WorldWideScience

Sample records for boil off

  1. Zero boil-off system testing

    Science.gov (United States)

    Plachta, D. W.; Johnson, W. L.; Feller, J. R.

    2016-03-01

    Cryogenic propellants such as liquid hydrogen (LH2) and liquid oxygen (LO2) are a part of NASA's future space exploration plans due to their high specific impulse for rocket motors of upper stages. However, the low storage temperatures of LH2 and LO2 cause substantial boil-off losses for long duration missions. These losses can be eliminated by incorporating high performance cryocooler technology to intercept heat load to the propellant tanks and modulating the cryocooler temperature to control tank pressure. The technology being developed by NASA is the reverse turbo-Brayton cycle cryocooler and its integration to the propellant tank through a distributed cooling tubing network coupled to the tank wall. This configuration was recently tested at NASA Glenn Research Center in a vacuum chamber and cryoshroud that simulated the essential thermal aspects of low Earth orbit, its vacuum and temperature. This test series established that the active cooling system integrated with the propellant tank eliminated boil-off and robustly controlled tank pressure.

  2. Zero Boil Off System for Cryogen Storage Project

    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...

  3. 46 CFR 154.705 - Cargo boil-off as fuel: General.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Cargo boil-off as fuel: General. 154.705 Section 154.705... Pressure and Temperature Control § 154.705 Cargo boil-off as fuel: General. (a) Each cargo boil-off fuel system under § 154.703(c) must meet §§ 154.706 through 154.709. (b) The piping in the cargo boil-off...

  4. Boiling water reactor off-gas systems evaluation

    International Nuclear Information System (INIS)

    An evaluation of the off-gas systems for all 25 operating Boiling Water Reactors (BWR) was made to determine the adequacy of their design and operating procedures to reduce the probability of off-gas detonations. The results of the evaluations are that, of the 25 operable units, 13 meet all the acceptance criteria. The other 12 units do not have the features needed to meet the criteria, but have been judged to have, or are committed to provide, features which give reasonable assurance that the potential for external off-gas detonations is minimized. The 12 units which did not originally meet the criteria are aware of the potential hazards associated with off-gas detonations and have agreed to take action to minimize the probability of future detonations

  5. Results of boil-off experiment QUENCH-11

    International Nuclear Information System (INIS)

    The QUENCH experiments are to investigate the hydrogen source term resulting from the water injection into an uncovered core of a Light-Water Reactor (LWR). The typical QUENCH test bundle consists of 21 fuel rod simulators with a total length of approximately 2.5 m. Boil-off experiment QUENCH-11 was performed at Forschungszentrum Karlsruhe (Karlsruhe Research Center) on 8 December, 2005 as the second of two experiments in the frame of the EC-supported LACOMERA program. The experiment focused on studying bundle behavior during boil-off and subsequent quenching at a small water injection rate. It was proposed by INRNE Sofia (Bulgarian Academy of Sciences) and defined together with the Karlsruhe Research Center. The analytical support by using the SCDAP/RELAP5 mod 3.2.irs and ASTEC codes for the preparation of the entire test was essential in conducting the test. A steady boil-off and a corresponding top-down uncovery of the test bundle was achieved by applying power from an electric auxiliary heater at the bundle bottom in addition to the electric bundle power. An additional outer heating system compensated heat losses that would lead to a reduction in boiling off the covered part of the bundle. When the water level had fallen to -70 mm elevation water was injected into the lower plenum at a rate of ca. 1 g/s enabling a nearly stable water level and an extension of the boil-off phase. Quenching of the bundle was performed at a maximum measured bundle temperature of 2040 K with a rather low mass flow rate of water, i.e. 18 (17+1) g/s, compared to the standard water injection rate of approx. 50 g/s. The conditions led to an enhanced cladding and shroud oxidation, quite similar to standard conditions of forced-convection steam flow. In the upper part the test bundle was significantly degraded by oxidation and melt formation. The total generation of hydrogen measured by the mass spectrometer was 141 g, of which 132 g, i.e. more than 90 % of the total, was produced during

  6. 46 CFR 154.706 - Cargo boil-off as fuel: Fuel lines.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Cargo boil-off as fuel: Fuel lines. 154.706 Section 154... Equipment Cargo Pressure and Temperature Control § 154.706 Cargo boil-off as fuel: Fuel lines. (a) Gas fuel lines must not pass through accommodation, service, or control spaces. Each gas fuel line...

  7. 46 CFR 154.709 - Cargo boil-off as fuel: Gas detection equipment.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Cargo boil-off as fuel: Gas detection equipment. 154.709 Section 154.709 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) CERTAIN BULK DANGEROUS... Equipment Cargo Pressure and Temperature Control § 154.709 Cargo boil-off as fuel: Gas detection...

  8. 46 CFR 154.708 - Cargo boil-off as fuel: Valves.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Cargo boil-off as fuel: Valves. 154.708 Section 154.708 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) CERTAIN BULK DANGEROUS CARGOES SAFETY... Pressure and Temperature Control § 154.708 Cargo boil-off as fuel: Valves. (a) Gas fuel lines to the...

  9. 46 CFR 154.707 - Cargo boil-off as fuel: Ventilation.

    Science.gov (United States)

    2010-10-01

    ... 46 Shipping 5 2010-10-01 2010-10-01 false Cargo boil-off as fuel: Ventilation. 154.707 Section 154.707 Shipping COAST GUARD, DEPARTMENT OF HOMELAND SECURITY (CONTINUED) CERTAIN BULK DANGEROUS CARGOES... Equipment Cargo Pressure and Temperature Control § 154.707 Cargo boil-off as fuel: Ventilation. (a)...

  10. Boils

    Science.gov (United States)

    ... the boil is very bad or comes back. Antibacterial soaps and creams cannot help much once a boil ... following may help prevent the spread of infection: Antibacterial soaps Antiseptic (germ-killing) washes Keeping clean (such as ...

  11. Boils

    Science.gov (United States)

    ... or recurrent boils, which are usually due to Staph infections. The bacteria are picked up somewhere and then ... version of boils is folliculitis . This is an infection of hair follicles, usually with Staph bacteria. These often itch more than hurt. The ...

  12. Problem of Boil - off in LNG Supply Chain

    OpenAIRE

    Dobrota, Đorđe; Lalić, Branko; Komar, Ivan

    2013-01-01

    This paper examines the problem of evaporation of Liquefied Natural Gas (LNG) occurring at different places in the LNG supply chain. Evaporation losses in the LNG supply chain are one of the key factors for LNG safety, technical and economic assessment. LNG is stored and transported in tanks as a cryogenic liquid, i.e. as a liquid at a temperature below its boiling point at near atmospheric pressure. Due to heat entering the cryogenic tank during storage and transportatio...

  13. Numerical Modeling of Propellant Boil-Off in a Cryogenic Storage Tank

    Science.gov (United States)

    Majumdar, A. K.; Steadman, T. E.; Maroney, J. L.; Sass, J. P.; Fesmire, J. E.

    2007-01-01

    A numerical model to predict boil-off of stored propellant in large spherical cryogenic tanks has been developed. Accurate prediction of tank boil-off rates for different thermal insulation systems was the goal of this collaboration effort. The Generalized Fluid System Simulation Program, integrating flow analysis and conjugate heat transfer for solving complex fluid system problems, was used to create the model. Calculation of tank boil-off rate requires simultaneous simulation of heat transfer processes among liquid propellant, vapor ullage space, and tank structure. The reference tank for the boil-off model was the 850,000 gallon liquid hydrogen tank at Launch Complex 39B (LC- 39B) at Kennedy Space Center, which is under study for future infrastructure improvements to support the Constellation program. The methodology employed in the numerical model was validated using a sub-scale model and tank. Experimental test data from a 1/15th scale version of the LC-39B tank using both liquid hydrogen and liquid nitrogen were used to anchor the analytical predictions of the sub-scale model. Favorable correlations between sub-scale model and experimental test data have provided confidence in full-scale tank boil-off predictions. These methods are now being used in the preliminary design for other cases including future launch vehicles

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Aksan, S.N.; Stierli, F.; Analytis, G.T. [Paul Scherrer Inst. (PSI), Villigen (Switzerland). Lab. for Thermal-Hydraulics

    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.

  16. Bubble Lift-off Diameter and frequency in a Vertical Subcooled Boiling Flow

    International Nuclear Information System (INIS)

    Together with an active nucleation site density and a bubble detachment frequency, the bubble detachment diameter determines the evaporative heat flux in commercial CFD codes. Also, an increase of an interfacial area concentration by a wall boiling nucleation, i.e., the boiling source term in an interfacial area transport equation (IATE), is expressed by the above three terms. Several studies were performed to investigate the bubble diameters in the forced convective boiling flows. However, the database is still insufficient and the applicability of the suggested models was not thoroughly examined against the existing database. In the present study, the bubble behaviors were captured using a highspeed digital video camera for a forced convective subcooled boiling flow in a vertical annulus. Bubble liftoff diameter and bubble nucleation frequency was quantified by analyzing the captured images. Also, the prediction capability of the models for the bubble lift-off diameter was evaluated against the experimental data of the present work and literature

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-12-31

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

  18. Boil off gas (BOG) management in Spanish liquid natural gas (LNG) terminals

    Energy Technology Data Exchange (ETDEWEB)

    Querol, E.; Gonzalez-Regueral, B.; Garcia-Torrent, J.; Garcia-Martinez, M.J. [Departamento de Ingenieria Quimica y Combustibles, Escuela Tecnica Superior de Ingenieros de Minas, Universidad Politecnica de Madrid, c. Alenza 4, 28003 Madrid (Spain)

    2010-11-15

    Spain is a country with six LNG terminals in operation and three more scheduled for 2011. At the same time an increasing number of LNG tanks are under construction to compensate the Spanish lack of underground storage. A method for evaluating the daily boil off generated is presented in this paper. This method is applied to evaluate the increase of BOG to be handle by LNG terminals in 2016, studying the best commercially available solution to be installed. Finally, as a solution to tackle with the BOG a cogeneration plant is suggested. This option will reduce terminal's operational costs increasing its availability. (author)

  19. Modelling of Boil-Off Gas in LNG Tanks: A Case Study

    OpenAIRE

    Sheikh Zahidul Islam; Ebenezer Adom; Xianda Ji

    2010-01-01

    This paper focuses on the effect of pressure and heat leakages on Boil-off Gas (BOG) in Liquefied Natural Gas (LNG) tanks. The Lee-Kesler-Plocker (LKP) and the Starling modified Benedict-Webb-Rubin (BWRS) empirical models were used to simulate the compressibility factor, enthalpy and hence heat leakage at various pressures to determine the factors that affect the BOG in typical LNG tanks of different capacities. Using a case study data the heat leakage of 140,000kl, 160,00kl, 180,000kl and 20...

  20. Analysis of Spent Fuel Assembly Thermal Behaviors in Boil-off Accident Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hye-Min; Chun, Tae-Hyun; Kim, Sun-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    The spent fuel pool (SFP) accidents would occur due to many different postulated scenarios, for example a SBO (Station Black Out) at SFP storage or an attack from external factor. In this study, we focused on the SFP boil off accident and analyzed the thermal behaviors of spent fuels following this accident, using MELCOR 1.8.6. version. MELCOR, originally the severe accident code, has been developed to also be appropriate to the SFP accident. This paper provides the spent fuel heatup characteristics in terms of decay heat, water level and fuel arrangement. The SFP model is based on 17x17 PWR assembly designed by Westinghouse. Spent fuel coolability has been analyzed with single and 1x4 assembly MELCOR models in the case of boil-off accident. It was shown that the low powered spent fuel assembly could be more vulnerable in the partial loss of coolant inventory because of lack of steam cooling and more fuel being uncovered. In addition, it was found that minimum water level has to be maintained above half of assembly height so as not to experience fuel failure, which depends on decay heat power.

  1. Bubble Lift-off Diameter and Nucleation Frequency in Vertical Subcooled Boiling Flow

    International Nuclear Information System (INIS)

    A series of experiments was carried out to investigate the bubble nucleation to lift-off phenomena for subcooled boiling flow in a vertical annulus channel. A high speed digital video camera was used to capture the dynamics of bubbles. The bubble lift-off diameter and the bubble nucleation frequency were evaluated in terms of heat flux, mass flux, and degree of subcooling. The fundamental features of the lift-off diameter and the nucleation frequency (i.e., the variations across nucleation sites and the dependence on the flow and heat flux conditions) were addressed based on the present observation. A database for the bubble lift-off diameter was built by gathering and summarizing the data of Prodanovic et al., Situ et al., and the present experiments. We evaluated the predictive capabilities of Unal's model, Situ et al.'s model, and Prodanovic et al.'s correlation against the database. We obtained the best prediction results through modifying the wall superheat correlation in Unal's model. In addition, we suggested a new correlation for a combined parameter of the bubble nucleation frequency and the bubble lift-off diameter

  2. Bubble lift-off diameter and nucleation frequency in vertical subcooled boiling flow

    International Nuclear Information System (INIS)

    A series of experiments were carried out to investigate phenomena from bubble nucleation to lift-off for a subcooled boiling flow in a vertical annulus channel. A high-speed digital video camera was used to capture the bubble dynamics. The bubble lift-off diameter and bubble nucleation frequency were evaluated in terms of heat flux, mass flux, and degree of subcooling. The fundamental features of the lift-off diameter and nucleation frequency (i.e., the variations across nucleation sites and the dependence on the flow and heat flux conditions) were addressed based on the present observation. A database for the bubble lift-off diameter was built by gathering and summarizing the data of Prodanovic et al., Situ et al., and the present experiments. We evaluated the predictive capabilities of Unal's model, Situ et al.'s model, and Prodanovic et al.'s correlation against the database. We obtained the best prediction results by modifying the wall superheat correlation in Unal's model. In addition, we suggested a new correlation for a combined parameter of the bubble nucleation frequency and bubble lift-off diameter. (author)

  3. Exergetic Optimization of a Refrigeration Cycle for Re-Liquefaction of LNG Boil-Off Gas

    Directory of Open Access Journals (Sweden)

    Mojtaba Babaelahi

    2010-11-01

    Full Text Available The development of liquefaction process for liquefied natural gas boil-off re-liquefaction plants will be addressed to provide an environmentally friendly and cost effective solution for gas transport. Onboard boil-off gas (BOG re-liquefaction is a new technology that liquefies BOG and returns it to the cargo tanks instead of burning it. Exergetic efficiency optimization for cryogenic refrigeration cycle for re-liquefaction of LNG boil-off gas is performed. Thermodynamic modeling has been performed based on the energy and exergy analyses. Objective problem is developed based on maximization of the plant exergetic efficiency and selected decision variables and constraints. Optimization process is performed using MATLAB genetic algorithm optimization

  4. Analysis of subcooled boiling flow with one-group interfacial area transport equation and bubble lift-off model

    International Nuclear Information System (INIS)

    To enhance the multi-dimensional analysis capability for a subcooled boiling two-phase flow, the one-group interfacial area transport equation was improved with a source term for the bubble lift-off. It included the bubble lift-off diameter model and the lift-off frequency reduction factor model. The bubble lift-off diameter model took into account the bubble's sliding on a heated wall after its departure from a nucleate site, and the lift-off frequency reduction factor was derived by considering the coalescences of the sliding bubbles. To implement the model, EAGLE (elaborated analysis of gas-liquid evolution) code was developed for a multi-dimensional analysis of two-phase flow. The developed model and EAGLE code were validated with the experimental data of SUBO (subcooled boiling) and SNU (Seoul National University) test, where the subcooled boiling phenomena in a vertical annulus channel were observed. Locally measured two-phase flow parameters included a void fraction, interfacial area concentration, and bubble velocity. The results of the computational analysis revealed that the interfacial area transport equation with the bubble lift-off model showed a good agreement with the experimental results of SUBO and SNU. It demonstrates that the source term for the wall nucleation by considering a bubble sliding and lift-off mechanism enhanced the prediction capability for the multi-dimensional behavior of void fraction or interfacial area concentration in the subcooled boiling flow. From the point of view of the bubble velocity, the modeling of an increased turbulence induced by boiling bubbles at the heated wall enhanced the prediction capability of the code.

  5. Annular-intermittent flow regime transition model and its application to boil-off pattern transition and dryout model

    International Nuclear Information System (INIS)

    A model is developed to describe the transition of annular flow to intermittent flow in a vertical two-phase flow system. The instability of the disturbance wave, which is a dominant wave shape at the boundary between annular flow and intermittent flow, is considered as the governing mechanism and this instability is described by the concept of hyperbolicity breaking in the characteristic equation. The developed model is validated by comparing its predictions of gas superficial velocity for the transition with the experimental data available from the literature, and comparing those with the predictions of the other correlations. The comparison results show that the developed model gives better predictions for the transition condition than the existing correlations and the effects of fluid properties, geometry and liquid flow rate on the transition are well considered by the developed model. It is found that the predictions of the developed model have much smaller bias than those of the other correlations; the average of the prediction error is 3% for the present model. The standard deviation of the prediction errors of the present model reaches 28%, which is the smallest among the models compared here. Through the core uncovery experiments, it has been known that the low power and high power core boil-off patterns are observed in the high pressure core uncovery following a small-break loss-of-coolant accident. The developed model for the annular to intermittent flow regime transition was applied to the classification of low power boil-off and high power boil-off patterns. At first, the applicability of the developed criterion to the rod-bundle geometry is demonstrated using the flow pattern transition data taken by Bergles et al. and Venkateswararao. It is shown that the developed criterion well predicts the boundary between low power boil-off and high power boil-off through the comparisons of the predicted annular to intermittent flow transition conditions with

  6. Experimental verification of the horizontal steam generator boil-off transfer degradation at natural circulation

    Energy Technology Data Exchange (ETDEWEB)

    Hyvaerinen, J. [Finnish Centre for Radiation and Nuclear Safety, Helsinki (Finland); Kouhia, J. [VTT Energy, Lappeenranta (Finland)

    1997-12-31

    The presentation summarises the highlights of experimental results obtained for VVER type horizontal steam generator heat transfer, primary side flow pattern, and mixing in the hot collector during secondary side boil-off with primary at single-phase natural circulation. The experiments were performed using the PACTEL facility with Large Diameter (LD) steam generator models, with collector instrumentation designed specifically for these tests. The key findings are as follows: (1) the primary to secondary heat transfer degrades as the secondary water inventory is depleted, following closely the wetted tube area; (2) a circulatory flow pattern exists in the tube bundle, resulting in reversed flow (from cold to the hot collector) in the lower part of the tube bundle, and continuous flow through the upper part, including the tubes that have already dried out; and (3) mixing of the hot leg flow entering the hot collector and reversed, cold, tube flow remains confined within the collector itself, extending only a row or two above the elevation at which tube flow reversal has taken place. 6 refs.

  7. Cryogenic system for collecting noble gases from boiling water reactor off-gas

    International Nuclear Information System (INIS)

    In boiling water reactors, noncondensible gases are expelled from the main condenser. This off-gas stream is composed largely of radiolytic hydrogen and oxygen, air in-leakage, and traces of fission product krypton and xenon. In the Air Products' treatment system, the stoichiometric hydrogen and oxygen are reacted to form water in a catalytic recombiner. The design of the catalytic recombiner is an extension of industrial gas technology developed for purification of argon and helium. The off-gas after the recombiner is processed by cryogenic air-separation technology. The gas is compressed, passed into a reversing heat exchanger where water vapor and carbon dioxide are frozen out, further cooled, and expanded into a distillation column where refrigeration is provided by addition of liquid nitrogen. More than 99.99 percent of the krypton and essentially 100 percent of the xenon entering the column are accumulated in the column bottoms. Every three to six months, the noble-gas concentrate accumulated in the column bottom is removed as liquid, vaporized, diluted with steam, mixed with hydrogen in slight excess of oxygen content, and fed to a small recombiner where all the oxygen reacts to form water. The resulting gas stream, containing from 20 to 40 percent noble gases, is compressed into small storage cylinders for indefinite retention or for decay of all fission gases except krypton-85, followed by subsequent release under controlled conditions and favorable meteorology. This treatment system is based on proven technology that is practiced throughout the industrial gas industry. Only the presence of radioactive materials in the process stream and the application in a nuclear power plant environment are new. Adaptations to meet these new conditions can be made without sacrificing performance, reliability, or safety

  8. Sub-channel analysis by RELAP5 system code of boil-off experiment (Test 5002) with NEPTUN facility

    Energy Technology Data Exchange (ETDEWEB)

    Petruzzi, A. [Pennsylvania State Univ., Dept. of Mechanical and Nuclear Engineering, University Park, Pennsylvania (United States)]. E-mail: axp46@psu.edu; Bousbia Salah, A.; D' Auria, F. [Univ. of Pisa, Dipartimento di Ingegneria Meccanica, Nucleare d della Produzione, Pisa (Italy)]. E-mail: b.salah@ing.unipi.it; f.dauria@ing.unipi.it

    2004-07-01

    This paper presents the results of RELAP5/Mod3.2 system thermalhydraulic code using the sub-channel analysis approach in predicting the NEPTUN separate effect boil off experiments. The boil off tests were conducted in order to simulate the consequences of loss of coolant inventory leading to uncovery and heat up of fuel elements of a nuclear reactor core. In this framework, the NEPTUN low pressure test N{sup o}5002 has been considered. A reference case was run, and the overall data comparison shows good agreement between calculated and experimental thermalhydraulic parameters. A series of sensitivity analyses were also performed in order to assess the code prediction capabilities. The obtained results were almost satisfactory and demonstrate, as well, the reasonable success of the 'sub-channel analysis' approach adopted in the present context for a system thermalhydraulic code. (author)

  9. Modelling of Boil-Off Gas in LNG Tanks: A Case Study

    Directory of Open Access Journals (Sweden)

    Sheikh Zahidul Islam

    2010-08-01

    Full Text Available This paper focuses on the effect of pressure and heat leakages on Boil-off Gas (BOG in Liquefied Natural Gas (LNG tanks. The Lee-Kesler-Plocker (LKP and the Starling modified Benedict-Webb-Rubin (BWRS empirical models were used to simulate the compressibility factor, enthalpy and hence heat leakage at various pressures to determine the factors that affect the BOG in typical LNG tanks of different capacities. Using a case study data the heat leakage of 140,000kl, 160,00kl, 180,000kl and 200,000kl LNG tanks were analyzed using the LKP and BWRS models. The heat leakage of LNG tanks depends on the structure of tanks, and the small tanks lose heatto the environment due to their large surface area to volume ratio. As the operation pressure was dropped to 200mbar, all four of the LNG tanks’ BOG levels reached 0.05vol%/day. In order to satisfy the BOG design requirement, the operating pressure of the four large LNG tanks in the case study was maintained above 200mbar. Thus, the operating pressure impacts BOG on LNG tanks, but this effect is limited under the extreme high operation pressure. An attempt was made to determine the relationship between the compositions of LNGand BOG; one been combustible and the other non-combustible gases. The main component of combustible gas was methane, and nitrogen was of non-combustible gases. The relationship between BOG and methane compositions was that, as the methane fraction increases in the LNG, the BOG volume also increases. In general, results showed a direct correlation between BOG and operating pressure. The study also found that larger LNG tanks have less BOG; however as the operation pressure is increased the differences in the quantity of BOGamong the four tanks decreased.

  10. Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

    Energy Technology Data Exchange (ETDEWEB)

    Korteniemi, V.; Haapalehto, T. [Lappeenranta Univ. of Technology (Finland); Puustinen, M. [VTT Energy, Lappeenranta (Finland)

    1995-09-01

    Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied.

  11. Reflooding and boil-off experiments in a VVER-440 like rod bundle and analyses with the CATHARE code

    International Nuclear Information System (INIS)

    Several experiments were performed with the VEERA facility to simulate reflooding and boil-off phenomena in a VVER-440 like rod bundle. The objective of these experiments was to get experience of a full-scale bundle behavior and to create a database for verification of VVER type core models used with modern thermal-hydraulic codes. The VEERA facility used in the experiments is a scaled-down model of the Russian VVER-440 type pressurized water reactors used in Loviisa, Finland. The test section of the facility consists of one full-scale copy of a VVER-440 reactor rod bundle with 126 full-length electrically heated rod simulators. Bottom and top-down reflooding, different modes of emergency core cooling (ECC) injection and the effect of heating power on the heat-up of the rods was studied. In this paper the results of calculations simulating two reflood and one boil-off experiment with the French CATHARE2 thermal-hydraulic code are also presented. Especially the performance of the recently implemented top-down reflood model of the code was studied

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

    International Nuclear Information System (INIS)

    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

  13. Investigation on the mechanisms of bubble lift-off from a vertical heated surface in subcooled pool boiling

    International Nuclear Information System (INIS)

    Experiments were conducted to elucidate the mechanisms of bubble lift-off from a vertical heated surface in subcooled pool boiling. The experiments were carried out at atmospheric pressure and distilled water was used as the test fluid. A high speed camera was used to observe the bubble behavior. Main experimental parameters were the static contact angle of the heated surface, wall heat flux and the liquid subcooling. In order to observe the bubble behavior clearly, the experiments were performed in the isolated bubble regime. In the present work, the following three typical types of bubble behavior were observed: (1) adhesion to the heated surface, (2) sliding along the heated surface, and (3) lift-off from the heated surface. The effects of the above-mentioned experimental parameters on the bubble behavior was clarified. In addition, the bubble diameter and the bubble migration velocity in the lateral direction at the moment of lift-off were investigated in detail. Assuming that the superheated layer thickness and the bubble growth rate are the appropriate scaling factors for the bubble size and the bubble migration velocity, respectively, semi-empirical correlations were developed. (author)

  14. Analysis and efficiency enhancement of a boil-off gas reliquefaction system with cascade cycle on board LNG carriers

    International Nuclear Information System (INIS)

    Highlights: • An LNG boil-off gas reliquefaction plant on board LNG carriers is improved. • Relevant improvements deals with a study on BOG–C2H4–C3H6 cascade system. • A novel design is proposed to reduce power consumption and COP improvement. • Efficiency improvement by BOG cold energy recovery and compression heat rejection. • Efficiency increase operating in parallel with the engine fuel gas supply system. - Abstract: In this paper, an LNG boil-off gas (BOG) reliquefaction plant operating in accordance with cascade vapor compression cycles, using propylene and ethylene as refrigerants, on board LNG carriers is investigated. As consequence of the analysis results, a new and original design is proposed to reduce power consumption and improve its exergy efficiency. Through energy and exergy analysis, a thermodynamic model is carried out to analyse and evaluate operating conditions as well as to obtain performance values such as the Coefficient of Performance (COP), exergy efficiency, irreversibilities and specific energy consumption. The thermodynamic analysis is performed using the Engineering Equation Solver (EES) software environment. The results of the improved design implemented on the reliquefaction plant for LNG tank conditions of -160.82 °C, a plant BOG input temperature of −125 °C and 25 °C seawater, give COP values of 0.22 and an exergetic efficiency of 37%, such values being 22.22% and 19.35% greater than the original design. The specific energy consumption decreases 14.66% to 0.64 kW h per kg/s of natural BOG. The proposal for improving efficiency is founded on BOG cold energy recovery and BOG compression heat rejection with cooling water in the intercoolers

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

    International Nuclear Information System (INIS)

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

  16. Steady-state and dynamic simulation study on boil-off gas minimization and recovery strategies at LNG exporting terminals

    Science.gov (United States)

    Kurle, Yogesh

    Liquefied natural gas (LNG) is becoming one of the prominent clean energy sources with its abundance, high calorific value, low emission, and price. Vapors generated from LNG due to heat leak are called boil-off gas (BOG). As world-wide LNG productions are increasing fast, BOG generation and handling problems are becoming more critical. Also, due to stringent environmental regulations, flaring of BOG is not a viable option. In this study, typical Propane-and-Mixed-Refrigerant (C3-MR) process, storage facilities, and loading facilities are modeled and simulated to study BOG generation at LNG exporting terminals, including LNG processing, storage, and berth loading areas. Factors causing BOG are presented, and quantities of BOG generated due to each factor at each location are calculated under different LNG temperatures. Various strategies to minimize, recover, and reuse BOG are also studied for their feasibility and energy requirements. Rate of BOG generation during LNG loading---Jetty BOG (JBOG)---changes significantly with loading time. In this study, LNG vessel loading is simulated using dynamic process simulation software to obtain JBOG generation profile and to study JBOG recovery strategies. Also, fuel requirements for LNG plant to run steam-turbine driven compressors and gas-turbine driven compressors are calculated. Handling of JBOG generated from multiple loadings is also considered. The study would help proper handling of BOG problems in terms of minimizing flaring at LNG exporting terminals, and thus reducing waste, saving energy, and protecting surrounding environments.

  17. NEPTUN-III reflooding and boil-off experiments with an LWHCR fuel rod bundle simulator: experimental results and initial code assessment efforts

    International Nuclear Information System (INIS)

    The NEPTUN test facility at Wuerenlingen has been modified to enable LWHCR-representative reflooding and boil-off experiments to be carried out. Results from a first series of forced feed reflooding tests, simulating cold-leg injection, are presented for a range of values of the parameters flooding rate, rod power and initial temperature. Rewetting of the LWHCR fuel bundle simulator was found to be possible in each case. Analysis of the NEPTUN-III reflooding experiments with RELAP5/MOD2 yield discrepant results and it has been shown, in the context of calculcations of the boil-off experiments, that some LWHCR-specific models and correlations need to be developed. (author)

  18. Liquid helium boil-off measurements of heat leakage from sinter-forged BSCCO current leads under DC and AC conditions

    International Nuclear Information System (INIS)

    Liquid helium boil-off experiments are conducted to determine the heat leakage rate of a pair of BSCCO 2223 high-temperature superconductor current leads made by sinter forging. The experiments are carried out in both DC and AC conditions and with and without an intermediate heat intercept. Current ranges are from 0-500 A for DC tests and 0-1,000 Arms for AC tests. The leads are self-cooled. Results show that magnetic hysteresis (AC) losses for both the BSCCO leads and the low-temperature superconductor current jumper are small for the current range. It is shown that significant reduction in heat leakage rate (liquid helium boil-off rate) is realized by using the BSCCO superconductor leads. At 100 A, the heat leakage rate of the BSCCO/copper binary lead is approximately 29% of that of the conventional copper lead. Further reduction in liquid helium boil-off rate can be achieved by using an intermediate heat intercept. For example, at 500 K, the heat leakage rate of the BSCCO/copper binary lead is only 7% of that of the conventional copper lead when an intermediate heat intercept is used

  19. Liquid helium boil-off measurements of heat leakage from sinter-forged BSCCO current leads under DC and AC conditions

    Energy Technology Data Exchange (ETDEWEB)

    Cha, Y.S.; Niemann, R.C.; Hull, J.R.; Youngdahl, C.A.; Lanagan, M.T. [Argonne National Lab., IL (United States); Nakade, M.; Hara, T. [Tokyo Electric Power Co., Yokohama (Japan)

    1995-06-01

    Liquid helium boil-off experiments are conducted to determine the heat leakage rate of a pair of BSCCO 2223 high-temperature superconductor current leads made by sinter forging. The experiments are carried out in both DC and AC conditions and with and without an intermediate heat intercept. Current ranges are from 0-500 A for DC tests and 0-1,000 A{sub rms} for AC tests. The leads are self-cooled. Results show that magnetic hysteresis (AC) losses for both the BSCCO leads and the low-temperature superconductor current jumper are small for the current range. It is shown that significant reduction in heat leakage rate (liquid helium boil-off rate) is realized by using the BSCCO superconductor leads. At 100 A, the heat leakage rate of the BSCCO/copper binary lead is approximately 29% of that of the conventional copper lead. Further reduction in liquid helium boil-off rate can be achieved by using an intermediate heat intercept. For example, at 500 K, the heat leakage rate of the BSCCO/copper binary lead is only 7% of that of the conventional copper lead when an intermediate heat intercept is used.

