976-01-01

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

  8. Critical heat flux in flow boiling in microchannels

    CERN Document Server

    Saha, Sujoy Kumar

    2015-01-01

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

  9. 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)

  10. Chemical composition and nutritional value of boiled Christmas ...

    African Journals Online (AJOL)

    A study was conducted to determine the chemical composition and the nutritive value of boiled Christmas bush (Alchornea cordifolia) for starter broiler chickens. Dried Christmas bush fruits (Capsules + seed) were boiled for 30 minutes, sundried and ground into meal. The meal was analyzed for proximate composition and ...

  11. Little low-power boiling never hurt anybody

    International Nuclear Information System (INIS)

    Dunn, F.E.

    1985-01-01

    Failures in the shutdown heat removal system of an LMFBR might lead to flow stagnation and coolant boiling in the reactor core. At normal operating power, the onset of sodium boiling will lead to film dryout and melting of the cladding and fuel within a few seconds. On the other hand, both calculations and currently available experimental data indicate that at heat fluxes corresponding to decay heat power levels, boiling leads to improved heat removal; and it limits the temperature rise in the fuel pins. Therefore, when setting safety criteria for decay heat removal systems, there is no reason to preclude sodium boiling per se because of heat removal considerations. As an example that illustrates the beneficial impact of coolant boiling, a case involving temporary loss of feedwater and staggered pump failures in a hypothetical, 1000-MWe loop-type reactor was run in the SASSYS-1 code

  12. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-15

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

  14. Changes provoked by boiling, steaming and sous-vide cooking in the lipid and volatile profile of European sea bass.

    Science.gov (United States)

    Nieva-Echevarría, Bárbara; Manzanos, María J; Goicoechea, Encarnación; Guillén, María D

    2017-09-01

    This study aims to shed light on the changes provoked by boiling, steaming and sous-vide cooking on the lipids and volatile profile of farmed and wild European sea bass meat. None of the cooking techniques provoked changes due to hydrolytic or oxidation processes detectable by 1 H NMR on sea bass lipids. The lipid profile of main and minor lipidic components was maintained after cooking. However, study by SPME-GC/MS evidenced that steaming and sous-vide cooking modified the volatile profile of sea bass meat, especially in farmed specimens. The compounds generated came from the occurrence, to a very small extent, of lipid and protein degradation. By contrast, boiling scarcely modified the initial characteristics of raw sea bass. Thus, from a sensory point of view and considering the odour-active compounds generated, steaming and sous-vide cooking provoked more noticeable changes than boiling, especially in farmed sea bass meat. Copyright © 2017. Published by Elsevier Ltd.

  15. The initial safe range of motion of the ankle joint after three methods of internal fixation of simulated fractures of the medial malleolus.

    Science.gov (United States)

    Shimamura, Yoshio; Kaneko, Kazuo; Kume, Kazuhiko; Maeda, Mutsuhiro; Iwase, Hideaki

    2006-07-01

    Previous studies have demonstrated the safe passive range of ankle motion for inter-bone stiffness after internal fixation under load but there is a lack of information about the safe range of ankle motion for early rehabilitation in the absence of loading. The present study was designed to assess the effect of ankle movement on inter-bone displacement characteristics of medial malleolus fractures following three types of internal fixation to determine the safe range of motion. Five lower legs obtained during autopsy were used to assess three types of internal fixation (two with Kirschner-wires alone; two with Kirschner-wires plus tension band wiring; and, one with an AO/ASIF malleolar screw alone). Following a simulated fracture by sawing through the medial malleolus the displacement between the fractured bone ends was measured during a passive range of movement with continuous monitoring using omega (Omega) shaped transducers and a biaxial flexible goniometer. Statistical analysis was performed with repeated measures analysis of variance. Inter-bone displacement was not proportional to the magnitude of movement throughout the range of ankle motion as, when separation exceeded 25 microm, there was increasingly wide separation as plantar-flexion or dorsal-flexion was increased. There was no statistical significant difference between the small amount of inter-bone displacement observed with three types of fixation within the safe range of dorsal-flexion and plantar-flexion for early rehabilitation. However the inter-bone separation when fixation utilized two Kirschner-wires alone tended to be greater than when using the other two types of fixation during dorsal-flexion and eversion. The present study revealed a reproducible range of ankle motion for early rehabilitation which was estimated to be within the range of 20 degrees of dorsal-flexion and 10 degrees of plantar-flexion without eversion. Also, internal fixation with two Kirschner-wires alone does not seem to

  16. Non-isothermal desorption and nucleate boiling in a water-salt droplet LiBr

    Directory of Open Access Journals (Sweden)

    Misyura Sergey Ya.

    2018-01-01

    Full Text Available Experimental data on desorption and nucleate boiling in a droplet of LiBr-water solution were obtained. An increase in salt concentration in a liquid-layer leads to a considerable decrease in the rate of desorption. The significant decrease in desorption intensity with a rise of initial mass concentration of salt has been observed. Evaporation rate of distillate droplet is constant for a long time period. At nucleate boiling of a water-salt solution of droplet several characteristic regimes occur: heating, nucleate boiling, desorption without bubble formation, formation of the solid, thin crystalline-hydrate film on the upper droplet surface, and formation of the ordered crystalline-hydrate structures during the longer time periods. For the final stage of desorption there is a big difference in desorption rate for initial salt concentration, C0, 11% and 51%. This great difference in the rate of desorption is associated with significantly more thin solution film for C0 = 11% and higher heat flux.

  17. Comparison of direct boiling method with commercial kits for extracting fecal microbiome DNA by Illumina sequencing of 16S rRNA tags.

    Science.gov (United States)

    Peng, Xin; Yu, Ke-Qiang; Deng, Guan-Hua; Jiang, Yun-Xia; Wang, Yu; Zhang, Guo-Xia; Zhou, Hong-Wei

    2013-12-01

    Low cost and high throughput capacity are major advantages of using next generation sequencing (NGS) techniques to determine metagenomic 16S rRNA tag sequences. These methods have significantly changed our view of microorganisms in the fields of human health and environmental science. However, DNA extraction using commercial kits has shortcomings of high cost and time constraint. In the present study, we evaluated the determination of fecal microbiomes using a direct boiling method compared with 5 different commercial extraction methods, e.g., Qiagen and MO BIO kits. Principal coordinate analysis (PCoA) using UniFrac distances and clustering showed that direct boiling of a wide range of feces concentrations gave a similar pattern of bacterial communities as those obtained from most of the commercial kits, with the exception of the MO BIO method. Fecal concentration by boiling method affected the estimation of α-diversity indices, otherwise results were generally comparable between boiling and commercial methods. The operational taxonomic units (OTUs) determined through direct boiling showed highly consistent frequencies with those determined through most of the commercial methods. Even those for the MO BIO kit were also obtained by the direct boiling method with high confidence. The present study suggested that direct boiling could be used to determine the fecal microbiome and using this method would significantly reduce the cost and improve the efficiency of the sample preparation for studying gut microbiome diversity. © 2013 Elsevier B.V. All rights reserved.

  18. 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

  19. An improved nucleate boiling design equation

    International Nuclear Information System (INIS)

    Basu, D.K.; Pinder, K.L.

    1976-01-01

    The effect of varying ΔT, the primary variable, on the value of heat transfer coefficient (h) in nucleate boiling is discussed. The three-parameter quadratic equation, h=P 1 + P 2 (ΔT) + P 3 (ΔT) 2 (where the constants, P 1 ,P 2 ,P 3 are functions of pressure, liquid properties and surface properties of the heater) is suggested. Ten sets of data at atmospheric pressure from six different workers and two more sets for pressure variation have been tested. The above quadratic equation fits the experimental data better than the existing two-parameter power relation, h=C(ΔT)sup(n) (where C is constant). The values of the three coeffcients in the quadratic equations are dependent on pressure, liquid properties and surface properties. A generalized empirical equation has been derived, which fits the selected pressure data well. (author)

  20. Heat transfer coeffcient for boiling carbon dioxide

    DEFF Research Database (Denmark)

    Knudsen, Hans Jørgen Høgaard; Jensen, Per Henrik

    1997-01-01

    Heat transfer coefficient and pressure drop for boiling carbon dioxide (R744) flowing in a horizontal pipe has been measured. The pipe is heated by condensing R22 outside the pipe. The heat input is supplied by an electrical heater wich evaporates the R22. With the heat flux assumed constant over...... the whole surface and with measured temperature difference between the inner surface and the evaporation temperature a mean heat transfer coefficient is calculated. The calculated heat transfer coefficient has been compared with the Chart Correlation of Shah. The Chart Correlation predicts too low heat...... transfer coefficient but the ratio between the measured and the calculated heat transfer coefficient is nearly constant and equal 1.9. With this factor the correlation predicts the measured data within 14% (RMS). The pressure drop is of the same order as the measuring uncertainty and the pressure drop has...

  1. Hydrogen water chemistry for boiling water reactors

    International Nuclear Information System (INIS)

    Cowan, R.L.; Cowan, R.L.; Kass, J.N.; Law, R.J.

    1985-01-01

    Hydrogen Water Chemistry (HWC) is now a practical countermeasure for intergranular stress corrosion cracking (IGSCC) susceptibility of reactor structural materials in Boiling Water Reactors (BWRs). The concept, which involves adding hydrogen to the feedwater to suppress the formation of oxidizing species in the reactor, has been extensively studied in both the laboratory and in several operating plants. The Dresden-2 Unit of Commonwealth Edison Company has completed operation for one full 18-month fuel cycle under HWC conditions. The specifications, procedures, equipment, instrumentation and surveillance programs needed for commercial application of the technology are available now. This paper provides a review of the benefits to be obtained, the side affects, and the special operational considerations needed for commercial implementation of HWC. Technological and management ''Lessons Learned'' from work conducted to date are also described

  2. HEDL W-1 SLSF experiment LOPI transient and boiling test results

    International Nuclear Information System (INIS)

    Henderson, J.M.; Wood, S.A.; Rothrock, R.B.

    1980-01-01

    The W-1 Sodium Loop Safety Facility (SLSF) experiment was designed to study the heat release characteristics of fast reactor fuel pins under Loss-of-Piping-Integrity (LOPI) accident conditions and determine stable sodium boiling initiation and recovery limits in a prototypic fuel pin bundle array. The results of the experiment address major second level of assurance (LOA-2) safety issues and provide increased insight and understanding of phenomena that would inherently terminate hypothesized accidents with only limited core damage. The irradiation phase of the experiment, consisting of thirteen individual transients, was performed between May 27 and July 20, 1979. The final transient produced approximately two seconds of coolant boiling, cladding dryout, and incipient fuel pin failure. The facility and test hardware performed as designed, allowing completion of all planned tests and achievement of all test objectives

  3. A numerical investigation of electrohydrodynamic (EHD) effects on bubble deformation under pseudo-nucleate boiling conditions

    International Nuclear Information System (INIS)

    Zu, Y.Q.; Yan, Y.Y.

    2009-01-01

    In this article, the electrohydrodynamic (EHD) effects on nucleate boiling are studied by developing a numerical modelling of EHD effect on bubble deformation in pseudo-nucleate boiling conditions. The volume of fluid (VOF) method is employed to track the interface between the gas-liquid two phases; the user-defined code is written and added to the commercial software FLUENT to solve the electric field and the corresponding electric body force. On this basis, the model is applied to study the EHD effects on heat transfer and fluid flows. An initial air bubble surrounded by liquid CCl 4 and attached to a horizontal superheated wall under the action of electric field is studied. The results of the EHD effect on bubble shape evolution are compared with those of available experiments showing good agreement. The mechanism of EHD enhancement of heat transfer and the EHD induced phenomena including bubble elongation and detachment are analyzed in detail.

  4. Using largest Lyapunov exponent to confirm the intrinsic stability of boiling water reactors

    International Nuclear Information System (INIS)

    Gavilian-Moreno, Carlos; Espinosa-Paredes, Gilberto

    2016-01-01

    The aim of this paper is the study of instability state of boiling water reactors with a method based in largest Lyapunov exponents (LLEs). Detecting the presence of chaos in a dynamical system is an important problem that is solved by measuring the LLE. Lyapunov exponents quantify the exponential divergence of initially close state-space trajectories and estimate the amount of chaos in a system. This method was applied to a set of signals from several nuclear power plant (NPP) reactors under commercial operating conditions that experienced instabilities events, apparently each of a different nature. Laguna Verde and Forsmark NPPs with in-phase instabilities, and Cofrentes NPP with out-of-phases instability. This study presents the results of intrinsic instability in the boiling water reactors of three NPPs. In the analyzed cases the limit cycle was not reached, which implies that the point of equilibrium exerts influence and attraction on system evolution

  5. Using largest Lyapunov exponent to confirm the intrinsic stability of boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Gavilian-Moreno, Carlos [Iberdrola Generacion, S.A., Cofrentes Nuclear Power Plant, Project Engineering Department, Paraje le Plano S/N, Valencia (Spain); Espinosa-Paredes, Gilberto [Area de ingeniera en Recursos Energeticos, Universidad Autonoma Metropolitana-Iztapalapa, Mexico city (Mexico)

    2016-04-15

    The aim of this paper is the study of instability state of boiling water reactors with a method based in largest Lyapunov exponents (LLEs). Detecting the presence of chaos in a dynamical system is an important problem that is solved by measuring the LLE. Lyapunov exponents quantify the exponential divergence of initially close state-space trajectories and estimate the amount of chaos in a system. This method was applied to a set of signals from several nuclear power plant (NPP) reactors under commercial operating conditions that experienced instabilities events, apparently each of a different nature. Laguna Verde and Forsmark NPPs with in-phase instabilities, and Cofrentes NPP with out-of-phases instability. This study presents the results of intrinsic instability in the boiling water reactors of three NPPs. In the analyzed cases the limit cycle was not reached, which implies that the point of equilibrium exerts influence and attraction on system evolution.

  6. Using Largest Lyapunov Exponent to Confirm the Intrinsic Stability of Boiling Water Reactors

    Directory of Open Access Journals (Sweden)

    Carlos J. Gavilán-Moreno

    2016-04-01

    Full Text Available The aim of this paper is the study of instability state of boiling water reactors with a method based in largest Lyapunov exponents (LLEs. Detecting the presence of chaos in a dynamical system is an important problem that is solved by measuring the LLE. Lyapunov exponents quantify the exponential divergence of initially close state-space trajectories and estimate the amount of chaos in a system. This method was applied to a set of signals from several nuclear power plant (NPP reactors under commercial operating conditions that experienced instabilities events, apparently each of a different nature. Laguna Verde and Forsmark NPPs with in-phase instabilities, and Cofrentes NPP with out-of-phases instability. This study presents the results of intrinsic instability in the boiling water reactors of three NPPs. In the analyzed cases the limit cycle was not reached, which implies that the point of equilibrium exerts influence and attraction on system evolution.

  7. Shock-induced heating and millisecond boiling in gels and tissue due to high intensity focused ultrasound

    Science.gov (United States)

    Canney, Michael S.; Khokhlova, Vera A.; Bessonova, Olga V.; Bailey, Michael R.; Crum, Lawrence A.