  20. Oscillate Boiling

    CERN Document Server

    Li, Fenfang; Nguyen, Dang Minh; Ohl, Claus-Dieter

    2016-01-01

    We report about an intriguing boiling regime occurring for small heaters embedded on the boundary in subcooled water. The microheater is realized by focusing a continuous wave laser beam to about $10\\,\\mu$m in diameter onto a 165\\,nm-thick layer of gold, which is submerged in water. After an initial vaporous explosion a single bubble oscillates continuously and repeatably at several $100\\,$kHz. The microbubble's oscillations are accompanied with bubble pinch-off leading to a stream of gaseous bubbles into the subcooled water. The self-driven bubble oscillation is explained with a thermally kicked oscillator caused by the non-spherical collapses and by surface pinning. Additionally, Marangoni stresses induce a recirculating streaming flow which transports cold liquid towards the microheater reducing diffusion of heat along the substrate and therefore stabilizing the phenomenon to many million cycles. We speculate that this oscillate boiling regime may allow to overcome the heat transfer thresholds observed dur...

  1. A high-sensitivity magnetocardiography system with a divided gradiometer array inside a low boil-off Dewar

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Y H; Yu, K K; Kim, J M; Kwon, H; Kim, K, E-mail: yhlee@kriss.re.k [Korea Research Institute of Standards and Science, 1 Doryong, Yuseong, Daejeon 305-600 (Korea, Republic of)

    2009-11-15

    We fabricated a low-noise 64-channel first-order axial gradiometer system for measuring magnetocardiography (MCG) signals. The key technical features of the system are the compact structure of the gradiometer, division of the sensor array plate, direct mounting of the sensor plates into the Dewar bottom, reduced neck diameter of the liquid He Dewar, and compact readout electronics. To make the refill interval of liquid He longer, the distance between the compensation coil of the gradiometer and the input coil pads of the superconducting quantum interference device (SQUID) was reduced to 20 mm. By using direct ultrasonic bonding of Nb wires between the pickup coil wires and input coil pads, the superconductive connection structure became simple. The baseline of the first-order gradiometer is 70 mm, a little longer than for typical conventional axial gradiometers, to provide a larger signal amplitude for deep sources. The 64-channel gradiometer array consists of four blocks, and each block is fixed separately onto the bottom of the Dewar. The neck diameter of the He Dewar (192 mm) is smaller than the bottom diameter (280 mm) in which the gradiometers are distributed. The average boil-off rate of the Dewar is 3 l per day when the 64-channel system is in operation every day. Double relaxation oscillation SQUIDs (DROSs) having large flux-to-voltage transfer coefficients were used to operate SQUIDs via compact electronics. The magnetically shielded room (MSR) has a wall thickness of 80 mm, and consists of two layers of permalloy and one layer of aluminum. When the 64-channel system was installed inside the MSR, the field noise level of the system was about 3.5 fT{sub rms} Hz{sup -1/2} at 100 Hz. MCG measurements with high signal quality were done successfully using the developed system. In addition to the parameter analysis method, we developed software for the three-dimensional imaging of the myocardial current on a realistic image of the heart based on the anatomical

  2. Mechanisms of heat transfer in the uncovered region of a bundle during the boil-off transient

    International Nuclear Information System (INIS)

    The small break accident which occurred at the TMI-2 plant resulted in partial uncovery of the core. To study the thermal-hydraulic phenomena in the uncovered portion of the core, tests were conducted from the EPRI/SUNY Buffalo 3 x 3 rod bundle. Observations from motion pictures and test data show that liquid entrainment and liquid fallback occur in the upper rod bundle region during the early stage of the boiling dry transient. The liquid entrainment and liquid fallback are the results of flow restrictions in the upper bundle tie-plate and spacer grids. The presence of liquid droplets during the entrainment and the fallback greatly influenced the heat transfer in the uncovered portion of the bundle

  3. Boils (Furunculosis)

    Science.gov (United States)

    ... boil starts to drain, wash the area with antibacterial soap and apply some triple antibiotic ointment and a ... avoid spreading the infection to others. Use an antibacterial soap on boil-prone areas when showering, and dry ...

  4. Calculation of the radiation dose from an upset condition in the off-gas system for a boiling water power reactor

    International Nuclear Information System (INIS)

    A study has been made which considered the upset conditions to result in a rupture of the delay line or charcoal adsorber portions of the radioactive off-gas treatment system for a boiling water power reactor. Radiation dose calculations were made for an individual at a 300-meter boundary fence. The doses calculated were the whole body immersion dose and the thyroid, bone and lung doses due to inhalation. The relationship between the various operating and upset parameters of the off-gas system and the radiation doses were investigated. A semi-infinite cloud model with a ground level release was assumed. For a delay line rupture, the calculated gamma immersion dose varies from a high of 9 rad for a break at the condenser to a low of 0.2 rad for a break at the maximum end of a 300-minute delay line. The thyroid dose from inhalation of radioiodine was calculated to vary from 3 to 6 millirem for a delay line rupture and to be 0.6 rem for a charcoal bed rupture. The highest gamma immersion dose from a charcoal adsorber bed rupture was calculated to be 1.5 rad for the low flow rate condition with either an ambient or chilled bed system. Curves have been constructed which show the variation of the calculated doses with the various input parameters. (U.S.)

  5. Mechanistic modeling for ring-type boiling water reactor fuel spacer design (3) run-off effect and model formulation

    International Nuclear Information System (INIS)

    The fuel spacer is one of the components of a fuel rod bundle and its role is to maintain an appropriate rod-to-rod clearance. The fuel spacer influences the liquid film flow distribution in the fuel rod bundle, so that the spacer geometry has a strong effect on thermal hydraulic characteristics of BWR such as critical power and pressure drop in the fuel bundle. In this paper, liquid film flow characteristics were experimentally investigated in a circular channel with ring-type spacers using air and water as test fluids in order to be compared with the analytical results introducing the mechanistic spacer model. The spacer model was composed of three effects; the drift flow effect, the narrow channel effect and the run-off effect. The drift flow effect and the narrow channel effect were discussed in the previous works and the run-off effect is done in this paper. This paper shows the formulation around the spacer with use of the three effects. The proposed model explains well the experimental results of liquid film flow rates and thickness carried out in reference to the spacer thickness, the gap between the channel wall and the spacer

  6. How To Boil the Perfect Egg

    Institute of Scientific and Technical Information of China (English)

    小雨

    2007-01-01

    A British inventor says he has cracked(破解)the age-old riddle(难题)of how to boil the perfect egg,get rid of(摆脱)the water. Simon Rhymes uses powerful light bulbs instead of boiling water to cook the egg. The gadget(小发明)does the job in six minutes,and then chons off(削)the top of

  7. Key technology analysis of boil-off control study on cryogenic propellant long-term application on orbit%低温推进剂长时间在轨的蒸发量控制关键技术分析

    Institute of Scientific and Technical Information of China (English)

    胡伟峰; 申麟; 彭小波; 于海鹏

    2011-01-01

    Cryogenic propellant is difficult to store for low boiling point and easy transpiration. In order to realize long-term application on orbit, boil-off must be controlled. The foreign cryogenic propellant application needs was introduced, and the control system technical schemes was studied. The key technologies of boil-off control on cryogenic propellant long-term application on orbit were also analyzed by passivity protection and active refrigeration. Finally, the boil-off control technology of cryogenic propellant were summarized and prospected.%介绍了国外低温推进剂应用需求,从被动防护和主动制冷等方面对低温推进剂长时间在轨蒸发量控制关键技术进行了归纳分析,对国外低温推进剂长时间在轨蒸发量控制系统技术方案进行了研究,对低温推进剂蒸发量控制技术进行了总结和展望.

  8. Boil off gas treatment process for ethylene cryogenic storage system%乙烯低温储存系统汽化气气体处理工艺

    Institute of Scientific and Technical Information of China (English)

    肖观秀

    2012-01-01

    With the large-scale development of cryogenic storage system in China, the optimization of boil off gas (BOG) treatment process design plays an increasing role in the energy saving and cost reducing of ethylene cryogenic storage system. Firstly, four different BOG gas-compression refrigeration cycle treatment processes were introduced. And then the energy conservation principle was analyzed in theory by pressure-enthalpy figure. The refrigeration coefficients and energy consumptions of four different processes were contrasted based on the actual data. The results show that two throttles with intermediate cooling process is more energy-efficient. Some applicable suggestions on optimization of BOG treatment process and energy saving were submitted.%随着我国低温储存系统应用的大型化发展,优化乙烯低温储存系统汽化气(BOG)气体处理工艺对降低装置运行能耗、节省投资的作用逐步增大.首先介绍了BOG气体的4种不同压缩制冷循环处理工艺;然后通过压焓图从理论上对其节能原理进行分析;同时根据实际数据对比不同工艺过程的制冷系数和能量消耗;最后总结出二次节流中间冷却工艺更为节能,为优化乙烯低温储存BOG气体处理工艺,实现节能降耗提出了适用性建议.

  9. 液氢贮箱零蒸发数值模拟与分析%Numerical simulation and analysis of zero boil-off in a liquid hydrogen storage tank

    Institute of Scientific and Technical Information of China (English)

    冶文莲; 王小军; 王田刚; 王丽红; 张晓曦

    2012-01-01

    A numerical study was performed for a steady - state analysis of fluid flow in a zero boil - off liquid hydrogen storage tank using nozzle injection displacement mixing in microgravity environment based on CFD technique. Meanwhile, 2D axi-symmetric model was applied to predict the distributions of temperature and velocity in different geometry parameters. The results show that the length of nozzle head, inlet diameter and nozzle orifice diameter, et al, play an important role in the temperature distributions and velocity fields inside the tank. Both the maximum and mean temperature decrease, respectively, as the length of nozzle head, inlet diameter increase. However, the effect of nozzle orifice diameter on temperature distributions is not obvious. The results above reveal that thermal performance of the system can be improved by increasing the inlet diameter and properly selecting the length of the nozzle head.%采用CFD技术,对处于微重力下的零蒸发(ZBO)液氢贮箱内采用喷嘴棒强迫混合的流场进行稳态数值模拟,建立了二维轴对称模型,预测了在不同几何参数下贮箱内温度场及速度场分布情况.研究表明,喷嘴棒伸入贮箱长度、入口直径等因素均会对系统内温度场产生影响.贮箱内平均温度和最大温度随喷嘴棒长度和入口直径的增大而减小,而喷口直径对贮箱内温度场影响不明显.由上述可以看出,通过增大入口直径、选择合适的喷嘴棒伸入长度,可以改善系统性能.

  10. Aspects of subcooled boiling

    Energy Technology Data Exchange (ETDEWEB)

    Bankoff, S.G. [Northwestern Univ., Evanston, IL (United States)

    1997-12-31

    Subcooled boiling boiling refers to boiling from a solid surface where the bulk liquid temperature is below the saturation temperature (subcooled). Two classes are considered: (1) nucleate boiling, where, for large subcoolings, individual bubbles grow and collapse while remaining attached to the solid wall, and (2) film boiling, where a continuous vapor film separates the solid from the bulk liquid. One mechanism by which subcooled nucleate boiling results in very large surface heat transfer coefficient is thought to be latent heat transport within the bubble, resulting from simultaneous evaporation from a thin residual liquid layer at the bubble base, and condensation at the polar bubble cap. Another is the increased liquid microconvection around the oscillating bubble. Two related problems have been attacked. One is the rupture of a thin liquid film subject to attractive and repulsive dispersion forces, leading to the formation of mesoscopic drops, which then coalesce and evaporate. Another is the liquid motion in the vicinity of an oscillating contact line, where the bubble wall is idealized as a wedge of constant angle sliding on the solid wall. The subcooled film boiling problem has been attacked by deriving a general long-range nonlinear evolution equation for the local thickness of the vapor layer. Linear and weakly-nonlinear stability results have been obtained. A number of other related problems have been attacked.

  11. 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)

  12. Radiolysis of boiling water

    Science.gov (United States)

    Yang, Shuang; Katsumura, Yosuke; Yamashita, Shinichi; Matsuura, Chihiro; Hiroishi, Daisuke; Lertnaisat, Phantira; Taguchi, Mitsumasa

    2016-06-01

    γ-radiolysis of boiling water has been investigated. The G-value of H2 evolution was found to be very sensitive to the purity of water. In high-purity water, both H2 and O2 gases were formed in the stoichiometric ratio of 2:1; a negligible amount of H2O2 remained in the liquid phase. The G-values of H2 and O2 gas evolution depend on the dose rate: lower dose rates produce larger yields. To clarify the importance of the interface between liquid and gas phase for gas evolution, the gas evolution under Ar gas bubbling was measured. A large amount of H2 was detected, similar to the radiolysis of boiling water. The evolution of gas was enhanced in a 0.5 M NaCl aqueous solution. Deterministic chemical kinetics simulation elucidated the mechanism of radiolysis in boiling water.

  13. Spectral analysis of boiling sound

    International Nuclear Information System (INIS)

    Experimental apparatus, measurements and spectral analysis of boiling sound are described as observed in subcooled boiling of water on a Pt-wire. The results indicate the existence of a strong relation between the intensity and the average frequency of the boiling sound vs heat flux. (author)

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

    International Nuclear Information System (INIS)

    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)

  15. 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.

  16. Natural Circulation with Boiling

    International Nuclear Information System (INIS)

    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

  17. Revision of nucleated boiling mechanisms

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.)

  19. Advanced boiling water reactor

    International Nuclear Information System (INIS)

    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)

  20. Dispersed flow film boiling

    International Nuclear Information System (INIS)

    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

  1. Bandages of boiled potato peels.

    Science.gov (United States)

    Patil, A R; Keswani, M H

    1985-08-01

    The use of potato peels as a dressing for burn wounds has been reported previously. A technique of preparing bandage rolls with boiled potato peels is now presented, which makes dressing of a burn wound more convenient. PMID:4041947

  2. Development of thermohydraulic codes for modeling liquid metal boiling in LMR fuel subassemblies

    International Nuclear Information System (INIS)

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

    International Nuclear Information System (INIS)

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

  4. 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.

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    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

  7. 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.

  8. Molecular geometry and boiling related thermodynamic properties

    International Nuclear Information System (INIS)

    Highlights: ► Molecular geometric factors were found to be important determinants for boiling entropy and thus the boiling temperature. ► Only four molecular geometric factors were used in the study. ► A group contribution method was used to calculate enthalpy of boiling. ► The proposed method is simple and the estimations are in good agreement with experimental values. - Abstract: Boiling related thermodynamic properties are important parameters in research. In this study, a model integrating both additive groups and non-additive molecular geometric factors has been developed for the calculation of boiling enthalpy, entropy and temperature. The calculated values are in good agreement with the measured values of 470 compounds. This model provides a simple and accurate estimation of enthalpy of boiling, entropy of boiling and boiling temperatures with absolute average errors of 0.62 kJ/mol, 1.15 J/K · mol and 7.13 K respectively.

  9. Numerical Modeling and Investigation of Boiling Phenomena

    OpenAIRE

    Kunkelmann, Christian

    2011-01-01

    The subject of the present thesis is the numerical modeling and investigation of boiling phenomena. The heat transfer during boiling is highly efficient and therefore used for many applications in power generation, process engineering and cooling of high performance electronics. The precise knowledge of particular boiling processes, their relevant parameters and limitations is of utmost importance for an optimized application. Therefore, the fundamentals of boiling heat transfer have been...

  10. The nucleate pool boiling dilemma

    International Nuclear Information System (INIS)

    It is shown that the scatter of experimental data is due to the history and machining finish of the heated surface. All experimental pool boiling data published to date, which does not specify precisely the characteristics of the heated surface cannot be expected to provide reliable design information. (U.K.)

  11. Atmospheres of low-mass planets: the "boil-off"

    OpenAIRE

    Owen, James. E.; Wu, Yanqin

    2015-01-01

    We show that, for a low-mass planet that orbits its host star within a few tenths of an AU (like the majority of the {\\it Kepler} planets), the atmosphere it was able to accumulate while embedded in the proto-planetary disk may not survive unscathed after the disk disperses. This gas envelope, if more massive than a few percent of the core (with a mass below $10 M_\\oplus$), has a cooling time that is much longer than the time-scale on which the planet exits the disk. As such, it could not hav...

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

    International Nuclear Information System (INIS)

    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

  13. Detonating gas in boiling water reactors

    International Nuclear Information System (INIS)

    The radiation in the core region of Boiling Water Reactors (BWRs) decomposes a small fraction of the coolant into hydrogen and oxygen, a phenomenon termed radiolysis. The radiolysis gas partitions to the steam during boiling. A 1000 MWe BWR produces around 1.5 tons of steam, containing 25 grams of radiolysis gas, per second. Practically all of the radiolysis gas is carried to the condenser and is taken care of by the condenser evacuation system and the off-gas system. The operation of these systems has been largely trouble-free. Radiolysis gas may also accumulate when stagnant steam condenses in pressurized pipes and components as a result of heat loss. Under certain circumstances a burnable mixture of hydrogen, oxygen and steam may form. Occasionally, the accumulated radiolysis gas has ignited. These incidents typically result in deformation of the components involved, but overpressure bursts have also occurred. Radiolysis gas accumulation in steam systems was largely overlooked by BWR designers (a likely technical reason for this is given in the report) and the problem had to be addressed by utilities. Even though the problem was recognized two decades ago, the counter-measures of today seem not always to be sufficient. Pipe-burst incidents in a German and a Japanese BWR recently attracted attention. Also, damage to a pilot valve in the steam relief system of a Swedish BWR forced a reactor shut-down during 2002. The recent incidents indicate that counter-measures against radiolysis gas accumulation in BWRs should be reviewed, perhaps also improved. The present report provides a short compilation of basic information related to radiolysis gas accumulation in BWRs. It is hoped that the compilation may prove useful to utilities and regulators reviewing the problem

  14. Bubble transport in subcooled flow boiling

    Science.gov (United States)

    Owoeye, Eyitayo James

    Understanding the behavior of bubbles in subcooled flow boiling is important for optimum design and safety in several industrial applications. Bubble dynamics involve a complex combination of multiphase flow, heat transfer, and turbulence. When a vapor bubble is nucleated on a vertical heated wall, it typically slides and grows along the wall until it detaches into the bulk liquid. The bubble transfers heat from the wall into the subcooled liquid during this process. Effective control of this transport phenomenon is important for nuclear reactor cooling and requires the study of interfacial heat and mass transfer in a turbulent flow. Three approaches are commonly used in computational analysis of two-phase flow: Eulerian-Lagrangian, Eulerian-Eulerian, and interface tracking methods. The Eulerian- Lagrangian model assumes a spherical non-deformable bubble in a homogeneous domain. The Eulerian-Eulerian model solves separate conservation equations for each phase using averaging and closure laws. The interface tracking method solves a single set of conservation equations with the interfacial properties computed from the properties of both phases. It is less computationally expensive and does not require empirical relations at the fluid interface. Among the most established interface tracking techniques is the volume-of-fluid (VOF) method. VOF is accurate, conserves mass, captures topology changes, and permits sharp interfaces. This work involves the behavior of vapor bubbles in upward subcooled flow boiling. Both laminar and turbulent flow conditions are considered with corresponding pipe Reynolds number of 0 -- 410,000 using a large eddy simulation (LES) turbulence model and VOF interface tracking method. The study was performed at operating conditions that cover those of boiling water reactors (BWR) and pressurized water reactors (PWR). The analysis focused on the life cycle of vapor bubble after departing from its nucleation site, i.e. growth, slide, lift-off, rise

  15. On-line system for monitoring of boiling in nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    The performance of the boiling detection system has been tested on boiling signals coming from the research reactor HOR during experiments with the NIOBE boiling setup. Several detection methods utilizing frequency domain analysis have been tested both on- and off-line. Results of these methods indicate that boiling detection is possible in real-time even in the incipient stage of the boiling. Both DC and AC components of the in-core and ex-core neutron detector signals can be used for boiling detection; these two components provide complementary information. Advanced signal analysis application to the DC signals may give information about the dynamic changes of the reactor, provided that the changes of the signal exceed the inherent noise of the measured channel. At the same time, AC signal analysis will characterize the changes even in the inherent signal fluctuation level. Boiling experiments of HOR and the methods implemented for signal analysis validates the techniques used for these experiments. (orig./HP)

  16. On-line system for monitoring of boiling in nuclear reactor fuel assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Tuerkcan, E. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Kozma, R. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands); Nabeshima, K. [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan); Verhoef, J.P. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands)

    1993-01-01

    The performance of the boiling detection system has been tested on boiling signals coming from the research reactor HOR during experiments with the NIOBE boiling setup. Several detection methods utilizing frequency domain analysis have been tested both on- and off-line. Results of these methods indicate that boiling detection is possible in real-time even in the incipient stage of the boiling. Both DC and AC components of the in-core and ex-core neutron detector signals can be used for boiling detection; these two components provide complementary information. Advanced signal analysis application to the DC signals may give information about the dynamic changes of the reactor, provided that the changes of the signal exceed the inherent noise of the measured channel. At the same time, AC signal analysis will characterize the changes even in the inherent signal fluctuation level. Boiling experiments of HOR and the methods implemented for signal analysis validates the techniques used for these experiments. (orig./HP)

  17. 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

  18. Thermosyphon boiling in vertical channels

    Science.gov (United States)

    Bar-Cohen, A.; Schweitzer, H.

    The thermal characteristics of ebullient cooling systems for VHSIC and VLSI microelectronic component thermal control are studied by experimentally and analytically investigating boiling heat transfer from a pair of flat, closely spaced, isoflux plates immersed in saturated water. A theoretical model for liquid flow rate through the channel is developed and used as a basis for correlating the rate of heat transfer from the channel walls. Experimental results for wall temperature as a function of axial location, heat flux, and plate spacing are presented. The finding that the wall superheat at constant imposed heat flux decreases as the channel is narrowed is explained with the aid of a boiling thermosiphon analysis which yields the mass flux through the channel.

  19. Numerical simulations of nucleate boiling in impinging jets: Applications in power electronics cooling

    Energy Technology Data Exchange (ETDEWEB)

    Narumanchi, Sreekant; Bharathan, Desikan; Hassani, Vahab [National Renewable Energy Laboratory, MS 1633, 1617 Cole Boulevard Golden, CO 80401-3393 (United States); Troshko, Andrey [ANSYS Inc., Fluid Business Unit, 10 Cavendish Court, Centerra Park Resource, Lebanon, NH 03766 (United States)

    2008-01-15

    Boiling jet impingement cooling is currently being explored to cool power electronics components. In hybrid vehicles, inverters are used for DC-AC conversion. These inverters involve a number of insulated-gate bipolar transistors (IGBTs), which are used as on/off switches. The heat dissipated in these transistors can result in heat fluxes of up to 200 W/cm{sup 2}, which makes the thermal management problem quite important. In this paper, turbulent jet impingement involving nucleate boiling is explored numerically. The framework for these computations is the CFD code FLUENT. For nucleate boiling, the Eulerian multiphase model is used. The numerical results for boiling water and R113 jets (submerged) are validated against existing experimental data in the literature. Some representative IGBT package simulations that use R134a as the cooling fluid are also presented. (author)

  20. Critical heat flux in subcooled flow boiling

    Science.gov (United States)

    Hall, David Douglas

    The critical heat flux (CHF) phenomenon was investigated for water flow in tubes with particular emphasis on the development of methods for predicting CHF in the subcooled flow boiling regime. The Purdue University Boiling and Two-Phase Flow Laboratory (PU-BTPFL) CHF database for water flow in a uniformly heated tube was compiled from the world literature dating back to 1949 and represents the largest CHF database ever assembled with 32,544 data points from over 100 sources. The superiority of this database was proven via a detailed examination of previous databases. The PU-BTPFL CHF database is an invaluable tool for the development of CHF correlations and mechanistic models that are superior to existing ones developed with smaller, less comprehensive CHF databases. In response to the many inaccurate and inordinately complex correlations, two nondimensional, subcooled CHF correlations were formulated, containing only five adjustable constants and whose unique functional forms were determined without using a statistical analysis but rather using the parametric trends observed in less than 10% of the subcooled CHF data. The correlation based on inlet conditions (diameter, heated length, mass velocity, pressure, inlet quality) was by far the most accurate of all known subcooled CHF correlations, having mean absolute and root-mean-square (RMS) errors of 10.3% and 14.3%, respectively. The outlet (local) conditions correlation was the most accurate correlation based on local CHF conditions (diameter, mass velocity, pressure, outlet quality) and may be used with a nonuniform axial heat flux. Both correlations proved more accurate than a recent CHF look-up table commonly employed in nuclear reactor thermal hydraulic computer codes. An interfacial lift-off, subcooled CHF model was developed from a consideration of the instability of the vapor-liquid interface and the fraction of heat required for liquid-vapor conversion as opposed to that for bulk liquid heating. Severe

  1. Duality of boiling systems and uncertainty phenomena

    Institute of Scientific and Technical Information of China (English)

    柴立合; 彭晓峰; 王补宣

    2000-01-01

    Interactions among dry patches at high heat flux are theoretically analyzed. The high heat flux boiling experiments on metal plate wall with different materials and thickness are correspondingly conducted. The duality of boiling system, i.e. hydrodynamic performance and self-organized performance is identified. A unified explanation of hydrodynamic models and dry patches models is given. The scatter and uncertainty in boiling data can be mainly attributed to the intrinsic duality, but not the sole surface effects. The present experimental results explain why the deviation point at high flux boiling is seen only on occasion and why the self-organization of dry patches is often ignored in available literature.

  2. 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...

  3. Heat transfer research on enhanced heating surfaces in flow boiling in a minichannel and pool boiling

    International Nuclear Information System (INIS)

    Graphical abstract: - Highlights: • Application of enhanced surfaces in boiling heat transfer. • Flow and pool boiling heat transfer on the heating surfaces with mini-recesses. • Minichannel (horizontal) with the enhanced heating wall. • Determination of heat transfer coefficients and boiling curves. • Comparative experimental data analysis for flow and pool boiling heat transfer. - Abstract: The paper focuses on the analysis of the enhanced surfaces in such applications as boiling heat transfer. The surfaces have similar geometric parameters for the surface development. Two testing measurement modules with enhanced heating surfaces are used independently, one for flow boiling and the other – for pool boiling research. The heating surfaces with mini-recesses which contact boiling liquid are made by spark erosion. Flow boiling is studied when FC-72 flows through a horizontally positioned minichannel and its bottom wall is heated. These experiments were carried out during under a pressure slightly higher than the atmospheric one. Pool boiling experiments were conducted with FC-72 at atmospheric pressure in the vessel using enhanced sample as the bottom heating surface. Comparison of results for flow and pool boiling indicates that obtained heat transfer coefficients are a few times higher for pool boiling in the boiling incipience conditions. There are basic differences in the local heat transfer coefficients during the development of flow boiling in a minichannel, depending on the location along the flow in the channel. In the subcooled boiling area, heat transfer coefficients are low. In developed boiling, they are high, but they decrease when the amount of vapour in the liquid–vapour mixture rises

  4. High Pressure Boiling Water Reactor

    International Nuclear Information System (INIS)

    Some four hundred Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) have been in operation for several decades. The presented concept, the High Pressure Boiling Water Reactor (HP-BWR) makes use of the operating experiences. HP-BWR combines the advantages and leaves out the disadvantages of the traditional BWRs and PWRs by taking in consideration the experiences gained during their operation. The best parts of the two traditional reactor types are used and the troublesome components are left out. HP-BWR major benefits are; 1. Safety is improved; -Gravity operated control rods -Large space for the cross formed control rods between fuel boxes -Bottom of the reactor vessel is smooth and is without penetrations -All the pipe connections to the reactor vessel are well above the top of the reactor core -Core spray is not needed -Internal circulation pumps are used. 2. Environment friendly; -Improved thermal efficiency, feeding the turbine with ∼340 oC (15 MPa) steam instead of ∼285 oC (7MPa) -Less warm water release to the recipient and less uranium consumption per produced kWh and consequently less waste is produced. 3. Cost effective, simple; -Direct cycle, no need for complicated steam generators -Moisture separators and steam dryers are inside the reactor vessel and additional separators and dryers can be installed inside or outside the containment -Well proved simple dry containment or wet containment can be used. (author)

  5. Boiling flow through diverging microchannel

    Indian Academy of Sciences (India)

    V S Duryodhan; S G Singh; Amit Agrawal

    2013-12-01

    An experimental study of flow boiling through diverging microchannel has been carried out in this work, with the aim of understanding boiling in nonuniform cross-section microchannel. Diverging microchannel of 4° of divergence angle and 146 m hydraulic diameter (calculated at mid-length) has been employed for the present study with deionised water as working fluid. Effect of mass flux (118–1182 kg/m2-s) and heat flux (1.6–19.2 W/cm2) on single and two-phase pressure drop and average heat transfer coefficient has been studied. Concurrently, flow visualization is carried out to document the various flow regimes and to correlate the pressure drop and average heat transfer coefficient to the underlying flow regime. Four flow regimes have been identified from the measurements: bubbly, slug, slug–annular and periodic dry-out/rewetting. Variation of pressure drop with heat flux shows one maxima which corresponds to transition from bubbly to slug flow. It is shown that significantly large heat transfer coefficient (up to 107 kW/m2-K) can be attained for such systems, for small pressure drop penalty and with good flow stability.

  6. Signal processing for boiling noise detection

    International Nuclear Information System (INIS)

    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

  7. Transition boiling heat transfer during reflooding transients

    International Nuclear Information System (INIS)

    Transition boiling heat transfer is characterized by a heat flux which declines as the heater wall temperature increases. Steady state transition boiling is also characterized by alternate periods of high and low heat transfer caused by intermittent wetting of the heated surface. In flow boiling, the reason for intermittent wetting depends on the volume fraction of vapor present. At high vapor volume fractions, annular flow exists during what is generally called the nucleate boiling region, and a thin liquid film is present on the surface. The remainder of the passage is filled with vapor carrying entrained droplets. Above the nucleate boiling region there is no liquid film, and heat is transferred to droplet-laden vapor. In the narrow transition boiling region between nucleate boiling and heat transfer to steam, the liquid film is present only part of the time. The intermittent wetting produces significant wall temperature oscillations. Recent phenomenologically based modeling of steady state transition boiling heat transfer at high vapor fractions has been successful in predicting the magnitude of both temperature oscillations and heat transfer rates. After a brief review of the steady state model, this note shows how the results of the steady state analysis for vertical surfaces may be used to obtain heat transfer rates during reflooding transients

  8. Development of two-phase flow CFD code (EAGLE) with interfacial area transport equation for analysis of subcooled boiling flow

    International Nuclear Information System (INIS)

    The interfacial area transport equation for a subcooled boiling flow is developed with a mechanistic model for the wall boiling source term. It includes the bubble lift-off diameter model and the lift-off frequency reduction factor model. Those models take into account a bubble's sliding on the heated wall after a departure from the nucleate site and the coalescences of sliding bubbles. To implement the model, the two-phase flow CFD code was developed, which is named as EAGLE (Elaborated Analysis of Gas-Liquid Evolution). The developed model and EAGLE code are validated by the experimental data of SUBO (Subcooled Boiling) facility. The computational analysis reveals that the interfacial area transport equation with the bubble lift-off diameter model agrees well with the experimental results. It presents that the source term for the wall nucleation enhanced the prediction capability for a multidimensional behavior of void fraction or interfacial area concentration. (authors)

  9. 21 CFR 872.6710 - Boiling water sterilizer.