    2009-01-01

    Nonlinear propagation causes high intensity ultrasound waves to distort and generate higher harmonics, which are more readily absorbed and converted to heat than the fundamental frequency. Although such nonlinear effects have previously been investigated and found not to significantly alter high intensity focused ultrasound (HIFU) treatments, two results reported here change this paradigm. One is that at clinically relevant intensity levels, HIFU waves not only become distorted but form shock waves in tissue. The other is that the generated shock waves heat the tissue to boiling in much less time than predicted for undistorted or weakly distorted waves. In this study, a 2-MHz HIFU source operating at peak intensities up to 25,000 W/cm2 was used to heat transparent tissue-mimicking phantoms and ex vivo bovine liver samples. Initiation of boiling was detected using high-speed photography, a 20-MHz passive cavitation detector, and fluctuation of the drive voltage at the HIFU source. The time to boil obtained experimentally was used to quantify heating rates and was compared to calculations using weak shock theory and the shock amplitudes obtained from nonlinear modeling and from measurements with a fiber optic hydrophone. As observed experimentally and predicted by calculations, shocked focal waveforms produced boiling in as little as 3 ms and the time to initiate boiling was sensitive to small changes in HIFU output. Nonlinear heating due to shock waves is therefore important to HIFU and clinicians should be aware of the potential for very rapid boiling since it alters treatments. PMID:20018433

  8. Research on boiling liquid expanding vapour explosions

    Energy Technology Data Exchange (ETDEWEB)

    McDevitt, C.A.; Steward, F.R.; Venart, J.E.S.

    A boiling liquid expanding vapor explosion (BLEVE) is due to rapid boiling and expansion, with no ignition or chemical reaction involved. Research is being conducted to examine such questions as under what conditions tanks and their contents undergo BLEVE, what are the characteristics of tanks affected by BLEVE, and what alterations in tank design can be made to minimize the likelihood of BLEVEs. Experiments have been done with both propane and freon, using commercially available one-liter propane cylinders. Outdoor tests were conducted and designed to have the tank fail at a particular set of internal conditions. High speed photography was used to record the explosion, and computerized monitoring equipment to record temperature and pressure data. Tests were run to attempt to determine the relationship between temperature and BLEVEs, and to test the possibility that the occurrence of a BLEVE depends on the amount of vapor that could be produced when the tank was ruptured. Discussion is made of the role of pressure waves and rarefaction waves in the explosion. It is concluded that the superheat temperature limit, theorized as the minimum temperature below which no BLEVE can occur, cannot be used to predict BLEVEs. It has been shown that BLEVEs can occur below this temperature. There appears to be a relationship between liquid temperature, liquid volume, and the energy required to drive the BLEVE. Fireballs may occur after a BLEVE of flammable material, but are not part of the tank destruction. Rupture location (vapor vs liquid space) appears to have no effect on whether a container will undergo a BLEVE. 7 refs., 7 figs., 1 tab.

  9. New Departure from Nucleate Boiling model relying on first principle energy balance at the boiling surface

    Science.gov (United States)

    Demarly, Etienne; Baglietto, Emilio

    2017-11-01

    Predictions of Departure from Nucleate Boiling have been a longstanding challenge when designing heat exchangers such as boilers or nuclear reactors. Many mechanistic models have been postulated over more than 50 years in order to explain this phenomenon but none is able to predict accurately the conditions which trigger the sudden change of heat transfer mode. This work aims at demonstrating the pertinence of a new approach for detecting DNB by leveraging recent experimental insights. The new model proposed departs from all the previous models by making the DNB inception come from an energy balance instability at the heating surface rather than a hydrodynamic instability of the bubbly layer above the surface (Zuber, 1959). The main idea is to modulate the amount of heat flux being exchanged via the nucleate boiling mechanism by the wetted area fraction on the surface, thus allowing a completely automatic trigger of DNB that doesn't require any parameter prescription. This approach is implemented as a surrogate model in MATLAB in order to validate the principles of the model in a simple and controlled geometry. Good agreement is found with the experimental data leveraged from the MIT Flow Boiling at various flow regimes. Consortium for Advanced Simulation of Light Water Reactors (CASL).

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

    International Nuclear Information System (INIS)

    Rahmani, R.

    1983-01-01

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

  11. Synthetic range profiling, ISAR imaging of sea vessels and feature extraction, using a multimode radar to classify targets: initial results from field trials

    CSIR Research Space (South Africa)

    Abdul Gaffar, MY

    2011-04-01

    Full Text Available tanazi@kacst.edu.sa, aazamil@kacst.edu.sa Abstract?This paper describes the design and working principles of an experimental multimode radar with a stepped-frequency Synthetic Range Profiling (SRP) and Inverse Synthetic Aperture Radar (ISAR...

  12. Initial response of small ground-dwelling mammals to forest alternative buffers along headwater streams in the Washington Coast Range, USA

    Science.gov (United States)

    Randall J. Wilk; Martin G. Raphael; Christopher S. Nations; Jeffrey D. Ricklefs

    2010-01-01

    We assessed the short-term effects of alternative designs of forested buffer treatments along headwater streams on small ground-dwelling mammals in managed forests in western Washington, USA. Over three summers (one pretreatment and two posttreatment), we trapped 19 mammalian species along 23 streams in the northern Coast Range. We compared faunal communities in...

  13. 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

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

    International Nuclear Information System (INIS)

    Sakurai, Akira

    1978-01-01

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

  15. 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.

  16. 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

  17. 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

  18. Numerical issues for liquid-metal boiling transient analysis

    International Nuclear Information System (INIS)

    Rowe, D.S.

    1986-01-01

    The large liquid-to-vapor density ratio of a boiling liquid-metal leads to a very abrupt change of the two-phase mixture density at the inception of boiling. Unfortunately, the strong dependence of mixture density on pressure leads to a key numerical issue that adversely affects the behavior of numerical solutions. The difficulties can be reduced by using techniques that acknowledge this functional behavior at the start of boiling. Some of the methods used include a spatially averaged density function, mathematical smoothing, and under relaxation. Nonequilibrium two-fluid models also seem to offer aid in obtaining reliable numerical solutions. (author)

  19. Quantification of tumour initiating effect of jute batching oil and its distillates over mouse skin.

    Science.gov (United States)

    Agarwal, R; Kumar, S; Shukla, Y; Antony, M; Mehrotra, N K

    1985-09-30

    In order to identify the tumour initiating constituent(s) of a mineral oil, jute batching oil (JBO), used in the processing of jute fibres, it was fractionally distilled in various boiling range fractions. The latter were then subjected to in vivo assessment of their aryl hydrocarbon hydroxylase (AHH) inducing potential in mouse epidermis. Fractions with almost similar AHH inducing potential were regrouped and studied for their tumour initiating potential over mouse skin following two-stage initiation-promotion protocol and using 12-O-tetradecanoyl phorbol-13-acetate (TPA) as tumour promoter. It was noticed that: (1) JBO as initiator, provoked local development of benign skin tumours over mouse back; (2) fractions of JBO boiling below 335 degrees C and above 399 degrees C accounted for most of the tumour initiating potential of the oil; (3) the histological features of the tumours (i.e. benign papillomas and keratoacanthomas) initiated by these fractions were similar to those developed after being initiated with unfractionated or reconstituted JBO; (4) removal of these fractions from JBO may be attempted which could decontaminate the batch oil from most of its tumorigenic components and make it safer for industrial use.

  20. 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.)

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  2. Nucleate boiling of halogenated coolants - correlation analysis; Ebulicao nucleada de refrigerantes halogenados: analise de correlacoes

    Energy Technology Data Exchange (ETDEWEB)

    Ribatski, Gherhardt; Jabardo, Jose M. Saiz [Sao Paulo Univ., Sao Carlos, SP (Brazil). Escola de Engenharia. Dept. de Engenharia Mecanica

    1998-07-01

    Present study has been focused on a literature of heat transfer under nucleate boiling conditions of halocarbon refrigerants and their mixtures with lubricating oil. Two kind of correlations regarding the heat transfer mechanism have been found: strictly empirical, based on a straight curve fitting of experimental data, and semi-empirical, based on the particular point of view of the author regarding the physical mechanism but still fitted with experimental data. As a general rule, it has been noted that correlations present significant discrepancies among each other, a result which mostly reflects the wide range of experimental conditions used as a reference. A similar trend has been observed with refrigerant/oil mixtures. Given the current status of halocarbon refrigerants for refrigeration applications, there is clearly a need for further research regarding the nucleate boiling phenomenon related to those compounds. (author)

  3. Crack propagation behavior of Ti-5Ta alloy in boiling nitric acid solution

    International Nuclear Information System (INIS)

    Motooka, Takafumi; Kiuchi, Kiyoshi

    1999-05-01

    The crack propagation behavior of Ti-5Ta alloy both in boiling nitric acid solution and in air at room temperature has been investigated. The crack growth rate of Ti-5Ta alloy was measured as a function of the stress intensity factor range. After the tests, the fracture surface morphology was observed by a scanning electron microscope and the crystallographic orientation was examined by X-ray diffraction analysis. Difference in the crack growth behavior was not observed in both environments. The crack growth rate in boiling nitric acid solution was similar to that in air at room temperature. Moreover, the crystallographic orientation of Ti-5Ta alloy had little effect on the fatigue behavior, because this alloy does not have the susceptibility to SCC in nitric acid solution. (author)

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  6. 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.)

  7. 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.

  8. 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)

  9. The Range of Initial 10Be/9Be Ratios in the Early Solar System: A Re-Assessment Based on Analyses of New CAIs and Melilite Composition Glass Standards

    Science.gov (United States)

    Dunham, E.; Wadhwa, M.; Liu, M.-C.

    2017-07-01

    We report a more accurate range of initial 10Be/9Be in CAIs including FUN CAI CMS-1 from Allende (CV3) and a new CAI from NWA 5508 (CV3) using melilite composition glass standards; we suggest 10Be is largely produced by irradiation in the nebula.

  10. Quantification of meat proportions by measuring DNA contents in raw and boiled sausages using matrix-adapted calibrators and multiplex real-time PCR.

    Science.gov (United States)

    Köppel, René; Eugster, Albert; Ruf, Jürg; Rentsch, Jürg

    2012-01-01

    The quantification of meat proportions in raw and boiled sausage according to the recipe was evaluated using three different calibrators. To measure the DNA contents from beef, pork, sheep (mutton), and horse, a tetraplex real-time PCR method was applied. Nineteen laboratories analyzed four meat products each made of different proportions of beef, pork, sheep, and horse meat. Three kinds of calibrators were used: raw and boiled sausages of known proportions ranging from 1 to 55% of meat, and a dilution series of DNA from muscle tissue. In general, results generated using calibration sausages were more accurate than those resulting from the use of DNA from muscle tissue, and exhibited smaller measurement uncertainties. Although differences between uses of raw and boiled calibration sausages were small, the most precise and accurate results were obtained by calibration with fine-textured boiled reference sausages.

  11. Void fraction prediction in saturated flow boiling

    International Nuclear Information System (INIS)

    Francisco J Collado

    2005-01-01

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

  12. 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

  13. 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.

  14. 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

  15. 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

  16. Experimental and theoretical study on transition boiling concerning downward-facing horizontal surface in confined space

    International Nuclear Information System (INIS)

    Zhao, D.W.; Su, G.H.; Tian, W.X.; Sugiyama, K.; Qiu, S.Z.

    2008-01-01

    Experimental study has been conducted to examine the pool boiling occurs on a relative large downward-facing round surface with a diameter of 300 mm in confined water pool at atmospheric pressure. An artificial neural network (ANN) has been trained successfully based on the experimental data for predicting Nusselt number of transition boiling in the present study. The input parameters of the ANN are wall superheat, ΔT w , the ratio of the gap size to the diameter of the heated surface, δ/D, Prandtl number and Rayleigh number. The output is Nusselt number, Nu. The results show that: Nu decreases with increasing ΔT w , and increases generally with an increase of δ/D. Nu increases with increasing Pr when gap size is smaller than 4.0 mm. And Nu decreases initially and then increases with increasing Pr as gap size bigger than 5.0 mm. The results also indicate that the influence of Grashof number, Gr, could be negligible. Finally, a new correlation was proposed to predict the transition boiling heat transfer under the present condition. The comparisons between the prediction of the new correlation and experimental data show a reasonable agreement

  17. Design of ultrasonically-activatable nanoparticles using low boiling point perfluorocarbons.

    Science.gov (United States)

    Sheeran, Paul S; Luois, Samantha H; Mullin, Lee B; Matsunaga, Terry O; Dayton, Paul A

    2012-04-01

    Recently, an interest has developed in designing biomaterials for medical ultrasonics that can provide the acoustic activity of microbubbles, but with improved stability in vivo and a smaller size distribution for extravascular interrogation. One proposed alternative is the phase-change contrast agent. Phase-change contrast agents (PCCAs) consist of perfluorocarbons (PFCs) that are initially in liquid form, but can then be vaporized with acoustic energy. Crucial parameters for PCCAs include their sensitivity to acoustic energy, their size distribution, and their stability, and this manuscript provides insight into the custom design of PCCAs for balancing these parameters. Specifically, the relationship between size, thermal stability and sensitivity to ultrasound as a function of PFC boiling point and ambient temperature is illustrated. Emulsion stability and sensitivity can be 'tuned' by mixing PFCs in the gaseous state prior to condensation. Novel observations illustrate that stable droplets can be generated from PFCs with extremely low boiling points, such as octafluoropropane (b.p. -36.7 °C), which can be vaporized with acoustic parameters lower than previously observed. Results demonstrate the potential for low boiling point PFCs as a useful new class of compounds for activatable agents, which can be tailored to the desired application. Copyright © 2012 Elsevier Ltd. All rights reserved.

  18. 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

  19. 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.

  20. An Analytical Approach for Relating Boiling Points of Monofunctional Organic Compounds to Intermolecular Forces

    Science.gov (United States)

    Struyf, Jef

    2011-01-01

    The boiling point of a monofunctional organic compound is expressed as the sum of two parts: a contribution to the boiling point due to the R group and a contribution due to the functional group. The boiling point in absolute temperature of the corresponding RH hydrocarbon is chosen for the contribution to the boiling point of the R group and is a…

  1. Feedback stabilisation of pool-boiling systems : for application in thermal management schemes

    NARCIS (Netherlands)

    Gils, van R.W.

    2012-01-01

    The research scope of this thesis is the stabilisation of unstable states in a pool-boiling system. Thereto, a compact mathematical model is employed. Pool-boiling systems serve as physical model for practical applications of boiling heat transfer in industry. Boiling has advantages over

  2. Superheating in nucleate boiling calculated by the heterogeneous nucleation theory

    International Nuclear Information System (INIS)

    Gerum, E.; Straub, J.; Grigull, U.

    1979-01-01

    With the heterogeneous nucleation theory the superheating of the liquid boundary layer in nucleate boiling is described not only for the onset of nuclear boiling but also for the boiling crisis. The rate of superheat depends on the thermodynamic stability of the metastable liquid, which is influenced by the statistical fluctuations in the liquid and the nucleation at the solid surface. Because of the fact that the cavities acting as nuclei are too small for microscopic observation, the size and distribution function of the nuclei on the surface necessary for the determination of the probability of bubble formation cannot be detected by measuring techniques. The work of bubble formation reduced by the nuclei can be represented by a simple empirical function whose coefficients are determined from boiling experiments. Using this the heterogeneous nucleation theory describes the superheating of the liquid. Several fluids including refrigerants, liquid gases, organic liquids and water were used to check the theory. (author)

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

    International Nuclear Information System (INIS)

    Schlueter, G.; Efferding, L.E.

    1973-01-01

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

  4. Correlations between boiling points and relative retention data for hydrocarbons

    NARCIS (Netherlands)

    Sojak, L.; Krupcik, J.; Rijks, J.A.

    1974-01-01

    An equation correlating retention indices, boiling points and activity coefficients is proposed. The equation can be applied not only to homologous series, but also to different classes of hydrocarbons.

  5. 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...

  6. Making Sense of Boiling Points and Melting Points

    Indian Academy of Sciences (India)

    GENERAL | ARTICLE. The boiling and melting points of a pure substance are char- ... bonds, which involves high energy and hence high temperatures. Among the .... with zero intermolecular force at all temperatures and pressures, which ...

  7. Film Boiling on Downward Quenching Hemisphere of Varying Sizes

    Energy Technology Data Exchange (ETDEWEB)

    Chan S. Kim; Kune Y. Suh; Joy L. Rempe; Fan-Bill Cheung; Sang B. Kim

    2004-04-01

    Film boiling heat transfer coefficients for a downward-facing hemispherical surface are measured from the quenching tests in DELTA (Downward-boiling Experimental Laminar Transition Apparatus). Two test sections are made of copper to maintain low Biot numbers. The outer diameters of the hemispheres are 120 mm and 294 mm, respectively. The thickness of all the test sections is 30 mm. The effect of diameter on film boiling heat transfer is quantified utilizing results obtained from the test sections. The measured data are compared with the numerical predictions from laminar film boiling analysis. The measured heat transfer coefficients are found to be greater than those predicted by the conventional laminar flow theory on account of the interfacial wavy motion incurred by the Helmholtz instability. Incorporation of the wavy motion model considerably improves the agreement between the experimental and numerical results in terms of heat transfer coefficient. In addition, the interfacial wavy motion and the quenching process are visualized through a digital camera.

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  9. 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

  10. 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.)

  11. A microgravity boiling and convective condensation experiment

    Science.gov (United States)

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

    1987-12-01

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

  12. SBWR: A simplified boiling water reactor

    International Nuclear Information System (INIS)

    Duncan, J.D.; Sawyer, C.D.; Lagache, M.P.

    1987-01-01

    An advanced light water reactor concept is being developed for possible application in the 1990's. The concept, known as SBWR is a boiling water reactor which uses natural circulation to provide flow to the reactor core. In an emergency, a gravity driven core cooling system is used. The reactor is depressurized and water from an elevated suppression pool flows by gravity to the reactor vessel to keep the reactor core covered. The concept also features a passive containment cooling system in which water flows by gravity to cool the suppression pool wall. No operator action is required for a period of at least three days. Use of these and other passive systems allows the elimination of emergency diesel generators, core cooling pumps and heat removal pumps which is expected to simplify the plant design, reduce costs and simplify licensing. The concept is being developed by General Electric, Bechtel and the Massachusetts Institute of Technology supported by the Electric Power Research Institute and the United States Department of Energy in the United States. In Japan, The Japan Atomic Power Company has a great interest in this concept

  13. Fundamentals of boiling water reactor (BWR)

    International Nuclear Information System (INIS)

    Bozzola, S.

    1982-01-01

    These lectures on fundamentals of BWR reactor physics are a synthesis of known and established concepts. These lectures are intended to be a comprehensive (even though descriptive in nature) presentation, which would give the basis for a fair understanding of power operation, fuel cycle and safety aspects of the boiling water reactor. The fundamentals of BWR reactor physics are oriented to design and operation. In the first lecture general description of BWR is presented, with emphasis on the reactor physics aspects. A survey of methods applied in fuel and core design and operation is presented in the second lecture in order to indicate the main features of the calculational tools. The third and fourth lectures are devoted to review of BWR design bases, reactivity requirements, reactivity and power control, fuel loading patterns. Moreover, operating limits are reviewed, as the actual limits during power operation and constraints for reactor physics analyses (design and operation). The basic elements of core management are also presented. The constraints on control rod movements during the achieving of criticality and low power operation are illustrated in the fifth lecture. Some considerations on plant transient analyses are also presented in the fifth lecture, in order to show the impact between core and fuel performance and plant/system performance. The last (sixth) lecture is devoted to the open vessel testing during the startup of a commercial BWR. A control rod calibration is also illustrated. (author)

  14. Boiling water reactor turbine trip (TT) benchmark

    International Nuclear Information System (INIS)

    2005-01-01

    In the field of coupled neutronics/thermal-hydraulics computation there is a need to enhance scientific knowledge in order to develop advanced modelling techniques for new nuclear technologies and concepts as well as for current applications. Recently developed 'best-estimate' computer code systems for modelling 3-D coupled neutronics/thermal-hydraulics transients in nuclear cores and for coupling core phenomena and system dynamics (PWR, BWR, VVER) need to be compared against each other and validated against results from experiments. International benchmark studies have been set up for this purpose. The present report is the second in a series of four and summarises the results of the first benchmark exercise, which identifies the key parameters and important issues concerning the thermalhydraulic system modelling of the transient, with specified core average axial power distribution and fission power time transient history. The transient addressed is a turbine trip in a boiling water reactor, involving pressurization events in which the coupling between core phenomena and system dynamics plays an important role. In addition, the data made available from experiments carried out at the Peach Bottom 2 reactor (a GE-designed BWR/4) make the present benchmark particularly valuable. (author)

  15. Self-Sustaining Thorium Boiling Water Reactors

    Directory of Open Access Journals (Sweden)

    Ehud Greenspan

    2012-10-01

    Full Text Available A thorium-fueled water-cooled reactor core design approach that features a radially uniform composition of fuel rods in stationary fuel assembly and is fuel-self-sustaining is described. This core design concept is similar to the Reduced moderation Boiling Water Reactor (RBWR proposed by Hitachi to fit within an ABWR pressure vessel, with the following exceptions: use of thorium instead of depleted uranium for the fertile fuel; elimination of the internal blanket; and elimination of absorbers from the axial reflectors, while increasing the length of the fissile zone. The preliminary analysis indicates that it is feasible to design such cores to be fuel-self-sustaining and to have a comfortably low peak linear heat generation rate when operating at the nominal ABWR power level of nearly 4000 MWth. However, the void reactivity feedback tends to be too negative, making it difficult to have sufficient shutdown reactivity margin at cold zero power condition. An addition of a small amount of plutonium from LWR used nuclear fuel was found effective in reducing the magnitude of the negative void reactivity effect and enables attaining adequate shutdown reactivity margin; it also flattens the axial power distribution. The resulting design concept offers an efficient incineration of the LWR generated plutonium in addition to effective utilization of thorium. Additional R&D is required in order to arrive at a reliable practical and safe design.

  16. 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

  17. Boiling water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    2003-01-01

    The International Atomic Energy Agency (IAEA) has established an activity in nuclear reactor simulation computer programs to assist its Member States in education. The objective is to provide, for a variety of advanced reactor types, insight and practice in their operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA arranges for the development and distribution of simulation programs and workshop material and sponsors workshops. The workshops are in two parts: techniques and tools for reactor simulator development; and the use of reactor simulators in education. Workshop material for the first part is covered in the IAEA publication: Training Course Series No. 12, 'Reactor Simulator Development' (2001). Course material for workshops using a WWER- 1000 simulator from the Moscow Engineering and Physics Institute, Russian Federation is presented in the IAEA publication: Training Course Series No. 21 'WWER-1000 Reactor Simulator' (2002). Course material for workshops using a pressurized water reactor (PWR) simulator developed by Cassiopeia Technologies Incorporated, Canada, is presented in the IAEA publication: Training Course Series No. 22 'Pressurized Water Reactor Simulator' (2003). This report consists of course material for workshops using a boiling water reactor (BWR) simulator. Cassiopeia Technologies Incorporated, developed the simulator and prepared this report for the IAEA

  18. 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

  19. Steady-state pool boiling heat transfer on nicr wire surface submerged in Al2O3 nano-fluids

    International Nuclear Information System (INIS)

    Dereje Shiferaw; Hyun Sun Park; Bal Raj Sehgal

    2005-01-01

    Full text of publication follows: nano-fluids, or conventional liquids, e.g., water, with small concentration of nano-particles uniformly suspended, have attracted attention as a new heat transport medium with enhanced thermo-physical properties. Up to the present, only exploratory experiments on nano-fluids have been reported. Das et al (Int. J. Heat Mass Transfer 43, pp 3701-3707, 2003) conducted boiling experiments with water containing 38 nm Al 2 O 3 nano-particles. They observed deterioration in the nucleate boiling heat transfer due to the deposition of nano-particles. Boiling experiments conducted by Vassallo et al (Int. J. Heat Mass Transfer 47, pp 407-411, 2004) using silica nano-fluid using 0.4 mm diameter NiCr wire showed three times higher critical heat flux (CHF) and the wire traversed the film boiling region before it failed. Another independent experiment performed on 1 cm 2 square plate with a very low concentration of nano-particles ranging from 0.01 to 0.05 g/liter and at under pressure (2.89 psia), nano-fluids resulted in drastic 2∼3 times enhancement of the CHF (You and Kim, Appl. Phys. Lett. 83. No 16, 2003). However in all the aforementioned studies no appropriate explanation of the CHF enhancement has been advanced. The measured 2-3 times higher critical heat flux for very dilute nano-fluids may have high significance if such nano-fluids could be employed in heat transport systems. Recently, we investigated the effect of nano-particles on film boiling, which governs heat transfer during accident conditions in a reactor plant, e.g., in coolability of a degraded core, or a particulate debris bed or a core melt, and in steam explosions. Our previous experiments performed on film boiling in nano-fluids having larger concentrations of 5, 10, and 20 g/liter than those in You's experiments showed that the nano-fluids lower the film boiling temperature, decrease the film boiling heat transfer and provide a much thicker and more stable film than

  20. Low-Temperature Synchrotron Photoionization Study of 2-Methyl-3-buten-2-ol (MBO) Oxidation Initiated by O(3P) Atoms in the 298-650 K Range.

    Science.gov (United States)

    Fathi, Yasmin; Price, Chelsea; Meloni, Giovanni

    2017-04-20

    This work studies the oxidation of 2-methyl-3-buten-2-ol initiated by O( 3 P) atoms. The oxidation was investigated at room temperature, 550, and 650 K. Using the synchrotron radiation from the Advanced Light Source (ALS) of the Lawrence Berkley National Laboratory, reaction intermediates and products were studied by multiplexed photoionization mass spectrometry. Mass-to-charge ratios, kinetic time traces, photoionization spectra, and adiabatic ionization energies for each primary reaction species were obtained and used to characterize their identity. Using electronic structure calculations, potential energy surface scans of the different species produced throughout the oxidation were examined and presented in this paper to further validate the primary chemistry occurring. Branching fractions of primary products at all three temperatures were also provided. At room temperature only three primary products formed: ethenol (26.6%), acetaldehyde (4.2%), and acetone (53.4%). At 550 and 650 K the same primary products were observed in addition to propene (5.1%, 11.2%), ethenol (18.1%, 2.8%), acetaldehyde (8.9%, 5.7%), cyclobutene (1.6%, 10.8%), 1-butene (2.0%, 10.9%), trans-2-butene (3.2%, 23.1%), acetone (50.4%, 16.8%), 3-penten-2-one (1.0%, 11.5%), and 3-methyl-2-butenal (0.9%, 2.5%), where the first branching fraction value in parentheses corresponds to the 550 K data. At the highest temperature, a small amount of propyne (1.0%) was also observed.

  1. 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

  2. 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)

  3. 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

  4. 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

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

    CERN Document Server

    Saha, Sujoy Kumar

    2016-01-01

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

  6. Diet quality and six-year risk of overweight and obesity among mid-age Australian women who were initially in the healthy weight range.

    Science.gov (United States)

    Aljadani, Haya M; Patterson, Amanda J; Sibbritt, David; Collins, Clare E

    2016-04-01

    Issue addressed The present study investigated the association between diet quality, measured using the Australian Recommended Food Score (ARFS), and 6-year risk of becoming overweight or obese in mid-age women from the Australian Longitudinal Study of Women's Health (ALSWH). Methods Women (n=1107) aged 47.6-55.8 years who were a healthy weight (body mass index (BMI) between ≤18.5 and healthy weight range and those who became overweight or obese at follow-up was 35.3±8.1 and 34.3±8.8, respectively. There was no relationship between baseline ARFS and risk of becoming overweight or obese over 6 years. Women who were smokers were more likely to become overweight or obese (odds ratio 1.5; 95% confidence interval 1.11-2.09; P=0.008). Conclusions Poor diet quality was common among mid-age women of a healthy weight in the ALSWH. Higher diet quality was not associated with the risk of overweight or obesity after 6 years, yet smoking status was. So what? Better diet quality alone will not achieve maintenance of a healthy weight, but should be encouraged to improve other health outcomes.

  7. 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

  8. Initial Cladding Condition

    International Nuclear Information System (INIS)

    Siegmann, E.

    2000-01-01

    The purpose of this analysis is to describe the condition of commercial Zircaloy clad fuel as it is received at the Yucca Mountain Project (YMP) site. Most commercial nuclear fuel is encased in Zircaloy cladding. This analysis is developed to describe cladding degradation from the expected failure modes. This includes reactor operation impacts including incipient failures, potential degradation after reactor operation during spent fuel storage in pool and dry storage and impacts due to transportation. Degradation modes include cladding creep, and delayed hydride cracking during dry storage and transportation. Mechanical stresses from fuel handling and transportation vibrations are also included. This Analysis and Model Report (AMR) does not address any potential damage to assemblies that might occur at the YMP surface facilities. Ranges and uncertainties have been defined. This analysis will be the initial boundary condition for the analysis of cladding degradation inside the repository. In accordance with AP-2.13Q, ''Technical Product Development Planning'', a work plan (CRWMS M andO 2000c) was developed, issued, and utilized in the preparation of this document. There are constraints, caveats and limitations to this analysis. This cladding degradation analysis is based on commercial Pressurized Water Reactor (PWR) fuel with Zircaloy cladding but is applicable to Boiling Water Reactor (BWR) fuel. Reactor operating experience for both PWRs and BWRs is used to establish fuel reliability from reactor operation. It is limited to fuel exposed to normal operation and anticipated operational occurrences (i.e. events which are anticipated to occur within a reactor lifetime), and not to fuel that has been exposed to severe accidents. Fuel burnup projections have been limited to the current commercial reactor licensing environment with restrictions on fuel enrichment, oxide coating thickness and rod plenum pressures. The information provided in this analysis will be used in

  9. Instrumentation availability during severe accidents for a boiling water reactor with a Mark I containment

    International Nuclear Information System (INIS)

    Arcieri, W.C.; Hanson, D.J.