    Science.gov (United States)

    2010-04-01

    ... 21 Food and Drugs 8 2010-04-01 2010-04-01 false Boiling water sterilizer. 872.6710 Section 872...) MEDICAL DEVICES DENTAL DEVICES Miscellaneous Devices § 872.6710 Boiling water sterilizer. (a) Identification. A boiling water sterilizer is an AC-powered device that consists of a container for boiling...

  10. Computations of film boiling. Part II: multi-mode film boiling

    Energy Technology Data Exchange (ETDEWEB)

    Esmaeeli, A.; Tryggvason, G. [Worcester Polytechnic Institute, MA (United States). Mechanical Engineering Department

    2004-12-01

    Film boiling on horizontal periodic surfaces is investigated by direct numerical simulations. A front tracking/finite difference technique is used to solve the momentum and the energy equations in both phases and to account for inertia, viscosity, and surface deformation. Effect of the unit cell size W on the interface dynamics, heat transfer, and fluid flow is studied for different wall superheats. The simulations are carried out over sufficiently long times to capture several bubble release cycles ands to evaluate the quasi steady-state Nusselt number (Nu). While instantaneous Nusselt number will change as result of a change in the system size, statistically steady-state Nusselt number remains almost the same. Simulations of two-dimensional systems in large unit cells, 5{lambda}{sub d2} < W < 10{lambda}{sub d2}, show a distribution of bubble spacing in the range of 0.61{lambda}{sub d2}-1.46{lambda}{sub d2}. At relatively low superheats (Ja {<=} 0.064) the bubbles are released periodically from the vapor film, but at intermediate superheats (0.064 < Ja < 2.13) permanent vapor jets are formed with no bubble break off. At sufficiently high superheats, the vapor jets start to interact. It is shown that the average bubble spacing does not change with changes in the wall superheat. (author)

  11. Using Boiling for Treating Waste Activated Sludge

    Institute of Scientific and Technical Information of China (English)

    2002-01-01

    In this work we investigated the feasibility of using short time, low superheat boiling to treat biological sludge. The treated sludge exhibited reduced filterability and enhanced settleability. The boiling treatment released a large amount of extra-cellular polymers (ECPs) from the solid phase and reduced the microbial density levels of the total coliform bacteria and the heterotrophic bacteria. A diluted sludge is preferable for its high degree of organic hydrolysis and sufficient reduction in microbial density levels.

  12. Boiling of subcooled water in forced convection

    International Nuclear Information System (INIS)

    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/cm2), 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)

  13. Consumers' perception and acceptance of boiled and fermented sausages from strongly boar tainted meat.

    Science.gov (United States)

    Meier-Dinkel, Lisa; Gertheiss, Jan; Schnäckel, Wolfram; Mörlein, Daniel

    2016-08-01

    Characteristic off-flavours may occur in uncastrated male pigs depending on the accumulation of androstenone and skatole. Feasible processing of strongly tainted carcasses is challenging but gains in importance due to the European ban on piglet castration in 2018. This paper investigates consumers' acceptability of two sausage types: (a) emulsion-type (BOILED) and (b) smoked raw-fermented (FERM). Liking (9 point scales) and flavour perception (check-all-that-apply with both, typical and negatively connoted sensory terms) were evaluated by 120 consumers (within-subject design). Proportion of tainted boar meat (0, 50, 100%) affected overall liking of BOILED, F (2, 238)=23.22, P<.001, but not of FERM sausages, F (2, 238)=0.89, P=.414. Consumers described the flavour of BOILED-100 as strong and sweaty. In conclusion, FERM products seem promising for processing of tainted carcasses whereas formulations must be optimized for BOILED in order to eliminate perceptible off-flavours. Boar taint rejection thresholds may be higher for processed than those suggested for unprocessed meat cuts. PMID:27038338

  14. Surface boiling of superheated liquid

    International Nuclear Information System (INIS)

    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

  15. 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.

  16. Experimental study on transient boiling heat transfer

    International Nuclear Information System (INIS)

    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. Lattice Boltzmann modeling of boiling heat transfer: The boiling curve and the effects of wettability

    CERN Document Server

    Li, Q; Francois, M M; He, Y L; Luo, K H

    2015-01-01

    A hybrid thermal lattice Boltzmann (LB) model is presented to simulate thermal multiphase flows with phase change based on an improved pseudopotential LB approach [Q. Li, K. H. Luo, and X. J. Li, Phys. Rev. E 87, 053301 (2013)]. The present model does not suffer from the spurious term caused by the forcing-term effect, which was encountered in some previous thermal LB models for liquid-vapor phase change. Using the model, the liquid-vapor boiling process is simulated. The boiling curve together with the three boiling stages (nucleate boiling, transition boiling, and film boiling) is numerically reproduced in the LB community for the first time. The numerical results show that the basic features and the fundamental characteristics of boiling heat transfer are well captured, such as the severe fluctuation of transient heat flux in the transition boiling and the feature that the maximum heat transfer coefficient lies at a lower wall superheat than that of the maximum heat flux. Furthermore, the effects of the he...

  18. How does surface wettability influence nucleate boiling?

    Science.gov (United States)

    Phan, Hai Trieu; Caney, Nadia; Marty, Philippe; Colasson, Stéphane; Gavillet, Jérôme

    2009-05-01

    Although the boiling process has been a major subject of research for several decades, its physics still remain unclear and require further investigation. This study aims at highlighting the effects of surface wettability on pool boiling heat transfer. Nanocoating techniques were used to vary the water contact angle from 20° to 110° by modifying nanoscale surface topography and chemistry. The experimental results obtained disagree with the predictions of the classical models. A new approach of nucleation mechanism is established to clarify the nexus between the surface wettability and the nucleate boiling heat transfer. In this approach, we introduce the concept of macro- and micro-contact angles to explain the observed phenomenon. To cite this article: H.T. Phan et al., C. R. Mecanique 337 (2009).

  19. Film boiling heat transfer during reflood process

    International Nuclear Information System (INIS)

    From Westinghouse's Full Length Emergency Cooling Heat Transfer (FLECHT) test data and the previous studies on the film boiling, local subcooling is found to be a dominant factor in the film boiling heat transfer, existing in the reflood process. By experiment, the correlation was obtained between saturated film boiling heat transfer coefficient h sub(c), sat and subcooled h sub(c), sub. The h sub(c), sat is similar to Bromley's expression, but the value differs from his. The ratio of h sub(c') sub to h sub(c') sat is expressed with the local coolant subcooling T sub(sub) (0C) as h sub(c') sub/h sub(c') sat = 1 + 0.025 ΔT sub(sub). The results in experiment are predicted by this formula with error +- 20%. (auth.)

  20. Thermodynamics of Flow Boiling Heat Transfer

    Science.gov (United States)

    Collado, F. J.

    2003-05-01

    Convective boiling in sub-cooled water flowing through a heated channel is essential in many engineering applications where high heat flux needs to be accommodated. It has been customary to represent the heat transfer by the boiling curve, which shows the heat flux versus the wall-minus-saturation temperature difference. However it is a rather complicated problem, and recent revisions of two-phase flow and heat transfer note that calculated values of boiling heat transfer coefficients present many uncertainties. Quite recently, the author has shown that the average thermal gap in the heated channel (the wall temperature minus the average temperature of the coolant) was tightly connected with the thermodynamic efficiency of a theoretical reversible engine placed in this thermal gap. In this work, whereas this correlation is checked again with data taken by General Electric (task III) for water at high pressure, a possible connection between this wall efficiency and the reversible-work theorem is explored.

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

    International Nuclear Information System (INIS)

    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)

  2. Evaluation of heat transfer at dispersed film boiling region and reflood blockage by downcomer boiling

    International Nuclear Information System (INIS)

    The large nuclear power plant like APR1400 have a emergency core cooling system (ECCS) for large break loss of coolant accident (LBLOCA). To evaluate the cooling capacity of ECCS, it is important to analysis the heat transfer at dispersed film boiling region and to evaluate the amount of reflood. During the reflood, boiling occurs at the downcomer of vessel and the boiling play a role of blockage and hence the amount of inflow is reduced. Therefore, the phenomena also will be evaluated. This study is composed of three key subjects. One is the study about the heat transfer at dispersed film boiling. In this subject, the final goal is to develop a boiling model. For this, we will analysis the experimental results and other correlation. and the mechanisms will be also compared with each other. The new model will be developed including recent experimental results. Second one is to do a experimental works about the amount of inflow under the downcomer boiling simulation. In this experiment, the air-water is working fluid. the bubble dynamics, pressure drop of two phase flow, blockage effect and etc. will be observed. and some of these will be quantified. Third subjects is to control the boiling heat transfer coefficient. Here, the method will be various surface treatment.

  3. 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.

  4. Water boiling kinetic in rapid decompression

    International Nuclear Information System (INIS)

    This study entering in the frame of a CEA, EDF and Framatome collaboration, has for objective to modelize two-phase flows in case of PWR Loca. The objective is to find, by taking in account the all imbalances, a formulation for the mass transfer at the interface water-vapor by the study of water boiling phenomenon in case of fast decompression such as a primary circuit break. In this accident, the estimation of boiling speeds in an essential parameter for determining the break discharge which conditions the safety systems design

  5. Fuel assembly for a boiling water reactor

    International Nuclear Information System (INIS)

    The fuel assembly of a boiling water reactor contains a number of vertical fuel rods with their lower ends against a bottom tie plate. The rods are positioned by spacers, which are fixed to the canning. The upward motion is reduced by the top plate of a special design. (G.B.)

  6. 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...

  7. 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...

  8. Classic and Hard-Boiled Detective Fiction.

    Science.gov (United States)

    Reilly, John M.

    Through an analysis of several stories, this paper defines the similarities and differences between classic and hard-boiled detective fiction. The characters and plots of three stories are discussed: "The Red House" by A. A. Milne; "I, The Jury" by Mickey Spillane; and "League of Frightened Men" by Rex Stout. The classic detective story is defined…

  9. CHF prediction model research for flat and hypervapotron heating surface in subcooled flow nucleate boiling

    International Nuclear Information System (INIS)

    Highlights: • Proposed CHF prediction model is based on hot spot model. • Improved prediction model is more convenient for practical application. • Good CHF prediction on flat and hypervapotron experiment. • CHF is inversely proportional to bubble departure diameter. - Abstract: Based on fundamental nucleation boiling theory, nucleation site distribution and hot spot model, a theoretical critical heat flux (CHF) prediction model in subcooled flow boiling is presented in this paper. With the increase of heat flux, nucleation site density will finally approach the limit of site population. When the critical number of nucleation sites is reached in an area, it will cut off the supply of liquid for the central site, and then CHF emerges. In this paper, an improved hot spot model is proposed, which uses statistical bubble departure diameter and standard square shape. CHF prediction can be reasonably achieved based on fully developed boiling heat transfer curve, nucleation site density and distribution. Some related experimental data are referred and R134a subcooled flow boiling data on flat and hypervapotron test section are reported. Comparison results showed a good prediction of CHF

  10. Prediction of a subcooled boiling flow with mechanistic wall boiling and bubble size models

    International Nuclear Information System (INIS)

    Subcooled boiling is one of the crucial phenomena for the design, operation and safety analysis of a nuclear power plant. In recent years, developers of multiphase CFD (Computational Fluid Dynamics) codes focused their development activity on the mechanistic prediction of DNB (Departure from Nucleate Boiling) in PWR. Wall boiling model is one of the key parameters for this purpose. In order to enhance prediction capability of the subcooled boiling flow, an advanced wall boiling model consisting of a mechanistic bubble departure model (Klausner et al., 1993), Hibiki et al.'s (2009) active nucleate site model and Cole's bubble departure frequency model was explored for the CFD code. To ensure a wide range applicability of the advanced wall boiling model, each model was evaluated separately according to the flow conditions such as pressure, temperature and flow rate. Finally, the advanced wall boiling model was implemented into the STAR-CD as a form of user FORTRAN file. One of the other important parameters for an accurate prediction of the subcooled boiling flow is bubble size which governs interfacial transfer terms between two phases. In this study, the S-gamma model, which was developed for the STAR-CD (Lo, 2006), was applied as a bubble size model. For the validation of the present wall boiling and bubble size models, benchmark calculations were carried out against SUBO and DEBORA subcooled boiling flow data. Working fluid of SUBO test is steam/water and its pressure condition is about 2 bars. In contrast to this, working fluid of DEBORA test is Refrigerant-12 (R-12) and phasic density ratio of the tests is equivalent to that of steam/water around 90 to 170 bars. Therefore, present benchmark calculation covers wide range pressure condition of steam/water. The calculation results confirms that the new mechanistic wall boiling and bubble size models follow well the tendency on the change of flow conditions and they can be applicable to the wide range of flow

  11. A dry-spot model of critical heat flux applicable to both pool boiling and subcooled forced convection boiling

    International Nuclear Information System (INIS)

    A study has been performed to predict CHF in pool boiling and subcooled forced convection boiling using the dry-spot model presented by the authors and existing correlations for heat transfer coefficient, active site density and bubble departure diameter in nucleate boiling. Comparisons of the model predictions with experimental data for pool boiling of water and subcooled upward forced convection boiling of water in vertical, uniformly-heated round tubes have been performed and the parametric trends of CHF have been investigated. The results of the present study strongly support the validity of physical feature of the present model on the CHF mechanism in pool boiling and subcooled forced convection boiling. To improve the prediction capability of the present model, further works on active site density, bubble departure diameter and suppression factor in subcooled boiling are needed

  12. Development of CFD Code for Subcooled Boiling Two-Phase Flow with Modeling of the Interfacial Area Transport Equation

    International Nuclear Information System (INIS)

    The interfacial area transport equation for the subcooled boiling flow was developed with a mechanistic model for the wall boiling source term. It included the bubble lift-off diameter model and lift-off frequency reduction factor model. To implement the model, the two-phase flow CFD code was developed, which was named as EAGLE (Elaborated Analysis of Gas-Liquid Evolution). The developed model and EAGLE code was validated the experimental data of SUBO and SNU facilities. The computational analysis revealed that the interfacial area transport equation with the bubble lift-off diameter model agreed well with the experimental results. It presents that the source term for the wall nucleation enhanced the prediction capability for a multi-dimensional behavior of void fraction or interfacial area concentration

  13. Nanotube Adsorption for the Capture and Re-liquefaction of Hydrogen Biol-Off During Tanker Transfer Operations Project

    Data.gov (United States)

    National Aeronautics and Space Administration — This proposal discloses an innovative, economically feasible technique to capture and re-liquefy the hydrogen boil-off by using carbon nanotube adsorption prior to...

  14. Rewetting an auto-wave change of boiling modes

    International Nuclear Information System (INIS)

    One studied heat exchange under auto-wave change of boiling metastable mode over to the stable one. One introduced a classification of boiling curves depending on temperature gradient, flow rate directions and temperature wave motion rate. One advanced the hypothesis according to which the essential change of heat exchange patterns within temperature wave range was caused by boiling local nonequilibrium state. One investigated experimentally heat exchange under water boiling in pipes under auto-wave change of boiling conditions within 3-10 MPa pressure range within wide range of steam contents and mass rates. Paper discusses application possibility of the derived data when simulating emergency conditions in reactor core

  15. 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

  16. Numerical simulation of subcooled flow boiling

    Science.gov (United States)

    Park, Won Cheol

    Sub-cooled flow boiling in a U-bend has been examined using numerical methods. An Eulerian/Eulerian mathematical description was used with a multiphase computational algorithm to predict several types of flows and to examine sub-cooled flow boiling. As a prelude to the study of sub-cooled boiling and two-phase flows, single-phase laminar and turbulent flows in a U-bend were investigated. Air-water bubbly up flow in a vertical straight duct followed by a U-bend with heat transfer was analyzed. In such a flow, as the flow develops through the U-bend the bubbles move from center and outer wall toward inner wall. After half way through the U-bend, the fluids do not have sufficient time for complete reorganization in the presence of centrifugal forces and the pressure gradients. After the U-bend, the bubbles finally reach the original distribution in about forty diameters. The heat transfer in the U-bend was also calculated and as expected heat transfer rate on the outer wall is higher than on the inner wall. For air-water bubbly two-phase flow, Nusselt numbers in the U-bend can be as high as 400 percent of the value in the straight duct on one of the walls. The method of partitioned wall heat flux was used to study sub-cooled flow boiling. For sub-cooled flow boiling in a U-bend, axial and lateral velocity distributions as well as quality and void fraction variations were analyzed. Computed axial and lateral variations of void fraction compare favorably with existing experimental data. As expected, the pressure drop for bubbly flow through the U-bend is larger than for single-phase flow by as much as fifty percent. Computed pressure drop for flow with phase change falls between the predictions of two different correlations in the literature, and thus seems reasonable. Predictions of heat transfer and void fraction under sub-cooled flow boiling using two-fluid models need better quantitative knowledge related to the mechanisms associated with bubble growth and

  17. Transition from natural convection or nucleate boiling regime to nucleate boiling or film boiling regime caused by a rapid pressure reduction in highly pressurized and subcooled water

    International Nuclear Information System (INIS)

    Transient boiling processes caused by exponentially decreasing system pressures with various decreasing pressure-reduction periods from the initial heat flux on a horizontal cylinder in a pool of highly subcooled water measured were divided into three groups for low and intermediate initial heat flux in natural convection regime and for high initial heat flux in nucleate boiling. The transitions from low initial heat flux values to stable nucleate boiling occurred independently of the pressure-reduction period values. The transitions from intermediate and high initial heat flux values to stable film boiling occurred for the short pressure-reduction period values, although those to stable nucleate boiling occurred for the long pressure-reduction period values. The mechanism of transient boiling process caused by an exponentially decreasing system pressure with a decreasing pressure-reduction period from an initial heat flux on a horizontal cylinder in a pool of highly subcooled water was clarified on the graph of α/q0.7 versus system pressure with the curves of corresponding fully developed nucleate boiling, incipient nucleate boiling due to unflooded cavities with vapor, and incipient heterogeneous spontaneous nucleation (HSN) due to flooded cavities without vapor. The transitions to stable nucleate boiling from the low initial heat flux values occurred independently of the pressure-reduction period values. The transitions from intermediate and high initial heat flux values in natural convection and nucleate boiling to stable film boiling occurred due to the HSN for short pressure-reduction period values; however those to stable nucleate boiling occurred for long pressure-reduction values. (author)

  18. Dynamic Bubble Behaviour during Microscale Subcooled Boiling

    Institute of Scientific and Technical Information of China (English)

    WANG Hao; PENG Xiao-Feng; David M.Christopher

    2005-01-01

    @@ Bubble cycles, including initiation, growth and departure, are the physical basis of nucleate boiling. The presentinvestigation, however, reveals unusual bubble motions during subcooled nucleate boiling on microwires 25 orl00μm in diameter. Two types of bubble motions, bubble sweeping and bubble return, are observed in theexperiments. Bubble sweeping describes a bubble moving back and forth along the wire, which is motion parallelto the wire. Bubble return is the bubble moving back to the wire after it has detached or leaping above thewire. Theoretical analyses and numerical simulations are conducted to investigate the driving mechanisms forboth bubble sweeping and return. Marangoni flow from warm to cool regions along the bubble interface is foundto produce the shear stresses needed to drive these unusual bubble movements.

  19. Boiling Heat Transfer in Circulating Fluidized Beds

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    A model is proposed to predict boiling heat transfer coefficient in a three-phase circulating fluidized bed (CFB), which is a new type of evaporation boiling means for enhancing heat transfer and preventing fouling. To verify the model, experiments are conducted in a stainless steel column with 39mm ID and 2.0m height, in which the heat transfer coefficient is measured for different superficial velocities, steam pressures, particle concentrations and materials of particle. As the steam pressure and particle concentrations increase, the heat transfer coefficient in the bed increases. The heat transfer coefficient increases with the liquid velocity but it exhibits a local minimum. The heat transfer coefficient is correlated with cluster renewed model and two-mechanism method. The prediction of the model is in good agreement with experimental data.

  20. Boiling Heat Transfer in Circulating Fluidized Beds

    Institute of Scientific and Technical Information of China (English)

    张利斌; 李修伦

    2001-01-01

    A model is proposed to predict boiling heat transfer coefficient in a three-phase circulating fluidized bed (CFB), which is a new type of evaporation boiling means for enhancing heat transfer and preventing fouling. To verify the model, experiments are conducted in a stainless steel column with 39 mm ID and 2.0 m height, in which the heat transfer coefficient is measured for different superficial velocities, steam pressures, particle concentrations and materials of particle. As the steam pressure and particle concentrations increase, the heat transfer coefficient in the bed increases. The heat transfer coefficient increases with the liquid velocity but it exhibits a local minimum.The heat transfer coefficient is correlated with cluster renewed model and two-mechanism method. The prediction of the model is in good agreement with experimental data.

  1. Enzyme engineering reaches the boiling point

    OpenAIRE

    Arnold, Frances H.

    1998-01-01

    The boiled enzyme was toppled as a standard enzymology control when researchers in the 1970s started uncovering enzymes that loved the heat (1). Identification of a variety of intrinsically hyperstable enzymes from hyperthermophilic organisms, with optimal growth temperatures of 100°C and above, has piqued academic curiosity (e.g., how do these proteins withstand such ‘‘extreme’’ conditions?) and generated considerable interest for their possible applications in biotechnology (2, 3). The real...

  2. Self-propelled film-boiling liquids

    CERN Document Server

    Linke, H; Melling, L D; Taormina, M J; Francis, M J; Dow-Hygelund, C C; Narayanan, V K; Taylor, R P; Stout, A

    2005-01-01

    We report that liquids perform self-propelled motion when they are placed in contact with hot surfaces with asymmetric (ratchet-like) topology. The pumping effect is observed when the liquid is in the film-boiling regime, for many liquids and over a wide temperature range. We propose that liquid motion is driven by a viscous force exerted by vapor flow between the solid and the liquid.

  3. Nucleate boiling of oxygen under reduced gravity

    Energy Technology Data Exchange (ETDEWEB)

    Kirichenko, Y.A.; Gladchenko, G.M.; Rusanov, K.V.

    1986-03-01

    Experimental results are presented on the coefficients of nucleate boiling heat transfer of oxygen under the conditions of low loading factors at elevated pressures. Based on the statistical distribution of the separation bubble radii and distances between the nucleation sites, a relation is obtained which provides a satisfactory description of the function ..cap alpha..(q) in case of deteriorated heat transfer at eta = g/g/sub n/ < 0.1.

  4. Nucleate boiling of oxygen under reduced gravity

    International Nuclear Information System (INIS)

    Experimental results are presented on the coefficients of nucleate boiling heat transfer of oxygen under the conditions of low loading factors at elevated pressures. Based on the statistical distribution of the separation bubble radii and distances between the nucleation sites, a relation is obtained which provides a satisfactory description of the function α(q) in case of deteriorated heat transfer at eta = g/g/sub n/ < 0.1

  5. Serious accidents on boiling water reactors (BWR)

    International Nuclear Information System (INIS)

    This short document describes, first, the specificities of boiling water reactors (BWRs) with respect to PWRs in front of the progress of a serious accident, and then, the strategies of accident management: restoration of core cooling, water injection, core flooding, management of hydrogen release, depressurization of the primary coolant circuit, containment spraying, controlled venting, external vessel cooling, erosion of the lower foundation raft by the corium). (J.S.)

  6. Self-Sustaining Thorium Boiling Water Reactors

    OpenAIRE

    Ehud Greenspan; Jasmina Vujic; Francesco Ganda; Arias, Francisco J.

    2012-01-01

    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 absorber...

  7. European simplified boiling water reactor (ESBWR) plant

    International Nuclear Information System (INIS)

    This paper covers innovative ideas which made possible the redesign of the US 660-MW Simplified Boiling Water Reactor (SBWR) Reactor Island for a 1,200-MW size reactor while actually reducing the building cost. This was achieved by breaking down the Reactor Island into multiple buildings separating seismic-1 from non-seismic-1 areas, providing for better space utilization, shorter construction schedule, easier maintainability and better postaccident accessibility

  8. Flow boiling test of GDP replacement coolants

    International Nuclear Information System (INIS)

    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, C4F10 and C4F8, 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/C4F10 mixture tests show boiling heat transfer coefficient degraded to a minimum value with about 25% C4F10 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

  9. Steady State Vapor Bubble in Pool Boiling

    Science.gov (United States)

    Zou, An; Chanana, Ashish; Agrawal, Amit; Wayner, Peter C.; Maroo, Shalabh C.

    2016-02-01

    Boiling, a dynamic and multiscale process, has been studied for several decades; however, a comprehensive understanding of the process is still lacking. The bubble ebullition cycle, which occurs over millisecond time-span, makes it extremely challenging to study near-surface interfacial characteristics of a single bubble. Here, we create a steady-state vapor bubble that can remain stable for hours in a pool of sub-cooled water using a femtosecond laser source. The stability of the bubble allows us to measure the contact-angle and perform in-situ imaging of the contact-line region and the microlayer, on hydrophilic and hydrophobic surfaces and in both degassed and regular (with dissolved air) water. The early growth stage of vapor bubble in degassed water shows a completely wetted bubble base with the microlayer, and the bubble does not depart from the surface due to reduced liquid pressure in the microlayer. Using experimental data and numerical simulations, we obtain permissible range of maximum heat transfer coefficient possible in nucleate boiling and the width of the evaporating layer in the contact-line region. This technique of creating and measuring fundamental characteristics of a stable vapor bubble will facilitate rational design of nanostructures for boiling enhancement and advance thermal management in electronics.

  10. Steady State Vapor Bubble in Pool Boiling.

    Science.gov (United States)

    Zou, An; Chanana, Ashish; Agrawal, Amit; Wayner, Peter C; Maroo, Shalabh C

    2016-01-01

    Boiling, a dynamic and multiscale process, has been studied for several decades; however, a comprehensive understanding of the process is still lacking. The bubble ebullition cycle, which occurs over millisecond time-span, makes it extremely challenging to study near-surface interfacial characteristics of a single bubble. Here, we create a steady-state vapor bubble that can remain stable for hours in a pool of sub-cooled water using a femtosecond laser source. The stability of the bubble allows us to measure the contact-angle and perform in-situ imaging of the contact-line region and the microlayer, on hydrophilic and hydrophobic surfaces and in both degassed and regular (with dissolved air) water. The early growth stage of vapor bubble in degassed water shows a completely wetted bubble base with the microlayer, and the bubble does not depart from the surface due to reduced liquid pressure in the microlayer. Using experimental data and numerical simulations, we obtain permissible range of maximum heat transfer coefficient possible in nucleate boiling and the width of the evaporating layer in the contact-line region. This technique of creating and measuring fundamental characteristics of a stable vapor bubble will facilitate rational design of nanostructures for boiling enhancement and advance thermal management in electronics. PMID:26837464

  11. Evaluation of onset of nucleate boiling models

    International Nuclear Information System (INIS)

    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)

  12. CFD simulation of DEBORA boiling experiments

    Science.gov (United States)

    Rzehak, Roland; Krepper, Eckhard

    2012-08-01

    In this work we investigate the present capabilities of computational fluid dynamics for wall boiling. The computational model used combines the Euler/Euler two-phase flow description with heat flux partitioning. This kind of modeling 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, bubble size and liquid temperature as well as axial profiles of wall temperature. After reviewing the theoretical and experimental basis of correlations used in the ANSYS CFX model used for the calculations, we give a careful assessment of the necessary recalibrations to describe the DEBORA tests. The basic CFX model is validated by a detailed comparison to the experimental data for two selected test cases. Simulations with a single set of calibrated parameters are found to give reasonable quantitative agreement with the data for several tests within a certain range of conditions and reproduce the observed tendencies correctly. Several model refinements are then presented each of which is designed to improve one of the remaining deviations between simulation and measurements. Specifically we consider a homogeneous MUSIG model for the bubble size, modified bubble forces, a wall function for turbulent boiling flow and a partial slip boundary condition for the liquid phase. Finally, needs for further model developments are identified and promising directions discussed.

  13. CFD for Subcooled Flow Boiling: Parametric Variations

    Directory of Open Access Journals (Sweden)

    Roland Rzehak

    2013-01-01

    Full Text Available 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 modeling was previously applied to boiling water under high pressure conditions relevant to nuclear power systems. Similar conditions in terms of the relevant nondimensional 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. Robust predictive capabilities of the modeling require that it is validated for a wide range of parameters. It is known that a careful calibration of correlations used in the wall boiling model is necessary to obtain agreement with the measured data. We here consider tests under a variety of conditions concerning liquid subcooling, flow rate, and heat flux. It is investigated to which extent a set of calibrated model parameters suffices to cover at least a certain parameter range.

  14. Study on film boiling heat transfer and minimum heat flux condition for subcooled boiling, 1

    International Nuclear Information System (INIS)

    In the present paper, film boiling heat transfer and minimum heat flux condition were experimentally studied for subcooled pool boiling of water at atmospheric pressure from a platinum sphere (D = 10 mm). Transient tests of subcooled boiling were conducted from an initial temperature of the sphere about 1500 K. Experimental parameters were liquid subcooling (0 ∼ 75 K) and the depth of immersion of the sphere (0.75 D ∼ 3.0 D). The obtained boiling curves indicated that the ideal depth was 1.8 D. Even for large subcooling conditions, the measured temperature at the minimum heat flux point for such a depth did not remarkably exceed the maximum superheat of water. Further, accounting for the density-viscosity ratio, the analytical equation for subcooled film boiling derived by Hamill and Baumeister was modified. This modified equation was in good agreement with experimental data of water and freon-11 under various conditions of surface geometry, size and system pressure. (author)

  15. Assessment of Boiling Model in a Computational Analysis of the Subcooled Boiling Flow

    International Nuclear Information System (INIS)

    Two-phase flow phenomena are known to be crucial for a nuclear reactor safety, such as a subcooled boiling at the downcomer during a Large-Break Loss-of- Coolant Accident (LBLOCA). For the analysis of a two-phase flow, the two-fluid model is considered as a state-of-the-art model which deals with the mass, momentum and energy of each phase. The interfacial area concentration (IAC), which is defined as the area of interface per unit mixture volume, is one of the most significant parameters in the two-fluid model. In order to resolve the problems of the conventional models for an IAC, an interfacial area transport equation has been developed for an adiabatic bubbly flow or nucleate boiling flow. For the investigation of a boiling flow with a dynamic modeling of the interfacial structure, this study focuses on the development of a computational fluid dynamics (CFD) code with implementing an interfacial area transport equation. A benchmark problem for the subcooled boiling flow is analyzed with the developed code so that the sensitivity on the boiling model can be analyzed

  16. Face Off

    OpenAIRE

    Busby, Lisa

    2010-01-01

    Face off explored and expanded the perceptions of where the DJ could be found by collaborating with the artist David Dixon to create an equipment set up and performance which was a sonic and structural intervention of his work Entangled Practice. This work is part of the Shit! I can DJ project which explores the fringes of what might be considered DJ practice, and seeks to explore and promote experimental DJ practices across fields.