    1992-02-01

    In support of the US Nuclear Regulatory Commission Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a Boiling Water Reactor with a Mark I containment. Results from this evaluation include: (1) the identification of plant conditions that would impact instrument performance and information needs during severe accidents; (2) the definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences; and (3) assessment of the availability of plant instrumentation during severe accidents

  10. 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

    Complex two-phase heat transfer phenomena such as nucleate boiling, critical heat flux, quenching and condensation govern the thermal performance of Light Water Reactors (LWRs) under normal operation and during transients/accidents. These phenomena are typically characterized by the presence of a liquid vapor- solid contact line on the surface from/to which the heat is transferred. For example, in nucleate boiling, a significant fraction of the energy needed for bubble growth comes from evaporation of a liquid meniscus, or microlayer, underneath the bubble itself. As the liquid vapor- solid line at the edge of the meniscus retreats, a circular dry patch in the middle of the bubble is exposed; the speed of the triple line retreat is a measure of the ability of the surface to transfer heat to the bubble. At very high heat fluxes, near the upper limit of the nucleate boiling regime, also known as Critical Heat Flux (CHF), the situation is characterized by larger dry areas on the surface, dispersed within an interconnected network of liquid menisci. In quenching heat transfer, which refers to the rapid cooling of a very hot object by immersion in a cooler liquid, the process is initially dominated by film boiling. In film boiling a continuous vapor film completely separates the liquid phase from the solid surface: however, as the temperature gets closer to the Leidenfrost point, intermittent and short-lived liquid-solid contacts occur at discrete locations on the surface, thus creating liquid vapor- solid interfaces once again. Ultimately, if bubble nucleation ensues at such contact points, the vapor film is disrupted and the heat transfer regime transitions from film boiling to transition boiling. Finally, in dropwise condensation, the phase transition from vapor to liquid occurs via formation of discrete droplets on the surface, and the resulting liquid-vapor-solid triple line is where heat transfer is most intense. To gain insight into and enable mechanistic

  11. 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)

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

    Science.gov (United States)

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

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

  13. 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

  14. 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].

  15. 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)

  16. 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

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

    International Nuclear Information System (INIS)

    Ueda, Tatsuhiro; Isayama, Yasushi

    1981-01-01

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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  19. Experiments of Pool Boiling Performance (Boiling Heat Transfer and Critical Heat Flux) on Designed Micro-Structures

    International Nuclear Information System (INIS)

    Kim, Seol Ha; Kang, Jun Young; Lee, Gi Chol; Kiyofumia, Moriyama; Kim, Moo Hwan; Park, Hyun Sun

    2015-01-01

    In general, the evaluation of the boiling performance mainly focuses on two physical parameters: boiling heat transfer (BHT) and critical heat flux (CHF). In the nuclear power plants, both BHT and CHF contribute the nuclear system efficiency and safety, respectively. In this study, BHT and CHF of the pool boiling on well-organized fabricated structured (micro scaled) surface has been evaluated. As a results, BHT change on microstructured surface shows strongly dependent on Pin-fin effect analysis. In terms of CHF, critical size of micro structure for CHF enhancement has been observed and analyzed based on the capillary wicking effect. In this study, BHT and CHF of the pool boiling on well-organized fabricated structured (micro scaled) surface has been evaluated. As a results, BHT change on microstructured surface shows strongly dependent on the roughness ratio. The extended heat transfer area contributes the boiling heat transfer increase on the structured surface, and its quantitative analysis has been performed. In terms of CHF, the critical size of micro structure for CHF enhancement has been observed and analyzed based on the capillary wicking effect. We suggested a capillary limit to CHF delay for modeling capillary induced liquid inflow through microstructured surfaces. The critical size of the capillary limit on the prepared structured surface, determined by a model, could be reasonable explanation points for the experimental results (optimal size for CHF delay). The present experimental results also showed clearly the critical size (10 - 20 μm) for CHF delay, predicted by capillary limit analysis. This study provides fundamental insight into BHT and CHF enhancement of structured surfaces, and an optimal design guide for the required CHF and boiling heat-transfer performance. Finally, this study can contribute the basic understanding of the boiling on designed microstructure surface, and it also suggest the optimal micro scaled structured surface of boiling

  20. Status of the advanced boiling water reactor and simplified boiling water reactor

    International Nuclear Information System (INIS)

    Smith, P.F.

    1992-01-01

    This paper reports that the excess of U.S. electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which are designed to ensure that the nuclear power option is available to utilities. Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14 point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other enabling conditions. GE is participating in this national effort and GE's family of advanced nuclear power plants feature two new reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the U.S. and worldwide. Both possess the features necessary to do so safely, reliably, and economically

  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. Direct numerical simulations of nucleate boiling flows of binary mixtures

    International Nuclear Information System (INIS)

    Didier Jamet; Celia Fouillet

    2005-01-01

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

  4. Future directions in boiling water reactor design

    International Nuclear Information System (INIS)

    Wilkins, D.R.; Hucik, S.A.; Duncan, J.D.; Sweeney, J.I.

    1987-01-01

    The Advanced Boiling Water Reactor (ABWR) is being developed by an international team of BWR manufacturers to respond to worldwide utility needs in the 1990's. Major objectives of the ABWR program are design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuver-ability; and reduced occupational exposure and radwaste. The ABWR incorporates the best proven features from BWR designs in Europe, Japan and the United States and application of leading edge technology. Key features of the ABWR are internal recirculation pumps; fine-motion, electrohydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling netwoek; pressure suppression containment with horizontal vents; cylindrical reinforced concrete containment; structural integration of the containment and reactor building; severe accident capability; state-of-the-art fuel; advanced trubine/generator with 52'' last stage buckets; and advanced radwaste technology. The ABWR is ready for lead plant application in Japan, where it is being developed as the next generation Japan standard BWR under the guidance and leadership of The Tokyo Electric Power Company, Inc. and a group of Japanese BWR utilities. In the United States it is being adapted to the needs of US utilities through the Electric Power Research Institute's Advanced LWR Requirements Program, and is being reviewed by the US Nuclear Regulatory Commission for certification as a preapproved US standard BWR under the US Department of Energy's ALWR Design Verification Program. These cooperative Japanese and US programs are expected to establish the ABWR as a world class BWR for the 1990's...... (author)

  5. Accident sequence analysis for a BWR [Boiling Water Reactor] during low power and shutdown operations

    International Nuclear Information System (INIS)

    Whitehead, D.W.; Hake, T.M.

    1990-01-01

    Most previous Probabilistic Risk Assessments have excluded consideration of accidents initiated in low power and shutdown modes of operation. A study of the risk associated with operation in low power and shutdown is being performed at Sandia National Laboratories for a US Boiling Water Reactor (BWR). This paper describes the proposed methodology for the analysis of the risk associated with the operation of a BWR during low power and shutdown modes and presents preliminary information resulting from the application of the methodology. 2 refs., 2 tabs

  6. 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

  7. 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

  8. Experimental investigation and correlation of two-phase frictional pressure drop of R410A-oil mixture flow boiling in a 5 mm microfin tube

    Energy Technology Data Exchange (ETDEWEB)

    Ding, Guoliang; Hu, Haitao; Huang, Xiangchao [Institute of Refrigeration and Cryogenics, Shanghai Jiaotong University, Shanghai 200240 (China); Deng, Bin [Institute of Heat Transfer Technology, Golden Dragon Precise Copper Tube Group Inc., Shanghai 200135 (China); Gao, Yifeng [International Copper Association, Shanghai Office, Shanghai 200020 (China)

    2009-01-15

    This study presents experimental two-phase frictional data for R410A-oil mixture flow boiling in an internal spiral grooved microfin tube with outside diameter of 5 mm. Experimental parameters include the evaporation temperature of 5 C, the mass flux from 200 to 400 kg m{sup -2} s{sup -1}, the heat flux from 7.46 to 14.92 kW m{sup -2}, the inlet vapor quality from 0.1 to 0.8, and nominal oil concentration from 0 to 5%. The test results show that the frictional pressure drop of R410A initially increases with vapor quality and then decreases, presenting a local maximum in the vapor quality range between 0.7 and 0.8; the frictional pressure drop of R410A-oil mixture increases with the mass flux, the presence of oil enhances two-phase frictional pressure drop, and the effect of oil on frictional pressure drop is more evident at higher vapor qualities where the local oil concentrations are higher. The enhanced factor is always larger than unity and increases with nominal oil concentration at a given vapor quality. The range of the enhanced factor is about 1.0-2.2 at present test conditions. A new correlation to predict the local frictional pressure drop of R410A-oil mixture flow boiling inside the internal spiral grooved microfin tube is developed based on local properties of refrigerant-oil mixture, and the measured local frictional pressure drop is well correlated with the empirical equation proposed by the authors. (author)

  9. Numerical investigation of nucleate pool boiling heat transfer

    Directory of Open Access Journals (Sweden)

    Stojanović Andrijana D.

    2016-01-01

    Full Text Available Multidimensional numerical simulation of the atmospheric saturated pool boiling is performed. The applied modelling and numerical methods enable a full representation of the liquid and vapour two-phase mixture behaviour on the heated surface, with included prediction of the swell level and heated wall temperature field. In this way the integral behaviour of nucleate pool boiling is simulated. The micro conditions of bubble generation at the heated wall surface are modelled by the bubble nucleation site density, the liquid wetting contact angle and the bubble grow time. The bubble nucleation sites are randomly located within zones of equal size, where the number of zones equals the nucleation site density. The conjugate heat transfer from the heated wall to the liquid is taken into account in wetted heated wall areas around bubble nucleation sites. The boiling curve relation between the heat flux and the heated wall surface temperature in excess of the saturation temperature is predicted for the pool boiling conditions reported in the literature and a good agreement is achieved with experimentally measured data. The influence of the nucleation site density on the boiling curve characteristic is confirmed. In addition, the influence of the heat flux intensity on the spatial effects of vapour generation and two-phase flow are shown, such as the increase of the swell level position and the reduced wetting of the heated wall surface by the heat flux increase. [Projekat Ministarstva nauke Republike Srbije, br. TR-33018 i br. OI-174014

  10. Converting higher to lower boiling hydrocarbons. [Australian patent

    Energy Technology Data Exchange (ETDEWEB)

    1937-06-16

    To transform or convert higher boiling hydrocarbons into lower boiling hydrocarbons for the production of motor fuel, the hydrocarbons are maintained in vapor phase until the desired conversion has been effected and the separation of the high from low boiling hydrocarbons is carried out by utilization of porous contact material with a preferential absorption for the former. The vapor is passed by supply line to a separator containing the porous material and heated to 750 to 950/sup 0/F for a few seconds, the higher boiling parts being retained by the porous material and the lower passing to a vent line. The latter is closed and the vapor supply cut off and an ejecting medium is passed through a line to carry the higher boiling parts to an outlet line from which it may be recycled through the apparatus. The porous mass may be regenerated by introducing medium from a line that carries off impurities to another line. A modified arrangement shows catalytic cracking apparatus through which the vaporized material is passed on the way to the separators.

  11. Single-bubble boiling under Earth's and low gravity

    Science.gov (United States)

    Khusid, Boris; Elele, Ezinwa; Lei, Qian; Tang, John; Shen, Yueyang

    2017-11-01

    Miniaturization of electronic systems in terrestrial and space applications is challenged by a dramatic increase in the power dissipation per unit volume with the occurrence of localized hot spots where the heat flux is much higher than the average. Cooling by forced gas or liquid flow appears insufficient to remove high local heat fluxes. Boiling that involves evaporation of liquid in a hot spot and condensation of vapor in a cold region can remove a significantly larger amount of heat through the latent heat of vaporization than force-flow cooling can carry out. Traditional methods for enhancing boiling heat transfer in terrestrial and space applications focus on removal of bubbles from the heating surface. In contrast, we unexpectedly observed a new boiling regime of water under Earth's gravity and low gravity in which a bubble was pinned on a small heater up to 270°C and delivered a heat flux up to 1.2 MW/m2 that was as high as the critical heat flux in the classical boiling regime on Earth .Low gravity measurements conducted in parabolic flights in NASA Boeing 727. The heat flux in flight and Earth's experiments was found to rise linearly with increasing the heater temperature. We will discuss physical mechanisms underlying heat transfer in single-bubble boiling. The work supported by NASA Grants NNX12AM26G and NNX09AK06G.

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

    International Nuclear Information System (INIS)

    Mori, Yasuo; Hijikata, Kunio; Iwata, Shoichiro

    1976-01-01

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

  13. Passive gamma analysis of the boiling-water-reactor assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Vo, D., E-mail: ducvo@lanl.gov [Los Alamos National Laboratory, Los Alamos, NM (United States); Favalli, A. [Los Alamos National Laboratory, Los Alamos, NM (United States); Grogan, B. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Jansson, P. [Uppsala University, Uppsala (Sweden); Liljenfeldt, H. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Mozin, V. [Lawrence Livermore National Laboratory, Livermore, CA (United States); Schwalbach, P. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg); Sjöland, A. [Swedish Nuclear Fuel and Waste Management Company, Stockholm (Sweden); Tobin, S.; Trellue, H. [Los Alamos National Laboratory, Los Alamos, NM (United States); Vaccaro, S. [European Atomic Energy Community (EURATOM), Luxemburg (Luxembourg)

    2016-09-11

    This research focused on the analysis of a set of stationary passive gamma measurements taken on the spent nuclear fuel assemblies from a boiling water reactor (BWR) using pulse height analysis data acquisition. The measurements were performed on 25 different BWR assemblies in 2014 at Sweden's Central Interim Storage Facility for Spent Nuclear Fuel (Clab). This study was performed as part of the Next Generation of Safeguards Initiative–Spent Fuel project to research the application of nondestructive assay (NDA) to spent fuel assemblies. The NGSI–SF team is working to achieve the following technical goals more easily and efficiently than in the past using nondestructive assay (NDA) measurements of spent fuel assemblies: (1) verify the initial enrichment, burnup, and cooling time of facility declaration; (2) detect the diversion or replacement of pins, (3) estimate the plutonium mass, (4) estimate the decay heat, and (5) determine the reactivity of spent fuel assemblies. The final objective of this project is to quantify the capability of several integrated NDA instruments to meet the aforementioned goals using the combined signatures of neutrons, gamma rays, and heat. This report presents a selection of the measured data and summarizes an analysis of the results. Specifically, trends in the count rates measured for spectral lines from the following isotopes were analyzed as a function of the declared burnup and cooling time: {sup 137}Cs, {sup 154}Eu, {sup 134}Cs, and to a lesser extent, {sup 106}Ru and {sup 144}Ce. From these measured count rates, predictive algorithms were developed to enable the estimation of the burnup and cooling time. Furthermore, these algorithms were benchmarked on a set of assemblies not included in the standard assemblies set used by this research team.

  14. Sound velocity in the coolant of boiling nuclear reactors

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Parshin, D.A.; Novikov, K.S.; Galivec, E.Yu.

    2009-01-01

    To prevent resonant interaction between acoustic resonance and natural frequencies of FE, FA and RI oscillations, it is necessary to determine the value of EACPO. Based on results of calculations of EACPO and natural frequencies of FR, FA and RI oscillations values, it would be possible to reveal the dynamical loadings on metal that are dangerous for the initiation of cracking process in the early stage of negative condition appearance. To calculate EACPO it is necessary to know the Speed Velocity in Coolant. Now we do not have any data about real values of such important parameter as pressure pulsations propagation velocity in two phase environments, especially in conditions with variations of steam content along the length of FR, with taking into account the type of local resistances, flow geometry etc. While areas of resonant interaction of the single-phase liquid coolant with equipment and internals vibrations are estimated well enough, similar estimations in the conditions of presence of a gas and steam phase in the liquid coolant are inconvenient till now. Paper presents results of calculation of velocity of pressure pulsations distribution in two-phase flow formed in core of RBMK-1000 reactors. Feature of the developed techniques is that not only thermodynamic factors and effect of a speed difference between water and steam in a two phase flow but also geometrical features of core, local resistance, non heterogeneity in the two phase environment and power level of a reactor are considered. Obtained results evidence noticeable decreasing of velocity propagation of pressure pulsations in the presence of steam actions in the liquids. Such estimations for real RC of boiling nuclear reactors with steam-liquid coolant are obtained for the first time. (author)

  15. 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

  16. 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.)

  17. 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

  18. The influence of surface roughness and solution concentration on pool boiling process in Diethanolamine aqueous solution

    Science.gov (United States)

    Khoshechin, Mohsen; Salimi, Farhad; Jahangiri, Alireza

    2018-04-01

    In this research, the effect of surface roughness and concentration of solution on bubble departing frequency and nucleation site density for pool boiling of water/diethanolamine (DEA) binary solution were investigated experimentally. In this investigation, boiling heat transfer coefficient, bubble departing frequency and nucleation site density have been experimentally investigated in various concentrations and heat fluxes. Microstructured surfaces with a wide range of well-defined surface roughness were fabricated, and a heat flux between 1.5-86 kW/m2 was achieved under atmospheric conditions. The Results indicated that surface roughness and concentration of solution increase the bubble departing frequency and nucleation site density with increasing heat flux. The boiling heat transfer coefficient in mixtures of water/DEA increases with increasing concentration of DEA in water. The experimental results were compared with predictions of several used correlations in the literatures. Results showed that the boiling heat transfer coefficients of this case study are much higher than the predicted values by major existing correlations and models. The excellent agreement for bubble departing frequency found between the models of Jackob and Fritz (1966) and experimental data and also the nucleation site density were in close agreement with the model of Paul (1983) data. f bubble departure frequency, 1/s or Hz N Number of nucleation sites per area per time R c Minimum cavity size, m D c critical diameter, m g gravitational acceleration, m/s2 ρ density, kg/m3 T temperature, °c ΔT temperature difference, °c d d vapor bubble diameter, m h fg enthalpy of vaporization, J/kg R Roughness, μm Ja Jakob number cp specific heat, J/kg °c Pr Prandtl number Ar Archimedes number h Heat transfer coefficient, J/(m2 °c) tg time it takes to grow a bubble, s q/A heat flux (kW/m2) tw time required to heat the layer, s gc Correction coefficient of incompatible units R a Surface

  19. Numerical simulation of pool boiling of a Lennard-Jones liquid

    KAUST Repository

    Inaoka, Hajime

    2013-09-01

    We performed a numerical simulation of pool boiling by a molecular dynamics model. In the simulation, a liquid composed of Lennard-Jones particles in a uniform gravitational field is heated by a heat source at the bottom of the system. The model successfully reproduces the change in regimes of boiling from nucleate boiling to film boiling with the increase of the heat source temperature. We present the pool boiling curve by the model, whose general behavior is consistent with those observed in experiments of pool boiling. © 2013 Elsevier B.V. All rights reserved.

  20. Numerical simulation of pool boiling of a Lennard-Jones liquid

    KAUST Repository

    Inaoka, Hajime; Ito, Nobuyasu

    2013-01-01

    We performed a numerical simulation of pool boiling by a molecular dynamics model. In the simulation, a liquid composed of Lennard-Jones particles in a uniform gravitational field is heated by a heat source at the bottom of the system. The model successfully reproduces the change in regimes of boiling from nucleate boiling to film boiling with the increase of the heat source temperature. We present the pool boiling curve by the model, whose general behavior is consistent with those observed in experiments of pool boiling. © 2013 Elsevier B.V. All rights reserved.

  1. Study on Boiling Heat Transfer Phenomenon in Micro-channels

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Namgyun [Inha Technical College, Incheon (Korea, Republic of)

    2017-09-15

    Recently, efficient heat dissipation has become necessary because of the miniaturization of devices, and research on boiling on micro-channels has attracted attention. However, in the case of micro-channels, the friction coefficient and heat transfer characteristics are different from those in macro-channels. This leads to large errors in the micro scale results, when compared to correlations derived from the macro scale. In addition, due to the complexity of the mechanism, the boiling phenomenon in micro-channels cannot be approached only by experimental and theoretical methods. Therefore, numerical methods should be utilized as well, to supplement these methods. However, most numerical studies have been conducted on macro-channels. In this study, we applied the lattice Boltzmann method, proposed as an alternative numerical tool to simulate the boiling phenomenon in the micro-channel, and predicted the bubble growth process in the channel.

  2. Boiling heat transfer on fins – experimental and numerical procedure

    Directory of Open Access Journals (Sweden)

    Orzechowski T.

    2014-03-01

    Full Text Available The paper presents the research methodology, the test facility and the results of investigations into non-isothermal surfaces in water boiling at atmospheric pressure, together with a discussion of errors. The investigations were conducted for two aluminium samples with technically smooth surfaces and thickness of 4 mm and 10 mm, respectively. For the sample of lower thickness, on the basis of the surface temperature distribution measured with an infrared camera, the local heat flux and the heat transfer coefficient were determined and shown in the form of a boiling curve. For the thicker sample, for which 1-D model cannot be used, numerical calculations were conducted. They resulted in obtaining the values of the local heat flux on the surface the invisible to the infrared, camera i.e. on the side on which the boiling of the medium proceeds.

  3. 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)

  4. 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.

  5. The concept and application of miniaturization boiling in cooling system

    International Nuclear Information System (INIS)

    Suhaimi Illias; Muhammad Asri Idris

    2009-01-01

    The purpose of this research is to study and examine the phenomena of miniaturization-boiling, which intensely scatters with a large number of minute liquid particles from a water droplet surface to the atmosphere, when the droplet collided with a heating surface. As the material of the heating surface, the following were used: stainless steel (SUS 303 A Cr=17%,Ni=8%), sapphire (Al 3 O 2 ), brass, copper and carbon plane. The material was heated in order to study the miniaturization-boiling and droplet bounding phenomena at a very high temperature (160 degree C- 420 degree C). The phenomenon was photographed by a high-speed camera (10,000 fps) from the horizontal direction. The nuclear fusion reactor needs a very severe cooling, heat removal cooling method by special boiling is lead to this research. (Author)

  6. 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.

  7. 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)

  8. Boiling points of the superheavy elements 117 and 118

    International Nuclear Information System (INIS)

    Takahashi, N.

    2001-01-01

    It has been shown that the relativistic effect on the electrons reveal in the heavy element region. What kind of changes will appear in the heavy elements because of the relativistic effects? Can we observe the changes? We observed that the boiling points of astatine and radon are lower than that extrapolated values from lighter elements in the same groups. Systematic behavior of the elements on the boiling point was examined and a new method for the estimation of the boiling points of the superheavy elements in the halogen and rare gases has been found. The estimated values of the elements 117 and 118 are 618 and 247 K, respectively which are considerably lower than those obtained until now. If these values are correct the production of the superheavy elements with heavy ions reaction may be affected. Further, the chemical properties may be fairly different from the lighter elements. (author)

  9. 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

  10. 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.)

  11. Experimental Study of Flow Boiling Heat Transfer in a Horizontal Microfin Tube

    OpenAIRE

    Yu, Jian; Koyama, Shigeru; Momoki, Satoru

    1995-01-01

    An experimental study on flow boiling heat transfer in a horizontal microfin tube is conducted with pure refrigerants HFC134a, HCFC123 and HCFC22 using a water-heated double-tube type test section. The test microfin tube is a copper tube having the following dimensions: 8.37mm mean inside diameter, 0.168mm fin height, 60fin number and 18 degree of helix angle. The local heat transfer coefficients for both counter and parallel flows are measured in a range of heat flux of 1 to 93W/m^2, mass ve...

  12. Burnout in the boiling of water and freon-113 on tubes with annular fins

    International Nuclear Information System (INIS)

    Rubin, I.R.; Pul'kin, I.N.; Roizen, L.I.

    1986-01-01

    This paper presents the results of numerical calculations of burnout heat flux associated with the boiling of Freon-113 and water on an annular fin of constant thickness which have been approximated by simple analytical relations. These are used to calculate the critical burnout parameters of tubes with an annular fin assembly. The calculated data may be used for the analysis of tubes with an annular fin assembly over a wide range of variation of the thermophysical properties of the material and geometrical parameters of the fin assembly

  13. Water boiling on the corium melt surface under VVER severe accident conditions

    International Nuclear Information System (INIS)

    Bechta, S.V.; Vitol, S.A.; Krushinov, E.V.; Granovsky, V.S.; Sulatsky, A.A.; Khabensky, V.B.; Lopukh, D.B.; Petrov, Y.B.; Pechenkov, A.Y.

    2000-01-01

    Experimental results are presented on the interaction of corium melt with water supplied on its surface. The tests were conducted in the 'Rasplav-2' experimental facility. Corium melt was generated by induction melting in the cold crucible. The following data were obtained: heat transfer at boiling water-melt surface interaction, gas and aerosol release, post-interaction solidified corium structure. The corium melt charge had the following composition, mass%: 60% UO 2+x -16% ZrO 2 -15% Fe 2 O 3 -6% Cr 2 O 3 -3% Ni 2 O 3 . The melt surface temperature ranged within 1920-1970 K. (orig.)

  14. Design and performance of General Electric boiling water reactor main steam line isolation valves

    International Nuclear Information System (INIS)

    Rockwell, D.A.; van Zylstra, E.H.

    1976-08-01

    An extensive test program has been completed by the General Electric Company in cooperation with the Commonwealth Edison Company on the basic design type of large main steam line isolation valves used on General Electric Boiling Water Reactors. Based on a total of 40 tests under simulated accident conditions covering a wide range of mass flows, mixture qualities, and closing times, it was concluded that the commercially available valves of this basic type will close completely and reliably as required. Analytical methods to predict transient effects in the steam line and valve after postulated breaks were refined and confirmed by the test program

  15. Corrosion of high purity Fe-Cr-Ni alloys in 13 N boiling nitric acid

    International Nuclear Information System (INIS)

    Ohta, Joji; Mayuzumi, Masami; Kusanagi, Hideo; Takaku, Hiroshi

    1998-01-01

    Corrosion in boiling nitric acid was investigated for high purity Fe-18%Cr-12%Ni alloys and type 304L stainless steels (SS). Owing to very low impurity concentration, the solution treated high purity alloys show almost no intergranular corrosion while the type 304L SS show severe intergranular corrosion. Both in the high purity alloys and type 304L SS, aging treatments ranging from 873 K to 1073 K for 1 h enhance intergranular corrosion. During the aging treatments, impurities should be segregated to the grain boundaries. The corrosion behaviors were discussed from a standpoint of impurity segregation to grain boundaries. This study is of importance for purex reprocessing of spent fuels

  16. Experimental study on dryout point of flow boiling in bilaterally heated narrow annular channel

    International Nuclear Information System (INIS)

    Wu Geping; Wu Aimin; Tian Wenxi; Li Hao; Jia Dounan; Su Guanghui; Qiu Suizheng

    2003-01-01

    This paper presents and experimental study of the dryout point of flow boiling in bilaterally heated narrow annular channel with 1.5 mm and 2 mm annular gap, respectively. The range of pressure is 2.0-4.0 MPa and that of mass flux is 40-80 kg/m 2 ·s. Kutajilagi equation which is adaptable to tubes is used to deal with the experimental data and an empirical equation is obtained. Again this empirical equation is amended, then an empirical equation of the dryout point suitable for narrow annular channel is obtained

  17. Atomistic modelling of evaporation and explosive boiling of thin film liquid argon over internally recessed nanostructured surface

    Energy Technology Data Exchange (ETDEWEB)

    Hasan, Mohammad Nasim, E-mail: nasim@me.buet.ac.bd.com; Shavik, Sheikh Mohammad, E-mail: shavik@me.buet.ac.bd.com; Rabbi, Kazi Fazle, E-mail: rabbi35.me10@gmail.com; Haque, Mominul, E-mail: mominulmarup@gmail.com [Department of Mechanical Engineering, Bangladesh University of Engineering & Technology (BUET) Dhaka-1000 (Bangladesh)

    2016-07-12

    Molecular dynamics (MD) simulations have been carried out to investigate evaporation and explosive boiling phenomena of thin film liquid argon on nanostructured solid surface with emphasis on the effect of solid-liquid interfacial wettability. The nanostructured surface considered herein consists of trapezoidal internal recesses of the solid platinum wall. The wetting conditions of the solid surface were assumed such that it covers both the hydrophilic and hydrophobic conditions and hence effect of interfacial wettability on resulting evaporation and boiling phenomena was the main focus of this study. The initial configuration of the simulation domain comprised of a three phase system (solid platinum, liquid argon and vapor argon) on which equilibrium molecular dynamics (EMD) was performed to reach equilibrium state at 90 K. After equilibrium of the three-phase system was established, the wall was set to different temperatures (130 K and 250 K for the case of evaporation and explosive boiling respectively) to perform non-equilibrium molecular dynamics (NEMD). The variation of temperature and density as well as the variation of system pressure with respect to time were closely monitored for each case. The heat flux normal to the solid surface was also calculated to illustrate the effectiveness of heat transfer for hydrophilic and hydrophobic surfaces in cases of both nanostructured surface and flat surface. The results obtained show that both the wetting condition of the surface and the presence of internal recesses have significant effect on normal evaporation and explosive boiling of the thin liquid film. The heat transfer from solid to liquid in cases of surface with recesses are higher compared to flat surface without recesses. Also the surface with higher wettability (hydrophilic) provides more favorable conditions for boiling than the low-wetting surface (hydrophobic) and therefore, liquid argon responds quickly and shifts from liquid to vapor phase faster in

  18. A high-fidelity approach towards simulation of pool boiling

    Energy Technology Data Exchange (ETDEWEB)

    Yazdani, Miad; Radcliff, Thomas; Soteriou, Marios; Alahyari, Abbas A. [United Technologies Research Center, East Hartford, Connecticut 06108 (United States)

    2016-01-15

    A novel numerical approach is developed to simulate the multiscale problem of pool-boiling phase change. The particular focus is to develop a simulation technique that is capable of predicting the heat transfer and hydrodynamic characteristics of nucleate boiling and the transition to critical heat flux on surfaces of arbitrary shape and roughness distribution addressing a critical need to design enhanced boiling heat transfer surfaces. The macro-scale of the phase change and bubble dynamics is addressed through employing off-the-shelf Computational Fluid Dynamics (CFD) methods for interface tracking and interphase mass and energy transfer. The micro-scale of the microlayer, which forms at early stage of bubble nucleation near the wall, is resolved through asymptotic approximation of the thin-film theory which provides a closed-form solution for the distribution of the micro-layer and its influence on the evaporation process. In addition, the sub-grid surface roughness is represented stochastically through probabilistic density functions and its role in bubble nucleation and growth is then represented based on the thermodynamics of nucleation process. This combination of deterministic CFD, local approximation, and stochastic representation allows the simulation of pool boiling on any surface with known roughness and enhancement characteristics. The numerical model is validated for dynamics and hydrothermal characteristics of a single nucleated bubble on a flat surface against available literature data. In addition, the prediction of pool-boiling heat transfer coefficient is verified against experimental measurements as well as reputable correlations for various roughness distributions and different surface orientations. Finally, the model is employed to demonstrate pool-boiling phenomenon on enhanced structures with reentrance cavities and to explore the effect of enhancement feature design on thermal and hydrodynamic characteristics of these surfaces.