  17. Stability monitoring for boiling water reactors

    Science.gov (United States)

    Cecenas-Falcon, Miguel

    1999-11-01

    A methodology is presented to evaluate the stability properties of Boiling Water Reactors based on a reduced order model, power measurements, and a non-linear estimation technique. For a Boiling Water Reactor, the feedback reactivity imposed by the thermal-hydraulics has an important effect in the system stability, where the dominant contribution to this feedback reactivity is provided by the void reactivity. The feedback reactivity is a function of the operating conditions of the system, and cannot be directly measured. However, power measurements are relatively easy to obtain from the nuclear instrumentation and process computer, and are used in conjunction with a reduced order model to estimate the gain of the thermal-hydraulics feedback using an Extended Kalman Filter. The reduced order model is obtained by estimating the thermal-hydraulic transfer function from the frequency-domain BWR code LAPUR, and the stability properties are evaluated based on the pair of complex conjugate eigenvalues. Because of the recursive nature of the Kalman Filter, an estimate of the decay ratio is generated every sampling time, allowing continuous estimation of the stability parameters. A test platform based on a nuclear-coupled boiling channel is developed to validate the capability of the BWR stability monitoring methodology. The thermal-hydraulics for the boiling channel is modeled and coupled with neutron kinetics to analyze the non-linear dynamics of the closed-loop system. The model uses point kinetics to study core-wide oscillations, and normalized modal kinetics are introduced to study out-of-phase oscillations. The coolant flow dynamics is dominant in the power fluctuations observed by in-core nuclear instrumentation, and additive white noise is added to the solution for the channel flow in the thermal-hydraulic model to generate noisy power time series. The operating conditions of the channel can be modified to accommodate a wide range of stability conditions

  18. Battery thermal management by boiling heat-transfer

    International Nuclear Information System (INIS)

    Highlights: • A thermal management scheme based on boiling heat-transfer is investigated. • Cooling capacity of the working fluid compared to that of air is investigated. • Battery gets fluid to boil, thus boiling heat-transfer occurs from battery to fluid. • Boiling process thermally homogenises the battery. • Boiling process can be influenced by pressure variation. - Abstract: In this study, the ability of a boiling process to thermally condition (homogenisation and cooling) batteries is investigated. Thereto, a series of experiments are performed and discussed. Subjects that are treated are the dielectric property of the proposed cooling fluid, its cooling capability compared to that of air, the ability of the boiling fluid to thermally homogenise a battery and the influence of pressure variation on the boiling process. It turns out that the proposed cooling fluid conducts no electricity, has good cooling characteristics compared to those of air and, when boiling, is able to thermally homogenise the battery. Furthermore, pressure variation seems to offer a good method to regulate the boiling process

  19. Boiling radial flow in fractures of varying wall porosity

    Energy Technology Data Exchange (ETDEWEB)

    Barnitt, Robb Allan

    2000-06-01

    The focus of this report is the coupling of conductive heat transfer and boiling convective heat transfer, with boiling flow in a rock fracture. A series of experiments observed differences in boiling regimes and behavior, and attempted to quantify a boiling convection coefficient. The experimental study involved boiling radial flow in a simulated fracture, bounded by a variety of materials. Nonporous and impermeable aluminum, highly porous and permeable Berea sandstone, and minimally porous and permeable graywacke from The Geysers geothermal field. On nonporous surfaces, the heat flux was not strongly coupled to injection rate into the fracture. However, for porous surfaces, heat flux, and associated values of excess temperature and a boiling convection coefficient exhibited variation with injection rate. Nucleation was shown to occur not upon the visible surface of porous materials, but a distance below the surface, within the matrix. The depth of boiling was a function of injection rate, thermal power supplied to the fracture, and the porosity and permeability of the rock. Although matrix boiling beyond fracture wall may apply only to a finite radius around the point of injection, higher values of heat flux and a boiling convection coefficient may be realized with boiling in a porous, rather than nonporous surface bounded fracture.

  20. Flow boiling heat transfer at low liquid Reynolds number

    International Nuclear Information System (INIS)

    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)

  1. Subcooled boiling model to simulate upward vertical flow boiling at low pressures

    International Nuclear Information System (INIS)

    A new model for upward vertical subcooled flow boiling at low pressure is proposed. The model considers the most relevant closure relationships of one-dimensional thermalhydraulic codes that are important for prediction of vapor contents in the channel: wall evaporation model, condensation model, flow regime transition criterion and drift-flux model. The new model was incorporated in the current version of the thermal-hydraulic computer code RELAP5/MOD3.2.2 Gamma. The modified code was validated against a number of published low-pressure subcooled boiling experiments, and in contrast to the current code, shows good agreement with experimental data. The presented analysis also leads to a better understanding of the basic mechanisms of subcooled flow boiling at low pressure.(author)

  2. Signing off

    Science.gov (United States)

    2001-01-01

    A new gadget for physics teachers everywhere Recently released onto the market we can now present the Deluxe Remote Control for use by teachers everywhere. It has several innovative features which should help hard pressed teachers organize their lives and their classrooms. At the top of the remote control, easily accessed, are three OFF buttons. These will help the physics teacher reintegrate with society at the end of the day, at weekends and in the holidays. In the important top left position is the 'Teacher Voice OFF' This allows the teacher to speak normally, rather than continue as if addressing a class of 30 across a noisy swimming pool. No less important, two other buttons switch off the teacher's organizing instinct (so that there is no inclination at all to organize any large group of people encountered on holiday into a line) and the teacher's analysing instinct (so that never again will you end up wondering why the waiter asked you that question rather than just ordering the wine). The class control feature allows the teacher to select at will from fully integrated fun, soft and stern modes. Switching time is less than one second, leading to effortless changes of mood in the classroom. In these times when records must be kept up to date teachers will value the 'mark by' feature. Most remotes have featured 'mark by weight' and the very old fashioned 'mark by worth' commands for some time (although this last one, actually evaluating whether a piece of work is good or not, is seldom used). The new breakthrough comes with the 'auto marking' feature for which the anticipated demand is colossal. Most teachers already use their Principal control on existing products. This remote has the normal mute, pause and, important for after-school staff meetings, fast forward functionality. Social interaction is a new concept in physics teacher remote controls. Most teachers have preferred the pause or off settings so these are still provided. The Formal setting is

  3. Regression analysis of post-CHF flow boiling data

    International Nuclear Information System (INIS)

    The successful application of statistical analysis in systematic investigations of heat transfer data for boiling water beyond the critical heat flux is described. Multiple linear regression analysis together with statistical tests of correlations and data were used in this study. Data from a number of experiments encompassing film and transition boiling in several geometries were correlated by boiling regime, by geometry, and in aggregate. Error estimates and uncertainty bounds were specified for all such correlations. (U.S.)

  4. Boiling of an Emulsion in a Yield Stress Fluid

    OpenAIRE

    Guéna, Geoffroy; Wang, Ji; D'Espinose, Jean-Baptiste; Lequeux, François; Talini, Laurence

    2010-01-01

    International audience We report the boiling behaviour of pentane emulsified in a yield stress fluid, a colloidal clay (Laponite) suspension. We have observed that a superheated state is easily reached: the emulsion, heated more than 50°C above the alkane boiling point, does not boil. Superheating is made possible by the suppression of heterogeneous nucleation in pentane, resulting from the emulsification process, a phenomenon evidenced decades ago in studies of the superheating of two pha...

  5. 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...

  6. Prospective Chemistry Teachers’ Understanding of Boiling: A Phenomenological Study

    OpenAIRE

    CANPOLAT, Nurtaç; PINARBAŞI, Tacettin

    2012-01-01

    This study investigates chemistry prospective teachers’ views regarding boiling phenomenon, and provides a concept analysis on the nature of boiling together with suggestions on how to teach boiling phenomenon in the light of literature and findings of this study. The sample of this study consists of 18 senior prospective chemistry teachers who attend chemistry teacher training program. Data were collected by discussions with the participants. The discussions were specifically focused on pros...

  7. Boiled Water Temperature Measurement System Using PIC Microcontroller

    OpenAIRE

    A.T.KARUPPIAH, AZHA. PERIASAMY, P.RAJKUMAR

    2013-01-01

    The measurement system for temperature of boiled water is a critical task in industry. In this paper we designed and implemented a PIC micro controller based boiled water temperature measurement system using PIC 18F452 and national semiconductors LM35 temperature sensor. The designing system is used to measure the tank I boiled water temperature value. If the temperature value reaches the set value high temperature relay board becomes ON to control the solenoid valve. The high temperature of ...

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

    International Nuclear Information System (INIS)

    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)

  9. Mass exchange calculation in a wall layer when water boiling

    International Nuclear Information System (INIS)

    Physical sense and mass exchange characteristics of liquid near-the-wall layer under boiling conditions were attempted to be stated. Equations of material and thermal balance were used to describe the mass exchange characteristics. Technique to calculate circulation ratio in the near-the-wall layer under boiling of under-heated and saturated water was suggested on the basis of the derived expressions. Comparison results of calculated and experimental data were analyzed for full-scale boiling

  10. Boiling Experiment Facility for Heat Transfer Studies in Microgravity

    Science.gov (United States)

    Delombard, Richard; McQuillen, John; Chao, David

    2008-01-01

    Pool boiling in microgravity is an area of both scientific and practical interest. By conducting tests in microgravity, it is possible to assess the effect of buoyancy on the overall boiling process and assess the relative magnitude of effects with regards to other "forces" and phenomena such as Marangoni forces, liquid momentum forces, and microlayer evaporation. The Boiling eXperiment Facility is now being built for the Microgravity Science Glovebox that will use normal perfluorohexane as a test fluid to extend the range of test conditions to include longer test durations and less liquid subcooling. Two experiments, the Microheater Array Boiling Experiment and the Nucleate Pool Boiling eXperiment will use the Boiling eXperiment Facility. The objectives of these studies are to determine the differences in local boiling heat transfer mechanisms in microgravity and normal gravity from nucleate boiling, through critical heat flux and into the transition boiling regime and to examine the bubble nucleation, growth, departure and coalescence processes. Custom-designed heaters will be utilized to achieve these objectives.

  11. Investigation of Enhanced Boiling Heat Transfer from Porous Surfaces

    Institute of Scientific and Technical Information of China (English)

    LinZhiping; MaTongze; 等

    1994-01-01

    Experimental investigations of boiling heat transfer from porous surfaces at atmospheric pressure were performne.The porous surfaces are plain tubes coverd with metal screens.V-shaped groove tubes covered with screens,plain tubes sintered with screens.and V-shaped groove tubes sintered with screens,The experimental results show that sintering metal screens around spiral V-shaped groove tubes can greatly improve the boiling heat transfer,The boiling hystesis was observed in the experiment.This paper discusses the mechanism of the boiling heat transfer from those kinds of porous surfaces stated above.

  12. The entrance effect on subcooled boiling in heated channels

    International Nuclear Information System (INIS)

    One of the major problems in the analysis of diabatic two-phase flows concerns the effect of thermodynamic nonequilibrium between the phases. In particular, this effect applies to forced-convection subcooled boiling in boiling water reactors (BWRs). An approach commonly used to evaluate the void distribution along reactor coolant channels is based on one-dimensional models of combined two-phase flow and boiling heat transfer. In the subcooled boiling region, the rate of phase change is governed mainly by the lateral transport of the vapor phase toward the subcooled liquid; thus, the related processes cannot be mechanistically modeled by one-dimensional, axially dependent models. Consequently, most existing subcooled boiling models are based on experimental correlations for parameters such as the onset of nucleate boiling (ONB) and the net vapor generation rate. This paper presents the results of analysis of subcooled boiling phenomena in the developing flow region of a boiling channel, based on a mechanistic two-dimensional, two-fluid model. The effect of turbulence has been accounted for by a k-ε model. The PHOENICS code was used to solve the governing mass, momentum, and energy conservation equations in both the nonboiling and boiling regions. The parameters calculated by the model include radially and axially dependent distributions of the local void fraction, temperatures and velocities of both phases, and the axial distribution of wall temperature

  13. Fuel assembly for a boiling water reactor

    International Nuclear Information System (INIS)

    A boiling water reactor fuel assembly is described which has vertical fuel rods and guide tubes positioned below the fuel rods and receiving control rod fingers and acting as water pipes, the guide tubes each being formed of a plurality of parts including a part secured to a grid plate positioned in the fuel assembly container, and low parts which fit into holes formed in the bottom of the fuel assembly. There is a flexible connection between the upper and lower parts of the guide tubes to allow for a certain tolerance in the procedure of manufacturing the various parts to allow insertion of the fuel rod bundle into the fuel assembly container

  14. Fuel recycling in boiling water reactors

    International Nuclear Information System (INIS)

    The present study confirms the feasibility of inserting mixed-oxid-fuel assemblies (MOX-FA) in boiling-water reactors in conjunction with reactivity-equivalent uranium-fuel assemblies. First, the established calculation methods were extended according to the specific MOX-uranium mutual interaction effects. Then, typical bundle-structures were analysed according to their neutron-physical features. The reactor-simulations show a non-critical behaviour with respect to limiting conditions and reactivity control. The variation of the isotopic composition and the plutonium content with its effects on the physical features was considered. (orig.) With 6 refs., 3 tabs., 29 figs

  15. 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

  16. Simulation of Boiling Water Reactor dynamics

    International Nuclear Information System (INIS)

    This master thesis describes a mathematical model of a boiling water reactor and address the dynamic behaviour of the neutron kinetics, boilding dynamics and pressur stability. The simulation have been done using the SIMNON-program. The meaning were that the result from this work possibly would be adjust to supervision methods suitable for application in computer systems. This master thesis in automatic control has been done at the Department of Automatic Control, Lund Institute of Technology. The initiative to the work came from Sydkraft AB. (author)

  17. 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

  18. Outline of advanced boiling water reactor

    International Nuclear Information System (INIS)

    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)

  19. Fuel assembly for a boiling water reactor

    International Nuclear Information System (INIS)

    A fuel assembly for a boiling water reactor comprises a plurality of fuel rods which constitute four partial bundles and are surrounded by a fuel channel system comprising one partial tube for each partial bundle. Each of the four partial bundles rests on a bottom tie plate and is positioned with respect to the others by means of a common top tie plate which is provided with a lifting loop which is sufficiently strong to be able to lift the four partial bundles simultaneously, a major part of the lifting force being transmitted to said bottom tie plates via a plurality of supporting fuel rods

  20. Theoretical investigations of the transition from bubble boiling to film boiling at forced convection

    International Nuclear Information System (INIS)

    The model laws for the initial film boiling at forced convection are realized in vertical tubes. The local conditions in the investigated area were regarded to be most effective and sufficient for the description. The theory was confirmed by experimental data. (orig.)

  1. Nucleate Boiling from Smooth and Rough Surfaces--Part 2: Analysis of Surface Roughness Effects on Nucleate Boiling

    OpenAIRE

    McHale, John P.; Garimella, Suresh V.

    2013-01-01

    The effect of surface roughness on nucleate boiling heat transfer is not clearly understood. This study is devised to conduct detailed heat transfer and bubble measurements during boiling on a heater surface with controlled roughness. This second of two companion papers presents an analysis of heat transfer and bubble ebullition in nucleate boiling with new measures of surface roughness: area ratio, surface mean normal angle, and maximum idealized surface curvature. An additional length scale...

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

    International Nuclear Information System (INIS)

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

  3. Unsteady heat transfer during subcooled film boiling

    Science.gov (United States)

    Yagov, V. V.; Zabirov, A. R.; Lexin, M. A.

    2015-11-01

    Cooling of high-temperature bodies in subcooled liquid is of importance for quenching technologies and also for understanding the processes initiating vapor explosion. An analysis of the available experimental information shows that the mechanisms governing heat transfer in these processes are interpreted ambiguously; a more clear-cut definition of the Leidenfrost temperature notion is required. The results of experimental observations (Hewitt, Kenning, and previous investigations performed by the authors of this article) allow us to draw a conclusion that there exists a special mode of intense heat transfer during film boil- ing of highly subcooled liquid. For revealing regularities and mechanisms governing intense transfer of energy in this process, specialists of Moscow Power Engineering Institute's (MPEI) Department of Engineering Thermal Physics conduct systematic works aimed at investigating the cooling of high-temperature balls made of different metals in water with a temperature ranging from 20 to 100°C. It has been determined that the field of temperatures that takes place in balls with a diameter of more than 30 mm in intense cooling modes loses its spherical symmetry. An approximate procedure for solving the inverse thermal conductivity problem for calculating the heat flux density on the ball surface is developed. During film boiling, in which the ball surface temperature is well above the critical level for water, and in which liquid cannot come in direct contact with the wall, the calculated heat fluxes reach 3-7 MW/m2.

  4. Boiling water reactor life extension monitoring

    International Nuclear Information System (INIS)

    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)

  5. Mitigation performance indicator for boiling water reactors

    International Nuclear Information System (INIS)

    All U.S. boiling water reactors (BWRs) inject hydrogen for mitigation of intergranular stress corrosion cracking (IGSCC), and most currently use or plan to use noble metals technology. The EPRI Boiling Water Reactor Vessels and Internals Project (BWRVIP) developed a Mitigation Performance Indicator (MPI) in 2006 to accurately depict to management the status of mitigation equipment and as a standardized way to show the overall health of reactor vessel internals from a chemistry perspective. It is a 'Needed' requirement in the EPRI BWR Water Chemistry Guidelines that plants have an MPI, and use of the BWRVIP MPI is a 'Good Practice'. The MPI is aligned with inspection relief criteria for reactor piping and internal components for U.S. BWRs. This paper discusses the history of the MPI, from its first use for plants operating with moderate hydrogen water chemistry (HWC-M) or Noble Metal Chemical Application (NMCA) + HWC to its more recent use for plants operating with On-Line NobleChem™ (OLNC) + HWC. Key mitigation parameters are discussed along with the technical bases for the indicators associated with the parameters. (author)

  6. Identification of dynamic basins in boiling fluxes

    International Nuclear Information System (INIS)

    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

  7. Boiling induced mixed convection in cooling loops

    International Nuclear Information System (INIS)

    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)

  8. 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)

  9. Visualization of supercritical fluid pseudo-boiling under forced convection

    International Nuclear Information System (INIS)

    Supercritical carbon dioxide flow has been visualized by a white light to inspect forced convection heat transfer. A 'pseudo-boiling' phenomenon which occurred in supercritical range for carbon dioxide flow was supposed to cause heat transfer deterioration. The effect of the pseudo-boiling phenomenon to the heat transfer has been investigated by the visualized images in this study. (author)

  10. 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.

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

    DEFF Research Database (Denmark)

    Munthe Andersen, J. G.; Dix, G. E.; Leonard, J. E.; Sun, K. H.

    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....

  12. Relief valve capacity operating under conditions of working fluid boiling

    International Nuclear Information System (INIS)

    To develop a method for the calculation of relief valve capacity in the mode of working fluid boiling the effect of flowing part geometry and working fluid initial state parameters on the valve capacity is studied experimentally. Dependences for the calculation of the valve capacity in the zones of one-phase coolant flow, its boiling at high underheatings and intensive evaporation are obtained

  13. Study of neutron noise physical model for reactor coolant boiling

    International Nuclear Information System (INIS)

    The neutron noise method has been used to monitoring reactor coolant boiling. Wach-Kosaly model has been used to interpret the neutron noise induced by coolant boiling. The equation based on the model is got and used for calculation. The physical variable with the relation of bubble's velocity is got from the calculated result (autopower spectral density)

  14. Flow boiling experiment study of OTSG in steady state

    International Nuclear Information System (INIS)

    Generally, Integrated Pressurized Water Reactor uses OTSG. In this paper OTSG's flow boiling laws is analyzed and experimented. More elaborate calculating model of two-phase flow section which is divided into different segments is used for theoretical analysis. Flow boiling experiment is operated in two different conditions. The results of experiment proves the calculation model rational and credible. (authors)

  15. Technique for technological calculation of critical flow of boiling water

    International Nuclear Information System (INIS)

    Average values of friction factor and mach number for a critical flow of boiling water are determined on the basis of computerized processing of experimental data. Empirical formula, relating these values, which can be used for technological calculations of critical conditions of boiling water flow through transport pipelines, is derived

  16. Flow boiling heat transfer in volumetrically heated packed bed

    International Nuclear Information System (INIS)

    Highlights: • The onset of nucleate boiling in the volumetrically heated packed bed is researched. • A correlation for predicting qONB is developed. • The effects on boiling heat transfer coefficient are investigated. - Abstract: The volumetrically heated packed bed has been widely utilized in modern industry. However, due to the variability and randomness of packed bed channels, flow boiling heat transfer characteristics becomes complex, and there are no published research regarding this topic. To study flow boiling heat transfer characteristics of volumetrically heated packed beds, electromagnetic induction heating method is used to heat oxidized carbon steel balls adopted to stack the packed bed, with water as coolant in the experiment. The experimental results indicate that heat flux at onset of nucleate boiling (ONB) increases as mass flux and inlet subcooling are increased. A new correlation is developed to predict the ONB heat flux qONB in volumetrically heated packed bed, the predictions by which agree well with the experimental data, and the deviation remains less than 15%. Subcooled flow boiling heat transfer coefficient (hsub) increases with increasing mass flux, and equilibrium quality is slightly affected by heat flux. The saturated flow boiling heat transfer coefficient (hsat) increases with mass flux and equilibrium quality when equilibrium quality is lower than about 0.05, while the nucleate boiling is suppressed when the equilibrium quality exceeds a certain value

  17. CFD modelling of subcooled flow boiling for nuclear engineering applications

    International Nuclear Information System (INIS)

    In this work a general-purpose CFD code CFX-5 was used for simulations of subcooled flow boiling. The subcooled boiling model, available in a custom version of CFX-5, uses a special treatment of the wall boiling boundary, which assures the grid invariant solution. The simulation results have been validated against the published experimental data [1] of high-pressure flow boiling in a vertical pipe covering a wide range of conditions (relevant to the pressurized water reactor). In general, a good agreement with the experimental data has been achieved. To adequately predict the lateral distribution of two-phase flow parameters, the modelling of two-phase flow turbulence and non-drag forces under wall boiling conditions have been also investigated in the paper. (author)

  18. Sensitivity Study for Wall Boiling Model in ANSYS CFX

    International Nuclear Information System (INIS)

    Because boiling heat transfer was crucial for the analysis of operation and safety of both nuclear reactors and conventional thermal power systems, extensive studies have been made to develop a variety of methods either to evaluate the boiling heat transfer coefficient or to assess the onset of critical heat flux (CHF) at various operating conditions of heating channels. Because of the limitation in grid resolution for a CFD simulation in comparison with the microscopic length scales of the wall boiling process, empirical closure for some underlying physical process is needed. The main objective of the present study is to conduct the sensitivity study for wall boiling related models in ANSYS CFX R.14 in order to examine the effect of model components on wall boiling heat transfer in an axis-symmetric vertical heated pipe

  19. Mechanistic Multidimensional Modeling of Forced Convection Boiling Heat Transfer

    Directory of Open Access Journals (Sweden)

    Michael Z. Podowski

    2009-01-01

    Full Text Available Due to the importance of boiling heat transfer in general, and boiling crisis in particular, for the analysis of operation and safety of both nuclear reactors and conventional thermal power systems, extensive efforts have been made in the past to develop a variety of methods and tools to evaluate the boiling heat transfer coefficient and to assess the onset of temperature excursion and critical heat flux (CHF at various operating conditions of boiling channels. The objective of this paper is to present mathematical modeling concepts behind the development of mechanistic multidimensional models of low-quality forced convection boiling, including the mechanisms leading to temperature excursion and the onset of CHF.

  20. Bubble spreading during the boiling crisis: modelling and experimenting in microgravity

    OpenAIRE

    Nikolayev, Vadim; Beysens, D.; Garrabos, Yves; Lecoutre, Carole; Chatain, D.

    2006-01-01

    International audience Boiling is a very efficient way to transfer heat from a heater to the liquid carrier. We discuss the boiling crisis, a transition between two regimes of boiling: nucleate and film boiling. The boiling crisis results in a sharp decrease in the heat transfer rate, which can cause a major accident in industrial heat exchangers. In this communication, we present a physical model of the boiling crisis based on the vapor recoil effect. Under the action of the vapor recoil ...

  1. A contribution to incipient boiling in the case of subcooled boiling with forced convection

    International Nuclear Information System (INIS)

    The literature gives contradictory statements about incipient subcooled boiling. To clear up these contradictions it seems important to study the effect of different thermo- and hydrodynamic parameters, like heating surface load, system pressure, local supercooling, and flowrate. Further influencing quantities investigated here are the concentration dissolved gases and the surface condition of the heat surface. To carry out the experimental investigations a measuring method which has already been used by Mayinger applied. With this method, incipient boiling can be determined as the first measurable heat transfer improvement in comparison with single-phase forced convection. Besides, photographs sould make it possible to give statements on the quantity and size of the bubbles on the heating surface. (orig./GL)

  2. Self-propelled film-boiling liquids

    Science.gov (United States)

    Linke, Heiner; Taormina, Michael; Aleman, Benjamin; Melling, Laura; Dow-Hygelund, Corey; Taylor, Richard; Francis, Matthew

    2006-03-01

    We report that liquids perform self-propelled motion when they are placed in contact with hot surfaces with asymmetric (ratchet-like) topology. Millimeter-sized droplets or slugs accelerate at rates up to 0.1 g and reach terminal velocities of several cm/s, sustained over distances up to a meter. The pumping effect is observed when the liquid is in the film-boiling regime, for many liquids and over a wide temperature range. We propose that liquid motion is driven by a viscous force exerted by vapor flow between the solid and the liquid. This heat-driven pumping mechanism may be of interest in cooling applications, eliminating the need for an additional power source.

  3. Hybrid modelling of a sugar boiling process

    CERN Document Server

    Lauret, Alfred Jean Philippe; Gatina, Jean Claude

    2012-01-01

    The first and maybe the most important step in designing a model-based predictive controller is to develop a model that is as accurate as possible and that is valid under a wide range of operating conditions. The sugar boiling process is a strongly nonlinear and nonstationary process. The main process nonlinearities are represented by the crystal growth rate. This paper addresses the development of the crystal growth rate model according to two approaches. The first approach is classical and consists of determining the parameters of the empirical expressions of the growth rate through the use of a nonlinear programming optimization technique. The second is a novel modeling strategy that combines an artificial neural network (ANN) as an approximator of the growth rate with prior knowledge represented by the mass balance of sucrose crystals. The first results show that the first type of model performs local fitting while the second offers a greater flexibility. The two models were developed with industrial data...

  4. Developments in predicting subcooled flow boiling CHF

    International Nuclear Information System (INIS)

    A two-phase flow model was developed to predict a critical heat flux (CHF) in the subcooled and low quality flow boiling. The CHF formula was derived from the local conservation equations of mass, energy and momentum, together with appropriate constitutive relations. The limiting transverse interchange of mass flux crossing the interface of the bubbly layer and core region is represented, in the local momentum conservation equation, by taking account of the convective shear effects due to the drag force on the wall-attached bubbles. Comparison between the predictions by the proposed model and the experimental CHF data from several sources shows good agreement over a wide range of flow conditions (2 P 20 MPa, 1 D 37.5 mm, 0.035 L 6 m, 450 G 7500 kg/m2s, exit 0.8). Also the model correctly accounts for the effects of flow variables

  5. Signing off

    Science.gov (United States)

    2001-09-01

    Physics Related Aptitude Test As the teacher shortage bites anyone with a degree in science expects to walk into a school and be received, with open arms, as a physics teacher. Are they really suitable? To help you decide Signing Off provides the following invaluable psychometric test. Extensively researched and, for single users only, it comes completely free to Physics Education subscribers! (Copies of this Physics Related Aptitude Test are available to credit-card customers from prat@realripoff.com priced #35 per client, 125 dollars to US customers.) This invaluable psychometric test has been extensively researched. Your first lesson of the new school year introduces the study of electricity. Do you: A Use the notes prepared by your predecessor. B Find a video on electricity and play it to the class. C Arrange a series of exciting practical demonstrations to stimulate the young inquiring mind. D Let the children design and make their own circuits to light flashlight bulbs. Your 14-year-olds have completed a written test on heat and energy. Do you: A Mark correct only the work of students who have written their names neatly at the top LEFT HAND corner, as required. B Only set multiple choice tests, so that the computer can mark them for you. C Mark carefully by hand, explaining in detail to each student exactly how and why they have made errors and adding encouraging comments with lots of praise. D Give out correct sets of answers and allow students to mark their own work. There is a staff social. Do you: A Ask for a definition of the term 'social'. B Ask for a web-based version. C Determine to go, so that you can discuss setting up cross-curricular links with colleagues. D Join the organizing committee. Who do you admire most? A Sir Isaac Newton. B Bill Gates. C Leonardo da Vinci. D Leonardo di Caprio. You are required to teach biology class. Your response is: A Denial. B To ask for an appropriate computer simulation. C To attend a specialized course for biology

  6. Signing off

    Science.gov (United States)

    2001-11-01

    . Appearing on two notes also raises questions about the effect on value of working in several countries. The idea is yet to be fully formulated, but it would be nice if it were exponential. Certainly the fact that New Zealand's hero Rutherford has been represented on the one hundred dollar note, valuing him at 28 Newtons, adds to the idea of an attenuation coefficient. There also seem to be transient effects on value, resulting from the personality of the physicist involved. It seems entirely appropriate that the mercurial Tesla should be represented by the ten billion dollar Yugoslavian note, which was nevertheless worth almost nothing. But of course any discussions of great physicists always involve Einstein. Amazingly he has been seen represented on the Israeli five-pound note, valuing him at about 0.08 Newtons. Before rushing off, in support of the great man, to prove that this is clearly a relativistic aberration, just pause. Perhaps calculating your salary in Einsteins could be really rather good for morale... More about physicists on money can be found at www2.physics.umd.edu/~redish/Money/ Philip Britton Head of Physics, Leeds Grammar School, UK

  7. Signal processing techniques for sodium boiling noise detection

    International Nuclear Information System (INIS)

    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

  8. How DNS could help understanding basic mechanisms of boiling crisis

    International Nuclear Information System (INIS)

    Full text of publication follows: Boiling crisis characterizes the upper bound of the heat transfer (critical heat flux, CHF) between a wall and a boiling liquid at low wall superheat. This limit is associated with the departure from nucleate boiling and the establishment of stable film boiling. The nature of the crisis and the mechanisms of the transition are still objects of study. The large scope of interacting phenomena at different scales involved in boiling processes at high heat fluxes prevents from getting precise information experimentally. As a consequence, a large variety of mechanistic models has been proposed that could not be attested experimentally. The influence of the physical parameters on the value of the critical heat flux is not clearly established and this limits the predictive use of existing correlations. One could remark that all physical phenomena affecting CHF affect nucleate wall boiling as well. Recent experimental results show bubble growth, local dryness of the wall and burnout to be closely linked. One can reasonably retain the hypothesis that boiling crisis is related to a crisis during a single bubble growth event. Models considering a mechanism of local crisis, and scaling the drying process with either thermal or mechanical considerations are reviewed. This study leads to a possible scenario consistent with experimental observations. We present computations using direct numerical simulation which are intended to be a first validation of the proposed mechanism. Next steps consist in studying the interaction between the proposed mechanism and the whole boiling process and its potentiality for being the triggering phenomenon of the boiling crisis in different configurations. (authors)

  9. Boiling heat transfer correlation on the outside of horizontal tube in a condenser

    International Nuclear Information System (INIS)

    diameter, expressed as the fraction of tube diameter and break-off bubble diameter. The experimental constant in the boiling heat transfer correlation was determined from test results. The boiling heat transfer correlation was able to predict the test results within an error of about 20% regardless of the difference of tube diameter for the tube O.D. range from 27.2 mm to 42.7 mm. (author)

  10. Subcooled boiling of nano-particle suspensions on Pt wires

    Institute of Scientific and Technical Information of China (English)

    LI Chunhui; WANG Buxuan; PENG Xiaofeng

    2004-01-01

    An experimental investigation is conducted to explore the subcooled boiling characteristics of nano-particle suspensions on Pt wires. Some phenomena are observed for the boiling of water-SiO2 nano-particle suspensions on Pt wires. The experiments show that there exist not any evident differences for boiling of pure water and of nano-particle suspensions at high heat fluxes. However, bubble overlap phenomenon can be easily found for nano-particle suspensions at low heat fluxes, which probably results from the increase of the attracter force between bubbles and of the bubble mass.