  19. Saturated Pool Boiling in Vertical Annulus with Reduced Outflow Area

    International Nuclear Information System (INIS)

    Kang, Myeong Gie

    2012-01-01

    The mechanisms of pool boiling heat transfer have been studied extensively to design efficient heat transfer devices or to assure the integrity of safety related systems. However, knowledge on pool boiling heat transfer in a confined space is still quite limited. The confined nucleate boiling is an effective technique to enhance heat transfer. Improved heat transfer might be attributed to an increase in the heat transfer coefficient due to vaporization from the thin liquid film on the heating surface or increased bubble activity. According to Cornwell and Houston, the bubbles sliding on the heated surface agitate environmental liquid. In a confined space a kind of pulsating flow due to the bubbles is created and, as a result very active liquid agitation is generated. The increase in the intensity of liquid agitation results in heat transfer enhancement. Sometimes a deterioration of heat transfer appears at high heat fluxes for confined boiling. The cause of the deterioration is suggested as active bubble coalescence. Recently, Kang published inflow effects on pool boiling heat transfer in a vertical annulus with closed bottoms. Kang regulated the gap size at the upper regions of the annulus and identified that effects of the reduced gaps on heat transfer become evident as the heat flux increases. This kind of geometry is found in an in-pile test section. Since more detailed analysis is necessary, effects of the outflow area on nucleate pool boiling heat transfer are investigated in this study. Up to the author's knowledge, no previous results concerning to this effect have been published yet

  20. Experimental study of mass boiling in a porous medium model

    International Nuclear Information System (INIS)

    Sapin, Paul

    2014-01-01

    This manuscript presents a pore-scale experimental study of convective boiling heat transfer in a two-dimensional porous medium. The purpose is to deepen the understanding of thermohydraulics of porous media saturated with multiple fluid phases, in order to enhance management of severe accidents in nuclear reactors. Indeed, following a long-lasting failure in the cooling system of a pressurized water reactor (PWR) or a boiling water reactor (BWR) and despite the lowering of the control rods that stops the fission reaction, residual power due to radioactive decay keeps heating up the core. This induces water evaporation, which leads to the drying and degradation of the fuel rods. The resulting hot debris bed, comparable to a porous heat-generating medium, can be cooled down by reflooding, provided a water source is available. This process involves intense boiling mechanisms that must be modelled properly. The experimental study of boiling in porous media presented in this thesis focuses on the influence of different pore-scale boiling regimes on local heat transfer. The experimental setup is a model porous medium made of a bundle of heating cylinders randomly placed between two ceramic plates, one of which is transparent. Each cylinder is a resistance temperature detector (RTD) used to give temperature measurements as well as heat generation. Thermal measurements and high-speed image acquisition allow the effective heat exchanges to be characterized according to the observed local boiling regimes. This provides precious indications precious indications for the type of correlations used in the non-equilibrium macroscopic model used to model reflooding process. (author) [fr

  1. A high-fidelity approach towards simulation of pool boiling

    International Nuclear Information System (INIS)

    Yazdani, Miad; Radcliff, Thomas; Soteriou, Marios; Alahyari, Abbas A.

    2016-01-01

    A novel numerical approach is developed to simulate the multiscale problem of pool-boiling phase change. The particular focus is to develop a simulation technique that is capable of predicting the heat transfer and hydrodynamic characteristics of nucleate boiling and the transition to critical heat flux on surfaces of arbitrary shape and roughness distribution addressing a critical need to design enhanced boiling heat transfer surfaces. The macro-scale of the phase change and bubble dynamics is addressed through employing off-the-shelf Computational Fluid Dynamics (CFD) methods for interface tracking and interphase mass and energy transfer. The micro-scale of the microlayer, which forms at early stage of bubble nucleation near the wall, is resolved through asymptotic approximation of the thin-film theory which provides a closed-form solution for the distribution of the micro-layer and its influence on the evaporation process. In addition, the sub-grid surface roughness is represented stochastically through probabilistic density functions and its role in bubble nucleation and growth is then represented based on the thermodynamics of nucleation process. This combination of deterministic CFD, local approximation, and stochastic representation allows the simulation of pool boiling on any surface with known roughness and enhancement characteristics. The numerical model is validated for dynamics and hydrothermal characteristics of a single nucleated bubble on a flat surface against available literature data. In addition, the prediction of pool-boiling heat transfer coefficient is verified against experimental measurements as well as reputable correlations for various roughness distributions and different surface orientations. Finally, the model is employed to demonstrate pool-boiling phenomenon on enhanced structures with reentrance cavities and to explore the effect of enhancement feature design on thermal and hydrodynamic characteristics of these surfaces

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

    International Nuclear Information System (INIS)

    Giustini, G.

    2015-01-01

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

  3. Effect of changing from the National Health and Nutritional Examination Survey III spirometry reference range to that of the Global Lung Initiative 2012 at Gold Coast Hospital and Health Service.

    Science.gov (United States)

    Embling, Laura A K; Zagami, Debbie; Sriram, Krishna Bajee; Gordon, Robert J; Sivakumaran, Pathmanathan

    2016-12-01

    The categorisation of lung disease into obstructive ventilatory defect (OVD) and tendency to a restrictive ventilatory defect (TRVD) patterns using spirometry is used to guide both prognostication and treatment. The effectiveness of categorisation depends upon having reference ranges that accurately represent the population they describe. The Global Lung Initiative 2012 (GLI 2012) has spirometry reference ranges drawn from the largest sample size to date. This study aimed to determine whether using spirometry reference ranges from the new GLI 2012 dataset, compared to the previously used National Health and Nutritional Examination Survey III (NHANES III) dataset, resulted in a change in diagnosis between OVD, TRVD and normal ventilatory pattern (NVP). Spirometry data were collected from 301 patients, aged 18-80 years, undergoing investigation at the Gold Coast Hospital and Health Service (GCHHS) throughout February and March 2014. OVD was defined as a forced expiratory volume in 1 second (FEV 1 ) divided by forced vital capacity (FVC) less than lower limit of normal (LLN). TRVD was defined as FEV 1 /FVC ≥ LLN, FEV 1 reference range resulted in a change in diagnosis of lung disease in 5.9% of the individuals included in this study. This variance in diagnosis when changing reference ranges should be taken into account by clinicians as it may affect patient management.

  4. Dimensional analysis of boiling heat transfer burnout conditions

    International Nuclear Information System (INIS)

    El-Mitwally, E.S.; Raafat, N.M.; Darwish, M.A.

    1979-01-01

    The first criteria in boiling water systems design, such as boiling water reactors, is that no burnout in the core is allowed to exist under any conditions of the reactor operation either during steady state operation or during any of the several postulated accidental transients, such as sudden interruption of coolant flow in the reactor core (due to pump failure or blockage of fuel channel). The aim of the present work is to obtain a correlation for the critical heat flux based on a theoretical study where the mechanism of burn out and the related hydrodynamic and heat transfer equations are considered. 8 refs

  5. Hysteresis of boiling for different tunnel-pore surfaces

    Directory of Open Access Journals (Sweden)

    Pastuszko Robert

    2015-01-01

    Full Text Available Analysis of boiling hysteresis on structured surfaces covered with perforated foil is proposed. Hysteresis is an adverse phenomenon, preventing high heat flux systems from thermal stabilization, characterized by a boiling curve variation at an increase and decrease of heat flux density. Experimental data were discussed for three kinds of enhanced surfaces: tunnel structures (TS, narrow tunnel structures (NTS and mini-fins covered with the copper wire net (NTS-L. The experiments were carried out with water, R-123 and FC-72 at atmospheric pressure. A detailed analysis of the measurement results identified several cases of type I, II and III for TS, NTS and NTS-L surfaces.

  6. 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)

  7. Simultaneous neutron radiography and infrared thermography measurement of boiling processes

    International Nuclear Information System (INIS)

    Murphy, J.H.; Glickstein, S.S.

    1997-01-01

    Boiling of water at 1 to 15 bar flowing upward within a narrow duct and a round test section was observed using both neutron radiography and infrared (IR) thermography. The IR readings of the test section outer wall temperatures show the effects of both fluid temperature and wall heat transfer coefficient variations, producing a difference between liquid and two phase regions. The IR images, in fact, appear very similar to the neutron images; both show clear indications of spatial and temporal variations in the internal fluid conditions during the boiling process

  8. The sudden coalescene model of the boiling crisis

    International Nuclear Information System (INIS)

    Carrica, P.M.; Clausse, A.

    1995-01-01

    A local two-phase flow integral model of nucleate boiling and crisis is presented. The model is based on average balances on a control volume, yielding to a set of three nonlinear differential equations for the local void fraction, bubble number density and velocity. Boiling crisis as critical heat flux is interpreted as a dynamic transition caused by the coalescence of bubbles near the heater. The theoretical dynamic model is compared with experimental results obtained for linear power ramps in a horizontal plate heater in R-113, showing an excellent qualitative agreement

  9. On Boiling of Crude Oil under Elevated Pressure

    Science.gov (United States)

    Pimenova, Anastasiya V.; Goldobin, Denis S.

    2016-02-01

    We construct a thermodynamic model for theoretical calculation of the boiling process of multicomponent mixtures of hydrocarbons (e.g., crude oil). The model governs kinetics of the mixture composition in the course of the distillation process along with the boiling temperature increase. The model heavily relies on the theory of dilute solutions of gases in liquids. Importantly, our results are applicable for modelling the process under elevated pressure (while the empiric models for oil cracking are not scalable to the case of extreme pressure), such as in an oil field heated by lava intrusions.

  10. 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

  11. The sudden coalescene model of the boiling crisis

    Energy Technology Data Exchange (ETDEWEB)

    Carrica, P.M.; Clausse, A. [Centro Atomico Bariloche and Instituto Balseiro, Bariloche (Argentina)

    1995-09-01

    A local two-phase flow integral model of nucleate boiling and crisis is presented. The model is based on average balances on a control volume, yielding to a set of three nonlinear differential equations for the local void fraction, bubble number density and velocity. Boiling crisis as critical heat flux is interpreted as a dynamic transition caused by the coalescence of bubbles near the heater. The theoretical dynamic model is compared with experimental results obtained for linear power ramps in a horizontal plate heater in R-113, showing an excellent qualitative agreement.

  12. 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)

  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. 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

  15. 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)

  16. 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))

  17. Variation of the Effectiveness of Hydrogen Water Chemistry in a Boiling Water Reactor during Startup Operations

    International Nuclear Information System (INIS)

    Yeh, Tsung-Kuang; Wang, Mei-Ya

    2012-09-01

    For mitigating intergranular stress corrosion cracking (IGSCC) in an operating boiling water reactor (BWR), the technology of hydrogen water chemistry (HWC) aiming at coolant chemistry improvement has been adopted worldwide. However, the hydrogen injection system employed in this technology was designed to operate only at power levels greater than 30% of the rated power or at coolant temperatures of greater than 450 deg. F. This system is usually in an idle and standby mode during a startup operation. The coolant in a BWR during a cold shutdown normally contains a relatively high level of dissolved oxygen from intrusion of atmospheric air. Accordingly, the structural materials in the primary coolant circuit (PCC) of a BWR could be exposed to a strongly oxidizing environment for a short period of time during a subsequent startup operation. At some plants, the feasibility of hydrogen water chemistry during startup operations has been studied, and its effectiveness on suppressing SCC initiation was evaluated. It is technically difficult to directly procure water chemistry data at various locations of an operating reactor. Accordingly, the impact of startup operation on water chemistry in the PCC of a BWR operating under normal water chemistry (NWC) or HWC can only be theoretically evaluated through computer modelling. In this study, a well-developed computer code DEMACE was used to investigate the variations in redox species concentration and in electrochemical corrosion potential (ECP) of components in the PCC of a domestic BWR during startup operations in the presence of HWC. Simulations were carried out for [H2] FW s ranging from 0.0 to 2.0 parts per million (ppm) and for power levels ranging from 2.5% to 11.3% during startup operations. Our analyses indicated that for power levels with steam generation in the core, a higher power level would tend to promote a more oxidizing coolant environment for the structural components and therefore lead to less HWC

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  19. 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

  20. Science 101: Why Does It Take Longer to Boil Potatoes at High Altitudes?

    Science.gov (United States)

    Robertson, Bill

    2017-01-01

    Why Does It Take Longer to Boil Potatoes at High Altitudes? This column provides background science information for elementary teachers. This month's issue looks at why water boils at different temperatures at different altitudes.

  1. Charge-Discharge Properties of the Surface-Modified ZrNi Alloy Electrode with Different Degrees of Boiling Alkaline Treatment

    Directory of Open Access Journals (Sweden)

    Akihiro Matsuyama

    2016-09-01

    Full Text Available Charge-discharge properties of the surface-modified ZrNi negative electrodes with different degrees of boiling alkaline treatment were investigated. The boiling alkaline treatment was performed by immersing the ZrNi electrode in a boiling 6 M KOH aqueous solution for 2 h or 4 h. The initial discharge capacity for the untreated ZrNi negative electrode was 21 mAh·g−1, but it was increased to 114 mAh·g−1 and 308 mAh·g−1 after the boiling alkaline treatments for 2 h and 4 h, respectively. The discharge capacity for the ZrNi negative electrode after the treatment for 2 h steadily increased with repeating charge-discharge cycles as well as that of the untreated electrode, whereas that for the ZrNi negative electrode after the 4 h treatment greatly decreased. The high rate of dischargeability was improved with an increase in the treatment period of time, and the charge-transfer resistance was drastically decreased. Scanning electron microscopy (SEM and electron dispersive X-ray spectroscopy demonstrated the ZrO2 passive layer on the ZrNi alloy surface was removed by the boiling alkaline treatment to form a porous morphology containing Ni(OH2, which can be reduced to Ni during charging, leading to the reduction of a barrier for the charge-discharge reactions.

  2. 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.

  3. Effect of Steaming and Boiling on the Antioxidant Properties and Biogenic Amines Content in Green Bean (Phaseolus vulgaris Varieties of Different Colours

    Directory of Open Access Journals (Sweden)

    Raffaella Preti

    2017-01-01

    Full Text Available Effects of boiling and steaming cooking methods were studied on total polyphenols, antioxidant capacity, and biogenic amines of three green bean varieties, purple, yellow, and green. The vegetables gave good values both for antioxidant capacity and for phenolics content, with the purple variety being the richest in healthful components. Both the heat treatments affected the antioxidant properties of these vegetables, with boiling that reduced the initial antioxidant capacity till 30% in the yellow variety, having the same trend for total polyphenols, with the major decrement of 43% in the green variety. On the contrary, biogenic amines significantly increased only after boiling in green and yellow variety, while purple variety did not show any changes in biogenic amines after cooking. The steaming method showed being better cooking approach in order to preserve the antioxidant properties of green beans varieties and to maintain the biogenic amines content at the lowest level.

  4. 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)

  5. Technology, safety and costs of decommissioning a Reference Boiling Water Reactor Power Station. Main report. Volume 1

    Energy Technology Data Exchange (ETDEWEB)

    Oak, H.D.; Holter, G.M.; Kennedy, W.E. Jr.; Konzek, G.J.

    1980-06-01

    Technology, safety and cost information is given for the conceptual decommissioning of a large (1100MWe) boiling water reactor (BWR) power station. Three approaches to decommissioning, immediate dismantlement, safe storage with deferred dismantlement and entombment, were studied to obtain comparisons between costs, occupational radiation doses, potential dose to the public and other safety impacts. It also shows the sensitivity of decommissioning safety and costs to the power rating of a BWR in the range of 200 to 1100 MWe.

  6. 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)

  7. Gravity influence on heat transfer rate in flow boiling

    NARCIS (Netherlands)

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

    2012-01-01

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

  8. 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)

  9. Delays due to gas diffusion in flash boiling nucleation

    International Nuclear Information System (INIS)

    Hanbury, W.T.; McCartney, W.S.

    1976-01-01

    A theoretical model to account for the time delay between decompression and nucleation in flash boiling is presented and analyzed. It shows that gas diffusion can be responsible for delayed nucleation when the critical radius for nucleation and the suspended particle size are of the same order of magnitude

  10. A highly stable microchannel heat sink for convective boiling

    International Nuclear Information System (INIS)

    Lu, Chun Ting; Pan Chin

    2009-01-01

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

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

    International Nuclear Information System (INIS)

    Rachedi, M.

    1986-01-01

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

  12. How long does it take to boil an egg? Revisited

    International Nuclear Information System (INIS)

    Buay, D; Foong, S K; Kiang, D; Kuppan, L; Liew, V H

    2006-01-01

    How long does it take to boil an egg? Theoretical prediction, based on a simple adaptation of the solution to the exact thermal diffusion equation for a sphere, is consistent with experiments. The experimental data are also used to estimate an average value for the thermal diffusivity of an egg

  13. Radioactive waste management practices with KWU-boiling water reactors

    International Nuclear Information System (INIS)

    Queiser, H.

    1976-01-01

    A Kraftwerk Union boiling water reactor is used to demonstrate the reactor auxiliary systems which are applied to minimize the radioactive discharge. Based on the most important design criteria the philosophy and function of the various systems for handling the off-gas, ventilation air, waste water and concentrated waste are described. (orig.) [de

  14. Geologic Controls of Sand Boil Formation at Buck Chute, Mississippi

    Science.gov (United States)

    2017-06-30

    26 5.1 Electrical resistivity tomography ...construction at Buck Chute in preparation to the 2011 Flood. ERDC/GSL TR-17-12 21 4 Methods 4.1 Electrical resistivity tomography The resistivity ...could contribute to sand boil formation is missed. Electrical Resistivity Tomography (ERT) is a subsurface investigation method that combines the

  15. Flow Boiling Critical Heat Flux in Reduced Gravity

    Science.gov (United States)

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

    2004-01-01

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

  16. Glycaemic Index Of Boiled Cocoyam And Stew | Alegbejo | Sahel ...

    African Journals Online (AJOL)

    Cocoyam can be processed in several ways. It contains digestible starch, protein and other valuable nutrients. Consumption of cocoyam is very high all over Nigeria. This study was undertaken to determine the glycaemic response of diabetic and healthy subjects to equal amounts of carbohydrate in the form of boiled ...

  17. How long does it take to boil an egg? Revisited

    Energy Technology Data Exchange (ETDEWEB)

    Buay, D [Natural Sciences and Science Education, National Institute of Education, Nanyang Technological University, 1, Nanyang Walk, Singapore 637616 (Singapore); Foong, S K [Natural Sciences and Science Education, National Institute of Education, Nanyang Technological University, 1, Nanyang Walk, Singapore 637616 (Singapore); Kiang, D [Department of Physics, Chinese University of Hong Kong, Shatin, New Territories, Hong Kong (China); Kuppan, L [Natural Sciences and Science Education, National Institute of Education, Nanyang Technological University, 1, Nanyang Walk, Singapore 637616 (Singapore); Centre for Research in Pedagogy and Practice, National Institute of Education, Nanyang Technological University, 1, Nanyang Walk, Singapore 637616 (Singapore); Liew, V H [Natural Sciences and Science Education, National Institute of Education, Nanyang Technological University, 1, Nanyang Walk, Singapore 637616 (Singapore)

    2006-01-01

    How long does it take to boil an egg? Theoretical prediction, based on a simple adaptation of the solution to the exact thermal diffusion equation for a sphere, is consistent with experiments. The experimental data are also used to estimate an average value for the thermal diffusivity of an egg.

  18. Statistical analysis of the BOIL program in RSYST-III

    International Nuclear Information System (INIS)

    Beck, W.; Hausch, H.J.

    1978-11-01

    The paper describes a statistical analysis in the RSYST-III program system. Using the example of the BOIL program, it is shown how the effects of inaccurate input data on the output data can be discovered. The existing possibilities of data generation, data handling, and data evaluation are outlined. (orig.) [de

  19. Investigation Status of Heat Exchange while Boiling Hydrocarbon Fuel

    Directory of Open Access Journals (Sweden)

    D. S. Obukhov

    2006-01-01

    Full Text Available The paper contains analysis of heat exchange investigations while boiling hydrocarbon fuel. The obtained data are within the limits of the S.S. Kutateladze dependence proposed in 1939. Heat exchange at non-stationary heat release has not been investigated. The data for hydrocarbon fuel with respect to critical density of heat flow are not available even for stationary conditions.

  20. Effects of Fermentation, Boiling and Roasting on Some ...

    African Journals Online (AJOL)

    The effects of processing methods such as fermentation, boiling and roasting on some micronutrients and antinutrient composition of jackfruit seed flour were evaluated. The mineral, vitamin and antinutrient composition of raw and processed jackfruit seed flours were determined using standard methods. Iron, calcium and ...

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

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  2. Effect of boiling and roasting on the proximate properties of ...

    African Journals Online (AJOL)

    Administrator

    2011-09-19

    Sep 19, 2011 ... Boiling is a common food processing method. As a thermal process .... Mean ± SEM for duplicate analysis; *the mean difference is significant at the 0.05 level. ... suggesting similar processing influence on the protein content of ...

  3. 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

  4. Boiling process in oil coolers on porous elements

    Directory of Open Access Journals (Sweden)

    Genbach Alexander A.

    2016-01-01

    Full Text Available Holography and high-speed filming were used to reveal movements and deformations of the capillary and porous material, allowing to calculate thermo-hydraulic characteristics of boiling liquid in the porous structures. These porous structures work at the joint action of capillary and mass forces, which are generalised in the form of dependences used in the calculation for oil coolers in thermal power plants (TPP. Furthermore, the mechanism of the boiling process in porous structures in the field of mass forces is explained. The development process of water steam formation in the mesh porous structures working at joint action of gravitational and capillary forces is investigated. Certain regularities pertained to the internal characteristics of boiling in cells of porous structure are revealed, by means of a holographic interferometry and high-speed filming. Formulas for calculation of specific thermal streams through thermo-hydraulic characteristics of water steam formation in mesh structures are obtained, in relation to heat engineering of thermal power plants. This is the first calculation of heat flow through the thermal-hydraulic characteristics of the boiling process in a reticulated porous structure obtained by a photo film and holographic observations.

  5. 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

  6. Mark I 1/5-scale boiling water reactor pressure suppression experiment facility report

    International Nuclear Information System (INIS)

    Altes, R.G.; Pitts, J.H.; Ingraham, R.F.; Collins, E.K.; McCauley, E.W.

    1977-01-01

    An accurate Mark I 1 / 5 -scale, boiling water reactor (BWR), pressure suppression facility was designed and constructed at Lawrence Livermore Laboratory (LLL) in 11 months. Twenty-seven air tests using the facility are described. Cost was minimized by utilizing equipment borrowed from other LLL programs. The total value of borrowed equipment exceeded the program's budget of $2,020,000. Substantial flexibility in the facility was used to permit independent variation in the drywell pressure-time history, initial pressure in the drywell and toroidal wetwells, initial toroidal wetwell water level and downcomer length, vent line flow resistance, and vent line flow asymmetry. The two- and three-dimensional sectors of the toroidal wetwell provided significant data

  7. 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

  8. 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

  9. Numerical Investigation of Microgravity Tank Pressure Rise Due to Boiling

    Science.gov (United States)

    Hylton, Sonya; Ibrahim, Mounir; Kartuzova, Olga; Kassemi, Mohammad

    2015-01-01

    The ability to control self-pressurization in cryogenic storage tanks is essential for NASAs long-term space exploration missions. Predictions of the tank pressure rise in Space are needed in order to inform the microgravity design and optimization process. Due to the fact that natural convection is very weak in microgravity, heat leaks into the tank can create superheated regions in the liquid. The superheated regions can instigate microgravity boiling, giving rise to pressure spikes during self-pressurization. In this work, a CFD model is developed to predict the magnitude and duration of the microgravity pressure spikes. The model uses the Schrage equation to calculate the mass transfer, with a different accommodation coefficient for evaporation at the interface, condensation at the interface, and boiling in the bulk liquid. The implicit VOF model was used to account for the moving interface, with bounded second order time discretization. Validation of the models predictions was carried out using microgravity data from the Tank Pressure Control Experiment, which flew aboard the Space Shuttle Mission STS-52. Although this experiment was meant to study pressurization and pressure control, it underwent boiling during several tests. The pressure rise predicted by the CFD model compared well with the experimental data. The ZBOT microgravity experiment is scheduled to fly on February 2016 aboard the ISS. The CFD model was also used to perform simulations for setting parametric limits for the Zero-Boil-Off Tank (ZBOT) Experiments Test Matrix in an attempt to avoid boiling in the majority of the test runs that are aimed to study pressure increase rates during self-pressurization. *Supported in part by NASA ISS Physical Sciences Research Program, NASA HQ, USA

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-06-15

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

  11. 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.

  12. 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

  13. 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.

  14. Use of ultrasonic waves in sub-cooled boiling

    International Nuclear Information System (INIS)

    Bartoli, Carlo; Baffigi, Federica

    2012-01-01

    This work focuses on the use of ultrasounds in heat transfer fields. Under particular conditions, ultrasonic waves induce a convection coefficient increase. This initial research work, indicates that there are some practical applications in the cooling of the latest generation electronic components. In the first part of this paper, some background on this subject is reported. The ultrasound's influence on heat transfer rate has been observed since the 60's: different authors studied the cooling effect due to ultrasonic waves from different heat transfer regimes. The most investigated configuration was a thin platinum wire immersed in water. Later, a bibliographic research on possible practical applications of ultrasounds was carried out. This research focused in particular on the issue for 3D highly integrated electronic components. For these systems the thermal problem is a major challenge, because they cannot exceed critical temperatures, after which they could be damaged irreversibly. On the basis of our experimental results, ultrasounds could represent a valid means to overcome these thermal problems. Finally, the paper presents a series of experiments performed in the Thermal-Fluid- Dynamic Lab at the Energy and Engineering Systems Department of University of Pisa. The experiments provide systematic evidence of ultrasonic waves effects, on free convection heat transfer, from a heated circular cylinder to sub-cooled water, at atmospheric pressure. Many variables involved in the heat transfer rise were tested, for example: the ultrasonic generator's power, the position of the heater inside the ultrasonic tank, the variation of the water sub-cooling degree, as function of the heat flux needed dissipating. The aim of the experiment was to find out the set of optimal conditions, in order to successively apply all the results to real packaging systems, as mentioned before. The maximum increase in the heat transfer coefficient, due to ultrasonic waves, was 57

  15. MRI monitoring of lesions created at temperature below the boiling point and of lesions created above the boiling point using high intensity focused ultrasound

    OpenAIRE

    Damianou, C.; Ioannides, K.; Hadjisavvas, V.; Mylonas, N.; Couppis, A.; Iosif, D.; Kyriacou, P. A.

    2010-01-01

    Magnetic Resonance Imaging (MRI) was utilized to monitor lesions created at temperature below the boiling point and lesions created at temperature above the boiling point using High Intensity Focused Ultrasound (HIFU) in freshly excised kidney, liver and brain and in vivo rabbit kidney and brain. T2-weighted fast spin echo (FSE) was proven as an excellent MRI sequence that can detect lesions with temperature above the boiling point in kidney. This advantage is attributed to the significant di...

  16. An assessment of in-tube flow boiling correlations for ammonia-water mixtures and their influence on heat exchanger size

    DEFF Research Database (Denmark)

    Kærn, Martin Ryhl; Modi, Anish; Jensen, Jonas Kjær

    2016-01-01

    on the required heat exchanger size (surface area)is investigated during numerical design. For this purpose, two case studies related to the use of the Kalina cycle are considered: a flue gas based heat recovery boiler for acombined cycle power plant and a hot oil based boiler for a solar thermal power plant......Heat transfer correlations for pool and flow boiling are indispensable for boiler design. The correlations for predicting in-tube flow boiling heat transfer ofammonia-water mixtures are not well established in the open literature and there is a lack of experimental measurements for the full range...... of composition, vapor qualities, fluid conditions, etc. This paper presents a comparison of several flow boiling heat transfer prediction methods (correlations) for ammonia-water mixtures. Firstly, these methods are reviewed and compared at various fluid conditions. The methods include: (1) the ammonia...

  17. Pool boiling performance of Novec{sup TM} 649 engineered fluid

    Energy Technology Data Exchange (ETDEWEB)

    Forrest, Eric; Buongiorno, Jacopo; McKrell, Thomas [Massachusetts Institute of Technology, Cambridge (United States). Dept. of Nuclear Science and Engineering; Hu, Lin-Wen [Massachusetts Institute of Technology, Cambridge (United States). Nuclear Reactor Lab.], e-mail: lwhu@mit.edu

    2009-07-01

    A new fluorinated ketone, C{sub 2}F{sub 5}C(O)CF(CF{sub 3}){sub 2}, is currently being considered as an environmentally friendly alternative for power electronics cooling applications due to its high dielectric strength and low global warming potential (GWP). Sold commercially by the 3M Company as Novec{sup TM} 649 Engineered Fluid, C{sub 2}F{sub 5}C(O)CF(CF{sub 3}){sub 2} exhibits very low acute toxicity while maintaining long-term stability. To assess the general two-phase heat transfer performance of Novec{sup TM} 649, pool boiling tests were conducted by resistively heating a 0.01 in. diameter nickel wire at the fluid's atmospheric saturation temperature of 49 deg C. The nucleate boiling heat transfer coefficient and critical heat flux (CHF) obtained for the fluorinated ketone compare favorably with results obtained for FC-72, a fluorocarbon widely used for the direct cooling of electronic devices. Initial results indicate that Novec{sup TM} 649 may prove to be a viable alternative to FC-72 and other halo alkanes for the cooling of high power density electronic devices. (author)

  18. Study of heat transfer in the heating wall during nucleate pool boiling

    International Nuclear Information System (INIS)

    Bergez, W.

    1991-12-01

    The subject of this these is to show the role of heat transfer in the wall during saturated pool boiling. This effect, usually neglected in the modelizations of boiling, can explain some behaviours of the ebullition cycle and of the activities of nucleation sites. Il has been found that the ebullition cycle can be described by two steps: (1) during bubble growth, the wall temperature decreases due to the evaporation of the micro-layer at the base of the bubble; (2) initial superheat is re-established mainly by radial heat conduction in the wall. It is then possible to account for the variations of the wall temperature displayed by liquid crystals put a the bottom of the heating surface, and for the influence of the contact angle on the heat transfer. In the case of the infinitely thick wall the main results are that the thermal transfer during the growth of the bubble depends on the thermal properties of both wall and liquid and that the time separating the detachment of a bubble and its replacement by a new one is proportional to the cross-section of the bubble and to the thermal diffusivity of the wall

  19. 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)

  20. Boiling and condensation in microfin tubes

    Science.gov (United States)

    Schlager, Lynn M.