  11. Microbiological Effectiveness of Disinfecting Water by Boiling in Rural Guatemala

    OpenAIRE

    Rosa, Ghislaine; Miller, Laura; Clasen, Thomas

    2010-01-01

    Boiling is the most common means of treating water in the home and the benchmark against which alternative point-of-use water treatment options must be compared. In a 5-week study in rural Guatemala among 45 households who claimed they always or almost always boiled their drinking water, boiling was associated with a 86.2% reduction in geometric mean thermotolerant coliforms (TTC) (N = 206, P < 0.0001). Despite consistent levels of fecal contamination in source water, 71.2% of stored water sa...

  12. Boiling of Refrigerant R-113. Three-dimensional numerical analysis

    International Nuclear Information System (INIS)

    In this paper a forced convective boiling of Refrigerant R-113 in a vertical annular channel has been simulated by the CFX-5 code. The employed subcooled boiling model uses a special treatment of the wall boiling boundary, which assures the grid invariant solution. The simulation results have been validated against the published experimental data. In general a good agreement with the experimental data has been achieved, which shows that the current model may be applied for the Refrigerant R-113 without significantly changing the model parameters. The influence of non-drag forces, bubble diameter size and interfacial drag model on the numerical results has been investigated as well. (author)

  13. Boiling Heat Transfer on Porous Surfaces with Vapor Channels

    Institute of Scientific and Technical Information of China (English)

    吴伟; 杜建华; 王补宣

    2002-01-01

    Boiling heat transfer on porous coated surfaces with vapor channels was investigated experimentally to determine the effects of the size and density of the vapor channels on the boiling heat transfer. Observations showed that bubbles escaping from the channels enhanced the heat transfer. Three regimes were identified: liquid flooding, bubbles in the channel and the bottom drying out region. The maximum heat transfer occurred for an optimum vapor channel density and the boiling heat transfer performance was increased if the channels were open to the bottom of the porous coating.

  14. Study on instability of natural circulation induced by subcooled boiling

    International Nuclear Information System (INIS)

    The best estimate system analysis code RELAP5 was used to analyze the natural circulation systems. The instability boundaries of one natural circulation system were obtained under different conditions. According to present results, most of the boundary points were found in the low subcooled boiling zone. The natural circulation systems can tolerate high subcooled boiling, and the disturbance of bubbles departing from the wall and condensing in the subcooled boiling region may be the inherent source to induce the instability, then the flow oscillations can become self-sustained and evolve because of the phase differences among system driving force, resistance and flow rate. (authors)

  15. 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…

  16. 40 CFR 180.1056 - Boiled linseed oil; exemption from requirement of tolerance.

    Science.gov (United States)

    2010-07-01

    ... 40 Protection of Environment 23 2010-07-01 2010-07-01 false Boiled linseed oil; exemption from... From Tolerances § 180.1056 Boiled linseed oil; exemption from requirement of tolerance. Boiled linseed... “boiled linseed oil.” This exemption is limited to use on rice before edible parts form....

  17. Visualization of boiling flow structure in a natural circulation boiling loop

    Energy Technology Data Exchange (ETDEWEB)

    Karmakar, Arnab; Paruya, Swapan, E-mail: swapanparuya@gmail.com

    2015-04-15

    Highlights: • Vapor–liquid jet flows in natural circulation boiling loop. • Flow patterns and their transitions during geysering instability in the loop. • Evaluation of the efficiency of the needle probe in detecting the vapor–liquid and boiling flow structure. - Abstract: The present study reports vapor–liquid jet flows, flow patterns and their transitions during geysering instability in a natural circulation boiling loop under varied inlet subcooling ΔT{sub sub} (30–50 °C) and heater power Q (4–5 kW). Video imaging, voltage measurement using impedance needle probe, measurement of local pressure and loop flow rate have been carried out in this study. Power spectra of the voltage, the pressure and the flow rate reveal that at a high ΔT{sub sub} the jet flows have long period (21.36–86.95 s) and they are very irregular with a number of harmonics. The period decreases and becomes regular with a decrease of ΔT{sub sub}. The periods of the jet flows at ΔT{sub sub} = 30–50 °C and Q = 4 kW are in close agreement with those obtained from the video imaging. The probe was found to be more efficient than the pressure sensor in detecting the jet flows within an uncertainty of 9.5% and in detecting a variety of bubble classes. Both the imaging and the probe consistently identify the bubbly flow/vapor-mushrooms transition or the bubbly flow/slug flow transition on decreasing ΔT{sub sub} or on increasing Q.

  18. Prediction of transition boiling heat transfer by artificial neural network

    International Nuclear Information System (INIS)

    Based on the capability of nonlinear mapping of artificial neural network, a neural network is presented to predict the transition boiling heat transfer in vertical annulus and vertical tube. The predicting results show good accordance with the experimental results

  19. Numerical model of post-DNB film boiling heat transfer

    International Nuclear Information System (INIS)

    It is proposed in this paper a physical model for the film boiling heat transfer. The corresponding mathematical descriptions are given in details and the heat transfer characteristic of post-DNB film boiling is analyzed. The numerical model of post-DNB film boiling heat transfer is obtained as the empirical value of the coefficient is determined by the experimental data. The numerical model is compared with the experimental data of different parameters and other numerical models, and the statistical deviations are calculated. The calculating results of the numerical model in this paper show good agreement with the experimental data, and the numerical model in this paper has comprehensive applicability compared with other numerical models. The effects of thermal-hydraulic parameters on the post-DNB film boiling heat transfer have been numerically researched using the numerical model in this paper. The calculating results are as same as the experimental results. (authors)

  20. Heat transfer phenomena related to the boiling crisis

    International Nuclear Information System (INIS)

    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)

  1. Transition from boiling to two-phase forced convection

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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

  4. 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.

  5. Boiling water reactor simulator. Workshop material

    International Nuclear Information System (INIS)

    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

  6. Theory of hydraulic stability of boiling channels

    International Nuclear Information System (INIS)

    A framework of boiling channel stability theory is analyzed. The fundamental equations are those of STABLE code: Three conservation laws of mass, energy and momentum applied to one-dimensional channel, together with Bankoff' slip and Marinelli-Nelson's pressure drop correlation. These equations are analyzed to yield ''Void Equation'', ''Linearized Void Equation'', ''Volume Conservation Law'' and the ''Flow Impedance'' R(s), defined by the dynamic response of pressure drop to the inlet flow. The impedance contains all the information such a stability index, dominant frequency and damping ratio. It is shown that R is a sum of the form R sub(IA) + N sub(F)-1R sub(D) + N sub(R)R sub(R) + N sub(OR), where N's are non-dimensional parameters and R's characteristic impedances determined by three kinds of parameters, N sub(X), N sub(s) and the power distribution parameter. Systematic edition of the characteristic impedances according to the non-dimensional parameters will reduce the need for case-by-case STABLE calculations. Hydraulic stability of BWR channels under constant system pressure, is a phenomenon with three parameters in view of complexity. Furthermore an analysis is conducted to confirm the above stability structure and three typical instabilities are identified. (auth.)

  7. Burnout heat flux in natural flow boiling

    International Nuclear Information System (INIS)

    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 50C. Obtained critical heat flux varied from 24.46 to 62.9 watts/cm2. 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

  8. Heat removal from a completely blocked breeder element by sodium boiling

    International Nuclear Information System (INIS)

    In order to find out how a fluid flow streaming off as a film depends on the gas phase being in counterflow, two experimental apparatures were designed and built, and measurements were made. For this purpose, the flow conditions occuring in a SNR breeder element completely blocked and boiled empty were simulated by a 31-rod bundle. A relation derived for the steady-state equilibrium between film and gas flow was used for the evaluation of the measured results. In the experiments, limiting curves could be established showing the maximum film rate flowing off as a function of the superposed gas flow rate. In addition the equilibrium parameters gas velocity, film velocity, and film thickness were determined for the material combination water/air. The influence of ths spacers on the film flow was also accounted for the series of the measurements carried out. It appeared that by mounting a honeycomb spacer in the test sections the maximum film rate flowing off is drastically reduced by about 75 to 85% for the same flow rate. Mounting a second spacer, however, had no further influence on the film flowing off. It was possible to explain this phenomenon by means of the course of the experiments. (orig.)

  9. Characteristics of phenomenon and sound in microbubble emission boiling

    International Nuclear Information System (INIS)

    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)

  10. Numerical Simulation of Pool Boiling from Reentrant Type Structured Surfaces

    OpenAIRE

    Dietl, Jochen

    2015-01-01

    Enhancement of heat transfer in pool boiling can be achieved by employing a structured surface. So called reentrant type surfaces, consisting of subsurface tunnels connected through pores with the pool, were found to strongly improve the performance of heat exchanger tubes. Although employed since decades, several of the processes within the tunnel are not understood and the presented models are not able to predict the different boiling modes. With the rapid development of numerical method...

  11. Hysteresis of boiling for different tunnel-pore surfaces

    OpenAIRE

    Pastuszko Robert; Piasecka Magdalena

    2015-01-01

    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 carrie...

  12. Flow Boiling in straight heated tube under microgravity conditions

    OpenAIRE

    Narcy, Marine; COLIN, Catherine

    2013-01-01

    Boiling two-phase flow can transfer large heat fluxes with small driving temperature differences, which is of great interest for the design of high-performance thermal management systems applied to space platforms and on-board electronics cooling in particular. However, such systems are designed using ground-based empirical correlations, which may not be reliable under microgravity conditions. Therefore, several two-phase flow (gas-liquid flow and boiling flow) experiments have been conducted...

  13. Design certification program of the simplified boiling water reactor

    International Nuclear Information System (INIS)

    General Electric (GE), the US Department of Energy, the Electric Power Research Institute (EPRI), and utilities are undertaking a cooperative program to enable advanced light water reactor (ALWR) designs to be certified by the US Nuclear Regulatory Commission (NRC). GE is seeking to certify two advanced plants; the Advanced Boiling Water Reactor (ABWR) and the Simplified Boiling Water Reactor (SBWR). Both plants use advanced features that build on proven BWR technology

  14. Modeling acid-gas generation from boiling chloride brines

    OpenAIRE

    Sonnenthal Eric; Spycher Nicolas; Zhang Guoxiang; Steefel Carl

    2009-01-01

    Abstract Background This study investigates the generation of HCl and other acid gases from boiling calcium chloride dominated waters at atmospheric pressure, primarily using numerical modeling. The main focus of this investigation relates to the long-term geologic disposal of nuclear waste at Yucca Mountain, Nevada, where pore waters around waste-emplacement tunnels are expected to undergo boiling and evaporative concentration as a result of the heat released by spent nuclear fuel. Processes...

  15. Adiabatic boiling of two-phase coolant in upward flow

    International Nuclear Information System (INIS)

    A mathematical model of the process of adiabatic boiling (self-condensation) of a two-phase coolant in upward (downward) flow is developed. The model takes account of changes in phase properties with static pressure decrease. The process is investigated numerically. Approximate analytical formulas for design calculations are obtained. It is shown that effects of adiabatic boiling (self-condensation) should be taken into account when calculating two-phase coolant flow in stretched vertical channels

  16. Development of an experimental apparatus for nucleate boiling analysis

    International Nuclear Information System (INIS)

    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)

  17. Thermodynamic crisis in boiling flow. Observation of the flicker noise

    International Nuclear Information System (INIS)

    The results of the experimental studies on both the characteristics of the boiling liquid flow (discharge, jet reactive force), emanating through a short channel, and the local pulsations in the flow are presented. The identified effects - the flow critical mode, sharp decrease in the value of the reactive force, pulsations with the 1/f spectrum (the flicker noise) are discussed with attraction of the notion on the boiling thermodynamic crisis

  18. Modeling of Heat Exchange with Developed Nucleate Boiling on Tenons

    OpenAIRE

    A. V. Оvsiannik

    2014-01-01

    The paper proposes a thermal and physical model for heat exchange processes with developed nucleate boiling on the developed surfaces (tenons) with various contours of heat transfer surface. Dependences for calculating convective heat exchange factor have been obtained on the basis of modeling representation. Investigations have shown that an intensity of convective heat exchange does not depend on tenon profile when boiling takes place on the tenons. The intensity is determined by operating ...

  19. Visualization of pool boiling on downward-facing convex surfaces

    International Nuclear Information System (INIS)

    Visualizations and quenching experiments were performed to investigate effect of material properties on pool boiling from downward-facing, convex stainless steel and copper surfaces in saturated water. Video images showed that more than one boiling regimes can co-exist on the surface. Maximum heat flux (MHF) occurred first at lowermost position, then propagated radially outward to higher inclination positions and its local value decreased with increased inclination. However, the wall superheats corresponding to MHF were independent of the local surface inclinations. MHF propagated ∼10 times slower on stainless-steel than on copper and was ∼12% and 40% lower on stainless-steel than on copper at θ = 0 degree and θ 7.91 degree, respectively. Results confirmed that transition boiling consisted of two distinct regions: high wall superheat, in which heat flux increased relatively slowly, and low wall superheat, in which heat flux increased precipitously with time. Nuclear boiling regime also consisted of two distinct regions: high heat flux nucleate boiling, in which heat flux decreased with increased inclination, and low heat flux nucleate boiling, in which heat flux increased with increased inclination

  20. Contamination of Foods Boiled in Containers of Different Materials

    International Nuclear Information System (INIS)

    The data presented in this communication deals with the contamination of milk and tomatoes juice, that are a daily consumed by a broad spectrum of Egyptian population when boiled in different types of utensils namely, glass, tin old and new aluminum, pottery, tefal, and stainless steel with heavy metals. The elements determined were sodium, calcium, potassium, iron, manganese, zinc, strontium, cobalt, cadmium, mercury, tin, rubidium, ytterbium and antimony. The technique used for the simultaneous determination of these elements was the instrumental neutron activation analysis. In the light of the obtained results, it was suggested that glass utensils are preferred for boiling milk on the others types of utensils. Also it was found that there is no danger when milk is boiled in other types of utensils since the concentration of both essential and hazardous metals were in the tolerable range. In case of boiling tomatoes juice in the aforementioned types of utensils, it was found that there was no distinct difference between the elemental content in the samples boiled in those types of utensils, thus it can be deduced that all these types of utensils can be used safety for boiling or cooking tomatoes juice

  1. Coupling of wall boiling with discrete population balance model

    International Nuclear Information System (INIS)

    A coupling between a polydisperse population balance method (Multiple Size Group Model - MUSIG) and the RPI wall boiling model for nucleate subcooled boiling has been implemented in ANSYS CFX. It allows more accurate prediction of the interfacial area density for mass, momentum and energy transfer between phases in comparison to the usual local-monodisperse bubble size assumption and underlying bulk bubble diameter correlations as they are commonly used in boiling flow applications like e.g. the prediction of subcooled nucleate boiling in rod bundles and fuel assemblies of PWR. The paper outlines the methodology of the coupled CFD model, which automatically avoids possible inconsistencies in the model formulation for the heated wall, when the generated steam bubbles on the heater surface are injected exactly in the bubble size class corresponding to the predicted bubble departure diameter. The coupling of the RPI wall boiling model and the MUSIG model has been implemented for both homogenous/inhomogeneous variants of the MUSIG model. The paper presents the validation of the coupled modeling approach for the well known test case of nucleate subcooled boiling of R113 refrigerant in a circular annulus with inner heated rod based on the experiments of Roy et al. ANSYS CFX results with the newly implemented approach as well as comparison to data and locally-monodisperse simulations are provided. (author)

  2. Film boiling of R-11 on liquid metal surfaces

    International Nuclear Information System (INIS)

    An interesting problem is the effect of an immiscible liquid heating surface on the process of film boiling. Such surfaces raise questions concerning interface stability to disturbances, effects of gas bubbling, and vapor explosions in layered systems. The specific motivation for this study was to investigate film boiling from a liquid surface with application to cooling of molten reactor core debris by an overlying pool of reactor coolant. To investigate this phenomenon, and apparatus consisting of a nominal six-inch diameter steel vessel to hold the liquid metal and boiling fluid was constructed; coolant reservoirs, heaters, controllers, and allied instrumentation were attached. A transient energy balance was performed on the liquid metal pool by a submerged assembly of microthermocouples in the liquid metal and an array of thermocouples on the wall of the test vessel. The thermocouple data were used to determine the boiling heat flux as well as the boiling superheat. On an average basis, the deviation between the prediction of the Berenson model and the experimental data was less than one percent when Berenson was corrected for thermal radiation effects. Evidence from visualization tests of R-11 in film boiling over molten metal pools to superheats in excess of 600 K supports this conclusion. 13 refs

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

    International Nuclear Information System (INIS)

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

  4. Passive Capillary Pumped Cryocooling System for Zero-Boil-Off Cryogen Storage Tanks Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Significant cost and weight savings of a space mission can be achieved by improving the cryogenic storage technology. Added cryogen mass due to the cryogen...

  5. Advanced, Long-Life Cryocooler Technology for Zero-Boil-Off Cryogen Storage Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Long-life, high-capacity cryocoolers are a critical need for future space systems utilizing stored cryogens. The cooling requirements for planetary and...

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  8. Development of sodium boiling model, 'SOBOIL'

    International Nuclear Information System (INIS)

    The objective of this research is to develop an algorithm for the sodium boiling modeling, essential to the KALIMER HCDA Analysis. The basic theory is based on the 'Muti-slug Ejection Model' in SAS2A. It allows a finite number of bubbles in a channel at any time, and a formed bubble fills the whole cross section of the coolant channel except for liquid film left on the cladding surface. The essence of the model is to estimate the bubble pressure and temperature by balancing the bubble state change with the energy transfer into the bubble at both the wall and interfaces. It assumes that the bubble is saturated with a uniform pressure and temperature, and a bubble is generated when the coolant temperature exceeds a specified superheat. The period of the algorithm development has been divided into two phases. The algorithm development and application during the first phase is limited to the active fuel region where relatively simple physical phenomena are anticipated under ULOHS accident conditions in KALIMER, because it is favorable to the verification of the basic algorithm. The main revision is made during the 2nd phase to take into account mass transfer between the liquid film and bubble, bubble collapse, and coalescence as a liquid slug diminishes. In conclusion, the model represents the anticipated physical phenomena reasonably. The bubble has grown rapidly in consistence with the previous model predictions. Instability, however, is found in the wall heat transfer because the bubble size exceeds a certain value, over which the homogeneous model is no longer valid. Therefore, the model is expected to be improved by taking account of the pressure drop due to vapor flow inside such a large bubble to extend its applicability

  9. Numerical simulation and artificial neural network modeling of natural circulation boiling water reactor

    International Nuclear Information System (INIS)

    Numerical simulation of natural circulation boiling water reactor is important in order to study its performance for different designs and under various off-design conditions. Numerical simulations can be performed by using thermal-hydraulic codes. Very fast numerical simulations, useful for extensive parametric studies and for solving design optimization problems, can be achieved by using an artificial neural network (ANN) model of the system. In the present work, numerical simulations of natural circulation boiling water reactor have been performed with RELAP5 code for different values of design parameters and operational conditions. Parametric trends observed have been discussed. The data obtained from these simulations have been used to train artificial neural networks, which in turn have been used for further parametric studies and design optimization. The ANN models showed error within ±5% for all the simulated data. Two most popular methods, multilayer perceptron (MLP) and radial basis function (RBF) networks, have been used for the training of ANN model. Sequential quadratic programming (SQP) has been used for optimization

  10. Study of film boiling collapse behavior during vapor explosion

    International Nuclear Information System (INIS)

    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.)

  11. Inspection of Pool Boiling with Superhydrophilic and Superhydrophobic Coating

    Energy Technology Data Exchange (ETDEWEB)

    Son, Gyumin; Moon, Sung Bo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

    2015-05-15

    In conventional nuclear power plants, increasing critical heat flux (CHF) margin by converting existing parts is economically meaningful since it means overall energy production increase without building additional power plants. There were researches to enhance margin from the very beginning of the commercialization of nuclear power plants and many efforts have led to current model of plants, optimized for both safety and production efficiency. Examples are mixing vane which is actually applied to plants nowadays, using nanofluids to enhance heat transfer coefficient (HTC), trying porous surfaces and so on. Takata et al. studied effects of surface wettability by using hydrophobic coating and observed enhanced nucleate boiling at coated surface regions. Betz et al. experimented superhydrophilic (SHPi), superhydrophobic (SHPo), and superbiphilic surfaces. Results indicate heat transfer coefficient enhancement due to increase of nucleation sites by hydrophobic regions and constrained diameter of growing bubbles by hydrophilic regions. Although it would be rough to apply their concept to real reactor coolant surface wall, understanding the possibility of enhanced boiling is meaningful. In this paper, SHPi and SHPo coatings were applied to wire at traditional pool boiling experiment by Nukiyama. By observing altered CHF margin and nucleate boiling, the effects of each coating and their tendencies are discussed. SHPi, SHPo and bare wire's pool boiling was investigated and their boiling graphs were discussed. SHPi shows enhancement in CHF while SHPo's case is more complicated since there were variables like partial CHF or micro scale bubbles. Additional experiment could be comparing HTC, checking whether hydrophobic wire's nucleate boiling enhancement can exceed the decreased CHF margin. More sophisticated method to remove unwanted bubbles should be considered such as using degassed water.

  12. Study of film boiling collapse behavior during vapor explosion

    Energy Technology Data Exchange (ETDEWEB)

    Yagi, Masahiro; Yamano, Norihiro; Sugimoto, Jun [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; 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.)

  13. Fuel performance in the Barsebeck boiling water reactors (Unit 1 and 2)

    International Nuclear Information System (INIS)

    Sydkraft is the largest privately owned utility in Sweden. It serves about 20% of the Swedish population with about 12 TWh of electric power per year, of which 64% is nuclear (1978 figures). The two identical 590 MWE ASEA-ATOM boiling water reactors in Barsebeck have been in operation since 1975 and 1977 respectively. Fission product activity in the primary circuits and in the off-gas systems is extremely low and indicate a near perfect fuel condition. Operating restrictions limiting the effect of pellet cladding interaction have been in use since initial start-up and testing. A few events involving rapid power increases above the preconditioned power level have occurred without causing fuel failures. It is believed that an analysis of power reactor operational transients, which did not cause fuel failures, can be useful to design more adequate and less conservative rules for the operation of nuclear reactor cores

  14. Boiling crisis and non-equilibrium drying transition

    CERN Document Server

    Nikolayev, Vadim

    2016-01-01

    Boiling crisis is the rapid formation of the quasi-continuous vapor film between the heater and the liquid when the heat supply exceeds a critical value. We propose a mechanism for the boiling crisis that is based on the spreading of the dry spot under a vapor bubble. The spreading is initiated by the vapor recoil force, a force coming from the liquid evaporation into the bubble. Since the evaporation intensity increases sharply near the triple contact line, the influence of the vapor recoil can be described as a change of the apparent contact angle. Therefore, for the most usual case of complete wetting of the heating surface by the liquid, the boiling crisis can be understood as a drying transition from complete to partial wetting. The state of nucleate boiling, which is boiling in its usual sense, is characterized by a very large rate of heat transfer from the heating surface to the bulk because the superheated liquid is carried away from the heating surface by the departing vapor bubbles. If the heating p...

  15. Heat transfer properties of organic coolants containing high boiling residues

    International Nuclear Information System (INIS)

    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 104 to 105. The data was correlated with the equation Nu = 0.015 Reb0.85 Prb0.4 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)

  16. Film boiling characteristics of potassium droplets on heated plate

    International Nuclear Information System (INIS)

    For providing background information on the possible vapor explosion in the event of a core disruptive accident of LMFBRs, an experiment was conducted on the film boiling characteristics of liquid metal potassium in association with the Leidenfrost phenomenon. In a steel container filled with Ar gas, K droplets were put on a joule-heated plate of 316-SS or Ta. The behaviors of droplet were observed by a camera and a color VTR through viewports. The experimental conditions were the Ar pressure 1 bar, the initial K temperature 350 -- 7600C, and the plate temperature 900 -- 1,2500C for 316-SS and 1,100 -- 1,6000C for Ta. Stable film boiling known as Leidenfrost phenomenon was observed for a high temperature condition of the plate, whereas an instantaneous break-up of droplet with extensive vaporization occurred for a low temperature. The heat transfer characteristics of film and transition boiling regions were obtaind by estimating the heat flux from the volumetric reducing rate of droplet due to vaporization. The results in the film boiling region showed an appreciably good agreement with the prediction based on Bromley's expression combined with the theory of Baumeister et al. The minimum film boiling temperature and heat flux were found to be about 1,3000C and 15 W/cm2, respectively, for a droplet size of 0.15 cm3. (author)

  17. 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.

  18. Enhanced boiling heat transfer in horizontal test bundles

    Energy Technology Data Exchange (ETDEWEB)

    Trewin, R.R.; Jensen, M.K.; Bergles, A.E.

    1994-08-01

    Two-phase flow boiling from bundles of horizontal tubes with smooth and enhanced surfaces has been investigated. Experiments were conducted in pure refrigerant R-113, pure R-11, and mixtures of R-11 and R-113 of approximately 25, 50, and 75% of R-113 by mass. Tests were conducted in two staggered tube bundles consisting of fifteen rows and five columns laid out in equilateral triangular arrays with pitch-to-diameter ratios of 1.17 and 1.5. The enhanced surfaces tested included a knurled surface (Wolverine`s Turbo-B) and a porous surface (Linde`s High Flux). Pool boiling tests were conducted for each surface so that reference values of the heat transfer coefficient could be obtained. Boiling heat transfer experiments in the tube bundles were conducted at pressures of 2 and 6 bar, heat flux values from 5 to 80 kW/m{sup 2}s, and qualities from 0% to 80%, Values of the heat transfer coefficients for the enhanced surfaces were significantly larger than for the smooth tubes and were comparable to the values obtained in pool boiling. It was found that the performance of the enhanced tubes could be predicted using the pool boiling results. The degradation in the smooth tube heat transfer coefficients obtained in fluid mixtures was found to depend on the difference between the molar concentration in the liquid and vapor.

  19. Study on saturated flow boiling heat transfer under vibration conditions

    International Nuclear Information System (INIS)

    The ability to predict void formation, void fraction and critical heat flux -CHF- in flow boiling under oscillatory flow and vibration conditions is important to the safety technology of nuclear reactor during earthquake. In the present study, the onset of nucleate boiling -ONB-, the point of net vapor generation -NVG- and CHF on saturated flow boiling under vibration conditions were investigated experimentally. Steady state experiments were conducted using a copper thin-film and subcooled water at 0.1 MPa. The liquid velocity was 0.27, 1.38, 3.20 and 4.07 m/s, respectively; the liquid subcooling was 0 K. A heater was made of a printed circuit board. A test section was a rectangular flow channel of 10 mm width and 10 mm height. The test heater was heated by Joule heating of d.c. current from a low-voltage high-current stabilizer. The heating rate of the heater was determined from supplied current and voltage. The temperature of the heater was obtained by referring to the measured electric resistance. The test section was arranged for horizontal position facing upward and for vertical position, respectively. For the vibration condition, the test section was set on a vibration table. The ONB was decided as an occurrence of the first boiling bubble. The critical heat flux was determined as that immediately before the heating surface physically burned-out. The CHF on saturated flow boiling under vibration conditions were investigated experimentally. (author)

  20. Wall function approach for boiling two-phase flows

    International Nuclear Information System (INIS)

    One of the important goals of the NURESIM project is to assess and improve the simulation capability of the three-dimensional two-fluid codes for prediction of local boiling flow processes. The boiling flow is strongly affected by local mechanisms in the turbulent boundary layer near the heated wall. Wall-to-fluid transfer models for boiling flow with the emphasis on near-wall treatment are being addressed in the paper. Since the computational grid of the 3D two-fluid models is too coarse to resolve the variable gradients in the near-wall region, the use of wall functions is a common approach to model the liquid velocity and temperature profile adjacent to the heated wall. The wall function model for momentum, based on the surface roughness analogy has been discussed and implemented in the NEPTUNECFD code. The model has been validated on several upward boiling flow experiments, differing in the geometry, working fluid and operating conditions. The simulations with the new wall function model show an improved prediction of flow parameters over the boiling boundary layer. Furthermore, a wall function model for the energy equation, based on enhanced two-phase wall friction has been derived and validated.

  1. Heater size effect on subcooled pool boiling of FC-72

    International Nuclear Information System (INIS)

    Extensive research has been conducted on pool boiling using heaters larger than the capillary length. For large heaters and/or high gravity conditions, boiling is dominated by buoyancy, and the heat transfer is heater size independent. Much less is known about boiling on small heaters and at low gravity levels. The ratio of heater size Lh to capillary length Lc is an important parameter in the determination of heater size dependence on heat transfer. As the ratio Lh/Lc decreases due to a decrease in either heater size or gravity, surface tension forces become dominant. It is proposed that transition from buoyancy to surface tension dominated boiling occurs when the heater size and bubble departure diameter are of the same order. Previous work in variable gravity with flat surfaces has shown that the heat transfer was heater size independent only when the ratio Lh/Lc was considerably larger than 1. An array of 96 platinum resistance heater elements in a 10 x 10 configuration with individual elements 0.7 x 0.7 mm2 in size was used to vary heater size and measure the heat transfer. The threshold value of Lh/Lc above which pool boiling is heater size independent was found to be about 2.8. (author)

  2. Improvements of fuel failure detection in boiling water reactors using helium measurements

    International Nuclear Information System (INIS)

    To certify a continuous and safe operation of a boiling water reactor, careful surveillance of fuel integrity is of high importance. The detection of fuel failures can be performed by off-line gamma spectroscopy of off-gas samples and/or by on-line nuclide specific monitoring of gamma emitting noble gases. To establish the location of a leaking fuel rod, power suppression testing can be used. The accuracy of power suppression testing is dependent on the information of the delay time and the spreading of the released fission gases through the systems before reaching the sampling point. This paper presents a method to improve the accuracy of power suppression testing by determining the delay time and gas spreading profile. To estimate the delay time and examine the spreading of the gas in case of a fuel failure, helium was injected in the feed water system at Forsmark 3 nuclear power plant. The measurements were performed by using a helium detector system based on a mass spectrometer installed in the off-gas system. The helium detection system and the results of the experiment are presented in this paper. (authors)

  3. 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.