    A general overview of microfin tubes and their applications is presented. Manufacturing processes, commercial availability, experimental heat transfer, and pressure drop data for various refrigerants (including alternative refrigerants and refrigerant-oil mixtures), physical mechanisms of enhancement, and the incorporation of microfin tubes in common heat exchanger configurations are discussed. Microfin tubes, also known by various trade names, are characterized by numerous small fins which typically spiral down the inside wall of tubes at angles ranging from 10 to 30 degrees. The number of fins ranges from 48 to 70 with typical fin heights of 0.12 to 0.30 mm (fin height generally less than 3 percent of the inside diameter of the tube). Fin shapes may vary and the inside surface area of microfin tubes is 10 to 70 percent greater than the area of equivalent smooth tubes. Heat transfer can be enhanced by up to a factor of three with microfin tubes.

  1. Thermal behavior in the transition region between nucleate and film boiling

    International Nuclear Information System (INIS)

    Adiutori, E.F.

    1991-01-01

    The prediction of post Critical Heat Flux (CHF) behavior is complicated by the highly nonlinear thermal behavior of boiling interfaces--ie by the nonlinear nature of the boiling curve. Nonlinearity in the boiling curve can and does cause thermal instability, resulting in temperature discontinuities. Thus the prediction of post CHF behavior requires the analysis of thermal stability. This in turn requires an accurate description of thermal behavior in transition boiling. This paper determines thermal behavior in transition boiling by analysis of literature data. It also describes design features which improve post CHF performance and are reported in the literature

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  3. Gas chromatographic simulated distillation-mass spectrometry for the determination of the boiling point distributions of crude oils

    Science.gov (United States)

    Roussis; Fitzgerald

    2000-04-01

    The coupling of gas chromatographic simulated distillation with mass spectrometry for the determination of the distillation profiles of crude oils is reported. The method provides the boiling point distributions of both weight and volume percent amounts. The weight percent distribution is obtained from the measured total ion current signal. The total ion current signal is converted to weight percent amount by calibration with a reference crude oil of a known distillation profile. Knowledge of the chemical composition of the crude oil across the boiling range permits the determination of the volume percent distribution. The long-term repeatability is equivalent to or better than the short-term repeatability of the currently available American Society for Testing and Materials (ASTM) gas chromatographic method for simulated distillation. Results obtained by the mass spectrometric method are in very good agreement with results obtained by conventional methods of physical distillation. The compositional information supplied by the method can be used to extensively characterize crude oils.

  4. Comparative performance of five Mexican plancha-type cookstoves using water boiling tests

    Directory of Open Access Journals (Sweden)

    Paulo Medina

    Full Text Available While plancha-type cookstoves are very popular and widely disseminated in Latin America, few peer review articles exist documenting their detailed technical performance. In this paper we use the standard Water Boiling Tests (WBT to assess the energy and emission performance of five plancha-type cookstoves disseminated in about 450 thousand Mexican rural homes compared to the traditional 3-stone fire (TSF. In the high-power phase, average modified combustion efficiencies (MCE for plancha-type stoves were 97±1% which was higher than TSF 93±4%, and reductions in CO and PM2.5 total emissions were on average 44%. Time to boil and specific fuel consumption, however, were increased in plancha-type stoves compared to the open fire as a result of the reduced overall thermal efficiency of the plancha during WBT. In the simmering phase, plancha-type stoves showed much more consistent performance reductions compared to the TSF. MCE for plancha stoves were on average 98±1% and 95±3% for the TSF, while reductions in CO and PM2.5 total emissions were on average 55%. In this phase 27% average savings in fuel use are achieved by plancha-type stoves. Removal of the plancha rings resulted in savings of specific fuel consumption (SFC, thermal efficiency (TE, and time to boil; however, CO and PM2.5 emissions increased significantly as flue air is drawn through the comal surface rather than through the combustion zone, resulting in suboptimal combustion conditions.International Workshop Agreement (IWA energy performance Tiers for plancha-type stoves ranged from 0 to 1. However, these results contrast sharply with the well documented reductions in fuel consumption during daily cooking activities achieved by these stoves. IWA indoor emissions Tiers are 4 for both PM2.5 and CO using locally measured values for fugitive emissions. Optimization of combustion chamber design on these stoves in Mexico is desirable to further reduce indoor emissions and to reduce the

  5. Prostate Health Index (Phi) and Prostate Cancer Antigen 3 (PCA3) significantly improve prostate cancer detection at initial biopsy in a total PSA range of 2-10 ng/ml.

    Science.gov (United States)

    Ferro, Matteo; Bruzzese, Dario; Perdonà, Sisto; Marino, Ada; Mazzarella, Claudia; Perruolo, Giuseppe; D'Esposito, Vittoria; Cosimato, Vincenzo; Buonerba, Carlo; Di Lorenzo, Giuseppe; Musi, Gennaro; De Cobelli, Ottavio; Chun, Felix K; Terracciano, Daniela

    2013-01-01

    Many efforts to reduce prostate specific antigen (PSA) overdiagnosis and overtreatment have been made. To this aim, Prostate Health Index (Phi) and Prostate Cancer Antigen 3 (PCA3) have been proposed as new more specific biomarkers. We evaluated the ability of phi and PCA3 to identify prostate cancer (PCa) at initial prostate biopsy in men with total PSA range of 2-10 ng/ml. The performance of phi and PCA3 were evaluated in 300 patients undergoing first prostate biopsy. ROC curve analyses tested the accuracy (AUC) of phi and PCA3 in predicting PCa. Decision curve analyses (DCA) were used to compare the clinical benefit of the two biomarkers. We found that the AUC value of phi (0.77) was comparable to those of %p2PSA (0.76) and PCA3 (0.73) with no significant differences in pairwise comparison (%p2PSA vs phi p = 0.673, %p2PSA vs. PCA3 p = 0.417 and phi vs. PCA3 p = 0.247). These three biomarkers significantly outperformed fPSA (AUC = 0.60), % fPSA (AUC = 0.62) and p2PSA (AUC = 0.63). At DCA, phi and PCA3 exhibited a very close net benefit profile until the threshold probability of 25%, then phi index showed higher net benefit than PCA3. Multivariable analysis showed that the addition of phi and PCA3 to the base multivariable model (age, PSA, %fPSA, DRE, prostate volume) increased predictive accuracy, whereas no model improved single biomarker performance. Finally we showed that subjects with active surveillance (AS) compatible cancer had significantly lower phi and PCA3 values (pphi and PCA3 comparably increase the accuracy in predicting the presence of PCa in total PSA range 2-10 ng/ml at initial biopsy, outperforming currently used %fPSA.

  6. 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.

  7. 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

  8. Acoustic analysis of sodium boiling stability tests using THORS bundle 6A

    International Nuclear Information System (INIS)

    Sheen, S.H.; Bobis, J.P.; Carey, W.M.

    1977-01-01

    Acoustic data from boiling stability tests on the THORS (Thermal-Hydraulic Out-of-Reactor Safety) facility are presented and discussed. The THORS sodium loop is a high temperature test facility that contains the bundle 6A, a full length stimulated fuel subassembly with nineteen electrically heated pins. Boiling stability tests on the THORS facility were designed to determine if a stable boiling region exists during the thermal hydraulic test at normal and off-normal conditions. Boiling was observed and the stable boiling region was determined. The acoustic data observed by three ANL sodium-immersible microphones have provided the following information: (1) the boiling signal is clearly observed and shows a correlation with the inlet flow fluctuations; (2) the signal level and the repetition rate of the boiling signal are directly related to the applied heat flux; (3) a typical boiling pulse consists of a high frequency signal due mainly to the bubble collapse and a low frequency (approximately 75 Hz) void oscillation; (4) a boiling pulse yields a frequency spectrum with significant amplitudes up to 80 KHz as compared with 4 KHz for background pulses; and (5) the frequency content of a boiling pulse can be mostly explained in terms of various resonance frequencies of the loop. The characterization of these data is pertinent to the design of sodium boiling detection systems

  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. 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

  11. Study of sodium boiling during a power transient in a heating duct. Investigation carried out within the scope of a study on the safety of fast neutron reactors

    International Nuclear Information System (INIS)

    Seiler, J.-M.

    1977-01-01

    This work comprises an experimental study and a physical and theoretical interpretation of sodium boiling during a power transient in conditions simulating a power excursion accident in a breeder reactor. The experimental study was performed on an 'out of pile' system with forced sodium convection. It enabled a detailed study to be made of the draining of the heating duct representing a part of a reactor assembly (expansion of the double-phase area on the heating part of the pin). The physical and theoretical interpretation rests on a very simple calculation model solely describing the heat transfers in the duct, assuming that the fluid stays liquid. The extension of the boiling is governed by the initial overheating and the boiling conditioned by thermal inertias of the walls. The extent of the overheating appears to depend on the degree of initial under-saturation. In the absence of overheating, the rate at which the double-phase front moves is controlled by the saturated boiling. This front is well localised when the central temperature gradient is significant. The draining rate is then directly proportional to the growth rate of the temperature in the liquid. The model employed in the CASPAR code for describing power excursion experiments in single liquid phase is described. This code makes it possible to calculate the heat transfers occurring in a forced system by a heating part, a coolant in movement and possibly a part representing an outer cover [fr

  12. Comparison of Boiling and Robotics Automation Method in DNA Extraction for Metagenomic Sequencing of Human Oral Microbes.

    Science.gov (United States)

    Yamagishi, Junya; Sato, Yukuto; Shinozaki, Natsuko; Ye, Bin; Tsuboi, Akito; Nagasaki, Masao; Yamashita, Riu

    2016-01-01

    The rapid improvement of next-generation sequencing performance now enables us to analyze huge sample sets with more than ten thousand specimens. However, DNA extraction can still be a limiting step in such metagenomic approaches. In this study, we analyzed human oral microbes to compare the performance of three DNA extraction methods: PowerSoil (a method widely used in this field), QIAsymphony (a robotics method), and a simple boiling method. Dental plaque was initially collected from three volunteers in the pilot study and then expanded to 12 volunteers in the follow-up study. Bacterial flora was estimated by sequencing the V4 region of 16S rRNA following species-level profiling. Our results indicate that the efficiency of PowerSoil and QIAsymphony was comparable to the boiling method. Therefore, the boiling method may be a promising alternative because of its simplicity, cost effectiveness, and short handling time. Moreover, this method was reliable for estimating bacterial species and could be used in the future to examine the correlation between oral flora and health status. Despite this, differences in the efficiency of DNA extraction for various bacterial species were observed among the three methods. Based on these findings, there is no "gold standard" for DNA extraction. In future, we suggest that the DNA extraction method should be selected on a case-by-case basis considering the aims and specimens of the study.

  13. Comparison of Boiling and Robotics Automation Method in DNA Extraction for Metagenomic Sequencing of Human Oral Microbes.

    Directory of Open Access Journals (Sweden)

    Junya Yamagishi

    Full Text Available The rapid improvement of next-generation sequencing performance now enables us to analyze huge sample sets with more than ten thousand specimens. However, DNA extraction can still be a limiting step in such metagenomic approaches. In this study, we analyzed human oral microbes to compare the performance of three DNA extraction methods: PowerSoil (a method widely used in this field, QIAsymphony (a robotics method, and a simple boiling method. Dental plaque was initially collected from three volunteers in the pilot study and then expanded to 12 volunteers in the follow-up study. Bacterial flora was estimated by sequencing the V4 region of 16S rRNA following species-level profiling. Our results indicate that the efficiency of PowerSoil and QIAsymphony was comparable to the boiling method. Therefore, the boiling method may be a promising alternative because of its simplicity, cost effectiveness, and short handling time. Moreover, this method was reliable for estimating bacterial species and could be used in the future to examine the correlation between oral flora and health status. Despite this, differences in the efficiency of DNA extraction for various bacterial species were observed among the three methods. Based on these findings, there is no "gold standard" for DNA extraction. In future, we suggest that the DNA extraction method should be selected on a case-by-case basis considering the aims and specimens of the study.

  14. Underwater Ranging

    OpenAIRE

    S. P. Gaba

    1984-01-01

    The paper deals with underwater laser ranging system, its principle of operation and maximum depth capability. The sources of external noise and methods to improve signal-to-noise ratio are also discussed.

  15. Experimental study on flow boiling heat transfer of LNG in a vertical smooth tube

    Science.gov (United States)

    Chen, Dongsheng; Shi, Yumei

    2013-10-01

    An experimental apparatus is set up in this work to study the upward flow boiling heat transfer characteristics of LNG (liquefied natural gas) in vertical smooth tubes with inner diameters of 8 mm and 14 mm. The experiments were performed at various inlet pressures from 0.3 to 0.7 MPa. The results were obtained over the mass flux range from 16 to 200 kg m-2 s-1 and heat fluxes ranging from 8.0 to 32 kW m-2. The influences of quality, heat flux and mass flux, tube diameter on the heat transfer characteristic are examined and discussed. The comparisons of the experimental heat transfer coefficients with the predicted values from the existing correlations are analyzed. The correlation by Zou et al. [16] shows the best accuracy with the RMS deviation of 31.7% in comparison with the experimental data.

  16. 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

  17. 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

  18. 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

  19. Bubble nucleation of R134A refrigerant in a pressurized flow boiling system

    Energy Technology Data Exchange (ETDEWEB)

    Murshed, S.M. Sohel; Vereen, Keon; Kumar, Ranganathan [University of Central Florida, Orlando, FL (United States). Dept. of Mechanical, Materials and Aerospace Engineering], e-mail: rnkumar@mail.ucf.edu

    2009-07-01

    The effect of heat flux and pressure on bubble nucleation of R134a refrigerant in a flow boiling system is experimentally studied. An experimental facility was built and an innovative concept of thermochromic liquid crystal (TLC) technique was introduced for the high resolution and accurate measurement of the overall heater surface temperature. The visualization and image recording process is performed by employing two synchronized high resolution and high speed cameras which simultaneously capture colored TLC images as well as bubble nucleation activities at high frame rates. Experiments were conducted at different high pressures ranging from 690 to 830 kPa and at different heat flux conditions in order to identify their influence on flow boiling performance specially bubbling event. Present results demonstrate that both the heat flux and pressure influence the bubble generation rate and size. For example, bubble generation frequency and size are found to increase with heat flux. An increase in pressure of 137 kPa (from 690 to 827 kPa) increased the bubble frequency and size about 32 Hz and 20 {mu}m, respectively. (author)

  20. Pool boiling with high heat flux enabled by a porous artery structure

    Science.gov (United States)

    Bai, Lizhan; Zhang, Lianpei; Lin, Guiping; Peterson, G. P.

    2016-06-01

    A porous artery structure utilizing the concept of "phase separation and modulation" is proposed to enhance the critical heat flux of pool boiling. A series of experiments were conducted on a range of test articles in which multiple rectangular arteries were machined directly into the top surface of a 10.0 mm diameter copper rod. The arteries were then covered by a 2.0 mm thickness microporous copper plate through silver brazing. The pool wall was fabricated from transparent Pyrex glass to allow a visualization study, and water was used as the working fluid. Experimental results confirmed that the porous artery structure provided individual flow paths for the liquid supply and vapor venting, and avoided the detrimental effects of the liquid/vapor counter flow. As a result, a maximum heat flux of 610 W/cm2 over a heating area of 0.78 cm2 was achieved with no indication of dryout, prior to reaching the heater design temperature limit. Following the experimental tests, the mechanisms responsible for the boiling critical heat flux and performance enhancement of the porous artery structure were analyzed.