  4. 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.)

  5. Visualization of pool boiling from complex surfaces with internal tunnels

    Directory of Open Access Journals (Sweden)

    Pastuszko Robert

    2012-04-01

    Full Text Available The paper presents experimental investigations of boiling heat transfer for a system of connected narrow horizontal and vertical tunnels. These extended surfaces, named narrow tunnel structure (NTS, can be applied to electronic element cooling. The experiments were carried out with ethanol at atmospheric pressure. The tunnel external covers were manufactured out of 0.1 mm thick perforated copper foil (hole diameters 0.5 mm, sintered with the mini-fins, formed on the vertical side of the 10 mm high rectangular fins and horizontal inter-fin surface. Visualization studies were conducted with a transparent structured model of joined narrow tunnels limited with the perforated foil. The visualization investigations aimed to formulate assumptions for the boiling model through distinguishing boiling types and defining all phases of bubble growth.

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

    International Nuclear Information System (INIS)

    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 (Al3O2), 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)

  7. The concept and application of miniaturisation boiling in cooling system

    International Nuclear Information System (INIS)

    The purpose of this research is to study and examine the phenomena of miniaturisation-boiling, which intensely scattered into a large number of minute liquid particles from a water droplet surface into the atmosphere, when the droplet collided with a heating surface. For the material of the heating surface, the following were used: stainless steel (SUS 303 A Cr= 17 %, Ni= 8 %), sapphire (Al2O3), brass, copper and carbon plane. The material was heated in order to study the miniaturisation-boiling and droplet bounding phenomena at a very high temperature (160 degree Celsius- 420 degree Celsius). The phenomenon was photographed by a high-speed camera (10000 fps) from the horizontal direction. The nuclear fusion reactor which needs a severe cooling, and heat removal cooling method through special boiling leads to this research. (author)

  8. On the dynamics of bubbles in boiling water

    International Nuclear Information System (INIS)

    Research highlights: → We devote this work to investigate the bubbles dynamics in boiling water. → A simple experiment of laser scattering was designed to obtain dynamical features. → Correlations and non-exponential distributions were found. → A simple model was able to describe several aspects of the system. - Abstract: We investigate the dynamics of many interacting bubbles in boiling water by using a laser scattering experiment. Specifically, we analyze the temporal variations of a laser intensity signal which passed through a sample of boiling water. Our empirical results indicate that the return interval distribution of the laser signal does not follow an exponential distribution; contrariwise, a heavy-tailed distribution has been found. Additionally, we compare the experimental results with those obtained from a minimalist phenomenological model, finding a good agreement.

  9. Boiling heat transfer in porous media composed of particles

    International Nuclear Information System (INIS)

    The boiling heat transfer in the porous media composed of spherical fuel elements exerts significant influences on the reactor's efficiency and safety. In the present study an experimental setup was designed and the boiling heat transfer in the porous media composed of spheres of regular distribution was investigated. Four spheres with diameters of 5mm, 6mm, 7mm and 8mm were used in the test sections. The greater particle diameter led to lower heat transfer coefficient, and resulted in higher wall superheat of original nucleation boiling. The variation of heat transfer coefficient was divided into three groups according to two-phase flow patterns and void fraction. A correlation of heat transfer coefficient was proposed with a mean relative deviation of ± 16%. (author)

  10. Numerical model of post-DNB transition boiling heat transfer

    International Nuclear Information System (INIS)

    In this paper a physical model for the transition boiling heat transfer is proposed. The corresponding mathematical descriptions are given in detail and the heat transfer characteristics of post-DNB transition boiling is analyzed. The numerical model of post-DNB transition boiling heat transfer is obtained as the empirical value of the coefficient is determined by the experimental data. The numerical model is compared with the experimental data of different parameters and other numerical models, and the statistical deviations are calculated. The calculating results of the numerical model in this paper show good agreement with the experimental data and the numerical model in this paper is with good applicability compared with other numerical models. (authors)

  11. Evaluation of N fuels boiling subchannel flow and DNB behavior

    International Nuclear Information System (INIS)

    An updated evaluation of the boiling and burnout performance of the N Reactor fuels is presented. Several new procedures and improvements for fuel analyses are described. These include the following items incorporated into the BOTHER computer program: subroutine FLOBAL which balances the flows among the boiling and nonboiling channels to equilibrate pressure drop; a new enthalpy imbalance factor for the MKIV fuel of 19.4% (95/95 basis); an updated method for calculating the two-phase pressure drop multiplier for each channel type, which includes both momentum and friction terms; and changes in the BOTHER code. The net effect of these changes is to predict DNBR limits before connecter boiling and at slightly higher flows and lower tube powers than predicted previously

  12. 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.

  13. Modeling of subcooled boiling in the vertical flow

    International Nuclear Information System (INIS)

    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)

  14. Pin cooling and dryout in steady local boiling

    International Nuclear Information System (INIS)

    A study is presented of pin cooling and dryout mechanisms in steady local boiling, with the particular objective of understanding the substantial dryout margins observed in the KNS local boiling experiments. Mechanisms for the entry of liquid into the voided region are discussed, and pin cooling by draining liquid films deduced to be likely. The conditions required for interruption of the film flow, and hence for dryout, are examined, with particular attention to vapour/liquid interactions causing film breakdown, inhibition of rewetting and film flooding. This leads to the hypothesis that dryout occurs when a critical vapour velocity is reached, which is shown to be consistent with the limited data on dryout conditions in steady boiling. (orig.)

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

    International Nuclear Information System (INIS)

    This paper reports in the experimental results concerning unsteady burnout phenomenon, based on unsteady boiling heat transfer data, burnout heat flux data and the data of changing pressure and water temperature in course of time. These data were acquired by unsteady heating of gas-liquid two phase flow. This experiment simulated the thermohydrodynamic conditions under the runaway power of a nuclear reactor. The following results have been clarified. The boiling with high heat flux showed the same heat transfer characteristics as the steady nuclear boiling curves under each flow condition. Under the conditions of low flow speed and high sub-cool degree, the unsteady burnout heat flux showed the extreme increase of the maximum heat flux owing to the shortening of the time constant. The generation of unsteady burnout phenomena is dominated by two phase flow conditions and by bubble behavior near the heat transfer surface owing to the change of heating conditions and flow conditions. (Tai, I.)

  16. Heat transfer correlation for saturated flow boiling of water

    International Nuclear Information System (INIS)

    The saturated flow boiling heat transfer of water (H2O, R718) is encountered in many applications such as compact heat exchangers and electronic cooling, for which an accurate correlation of evaporative heat transfer coefficients is necessary. A number of correlations for two-phase flow boiling heat transfer coefficients were proposed. However, their prediction accuracies for H2O are not satisfactory. This work compiles an H2O database of 1055 experimental data points from micro/mini-channels from nine independent studies, evaluates 41 existing correlations to provide a clue for developing a better correlation of saturated flow boiling heat transfer coefficients for H2O, and then proposes a new one. The new correlation incorporates a newly proposed dimensionless number and makes great progress in prediction accuracy. It has a mean absolute deviation of 10.1%, predicting 81.9% of the entire database within ±15% and 91.2% within ±20%, far better than the best existing one. Besides, it also works well for several other working fluids, such as R22, R134a, R410A and NH3 (ammonia, R717), being the best for R22, R410A and NH3 so far. - Highlights: • Compiles a database of 1055 data points of H2O flow boiling heat transfer. • Evaluates 41 correlations of flow boiling heat transfer coefficient. • Generalize approach for developing experiment-based correlation. • Propose a correlation of H2O flow boiling heat transfer in small channels. • The new correlation has a mean absolute deviation of 10.1% for the database

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  19. Immersion cooling nucleate boiling of high power computer chips

    International Nuclear Information System (INIS)

    Highlights: ► Experimental investigations of nucleate boiling of dielectric liquids on porous graphite (PG). ► Marked enhancements in nucleate boiling heat transfer coefficient and CHF. ► Critical heat flux (CHF) increases linearly with increased liquid subcooling. ► PG–Cu spreaders for cooling 10 × 10 computer chips remove up to 100 W. - Abstract: This paper presents experimental results of saturation and subcooled boiling of FC-72 and HFE-7100 dielectric liquids on uniformly heated, 10 × 10 mm porous graphite (PG) surfaces for potential applications to immersion cooling of high power computer chips. The experiments investigated the effects of surface inclination, from upward-facing (0°) to downward-facing (180°), and liquid subcooling from 0 to 30 K on nucleate boiling heat transfer coefficient and critical heat flux. The presented experimental data and correlations for natural convection of dielectric liquids on PG and plane surfaces are important for cooling chips while in the standby mode when surface heat flux 2. The experimental curves of the nucleate boiling heat transfer coefficient for FC-72 dielectric liquid in the upward-facing orientation are used in 3-D thermal analysis for sizing and quantifying the performance of copper (Cu), PG and PG–Cu composite spreaders for removing the dissipated thermal power by an underlying 10 × 10 mm computer chip with non-uniform heat dissipation. The 2 mm-thick spreaders are cooled by either saturation or 30 K subcooled nucleate boiling of FC-72 and the composite spreader consists of 0.4 mm-thick surface layer of PG and 1.6 mm-thick Cu substrate.

  20. A stability identification system for boiling water nuclear reactors

    International Nuclear Information System (INIS)

    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

  1. 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.

  2. On Boiling of Crude Oil under Elevated Pressure

    CERN Document Server

    Pimenova, Anastasiya V

    2015-01-01

    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.

  3. 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.

  4. Dimensional analysis of boiling heat transfer burnout conditions

    International Nuclear Information System (INIS)

    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. Prediction of void fraction in subcooled flow boiling

    International Nuclear Information System (INIS)

    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)

  6. 76 FR 14437 - Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of...

    Science.gov (United States)

    2011-03-16

    ... From the Federal Register Online via the Government Publishing Office ] NUCLEAR REGULATORY COMMISSION Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of... GE Hitachi Nuclear Energy (GEH) for the economic simplified boiling water reactor (ESBWR)...

  7. 76 FR 61118 - Meeting of the ACRS Subcommittee on Advanced Boiling Water Reactor; Notice of Meeting

    Science.gov (United States)

    2011-10-03

    ... Boiling Water Reactor; Notice of Meeting The ACRS Subcommittee on Advanced Boiling Water Reactor (ABWR... published in the Federal Register on October 21, 2010, (75 FR 65038-65039). Detailed meeting agendas...

  8. Minimum fluence for laser blow-off of thin gold films at 248 and 532 nm

    International Nuclear Information System (INIS)

    The minimum 248 nm, 25 ns, and 532 nm, 15 ns laser fluences required to blow off thin gold films from optical quartz have been measured as a function of film thickness. The films apparently blow off when the gold-quartz interface reaches the normal boiling point of gold. Even though the initial reflectivities at the two wavelengths are very different, the actual laser fluences required to blow off the films are very similar. While the reflectivities above the melting point appear to be very low, as expected, large decreases in the reflectivity at 532 nm may also occur prior to film melting

  9. Minimum fluence for laser blow-off of thin gold films at 248 and 532 nm

    Energy Technology Data Exchange (ETDEWEB)

    Baseman, R.J.; Froberg, N.M.; Andreshak, J.C.; Schlesinger, Z. (IBM Thomas J. Watson Research Center, Yorktown Heights, New York 10598 (USA))

    1990-04-09

    The minimum 248 nm, 25 ns, and 532 nm, 15 ns laser fluences required to blow off thin gold films from optical quartz have been measured as a function of film thickness. The films apparently blow off when the gold-quartz interface reaches the normal boiling point of gold. Even though the initial reflectivities at the two wavelengths are very different, the actual laser fluences required to blow off the films are very similar. While the reflectivities above the melting point appear to be very low, as expected, large decreases in the reflectivity at 532 nm may also occur prior to film melting.

  10. Determination of Boiling Range of Xylene Mixed in PX Device Using Artificial Neural Networks

    OpenAIRE

    Zhu, Ting; Zhu, Yuxuan; Yang, Hong; Li, Hao

    2014-01-01

    Determination of boiling range of xylene mixed in PX device is currently a crucial topic in the practical applications because of the recent disputes of PX project in China. In our study, instead of determining the boiling range of xylene mixed by traditional approach in laboratory or industry, we successfully established two Artificial Neural Networks (ANNs) models to determine the initial boiling point and final boiling point respectively. Results show that the Multilayer Feedforward Neural...

  11. Experimental Evidence of the Vapor Recoil Mechanism in the Boiling Crisis

    OpenAIRE

    Nikolayev, Vadim; Chatain, D.; Garrabos, Y.; Beysens, D.

    2006-01-01

    International audience Boiling crisis experiments are carried out in the vicinity of the liquid-gas critical point of H2. A magnetic gravity compensation setup is used to enable nucleate boiling at near critical pressure. The measurements of the critical heat flux that defines the threshold for the boiling crisis are carried out as a function of the distance from the critical point. The obtained power law behavior and the boiling crisis dynamics agree with the predictions of the vapor reco...

  12. Evaluation of Correlations of Flow Boiling Heat Transfer of R22 in Horizontal Channels

    OpenAIRE

    Zhanru Zhou; Xiande Fang; Dingkun Li

    2013-01-01

    The calculation of two-phase flow boiling heat transfer of R22 in channels is required in a variety of applications, such as chemical process cooling systems, refrigeration, and air conditioning. A number of correlations for flow boiling heat transfer in channels have been proposed. This work evaluates the existing correlations for flow boiling heat transfer coefficient with 1669 experimental data points of flow boiling heat transfer of R22 collected from 18 published papers. The top two corr...

  13. Experimental Investigation on Pool Boiling Heat Transfer With Ammonium Dodecyl Sulfate

    OpenAIRE

    Mr.P. Atcha Rao; Mr.V.V.Ramakrishna

    2015-01-01

    We have so many applications related to Pool Boiling. The Pool Boiling is mostly useful in arid areas to produce drinking water from impure water like sea water by distillation process. It is very difficult to distill the only water which having high surface tension. The surface tension is important factor to affect heat transfer enhancement in pool boiling. By reducing the surface tension we can increase the heat transfer rate in pool boiling. From so many years we are using surf...

  14. Research on radiation detectors, boiling transients, and organic lubricants

    Science.gov (United States)

    1974-01-01

    The accomplishments of a space projects research facility are presented. The subjects discussed are: (1) a study of radiation resistant semiconductor devices, (2) synthesis of high temperature organic lubricants, (3) departure from phase equilibrium during boiling transients, (4) effects of neutron irradiation on defect state in tungsten, and (5) determination of photon response function of NE-213 liquid scintillation detectors.

  15. Experiments on microgravity boiling heat transfer by using transparent heaters

    Energy Technology Data Exchange (ETDEWEB)

    Ohta, H. [Kyushu Univ., Fukuoka (Japan). Dept. of Energy and Mech. Eng.

    1997-11-01

    To clarify the relation between the liquid-vapor behavior and the heat transfer characteristics in the boiling phenomena, the structures of transparent heaters were developed for both flow boiling experiments and were applied to the microgravity environment realized by the parabolic flight of aircraft. In the flow boiling experiment, a transparent heated tube makes the heating, the observation of liquid-vapor behavior and the measurement of heat transfer data simultaneously possible. The heat transfer coefficient in the annular flow regime at moderate quality has distinct dependence on gravity provided that the mass velocity is not so high, while no noticeable gravity effect is seen at high quality and in the bubbly flow regime. The measured gravity effect was directly related to the behavior of annular liquid film observed through the transparent tube wall. In the pool boiling experiment, a structure of transparent heating surface realizes both the observation of the macrolayer or microlayer behavior from underneath and the measurements of local surface temperatures and the layer thickness. It was clarified in the microgravity experiments that no vapor stem exists but tiny bubbles are observed in the macrolayer underneath a large coalesced bubble at high heat flux. The heat flux evaluated by the heat conduction across the layer assumes less than 30% of the total to be transferred. The evaporation of the microlayers underneath primary bubbles just after the generation dominates the heat transfer in the microgravity, not only in the isolated bubble region but also in the coalesced bubble region. (orig.) 14 refs.

  16. Film boiling on downward quenching hemisphere of varying sizes

    International Nuclear Information System (INIS)

    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. (authors)

  17. Investigations on coolant boiling in research reactors. 2

    International Nuclear Information System (INIS)

    Subcooled boiling has been investigated systematically at the Rossendorf Research Reactor in the range between boiling onset and boiling crisis. This is of particular interest because in the core the direction of the coolant flow is opposite to the bubble buoyance of the bubbles - in contrast to power reactors. For this reason an experimental fuel assembly equipped with a throttle valve for coolant flow reduction and different detectors was built up and installed in the reactor core. Measurements of thermohydraulic parameters and noise signals from temperature, neutron flux and acoustic sources were subject of the investigations. Besides other results fluctuations of the void fraction induced by a standing wave of the two-phase flow in the coolant channel and the 24-Hz pressure fluctuations of the circulation pumps have been observed. It has been shown that the frequency of the standing wave is determined by the size of the boiling volume in the coolant channel and that this frequency therefore depends on the outlet temperature of the coolant. (author)

  18. STEAM TURBINES WITH A LOW-BOILING WORKING AGENT

    OpenAIRE

    Morozov, N.; Karasev, V.

    2010-01-01

    The subject of the article is the assembly of a steam-generator plant with a natural working agent. A method of calculation for steam turbines with a low-boiling working agent is offered, which accounts for the correlation between the adiabatic curve indication, pressure and temperature in the overheated vapor area.

  19. 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.

  20. 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.

  1. Electrochemical study of aluminum corrosion in boiling high purity water

    Science.gov (United States)

    Draley, J. E.; Legault, R. A.

    1969-01-01

    Electrochemical study of aluminum corrosion in boiling high-purity water includes an equation relating current and electrochemical potential derived on the basis of a physical model of the corrosion process. The work involved an examination of the cathodic polarization behavior of 1100 aluminum during aqueous oxidation.

  2. Experimental demonstration of contaminant removal from fractured rock by boiling.

    Science.gov (United States)

    Chen, Fei; Liu, Xiaoling; Falta, Ronald W; Murdoch, Lawrence C

    2010-08-15

    This study was conducted to experimentally demonstrate removal of a chlorinated volatile organic compound from fractured rock by boiling. A Berea sandstone core was contaminated by injecting water containing dissolved 1,2-DCA (253 mg/L) and sodium bromide (144 mg/L). During heating, the core was sealed except for one end, which was open to the atmosphere to simulate an open fracture. A temperature gradient toward the outlet was observed when boiling occurred in the core. This indicates that steam was generated and a pressure gradient developed toward the outlet, pushing steam vapor and liquid water toward the outlet. As boiling occurred, the concentration of 1,2-DCA in the condensed effluent peaked up to 6.1 times higher than the injected concentration. When 38% of the pore volume of condensate was produced, essentially 100% of the 1,2-DCA was recovered. Nonvolatile bromide concentration in the condensate was used as an indicator of the produced steam quality (vapor mass fraction) because it can only be removed as a solute, and not as a vapor. A higher produced steam quality corresponds to more concentrated 1,2-DCA removal from the core, demonstrating that the chlorinated volatile compound is primarily removed by partitioning into vapor phase flow. This study has experimentally demonstrated that boiling is an effective mechanism for CVOC removal from the rock matrix. PMID:20666474

  3. BORATING OF CARBON AND ALLOY STEEL IN BOILING LAYER

    Directory of Open Access Journals (Sweden)

    N. Koukhareva

    2012-01-01

    Full Text Available The paper describes how to obtain boride coatings on steel 20, 4X5MФС, X12M being treated in a boiling layer of metallothermic powder environment. Phase and chemical compositions, hardness and wear- resistance of boride coatings

  4. Corrosion fatigue behavior of zirconium in boiling nitric acid

    International Nuclear Information System (INIS)

    The corrosion fatigue behavior of zirconium in boiling nitric acid has been studied to evaluate the reliability of zirconium used in nuclear fuel reprocessing equipment. An apparatus designed for corrosion fatigue tests in boiling nitric acid was used. The crack growth rate of zirconium was measured as a function of the stress intensity factor using TDCB type specimens. After the tests, the fracture morphology was examined with a scanning electron microscope. The crack growth rate was influenced with the texture of specimens and the test environments. In air at room temperature, the crack growth rate at the longitudinal direction of specimens was faster than that of the transverse direction. Moreover, the crack growth rate in boiling nitric acid was more faster than that in air at room temperature. According to the fractographic examination, X-ray analysis, and so on, the observed results were interpreted with based on the crystal anisotropy on mechanical properties and the susceptibility to stress corrosion cracking in boiling nitric acid of zirconium. (author)

  5. Calculation of boiling model change wave propagation rate

    International Nuclear Information System (INIS)

    Approximate analytical expression for the boiling mode change wave front rate on the rod and on the plate was obtained. The influence of the Thomson effect and of heater orientation in the gravitational field was taken into account. Paper shows satisfactory agreement of the obtained ratios with the experimental data

  6. Investigation Status of Heat Exchange while Boiling Hydrocarbon Fuel

    OpenAIRE

    D. S. Obukhov

    2014-01-01

    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.

  7. Pressure drop of subcooled flow boiling in narrow tube

    International Nuclear Information System (INIS)

    The pressure drop of subcooled flow boiling in a narrow tube was investigated experimentally using water as a coolant. Experiments were conducted at nearly ambient pressure under the conditions: tube inside diameter: 1 and 3mm, tube length: 10∼100mm, and water mass velocity: 7000∼20000kg/m2s. The friction pressure drop ratio of subcooled flow boiling to non-heating water flow was examined by increasing the heat flux. The ratio begins to increase at the heat flux proposed by the Saha-Zuber correlation that the bubble begins to detach for 3mm inside diameter tube, though the heat flux is higher than the Saha-Zuber heat flux for 1mm tube. The ratio was further compared with the Bergles-Dormer correlation. The two phase friction multiplier of subcooled flow boiling was examined assuming the Ahmad void fraction and applied to the Lockhart-Martinelli (L-M) correlation. The abnormarity of the subcooled flow boiling in the case of 1mm inside diameter tube was confirmed in these discussions. (author)

  8. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs)....

  9. 76 FR 3540 - U.S. Advanced Boiling Water Reactor Aircraft Impact Design Certification Amendment

    Science.gov (United States)

    2011-01-20

    ... COMMISSION 10 CFR Part 52 RIN 3150-AI84 U.S. Advanced Boiling Water Reactor Aircraft Impact Design... the U.S. Advanced Boiling Water Reactor (ABWR) standard plant design to comply with the NRC's aircraft...--Design Certification Rule for the U.S. Advanced Boiling Water Reactor IV. Section-by-Section Analysis...

  10. Modeling acid-gas generation from boiling chloride brines

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Guoxiang; Spycher, Nicolas; Sonnenthal, Eric; Steefel, Carl

    2009-11-16

    This study investigates the generation of HCl and other acid gases from boiling calcium chloride dominated waters at atmospheric pressure, primarily using numerical modeling. The main focus of this investigation relates to the long-term geologic disposal of nuclear waste at Yucca Mountain, Nevada, where pore waters around waste-emplacement tunnels are expected to undergo boiling and evaporative concentration as a result of the heat released by spent nuclear fuel. Processes that are modeled include boiling of highly concentrated solutions, gas transport, and gas condensation accompanied by the dissociation of acid gases, causing low-pH condensate. Simple calculations are first carried out to evaluate condensate pH as a function of HCl gas fugacity and condensed water fraction for a vapor equilibrated with saturated calcium chloride brine at 50-150 C and 1 bar. The distillation of a calcium-chloride-dominated brine is then simulated with a reactive transport model using a brine composition representative of partially evaporated calcium-rich pore waters at Yucca Mountain. Results show a significant increase in boiling temperature from evaporative concentration, as well as low pH in condensates, particularly for dynamic systems where partial condensation takes place, which result in enrichment of HCl in condensates. These results are in qualitative agreement with experimental data from other studies. The combination of reactive transport with multicomponent brine chemistry to study evaporation, boiling, and the potential for acid gas generation at the proposed Yucca Mountain repository is seen as an improvement relative to previously applied simpler batch evaporation models. This approach allows the evaluation of thermal, hydrological, and chemical (THC) processes in a coupled manner, and modeling of settings much more relevant to actual field conditions than the distillation experiment considered. The actual and modeled distillation experiments do not represent

  11. Modeling acid-gas generation from boiling chloride brines

    International Nuclear Information System (INIS)

    This study investigates the generation of HCl and other acid gases from boiling calcium chloride dominated waters at atmospheric pressure, primarily using numerical modeling. The main focus of this investigation relates to the long-term geologic disposal of nuclear waste at Yucca Mountain, Nevada, where pore waters around waste-emplacement tunnels are expected to undergo boiling and evaporative concentration as a result of the heat released by spent nuclear fuel. Processes that are modeled include boiling of highly concentrated solutions, gas transport, and gas condensation accompanied by the dissociation of acid gases, causing low-pH condensate. Simple calculations are first carried out to evaluate condensate pH as a function of HCl gas fugacity and condensed water fraction for a vapor equilibrated with saturated calcium chloride brine at 50-150 C and 1 bar. The distillation of a calcium-chloride-dominated brine is then simulated with a reactive transport model using a brine composition representative of partially evaporated calcium-rich pore waters at Yucca Mountain. Results show a significant increase in boiling temperature from evaporative concentration, as well as low pH in condensates, particularly for dynamic systems where partial condensation takes place, which result in enrichment of HCl in condensates. These results are in qualitative agreement with experimental data from other studies. The combination of reactive transport with multicomponent brine chemistry to study evaporation, boiling, and the potential for acid gas generation at the proposed Yucca Mountain repository is seen as an improvement relative to previously applied simpler batch evaporation models. This approach allows the evaluation of thermal, hydrological, and chemical (THC) processes in a coupled manner, and modeling of settings much more relevant to actual field conditions than the distillation experiment considered. The actual and modeled distillation experiments do not represent

  12. Modeling acid-gas generation from boiling chloride brines

    Directory of Open Access Journals (Sweden)

    Sonnenthal Eric

    2009-11-01

    Full Text Available Abstract Background This study investigates the generation of HCl and other acid gases from boiling calcium chloride dominated waters at atmospheric pressure, primarily using numerical modeling. The main focus of this investigation relates to the long-term geologic disposal of nuclear waste at Yucca Mountain, Nevada, where pore waters around waste-emplacement tunnels are expected to undergo boiling and evaporative concentration as a result of the heat released by spent nuclear fuel. Processes that are modeled include boiling of highly concentrated solutions, gas transport, and gas condensation accompanied by the dissociation of acid gases, causing low-pH condensate. Results Simple calculations are first carried out to evaluate condensate pH as a function of HCl gas fugacity and condensed water fraction for a vapor equilibrated with saturated calcium chloride brine at 50-150°C and 1 bar. The distillation of a calcium-chloride-dominated brine is then simulated with a reactive transport model using a brine composition representative of partially evaporated calcium-rich pore waters at Yucca Mountain. Results show a significant increase in boiling temperature from evaporative concentration, as well as low pH in condensates, particularly for dynamic systems where partial condensation takes place, which result in enrichment of HCl in condensates. These results are in qualitative agreement with experimental data from other studies. Conclusion The combination of reactive transport with multicomponent brine chemistry to study evaporation, boiling, and the potential for acid gas generation at the proposed Yucca Mountain repository is seen as an improvement relative to previously applied simpler batch evaporation models. This approach allows the evaluation of thermal, hydrological, and chemical (THC processes in a coupled manner, and modeling of settings much more relevant to actual field conditions than the distillation experiment considered. The actual

  13. Phase relations and adiabats in boiling seafloor geothermal systems

    Science.gov (United States)

    Bischoff, James L.; Pitzer, Kenneth S.

    1985-11-01

    Observations of large salinity variations and vent temperatures in the range of 380-400°C suggest that boiling or two-phase separation may be occurring in some seafloor geothermal systems. Consideration of flow rates and the relatively small differences in density between vapors and liquids at the supercritical pressures at depth in these systems suggests that boiling is occurring under closed-system conditions. Salinity and temperature of boiling vents can be used to estimate the pressure-temperature point in the subsurface at which liquid seawater first reached the two-phase boundary. Data are reviewed to construct phase diagrams of coexisting brines and vapors in the two-phase region at pressures corresponding to those of the seafloor geothermal systems. A method is developed for calculating the enthalpy and entropy of the coexisting mixtures, and results are used to construct adiabats from the seafloor to the P-T two-phase boundary. Results for seafloor vents discharging at 2300 m below sea level indicate that a 385°C vent is composed of a brine (7% NaCl equivalent) in equilibrium with a vapor (0.1% NaCl). Brine constitutes 45% by weight of the mixture, and the fluid first boiled at approximately 1 km below the seafloor at 415°C, 330 bar. A 400°C vent is primarily vapor (88 wt.%, 0.044% NaCl) with a small amount of brine (26% NaCl) and first boiled at 2.9 km below the seafloor at 500°C, 520 bar. These results show that adiabatic decompression in the two-phase region results in dramatic cooling of the fluid mixture when there is a large fraction of vapor.

  14. 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)

  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. Experimental investigation into the effects of coolant additives on boiling phenomena in pressurized water reactors

    International Nuclear Information System (INIS)

    This study investigates the effects of coolant additives like boric acid on boiling phenomena in pressurized water reactors under conditions as realistic as possible. The effects covered range from subcooled boiling to critical boiling conditions (CHF). The focus of this project lies on flow boiling with up to 40 bar and 250 °C in order to generate a data basis for a possible extrapolation to reactor conditions. The results of the experiments are used to implement and validate new models into CFD-Codes in context to a nationwide German joint research project with the specific aim of improving CFD boiling-models. (author)

  17. Development of a mechanistic model for forced convection subcooled boiling

    Science.gov (United States)

    Shaver, Dillon R.

    The focus of this work is on the formulation, implementation, and testing of a mechanistic model of subcooled boiling. Subcooled boiling is the process of vapor generation on a heated wall when the bulk liquid temperature is still below saturation. This is part of a larger effort by the US DoE's CASL project to apply advanced computational tools to the simulation of light water reactors. To support this effort, the formulation of the dispersed field model is described and a complete model of interfacial forces is formulated. The model has been implemented in the NPHASE-CMFD computer code with a K-epsilon model of turbulence. The interfacial force models are built on extensive work by other authors, and include novel formulations of the turbulent dispersion and lift forces. The complete model of interfacial forces is compared to experiments for adiabatic bubbly flows, including both steady-state and unsteady conditions. The same model is then applied to a transient gas/liquid flow in a complex geometry of fuel channels in a sodium fast reactor. Building on the foundation of the interfacial force model, a mechanistic model of forced-convection subcooled boiling is proposed. This model uses the heat flux partitioning concept and accounts for condensation of bubbles attached to the wall. This allows the model to capture the enhanced heat transfer associated with boiling before the point of net generation of vapor, a phenomenon consistent with existing experimental observations. The model is compared to four different experiments encompassing flows of light water, heavy water, and R12 at different pressures, in cylindrical channels, an internally heated annulus, and a rectangular channel. The experimental data includes axial and radial profiles of both liquid temperature and vapor volume fraction, and the agreement can be considered quite good. The complete model is then applied to simulations of subcooled boiling in nuclear reactor subchannels consistent with the

  18. A New Theory of Nucleate Pool Boiling in Arbitrary Gravity

    Science.gov (United States)

    Buyevich, Y. A.; Webbon, Bruce W.

    1995-01-01

    Heat transfer rates specific to nucleate pool boiling under various conditions are determined by the dynamics of vapour bubbles that are originated and grow at nucleation sites of a superheated surface. A new dynamic theory of these bubbles has been recently developed on the basis of the thermodynamics of irreversible processes. In contrast to other existing models based on empirically postulated equations for bubble growth and motion, this theory does not contain unwarrantable assumptions, and both the equations are rigorously derived within the framework of a unified approach. The conclusions of the theory are drastically different from those of the conventional models. The bubbles are shown to detach themselves under combined action of buoyancy and a surface tension force that is proven to add to buoyancy in bubble detachment, but not the other way round as is commonly presumed. The theory ensures a sound understanding of a number of so far unexplained phenomena, such as effect caused by gravity level and surface tension on the bubble growth rate and dependence of the bubble characteristics at detachment on the liquid thermophysical parameters and relevant temperature differences. The theoretical predictions are shown to be in a satisfactory qualitative and quantitative agreement with observations. When being applied to heat transfer at nucleate pool boiling, this bubble dynamic theory offers an opportunity to considerably improve the main formulae that are generally used to correlate experimental findings and to design boiling heat removal in various industrial applications. Moreover, the theory makes possible to pose and study a great deal of new problems of essential impact in practice. Two such problems are considered in detail. One problem concerns the development of a principally novel physical model for the first crisis of boiling. This model allows for evaluating critical boiling heat fluxes under various conditions, and in particular at different

  19. An experimental study of flow boiling in a rectangular channel with offset strip fins

    International Nuclear Information System (INIS)

    An experimental study on saturated flow boiling heat transfer of R113 was performed in a vertical rectangular channel with offset strip fins. Two-phase pressure gradients and boiling heat transfer coefficients in an electrically heated test section were measured for the quality range of 0-0.6, mass flux range of 17-43 kg/m2 s and heat flux of 500-3000 W/m2. Two-phase frictional multiplier was determined as a function of Martinelli parameter. The two-phase forced convective component of the local boiling heat transfer coefficient was found to be well correlated with the Reynolds number factor. A superposition method for the flow boiling heat transfer coefficient that included the contribution of saturated nucleate boiling was verified also for flow boiling in a channel with offset strip fins. The predictions of local flow boiling heat transfer coefficients were found to be in good agreement with experimental data

  20. Development and Validation of Pressure Tube Deformation and Subcooled Boiling Models of the MARS Code for Safety Analyses of the Wolsong NPP's

    International Nuclear Information System (INIS)

    The MARS code is being considered by KINS(Korea Institute of Nuclear Safety) as a thermal-hydraulic regulatory auditing tool for nuclear power plants in South Korea. Because Korea currently has four operating units of the CANDU(Canadian Deuterium Uranium)-type reactor in Wolsong, analytic models such as the Wolsong pump model, the off-take model for arbitrary-angled branch pipes, the radiation heat transfer input model, and the subcooled boiling model have been implemented into the MARS code to extend its applicability into CANDU reactors as well as PWR's. This part of the research series presents verification and validation of the pressure tube deformation model and the Podowski subcooled boiling model

  1. Control of the boiling crisis: analysis of a model system

    International Nuclear Information System (INIS)

    Controlling the transition between the low (nucleate) and high temperature (film) regimes of boiling is a serious challenge for a number of technological applications. Based on the theoretical analysis of a simplified reaction-diffusion model, it has recently been shown that the transition towards the dangerous situation where the high temperature phase tends to invade the whole system requires a higher power in a periodically spatially modulated system than in an homogeneous system. We show here that the transition mechanisms between the various boiling regimes depend on the ratio between the periodicity length along the wire and the characteristic thermal diffusion length. We analyse theoretically a simple experimental setup aimed at testing these ideas. The heater consists of a thin wire, with an applied electric current, with alternatively low resistance and high resistance sections. We determine the gain in stability for a set of realistic values of the parameters. (authors)

  2. Boiling flow simulation in Neptune-CFD and Fluent codes

    International Nuclear Information System (INIS)

    This paper presents simulations of the convective boiling flow performed with NEPTUNE-CFD and FLUENT codes. The DEBORA experiments carried out at CEA Grenoble were used as an experimental data set. In these experiments, freon R12 flows upwards inside a vertical pipe. Radial profiles of the flow variables are measured at the end of the heated section. Seven DEBORA cases were selected for simulation. NEPTUNE-CFD code was used without modifications because it contains all necessary models. In FLUENT, an important part of the models has been implemented by programming in User Defined Functions. The comparison of the radial profiles of void fraction, liquid temperature, gas velocity and mean bubble diameter at the end of the heated section shows that both codes can provide reasonable results in boiling conditions. The presented work was carried out within the 6. Framework EC NURESIM project. NEPTUNE-CFD code is implemented in the NURESIM platform. (authors)

  3. Computations of film boiling. Part I: numerical method

    Energy Technology Data Exchange (ETDEWEB)

    Esmaeeli, A.; Tryggvason, G. [Worcester Polytechnic Institute, MA (United States). Mechanical Engineering Department

    2004-12-01

    A numerical method for direct simulations of boiling flows is presented. The method is similar to the front tracking/finite difference technique of Juric and Tryggvason [Int. J. Multiphase Flow 24 (1998) 387], where one set of conservation equations is used to represent the mass transfer, heat transfer, and fluid flow in the liquid and the vapor, but improves on their numerical technique by elimination of their iterative algorithm. The justification of the mathematical formulation is presented and the numerical method and the code is validated by comparison of the results with the exact solutions of a few analytical problems. A grid refinement test for film boiling on a horizontal surface shows the convergence of results. (author)

  4. Acoustic measurement of boiling instabilities in a solar receiver

    Energy Technology Data Exchange (ETDEWEB)

    Beattie, A. G.

    1980-11-01

    An acoustic technique was developed and used to search for boiling instabilities in the prototype receiver for the Barstow 10 MW Solar Thermal Pilot Plant. Instabilities, consisting of movements of the transition zone between regions of nucleate and film boiling, were observed. The periods of these fluctuations ranged between three and fifteen seconds with no indications of preferred frequencies. The peak to peak amplitudes of the fluctuations averaged 0.4 meters under steady state conditions at absorbed power levels between 2.0 and 3.2 MW. Transient fluctuations with amplitudes up to 2.0 meters were also seen. These transients usually lasted between 30 and 300 seconds. It was not possible to pinpoint the causes of these transients.

  5. Flow pattern characteristics of pool boiling in inclined confined spaces

    International Nuclear Information System (INIS)

    In this paper, visualization of bubble behavior and two-phase flow in inclined confined spaces are performed for near-saturated demineralized water at atmospheric pressure with gap sizes of 3 mm to 8 mm, and inclination angles of 0° to 300°. Based on the results, three boiling regimes are observed: isolated deformed bubbles, coalesced deformed bubbles and partial dry-out. A flow pattern map for confined pool boiling, based on the Bond number and a dimensionless number of heat flux, has been developed in order to determine the regimes. Objective criteria are proposed for the transitions between various regimes. It is shown that the transition criteria between isolated deformed bubbles and coalesced deformed bubbles is Q=0.16 and Q=0.55 between coalesced deformed bubbles and partial dry-out. It has to be pointed out that the gap structures and downward facing heating surfaces may be responsible for the special flow patterns. (authors)

  6. Critical heat flux maxima during boiling crisis on textured surfaces

    Science.gov (United States)

    Dhillon, Navdeep Singh; Buongiorno, Jacopo; Varanasi, Kripa K.

    2015-09-01

    Enhancing the critical heat flux (CHF) of industrial boilers by surface texturing can lead to substantial energy savings and global reduction in greenhouse gas emissions, but fundamentally this phenomenon is not well understood. Prior studies on boiling crisis indicate that CHF monotonically increases with increasing texture density. Here we report on the existence of maxima in CHF enhancement at intermediate texture density using measurements on parametrically designed plain and nano-textured micropillar surfaces. Using high-speed optical and infrared imaging, we study the dynamics of dry spot heating and rewetting phenomena and reveal that the dry spot heating timescale is of the same order as that of the gravity and liquid imbibition-induced dry spot rewetting timescale. Based on these insights, we develop a coupled thermal-hydraulic model that relates CHF enhancement to rewetting of a hot dry spot on the boiling surface, thereby revealing the mechanism governing the hitherto unknown CHF enhancement maxima.

  7. Numerical modeling of boiling heat transfer in porous media

    International Nuclear Information System (INIS)

    Theoretical models were developed and validated to investigate boiling heat transfer in porous layers with and without the presence of chimneys. The critical heat flux and distributions of temperature, liquid saturation, liquid and vapor pressures, and liquid and vapor velocities were predicted numerically under typical PWR conditions. The results indicate that a porous layer produces a higher heat transfer coefficient in the nucleate boiling regime, as is well-known, and could potentially yield a much higher critical heat flux than a plain surface does. Moreover, a chimney-type porous layer can have a better thermal performance, i.e., heat transfer coefficient and critical heat flux than a homogeneous one, primarily due to the presence of chimneys providing pathways for vapor to escape from the porous layer with less resistance

  8. Consequences of select boiling waste scenarios for waste feed delivery

    International Nuclear Information System (INIS)

    The purpose of this calculation is to quantify thermal phenomena and predict consequences for select double-shell tank (DST) heat-up scenarios for waste feed delivery safety applications. This work extends previous tank bump analyses [Epstein, et al., 2000; and Epstein, et al., 2000a] by focusing on selected scenarios in which extended loss of cooling leads to waste boiling and aerosol generation. The HADCRT code, Version 1.3, is used to provide a best-estimate, coupled simulation of both thermal-hydraulic and aerosol transport phenomena. The physical model employed considers boiling aerosol generation, aerosol agglomeration and gravitational sedimentation in the tank headspace, aerosol removal FR-om steam condensation, and aerosol transport FR-om the headspace to the environment and through ventilation flow paths. Thus, the net result of the calculation is the time history of aerosol release to the immediate vicinity of an underground DST

  9. Vertical clog flow heat transfer with nucleate boiling

    International Nuclear Information System (INIS)

    This paper presents a model for slug flow heat transfer which is modified by incorporating the effect of nucleate boiling. This modification has made the slug flow model more useful for practical situations where presence of nucleate boiling is generally a norm rather than an exception. The model is further improved by taking into account the previously ignored effect of vapor bubbles present in the liquid slug. These modifications not only make the slug flow model more realistic but also improve its predictive capabilities. This improvement is demonstrated by comparing the predictions of the current model and the previous model with the published slug flow data. The mean deviation between the prediction and the measured slug flow data for water in the range of 20 to 100 kW/m2 heat flux is 5.87% for a total 171 data points

  10. Evaluation of boiled potato peel as a wound dressing.

    Science.gov (United States)

    Dattatreya, R M; Nuijen, S; van Swaaij, A C; Klopper, P J

    1991-08-01

    In a series of experiments full thickness skin defects in 68 rats were covered with dressings made of boiled potato peels according to the method developed in Bombay. The wounds closed within 14 days and histologically complete repair of epidermis was found. The cork layer of the potato peel prevents dehydration of the wound and protects against exogenous agents. Experiments with homogenates revealed that a complete structure of the peel is necessary. Steroidal glycosides may have contributed to the favourable results. PMID:1930669

  11. Saturation conditions in elongated single-cavity boiling water targets

    OpenAIRE

    Steyn, G. F.; Vermeulen, C.

    2015-01-01

    Introduction It is shown that a very simple model reproduces the pressure versus beam current characteristics of elongated single-cavity boiling water targets for 18F production surprisingly well. By fitting the model calculations to measured data, values for a single free parameter, namely an overall heat-transfer coefficient, have been extracted for several IBA Nirta H218O targets. IBA recently released details on their new Nirta targets that have a conical shape, which constitutes an im...

  12. Superfluid helium boiling in porous structure under microgravity: model representation

    International Nuclear Information System (INIS)

    The results of model calculation of superfluid helium boiling under microgravity conditions are reported. The evolution of a vapour film on the cylindrical heater surface inside the porous thick walled structure is analyzed. The molecular-kinetic theory methods are used to describe heat and mass-transfer within helium interface. The equation of vapour-liquid interface motion is solved. The effect of experimental parameters on the vapour film properties is studied. The calculation data for microgravity and terrestrial conditions are compared

  13. Revisting the boiling of quark nuggets at nonzero chemical potential

    OpenAIRE

    Li, Ang; Liu, Tong; Gubler, Philipp; Xu, Ren-Xin

    2013-01-01

    The boiling of possible quark nuggets during the quark-hadron phase transition of the Universe at nonzero chemical potential is revisited within the microscopic Brueckner-Hartree-Fock approach employed for the hadron phase, using two kinds of baryon interactions as fundamental inputs. To describe the deconfined phase of quark matter, we use a recently developed quark mass density-dependent model with a fully self-consistent thermodynamic treatment of confinement. We study the baryon number li...

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

    OpenAIRE

    Orzechowski T.; Tyburczyk A.

    2014-01-01

    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 h...

  15. Calculation of limit cycle amplitudes in commercial boiling water reactors

    International Nuclear Information System (INIS)

    This paper describes an investigation of the dynamic behavior of a boiling water reactor (BWR) in the nonlinear region corresponding to linearly unstable conditions. A nonlinear model of a typical BWR was developed. The equations underlying this model represent a one-dimensional void reactivity feedback, point kinetics with a single delayed neutron group, fuel behavior, and recirculation loop dynamics (described by a single-node integral momentum equation)

  16. Analyses of quenching process during turn-off of plasma electrolytic carburizing on carbon steel

    International Nuclear Information System (INIS)

    Highlights: • Cooling rate of carburized steel at the end of PEC treatment is measured. • The quench hardening in the fast or slow turn-off mode hardly takes place. • Decrease of the surface roughness during slow turn-off process is found. • A slow turn-off mode is recommended to replace the conventional turn-off mode. - Abstract: Plasma electrolytic carburizing (PEC) under different turn-off modes was employed to fabricate a hardening layer on carbon steel in glycerol solution without stirring at 380 V for 3 min. The quenching process in fast turn-off mode or slow turn-off mode of power supply was discussed. The temperature in the interior of steel and electron temperature in plasma discharge envelope during the quenching process were evaluated. It was found that the cooling rates of PEC samples in both turn-off modes were below 20 °C/s, because the vapor film boiling around the steel sample reduced the cooling rate greatly in terms of Leidenfrost effect. Thus the quench hardening hardly took place, though the slow turn-off mode slightly decreased the surface roughness of PEC steel. At the end of PEC treatment, the fast turn-off mode used widely at present cannot enhance the surface hardness by quench hardening, and the slow turn-off mode was recommended in order to protect the electronic devices against a large current surge

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

    International Nuclear Information System (INIS)

    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

  18. Electrical control and enhancement of boiling heat transfer during quenching

    Science.gov (United States)

    Shahriari, Arjang; Hermes, Mark; Bahadur, Vaibhav

    2016-02-01

    Heat transfer associated with boiling degrades at elevated temperatures due to the formation of an insulating vapor layer at the solid-liquid interface (Leidenfrost effect). Interfacial electrowetting (EW) fields can disrupt this vapor layer to promote liquid-surface wetting. We experimentally analyze EW-induced disruption of the vapor layer and measure the resulting enhanced cooling during the process of quenching. Imaging is employed to visualize the fluid-surface interactions and understand boiling patterns in the presence of an electrical voltage. It is seen that EW fields fundamentally change the boiling pattern, wherein a stable vapor layer is replaced by intermittent wetting of the surface. Heat conduction across the vapor gap is thus replaced with transient convection. This fundamental switch in the heat transfer mode significantly accelerates cooling during quenching. An order of magnitude increase in the cooling rate is observed, with the heat transfer seen approaching saturation at higher voltages. An analytical model is developed to extract voltage dependent heat transfer rates from the measured cooling curve. The results show that electric fields can alter and tune the traditional cooling curve. Overall, this study presents an ultralow power consumption concept to control the mechanical properties and metallurgy, by electrically tuning the cooling rate during quenching.

  19. Measurement of film dynamics in a boiling liquid film

    International Nuclear Information System (INIS)

    Motivated by understanding the micro-hydrodynamics of boiling heat transfer and its critical heat flux (CHF), the present study investigates the boiling phenomenon in a liquid film whose dynamic thickness is recorded by a confocal optical sensor till micrometres, while the bubble dynamics of the boiling in the film is visualized by high-speed photography (100 fps). This paper is focused on statistical analysis of the thickness signals from the scoping tests from low heat flux till high heat flux (CHF). The dynamic thickness of the liquid film appears peak values, corresponding to the liquid film movements due to nucleation of bubble(s) and its growth and collapse. The maximum thickness decreases rapidly with increasing heat flux, but after 0.625 WM/m2 it keeps almost constant. It reduces again after 1.09 WM/m2 and finally reaches 105 μm prior to the CHF which occurs at 1.563 WM/m2 for the nano heater made of titanium. (author)

  20. Film boiling on porous layered brass sphere during quenching

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Jun-young; Kim, Seol Ha; Jo, Hangjin; Lee, Gi Cheol; Kiyofumi, Moriyama; Park, Hyun Sun [POSTECH, Pohang (Korea, Republic of); Kim, Moo Hwan [KOREA Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-05-15

    Fluid (liquid or gas) can afford to be permeable into porous layer on heat transfer surface and this phenomenon significantly affects phase-change heat transfer, especially boiling. The Corrosion Residual Unidentified Deposition (CRUD) which has generally micro-scaled pore geometry could have considered as porous layer and it was suggested that modification of heat transfer surface like CRUD can influence cooling rate during Loss-Of-Coolant Accident (LOCA) transient. Therefore, role of porous layer will be more emphasized at core-safety analysis, because, recently strategy of nuclear-fuel operation gradually becomes higher burn-up and longer cycle. As another aspect, study about film boiling has widely concerned due to its importance at core-coolability in LOCA, however, consideration of porous layer has relatively restricted because of difficulty of fabrication, excepting for horizontal surface. In this article, we briefly introduce experimental result of film boiling on porous layered surface during quenching. Laboratory-scaled quenching facility was applied and porous layer was fabricated by Electro-Chemical Deposition (ECD) method at spherical brass test section. We observed that the existence of porous layer on heat transfer surface considerable affected the cooling rate (t{sub cool,MPS}/t{sub cool,BBS}-12) during quenching in a saturated distilled water, therefore, it is expected that porous layer like CURD may have the potential able to affect LOCA transient.

  1. Film boiling characteristics of liquid metals in Leidenfrost phenomenon

    International Nuclear Information System (INIS)

    In a steel container filled with 1.0 bar argon gas, potassium or sodium droplets were put on a heated plate and observed by a camera and a VTR through viewports. In the case of potassium plate of 316-stainless steel and tantalum were used respectively for low and high temperatures and joule-heated by a DC power supply, while in the case of sodium a tantalum disk was induction-heated by a 20 kHz and 30 kW power supply. Stable film boiling was observed for a high temperature of the plate or disk, whereas an instantaneous breakup of droplet with extensive vaporization occurred for a low temperature. Whether or not a droplet did float on the plate was mainly governed by the interface temperature at the contact. The heat flux was estimated from volumetric reducing rate of droplet observed. For potassium the results in the film boiling region showed an appreciably good agreement with prediction based on the Bromley's expression combined with the theory of Baumeister et al. For sodium, the experimental results were slightly higher than the prediction. The minimum film boiling temperature and heat flux were roughly estimated to be about 13000C and 15W/cm2 for potassium and about 16000C and 45W/cm2 for sodium, respectively

  2. Phase field model for the study of boiling

    International Nuclear Information System (INIS)

    This study concerns both the modeling and the numerical simulation of boiling flows. First we propose a review concerning nucleate boiling at high wall heat flux and focus more particularly on the current understanding of the boiling crisis. From this analysis we deduce a motivation for the numerical simulation of bubble growth dynamics. The main and remaining part of this study is then devoted to the development and analyze of a phase field model for the liquid-vapor flows with phase change. We propose a thermodynamic quasi-compressible formulation whose properties match the one required for the numerical study envisaged. The system of governing equations is a thermodynamically consistent regularization of the sharp interface model, that is the advantage of the di use interface models. We show that the thickness of the interface transition layer can be defined independently from the thermodynamic description of the bulk phases, a property that is numerically attractive. We derive the kinetic relation that allows to analyze the consequences of the phase field formulation on the model of the dissipative mechanisms. Finally we study the numerical resolution of the model with the help of simulations of phase transition in simple configurations as well as of isothermal bubble dynamics. (author)

  3. Flow boiling heat transfer in mini-channels

    International Nuclear Information System (INIS)

    In view of practical significance of a correlation of heat transfer coefficient in the aspect of such applications as engineering design and prediction, some efforts towards correlating flow boiling heat transfer coefficients for mini-channels have been made in this study. Based on analyses of existing experimental investigations of flow boiling, it was found that liquid-laminar and gas-turbulent flow is a common feature in many applications of mini-channels. Traditional heat transfer correlations for saturated flow boiling were developed for liquid-turbulent and gas-turbulent flow conditions and thus may not be suitable in principle to be used to predict heat transfer coefficients in mini-channels when flow conditions are liquid-laminar and gas-turbulent. By considering flow conditions (laminar or turbulent) in the Reynolds number factor F and single-phase heat transfer coefficient hsp, the Chen correlation has been modified to be used for four flow conditions such as liquid-laminar and gas-turbulent one often occurring in mini-channels. A comparison of the newly developed correlation with various existing data for mini-channels shows a satisfactory agreement. In addition, an extensive comparison of existing general correlations with databases for mini-channels has also been made. (author)

  4. Boiling Heat Transfer Experiments by using Transparent Heated Microtube

    Science.gov (United States)

    Huang, Shih-Che; Kawanami, Osamu; Kawakami, Kazunari; Honda, Itsuro; Kawashima, Yousuke; Ohta, Haruhiko

    For detailed study of the relationship between boiling bubble behavior and inner wall temperature during flow boiling in microtubes, a transparent heated microtube, whose inner wall was coated with a thin gold film, was employed. Boiling behavior could be observed clearly, and the inner wall temperature of the tube was measured simultaneously with direct heating of the film. Ionized water was used as a test fluid. The experimental conditions were as follows: tube diameter, 1 mm; inlet liquid subcooling, 10 K; mass velocity, 100 kg/m2s and heat flux, up to 469 kW/m2 in the open system. As a result, the frequencies of fluctuation of the inner wall temperature and flow rate were divided into four regions. In addition, the fluctuation range of flow rate increased with increasing heat flux however, this fluctuation decreased drastically for heat flux over 212 kW/m2. The fluctuation of void fraction coincided with that of inner wall temperature.

  5. Two-phase boundary layer prediction in upward boiling flow

    International Nuclear Information System (INIS)

    In the present work, the numerical modelling of the two-phase turbulent boundary layer in upward boiling flow was investigated. First, non-dimensional liquid velocity and temperature profiles in the two-phase boundary layer were validated on the one-dimensional section of a pipe with prescribed radial void fraction profiles. Simulations were performed on a fine grid with a commercial code CFX-5 using the k-ω turbulence model. A significant deviation of results from the analytical single-phase and two-phase wall functions from the literature was observed. Second, a wall boiling model in a vertical heated pipe was simulated (CFX-5) on the coarse grid. Here the prediction of the two-phase thermal boudary layer was compared to the experimental data, k-ω calculation on the fine grid and against the singlephase analytical wall function. Again a major deviation against single-phase temperature wall function was obtained. Presented analyses suggest that the existing analytical velocity and temperature wall functions cannot be valid for the boiling boundary layer with the high void fraction on the wall. (author)

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

    International Nuclear Information System (INIS)

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

  7. Intermittent phenomena in the boiling two-phase boundary layer

    International Nuclear Information System (INIS)

    In order to investigate statistical properties of temperature fluctuation in a boiling two-phase boundary layer the corresponding intermittency functions, which describe liquid, vapour and interface region at an individual fixed point, have been defined. In water boiling on a horizontal surface the temperature fluctuation was measured with a microthermocouple and the signal was processed through the digital computer with the detector function specified for liquid, vapor and interface region. The results obtained confirm that the temperature fluctuation in the boiling two-phase layer can be divided into three parts corresponding to individual regions and that its statistical distribution depends on the properties of respective systems. It has also been shown that the temperature fluctuation in the interface region is determinative and corresponds to the temperature changes in the liquid layer surrounding vapor bubble growth. Amplitude distribution in the liquid region changes its form with the distance from the wall as a result of the change in intensity of turbulence at different distances. The probability density distribution in the vapor region shows very small amplitude fluctuation and is almost constant for all distances. (author)

  8. Rheological Properties and Structural Changes in Different Sections of Boiled Abalone Meat

    Institute of Scientific and Technical Information of China (English)

    GAO Xin; TANG Zhixu; ZHANG Zhaohui; Ogawa Hiroo

    2003-01-01

    Changes in tissue structures, rheological properties of cross- and vertical section boiled abalone meat were studied in relation to boiling time. The adductor muscle of abalone Haliotis discus which was removed from the shell, was boiled for 1, 2, and 3 h, respectively. Then it was cut up and separated into cross- and vertical section meat. When observed by a light microscope and a scanning electron microscope, structural changes in the myofibrils were greatest in the cross section meat compared with the vertical section meat. When boiling time was increased from 1 h to 3 h, the instantaneous modulus E0 and rupture strength of both section meat decreased gradually with increased boiling time, and no significant differences were observed between these two section meat for the same boiling time. When boiled for 1 h, the relaxation time of cross section meat was much longer than that of vertical section meat. There were no significant changes in the relaxation time of vertical section for different boiling time, but the relaxation time of cross section meat was reduced gradually with increasing boiling time. These results confirmed that the difference in rheological properties between the cross- and vertical section meat was mainly due to the denaturation level of myofibrils when heated for 1 h, as well as due to the changes in the amount of denatured proteins, and the manner in which the inner denatured protein components weve exchanged after boiling time was increased from 1 h to 3 h.

  9. Experimental investigation and mechanistic modelling of dilute bubbly bulk boiling

    International Nuclear Information System (INIS)

    During evaporation the geometric shape of the vapour is not described using thermodynamics. In bubbly flows the bubble shape is considered spheric with small diameters and changing into various shapes upon growth. The heat and mass transfer happens at the interfacial area. The forces acting on the bubbles depend on the bubble diameter and shape. In this work the prediction of the bubble diameter and/or bubble number density in bulk boiling was considered outside the vicinity of the heat input area. Thus the boiling effects that happened inside the nearly saturated bulk were under investigation. This situation is relevant for nuclear safety analysis concerning a stagnant coolant in the spent fuel pool. In this research project a new experimental set-up to investigate was built. The experimental set-up consists of an instrumented, partly transparent, high and slender boiling container for visual observation. The direct visual observation of the boiling phenomena is necessary for the identification of basic mechanisms, which should be incorporated in the simulation model. The boiling process has been recorded by means of video images and subsequently was evaluated by digital image processing methods, and by that data concerning the characteristics of the boiling process were generated for the model development and validation. Mechanistic modelling is based on the derivation of relevant mechanisms concluded from observation, which is in line with physical knowledge. In this context two mechanisms were identified; the growth/-shrink mechanism (GSM) of the vapour bubbles and sudden increases of the bubble number density. The GSM was implemented into the CFD-Code ANSYS-CFX using the CFX Expression Language (CEL) by calculation of the internal bubble pressure using the Young-Laplace-Equation. This way a hysteresis is realised as smaller bubbles have an increased internal pressure. The sudden increases of the bubble number density are explainable by liquid super

  10. Experimental investigation and mechanistic modelling of dilute bubbly bulk boiling

    Energy Technology Data Exchange (ETDEWEB)

    Kutnjak, Josip

    2013-06-27

    During evaporation the geometric shape of the vapour is not described using thermodynamics. In bubbly flows the bubble shape is considered spheric with small diameters and changing into various shapes upon growth. The heat and mass transfer happens at the interfacial area. The forces acting on the bubbles depend on the bubble diameter and shape. In this work the prediction of the bubble diameter and/or bubble number density in bulk boiling was considered outside the vicinity of the heat input area. Thus the boiling effects that happened inside the nearly saturated bulk were under investigation. This situation is relevant for nuclear safety analysis concerning a stagnant coolant in the spent fuel pool. In this research project a new experimental set-up to investigate was built. The experimental set-up consists of an instrumented, partly transparent, high and slender boiling container for visual observation. The direct visual observation of the boiling phenomena is necessary for the identification of basic mechanisms, which should be incorporated in the simulation model. The boiling process has been recorded by means of video images and subsequently was evaluated by digital image processing methods, and by that data concerning the characteristics of the boiling process were generated for the model development and validation. Mechanistic modelling is based on the derivation of relevant mechanisms concluded from observation, which is in line with physical knowledge. In this context two mechanisms were identified; the growth/-shrink mechanism (GSM) of the vapour bubbles and sudden increases of the bubble number density. The GSM was implemented into the CFD-Code ANSYS-CFX using the CFX Expression Language (CEL) by calculation of the internal bubble pressure using the Young-Laplace-Equation. This way a hysteresis is realised as smaller bubbles have an increased internal pressure. The sudden increases of the bubble number density are explainable by liquid super

  11. Dust-off

    OpenAIRE

    Maycroft, Neil; Cheang, Shu Lea

    2015-01-01

    The fan of a motherboard switches on and off intermittently. It blows household dust, removed from the inside of a computer carcass, into the air. The dust then settles onto the motherboard, to be blown off again. This continual movement of dust is contained in the piece. However, it should remind us that the ceaseless creation and motion of unconfined dust accompanies all stages of the e-waste journey.

  12. Study on pool-nucleate boiling heat transfer characteristics by using artificial cavities

    International Nuclear Information System (INIS)

    Pool boiling heat transfer experiments were performed by using the well-controlled/defined heat transfer surface for water. Uni-size and -shape artificial cavities were created on the mirror-finished silicon plate by utilizing the MEMS technology. Experimental results agreed well with what were predicted by the traditional boiling theory. The mirror finished surface showed only the tendency of natural circulation heat transfer. The artificial-cavity heat transfer surface followed the pool-nucleate boiling trend. The onset of the pool-nucleate boiling was well predicted by the traditional pool-nucleate boiling theory. These results indicated that the artificial cavities behave just like natural cavities. The results indicated the artificial cavities are quite useful and promising to examine the true features of complicated boiling that have been overshadowed by complicatedness. (authors)

  13. Acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor from autoregressive models

    Energy Technology Data Exchange (ETDEWEB)

    Geraldo, Issa Cherif [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Bose, Tanmoy [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Pekpe, Komi Midzodzi, E-mail: midzodzi.pekpe@univ-lille1.fr [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Cassar, Jean-Philippe [Laboratoire d’Automatique, Génie Informatique et Signal (LAGIS UMR CNRS 8219), Université Lille 1, Sciences et technologies, Avenue Paul Langevin, BP 48, 59651 Villeneuve d’Ascq CEDEX (France); Mohanty, A.R. [Indian Institute of Technology Kharagpur, Kharagpur 721302, West Bengal (India); Paumel, Kévin [CEA, DEN, Nuclear Technology Department, F-13108 Saint-Paul-lez-Durance (France)

    2014-10-15

    Highlights: • The work deals with sodium boiling detection in a liquid metal fast breeder reactor. • The authors choose to use acoustic data instead of thermal data. • The method is designed to not to be disturbed by the environment noises. • A real time boiling detection methods are proposed in the paper. - Abstract: This paper deals with acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor (LMFBR) based on auto regressive (AR) models which have low computational complexities. Some authors have used AR models for sodium boiling or sodium–water reaction detection. These works are based on the characterization of the difference between fault free condition and current functioning of the system. However, even in absence of faults, it is possible to observe a change in the AR models due to the change of operating mode of the LMFBR. This sets up the delicate problem of how to distinguish a change in operating mode in absence of faults and a change due to presence of faults. In this paper we propose a new approach for boiling detection based on the estimation of AR models on sliding windows. Afterwards, classification of the models into boiling or non-boiling models is made by comparing their coefficients by two statistical methods, multiple linear regression (LR) and support vectors machines (SVM). The proposed approach takes into account operating mode information in order to avoid false alarms. Experimental data include non-boiling background noise data collected from Phenix power plant (France) and provided by the CEA (Commissariat à l’Energie Atomique et aux énergies alternatives, France) and boiling condition data generated in laboratory. High boiling detection rates as well as low false alarms rates obtained on these experimental data show that the proposed method is efficient for boiling detection. Most importantly, it shows that the boiling phenomenon introduces a disturbance into the AR models that can be clearly detected.

  14. Upward Flow Boiling to DI-Water and Cuo Nanofluids Inside the Concentric Annuli

    OpenAIRE

    N. Vaeli; M. M. Sarafraz; Peyghambarzadeh, S. M.; F Hormozi

    2015-01-01

    In this work, flow boiling heat transfer coefficients of deionized water and copper oxide water-based nanofluids at different operating conditions have been experimentally measured and compared. The liquid flowed in an annular space. According to the experiments, two distinguished heat transfer regions with two different mechanisms can be seen namely forced convective and nucleate boiling regions. Results demonstrated that with increasing the applied heat flux, flow boiling heat transfer coef...

  15. Transient measurement of temperature oscillation during noisy film boiling in superfluid helium II

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    Noisy film boiling, which is characterized by a loud noise andsevere mechanical vibration, is a particular phenomenon of superfluid helium II (He II). Experiments have been conducted under various thermal conditions by varying the heating time th and the heat flux q, and the temperature oscillation during noisy film boiling is measured by the superconductor temperature sensors in order to understand the physical mechanism of noisy film boiling.

  16. Acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor from autoregressive models

    International Nuclear Information System (INIS)

    Highlights: • The work deals with sodium boiling detection in a liquid metal fast breeder reactor. • The authors choose to use acoustic data instead of thermal data. • The method is designed to not to be disturbed by the environment noises. • A real time boiling detection methods are proposed in the paper. - Abstract: This paper deals with acoustic monitoring of sodium boiling in a liquid metal fast breeder reactor (LMFBR) based on auto regressive (AR) models which have low computational complexities. Some authors have used AR models for sodium boiling or sodium–water reaction detection. These works are based on the characterization of the difference between fault free condition and current functioning of the system. However, even in absence of faults, it is possible to observe a change in the AR models due to the change of operating mode of the LMFBR. This sets up the delicate problem of how to distinguish a change in operating mode in absence of faults and a change due to presence of faults. In this paper we propose a new approach for boiling detection based on the estimation of AR models on sliding windows. Afterwards, classification of the models into boiling or non-boiling models is made by comparing their coefficients by two statistical methods, multiple linear regression (LR) and support vectors machines (SVM). The proposed approach takes into account operating mode information in order to avoid false alarms. Experimental data include non-boiling background noise data collected from Phenix power plant (France) and provided by the CEA (Commissariat à l’Energie Atomique et aux énergies alternatives, France) and boiling condition data generated in laboratory. High boiling detection rates as well as low false alarms rates obtained on these experimental data show that the proposed method is efficient for boiling detection. Most importantly, it shows that the boiling phenomenon introduces a disturbance into the AR models that can be clearly detected

  17. Pressure drop across micro-pin heat sinks under boiling conditions

    OpenAIRE

    Koşar, Ali; Kosar, Ali; Özdemir, Mehmed Rafet; Ozdemir, Mehmed Rafet; Keskinöz, Mehmet; Keskinoz, Mehmet

    2009-01-01

    Two-phase pressure drop was studied in four different micro pin fin heat sinks. Micro pin fin heat sinks used in the current studies were operated under boiling conditions using water and R-123 as working fluids. It was observed that once boiling was initiated severe temperature fluctuations and flow oscillations were recorded for three of the micro pin fin heat sinks, which was characterized as unstable boiling. Pressure drop signals were presented just before and after the unstable boili...

  18. Boiling heat transfer in thin liquid layers of mercury under magnetic field

    International Nuclear Information System (INIS)

    Experimental data are presented on the boiling heat transfer from a horizontal plane heater to liquid layers of mercury in the presence of a magnetic field of which direction is parallel to the direction of gravity. Increasing the magnetic flux density, the incipient boiling heat flux and burnout heat flux decrease in compare with those of no magnetic field. However, when the liquid layer is thin, the magnetic field affects them little. The visual study of mercury boiling experiment is also performed. (author)

  19. An analytical and experimental study of pool boiling with particular reference to additives

    International Nuclear Information System (INIS)

    An experimental investigation of nucleate boiling heat transfer and critical heat flux is presented for water and various aqueous solutions boiling from horizontal stainless steel tubes and flat strips at atmospheric pressure. An integral method solution for film boiling is given and compared with existing experimental data. Analytical solutions are also obtained for the temperature profiles with periodic internal heating of a flat plate and a cylinder. (author)

  20. Influence of surface topography in the boiling mechanisms

    International Nuclear Information System (INIS)

    Highlights: • Pool boiling heat transfer. • Use of micro-textured surfaces to enhance heat transfer. • Importance of the bubble dynamics and of the interaction mechanisms in the overall heat transfer efficiency. • Effect of the micro-textures on bubble dynamics as a way to enhance pool boiling heat transfer. - Abstract: The present paper addresses the qualitative and quantitative analysis of the pool boiling heat transfer over micro-structured surfaces. The surfaces are made from silicon chips, in the context of pool boiling heat transfer enhancement of immersion liquid cooling schemes for electronic components. The first part of the analysis deals with the effect of the liquid properties. Then the effect of surface micro-structuring is discussed, covering different configurations, from cavities to pillars being the latter used to infer on the potential profit of a fin-like configuration. The use of rough surfaces to enhance pool boiling mainly stands on the arguments that the surface roughness will increase the liquid–solid contact area, thus enhancing the convection heat transfer coefficient and will promote the generation of nucleation sites. However, one should not disregard bubble dynamics. Indeed, the results show a strong effect of bubble dynamics and particularly of the interaction mechanisms in the overall cooling performance of the pair liquid–surface. The inaccurate control of these mechanisms leads to the formation of large bubbles and strong vertical and horizontal coalescence effects promote the very fast formation of a vapor blanket, which causes a steep decrease of the heat transfer coefficient. This effect can be strong enough to prevail over the benefit of increasing the contact area by roughening the surface. For the micro-patterns used in the present work, the results evidence that one can reasonably determine guiding pattern characteristics to evaluate the intensity of the interaction mechanisms and take out the most of the

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  3. Experimental Evidence of the Vapor Recoil Mechanism in the Boiling Crisis

    CERN Document Server

    Nikolayev, Vadim; Garrabos, Y; Beysens, D

    2016-01-01

    Boiling crisis experiments are carried out in the vicinity of the liquid-gas critical point of H2. A magnetic gravity compensation setup is used to enable nucleate boiling at near critical pressure. The measurements of the critical heat flux that defines the threshold for the boiling crisis are carried out as a function of the distance from the critical point. The obtained power law behavior and the boiling crisis dynamics agree with the predictions of the vapor recoil mechanism and disagree with the classical vapor column mechanism.

  4. Experimental study on the pool boiling CHF enhancement using magnetic fluid

    International Nuclear Information System (INIS)

    This paper will describe the effects of magnetic fluid on CHF enhancement of pool boiling. In order to evaluate the effects as nanoparticle characteristic of magnetic fluid, we compared the CHF values of pool boiling experiment between magnetic fluid and other nanofluids with several volume concentrations. SEM(Scanning Electron Microscope) images were obtained to explain CHF enhancement through the effect of the deposited nanoparticles, which can change the surface wettability, during the pool boiling experiment. Lastly, the analysis for bubble formation in pool boiling using image processing was performed to demonstrate between the characteristics of bubble formation and CHF enhancement. (author)

  5. Two-dimensional simulation of the downcomer boiling experiment using the CUPID code

    International Nuclear Information System (INIS)

    For the analysis of transient two-phase flows in nuclear reactor components, a three-dimensional thermal hydraulics code, named CUPID, has been being developed. We simulated the DOBO (Downcomer Boiling) experiment in two-dimensions using the CUPID code to evaluate its two-phase flow models and verify its applicability to the downcomer boiling analysis. The simulation result showed that it can reproduce the important characteristics of the downcomer boiling, such as a flow pattern change and a circulation of liquid accelerated by bubbles. The two-phase flow models that require further improvement were identified as well for an enhanced prediction of the downcomer boiling. (author)

  6. Nucleate Boiling Heat Transfer of Nanofluids with Carbon Nanotubes on Plain and Low Fin Surfaces

    International Nuclear Information System (INIS)

    Nuclear power generation is being discussed in many countries as an alternative method to solving the world's energy crisis. For the safe operation of nuclear power plants, ways for increasing the critical heat flux (CHF) related to a loss of coolant accident are being investigated. In the event that the local heat flux exceeds the CHF, there is an abrupt shift in the boiling curve such that the nucleate boiling ceases and transition boiling and ultimately film boiling occur, finally resulting in a physical break down of the surface. Therefore, it is essential to maximize the CHF for the protection of nuclear power plants with maximum system performance. For the past decade, as a lot of research has been carried out for an improvement of the boiling heat transfer coefficients (HTCs) and CHF, new methods employing nano particles have been proposed. The objectives of this study are to measure the pool boiling HTCs of the water without and with carbon nanotubes (CNTs) on plain and low fin surfaces up to the CHF, and to analyze the effect of CNTs on both nucleate boiling HTCs and CHF. Pool boiling HTCs on all surfaces tested in water without and with CNTs increased as the heat flux increased, which is a typical trend in the pool boiling of pure fluids. For nanofluid with CNTs on low fin surfaces, the surface geometry and nano particles produced a double effect of increasing the CHFs

  7. Exploring the Limits of Boiling and Evaporative Heat Transfer Using Micro/Nano Structures

    OpenAIRE

    Lu, Ming-Chang

    2010-01-01

    This dissertation presents a study exploring the limits of phase-change heat transfer with the aim of enhancing critical heat flux (CHF) in pool boiling and enhancing thermal conductance in heat pipes. The state-of-the-art values of the CHF in pool boiling and the thermal conductance in heat pipes are about two orders of magnitudes smaller than the limits predicted by kinetic theory. Consequently, there seems to be plenty of room for improvement. Pool boiling refers to boiling at a surface im...

  8. Integrated particle imaging velocimetry and infrared thermometry for high resolution measurement of subcooled nucleate pool boiling

    International Nuclear Information System (INIS)

    High-resolution data of nucleate pool boiling are important for the development of mechanistic numerical models based on computation fluid dynamics (CFD). This paper describes an innovative experimental facility that allows time- and space- resolved measurement of a series of important boiling-relevant quantities, including velocity and temperature distributions around individual bubbles, bubble shape, departure diameter and frequency, all of which can be used for validation of CFD-based models of boiling heat transfer. The facility comprises a temperature controlled boiling cell, with an indium tin oxide (ITO) heater and integrated high-frequency Particle Imaging Velocimetry (PIV) and Infrared Thermometry (IR). Some representative results are reported. (author)

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

    International Nuclear Information System (INIS)

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

  10. Pool boiling in microgravity with a single specie system

    OpenAIRE

    Sagan, Michael; Colin, Catherine; Tanguy, Sébastien

    2012-01-01

    Pool boiling experiments in microgravity on the small copper plate of 1cm² have been performed in the SOURCE 2 experiment aboard the sounding Rocket Maser 12 launched on February 13th, 2012. The SOURCE 2 experiment is a small-scale tank devoted to the study of heat and mass transfers with a liquid refrigerant HFE7000 pressurised with its vapour. SOURCE 2 (SOUnding Rocket Compere Experiment) was developed in the frame of a French German space programme COMPERE (on the behaviour of propellant i...

  11. Digital control application for the advanced boiling water reactor

    International Nuclear Information System (INIS)

    The Advanced Boiling Water Reactor (ABWR) is a 1300 MWe class Nuclear Power Plant whose design studies and demonstration tests are being performed by the three manufacturers, General Electric, Toshiba and Hitachi, under requirement specifications from the Tokyo Electric Power Company. The goals are to apply new technology to the BWR in order to achieve enhanced operational efficiencies, improved safety measures and cost reductions. In the plant instrumentation and control areas, traditional analog control equipment and wire cables will be replaced by distributed digital microprocessor based control units communicating with each other and the control room over fiber optic multiplexed data buses

  12. Loss of coolant accident at boiling water reactors

    International Nuclear Information System (INIS)

    A revision is made with regard to the methods of thermohydraulic analysis which are used at present in order to determine the efficiency of the safety systems against loss of coolant at boiling water reactors. The object is to establish a program of work in the INEN so that the personnel in charge of the safety of the nuclear plants in Mexico, be able to make in a near future, independent valuations of the safety systems which mitigate the consequences of the above mentioned accident. (author)

  13. Safety systems and features of boiling and pressurized water reactors

    International Nuclear Information System (INIS)

    The safe operation of nuclear power plants (NPP) requires a deep understanding of the functioning of physical processes and systems involved. This study was carried out to present an overview of the features of safety systems of boiling and pressurized water reactors that are available commercially. Brief description of purposes and functions of the various safety systems that are employed in these reactors was discussed and a brief comparison between the safety systems of BWRs and PWRs was made in an effort to emphasize of safety in NPPs.(Author)

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

    International Nuclear Information System (INIS)

    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

  15. Flow pattern at critical condition in forced convection boiling

    International Nuclear Information System (INIS)

    An experimental investigation on flow pattern at critical condition (burnout) in forced convection boiling was carried out using R-113 as a working fluid. The test section was an internally heated vertical annular channel with a stainless-steel heater tube of 10 mm O. D. and a glass shroud of 22 mm I. D.. The flow pattern was identified by means of photographic observation and statistical nature of void fraction. Measurements were performed at the pressure 0.3 MPa, mass flux of 500 to 2000 kg/m2s, inlet subcooling of 0 to 58 K. (author)

  16. Contaminant Recovery during In-Situ Boiling in Rock

    Science.gov (United States)

    Chen, F.; Liu, X.; Falta, R. W.; Murdoch, L. C.

    2009-12-01

    In-situ boiling may be an effective mechanism for removing contaminants from tight rock matrix where they would otherwise be all but inaccessible. Heating the matrix above the boiling temperature and then depressurizing will induce boiling that leads to large gas-phase pressure gradients and a steam stripping effect that can remove the contaminants from the matrix. Despite the promise of this process, it has not yet been demonstrated in the field or laboratory, and the controlling parameters and limits of the process are poorly understood. The objective of this project is to characterize mass transfer during boiling in saturated rock. We built an experimental apparatus to heat cores (5cmx30cm) of contaminated rock in a pressurized vessel. The core was sealed in Teflon tube with metal end caps and wrapped with a strip heater. Additional heaters were located in the end caps. Sensors were placed on the surface and embedded within the core to monitor the temperature. An insulation layer covered the strip heater to minimize the heat loss. A recent test was conducted using Berea sandstone (18 millidarcy) initially saturated with de-aired water and contaminated by injecting 200ml (about 2 pore volumes) containing 200mg/L of 1,2-dichloroethane (1,2-DCA), 10 mg/L of chlorobenzene (CB), and 195 mg/L sodium bromide (NaBr). The solution was circulated and both inlet and outlet concentrations were monitored. After the contaminant injection, both the inlet and outlet valves were closed and the core was heated at a constant power of 31.3 watts. Pressure and temperature increased for 3 hours until temperatures exceeded 100 C. A valve on the outlet tube was opened and steam flow started immediately and was routed through a condenser. Concentrations of chlorinated solvents in the outflow increased abruptly to between 6 and 10 times the input concentration. The concentrations decreased after a few 10s of ml were recovered, and at least 80 to 90 percent of the contaminant masses were

  17. Simulation of the Lower Head Boiling Water Reactor Vessel in a Severe Accident

    International Nuclear Information System (INIS)

    The objective of this paper is the simulation and analysis of the Boiling Water Reactor (BWR) lower head during a severe accident. The Couple computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the failure of the Reactor Pressure Vessel (RPV). The release of molten material to the primary containment and the possibility of steam explosion may produce the failure of the primary containment with high radiological consequences. Then, it is important to have a detailed model in order to predict the behavior of the reactor vessel lower head in a severe accident. In this paper, a hypothetical simulation of a Loss of Coolant Accident (LOCA) with simultaneous loss of off-site power and without injection of cooling water is presented with the proposal to evaluate the temperature distribution and heatup of the lower part of the RPV. The SCDAPSIM/RELAP5 3.2 code was used to build the BWR model and conduct the numerical simulation

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

    International Nuclear Information System (INIS)

    For a comprehensive approach of boiling crisis phenomenon in order to get more reliable predictions of critical heat flux in PWR core, a flow pattern study is under progress at CEA GRENOBLE (in a joint program with Electricite de France: EdF). The first aim is to get experimental results on flow structure in the range of thermal hydraulic parameters involved in the core of a PWR (pressure up to 16 MPa, heat flux about 1 MW/m2, mass velocity up to 5000 kg/s/m2. As critical heat flux is a local phenomenon and is the result of the flow development, the data has to be measured from the beginning of boiling until boiling crisis, and from the bulk flow until the boundary layer close to the heating walls. Therefore, these results will be useful in modeling not only boiling crisis phenomenon but also condensation in subcooled boiling, coalescence, splitting up, mass and energy transfers at interfaces, and so on. In a first step, the test section is a vertical tube 19.2 mm internal diameter with an axial uniform heat flux over a 3.5m length. The study is performed on the DEBORA loop with Freon 12 as coolant fluid. We assume that basic boiling phenomena (and the knowledge we get about them) only depend on the fluid properties by means of dimensionless parameters but not on the fluid itself. In a first part, we briefly recall that interfacial detection is the most important parameter of a flow pattern study. Therefore, the use of probes able to measure the Phase Indicator Function (P.I.F.) is necessary. A first study of flow conditions shows that the flow pattern is essentially a bubbly one with vapor particles of low diameter (about 300 clm) and high velocity (up to 7 m/s). These criteria induce that a multiple optical probe is the most appropriate tool provided we improve the technology. We detail the way to obtain probes able to detect small particles at high velocity. Each fiber is stretched to get a tip of 10 Clm with the cladding kept on 50 μm length which defines the

  19. Effect of superheat and electric field on saturated film boiling

    Science.gov (United States)

    Pandey, Vinod; Biswas, Gautam; Dalal, Amaresh

    2016-05-01

    The objective of this investigation is to study the influence of superheat temperature and applied uniform electric field across the liquid-vapor interface during film boiling using a coupled level set and volume of fluid algorithm. The hydrodynamics of bubble growth, detachment, and its morphological variation with electrohydrodynamic forces are studied considering the medium to be incompressible, viscous, and perfectly dielectric at near critical pressure. The transition in interfacial instability behavior occurs with increase in superheat, the bubble release being periodic both in space and time. Discrete bubble growth occurs at a smaller superheat whereas vapor columns form at the higher superheat values. Destabilization of interfacial motion due to applied electric field results in decrease in bubble separation distance and increase in bubble release rate culminating in enhanced heat transfer rate. A comparison of maximum bubble height owing to application of different intensities of electric field is performed at a smaller superheat. The change in dynamics of bubble growth due to increasing superheat at a high intensity of electric field is studied. The effect of increasing intensity of electric field on the heat transfer rate at different superheats is determined. The boiling characteristic is found to be influenced significantly only above a minimum critical intensity of the electric field.

  20. Pool boiling performance of NovecTM 649 engineered fluid

    International Nuclear Information System (INIS)

    A new fluorinated ketone, C2F5C(O)CF(CF3)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 NovecTM 649 Engineered Fluid, C2F5C(O)CF(CF3)2 exhibits very low acute toxicity while maintaining long-term stability. To assess the general two-phase heat transfer performance of NovecTM 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 NovecTM 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)

  1. On-line monitoring of boiling crevice chemistry evolution

    International Nuclear Information System (INIS)

    In a locally restricted geometry on the secondary side of steam generator (SG) in a pressurized water reactor (PWR), impurities in bulk water can be concentrated by boiling process to extreme pH that may then accelerate the corrosion of tubing and adjacent materials. To simulate a real SG tubesheet crevice, a high temperature/high pressure (HT/HP) crevice simulation system was constructed. Primary water was pumped at a high flow rate through a 3/4'' outer-diameter tubing and a crevice section was made on the outer diameter (OD) side of the tubing. The simulated crevice area was monitored with thermocouples and electrodes for the measurement of temperature and electrochemical corrosion potential (ECP), respectively, in the crevice as well as free span. A secondary solution composed of 50 ppm Na and 200 ppb hydrogen (H2) was supplied at a flow rate of about 4 L/hr. In an open tubesheet crevice with 0.15 mm radial gap and 40 mm depth, axial distributions of temperature and ECP were measured as a function of time and available superheat. Sodium hydroxide (NaOH) concentration process in the crevice and the resultant evolution of crevice boiling regions were characterized from temperature and ECP data. Measured data for an open crevice showed a similar behavior to predictions by a thermodynamic equilibrium code. Magnetite-packed crevice had much longer time to reach a steady state than open crevice. (authors)

  2. Development and Capabilities of ISS Flow Boiling and Condensation Experiment

    Science.gov (United States)

    Nahra, Henry; Hasan, Mohammad; Balasubramaniam, R.; Patania, Michelle; Hall, Nancy; Wagner, James; Mackey, Jeffrey; Frankenfield, Bruce; Hauser, Daniel; Harpster, George; Nawrocki, David; Clapper, Randy; Kolacz, John; Butcher, Robert; May, Rochelle; Chao, David; Mudawar, Issam; Kharangate, Chirag R.; O'Neill, Lucas E.

    2015-01-01

    An experimental facility to perform flow boiling and condensation experiments in long duration microgravity environment is being designed for operation on the International Space Station (ISS). This work describes the design of the subsystems of the FBCE including the Fluid subsystem modules, data acquisition, controls, and diagnostics. Subsystems and components are designed within the constraints of the ISS Fluid Integrated Rack in terms of power availability, cooling capability, mass and volume, and most importantly the safety requirements. In this work we present the results of ground-based performance testing of the FBCE subsystem modules and test module which consist of the two condensation modules and the flow boiling module. During this testing, we evaluated the pressure drop profile across different components of the fluid subsystem, heater performance, on-orbit degassing subsystem, heat loss from different modules and components, and performance of the test modules. These results will be used in the refinement of the flight system design and build-up of the FBCE which is manifested for flight in late 2017-early 2018.

  3. Critical Heat Flux during Flow Boiling Experiment with Surfactant Solutions

    International Nuclear Information System (INIS)

    Some additives enhance heat transfer, although, the magnitude and mechanism of enhancement are not consistent or clearly understood. A low concentration of surfactant can also reduce the solution's surface tension considerably, and its level of reduction depends on the amount and type of surfactant present in solution. The surfactant concentrations are usually low enough that the addition of surfactant to water causes no significant change in saturation temperature and most other physical properties, except viscosity and surface tension. Reduced surface tension influences the activation of nucleation sites, bubble growth and dynamics, affecting the boiling heat transfer coefficient. Surfactants effect on CHF (Critical Heat Flux) was determined during flow boiling at atmospheric pressure in closed loop filled with water solutions of tri-sodium phosphate (TSP, Na3PO4.12H2O). TSP was added to the containment sump water to adjust pH level during accidents in nuclear power plants. CHF was measured for four water surfactant solutions at different mass fluxes (100 - 500 kg/m2sec) and two inlet subcooling temperatures (50 .deg. C and 75 .deg. C). Wettability was determined by measuring the contact angle at different concentration cases that will substantiate any CHF increase

  4. High Pressure Boiling Water Reactor HP-BWR

    International Nuclear Information System (INIS)

    Some four hundred Boiling Water Reactors (BWR) and Pressurized Water Reactors (PWR) have been in operation for several decades. The presented concept, the High Pressure Boiling Water Reactor (HP-BWR) makes use of the operating experiences. HP-BWR combines the advantages and leaves out the disadvantages of the traditional BWRs and PWRs by taking in consideration the experiences gained during their operation. The best parts of the two traditional reactor types are used and the troublesome components are left out. HP-BWR major benefits are; 1. Safety is improved; -Gravity operated control rods -Large space for the cross formed control rods between fuel boxes -Bottom of the reactor vessel is smooth and is without penetrations -All the pipe connections to the reactor vessel are well above the top of the reactor core -Core spray is not needed -Internal circulation pumps are used. 2. Environment friendly; -Improved thermal efficiency, feeding the turbine with ∼340 oC (15 MPa) steam instead of ∼285 oC (7MPa) -Less warm water release to the recipient and less uranium consumption per produced kWh and consequently less waste is produced. 3. Cost effective, simple; -Direct cycle, no need for complicated steam generators -Moisture separators and steam dryers are inside the reactor vessel and additional separators and dryers can be installed inside or outside the containment -Well proved simple dry containment or wet containment can be used. (author)

  5. SWR 1000: the Boiling Water Reactor of the future

    International Nuclear Information System (INIS)

    Siemens Power Generation Group (KWU) is currently developing - on behalf of and in close cooperation with the German nuclear utilities and with support from various European partners - Germany's next generation of boiling water reactor. This innovative design concept marks a new era in the successful tradition of boiling water reactor technology and is aimed, with an electric output of 1000 MW, at assuring competitive power generating costs compared lo large-capacity nuclear power plants as well as coal-fired stations, while at the same time meeting the highest of safety standards, including control of a core melt accident. This objective is met by replacing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burnup all contribute towards meeting this goal. In addition, a state-of-the-art materials concept featuring erosion-resistant materials and low-cobalt alloys as well as cobalt-free substitute materials ensures a low cumulative dose for operating and maintenance personnel and also minimizes radioactive waste. (author)

  6. SWR 1000: The new boiling water reactor power plant concept

    International Nuclear Information System (INIS)

    Siemens' Power Generation Group (KWU) is currently developing - on behalf of and in close co-operation with the German nuclear utilities and with support from various European partners - the boiling water reactor SWR 1000. This advanced design concept marks a new era in the successful tradition of boiling water reactor technology in Germany and is aimed, with an electric output of 1000 MW, at assuring competitive power generating costs compared to large-capacity nuclear power plants as well as coal-fired stations, while at the same time meeting the highest of safety standards, including control of a core melt accident. This objective is met by replacing active safety systems with passive safety equipment of diverse design for accident detection and control and by simplifying systems needed for normal plant operation on the basis of past operating experience. A short construction period, flexible fuel cycle lengths of between 12 and 24 months and a high fuel discharge burnup all contribute towards meeting this goal. The design concept fulfils international nuclear regulatory requirements and will reach commercial maturity by the year 2000. (author)

  7. Nucleate boiling pressure drop in an annulus: Book 5

    International Nuclear Information System (INIS)

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D2O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. Nineteen test series and a total of 178 tests were performed. Testing addressed the effects of: Heat flux; pressure; helium gas; power tilt; ribs; asymmetric heat flux. This document consists solely of the plato file index from 11/87 to 11/90

  8. Experimental Research on Flash Boiling Spray of Dimethyl Ether

    Institute of Scientific and Technical Information of China (English)

    Peng Zhang

    2014-01-01

    The high-speed digital imaging technique is applied to observe the developing process of flash boiling spray of dimethyl ether at low ambient pressure, and the effects of nozzle opening pressure and nozzle hole diameter on the spray shape, spray tip penetration and spray angle during the injection are investigated. The experimental results show that the time when the vortex ring structure of flash boiling spray forms and its developing process are determined by the combined action of the bubble growth and breakup in the spray and the air drag on the leading end of spray;with the enhancement of nozzle opening pressure, the spray tip penetration increases and the spray angle decreases. The influence of nozzle hole diameter on the spray tip penetration is relatively complicated, the spray tip penetration is longer with a smaller nozzle hole diameter at the early stage of injection, while the situation is just opposite at the later stage of injection. This paper establishes that the variation of spray angle is consistent with that of nozzle hole diameter.

  9. Radiolysis effects in sub-cooled nucleate boiling

    International Nuclear Information System (INIS)

    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 H2O2. 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 H2 (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 H2 (STP) kg-1 range. (author)

  10. Modelling of boiling bubbly flows using a polydisperse approach

    International Nuclear Information System (INIS)

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

  11. Thermodynamic study on redox reactions of boiling nitric acid solutions

    International Nuclear Information System (INIS)

    In order to understand corrosion of metals in nitric acid solutions, it is necessary to know the generation mechanism of high equilibrium potential in the solutions, especially under boiling conditions. Existing nitrogen oxides in nitric acid solutions were first analyzed by Raman spectroscopy and then existing amount of nitrogen oxides were examined by thermodynamic calculation using the SOLGASMIX software. The Raman spectroscopic analysis showed that the existing amount of un-dissociated HNO3 increased with increasing nitric acid concentration and solution temperature. The existing amount of NO2 also increased by thermal decomposition. The thermodynamic calculation showed that the important nitrogen oxides in nitric acid solutions are HNO3, NO3-, HNO2, NO2, and NO. The equilibrium potential of nitric acid solution is, however, mainly decided by the HNO3/HNO2 equilibrium. The thermodynamic calculation also suggested that the increased oxidization potential on the heat-transfer surface is attributed to the reduction of nitrous acid concentration by the thermal decomposition of nitrous acid on the surface and the continuous removal of decomposition product from the solutions by boiling babbles. (author)

  12. Drag reduction of flow boiling with polymer additives

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The drag-reducing effect of polymer additive aqueous solution was investigated in flow boiling, and the polymer additives were two kinds of polyacrylamide (PAM) with relative molecular mass about 2.56×106 and 8.55×106. The frictional pressure drop was calculated according to the measured total pressure drop. The results show that the flow drag of flow boiling is reduced by adding a small amount of PAM to water when heat flux is in the range of 15.1 kW*m-2 to 47.0 kW*m-2, when the mass fraction of PAM is higher than 2.0×10-5, the drag-reducing effect is obvious. Drag-reducing effect of PAM, whose relative molecular mass is 8.55×106, is slightly better than that of 2.56×106 at the same mass fraction, and the greater the flow rate of the additive solution, the better the effect of the drag reduction.

  13. Numerical modelling of boiling heat transfer in microchannels

    International Nuclear Information System (INIS)

    In this paper we report the results of our modelling studies on two-phase forced convection in microchannels using water as the fluid medium. The study incorporates the effects of fluid flow rate, power input and channel geometry on the flow resistance and heat transfer from these microchannels. Two separate numerical models have been developed assuming homogeneous and annular flow boiling. Traditional assumptions like negligible single-phase pressure drop or fixed inlet pressure have been relaxed in the models making analysis more complex. The governing equations have been solved from the grass-root level to predict the boiling front, pressure drop and thermal resistance as functions of exit pressure and heat input. The results of both the models are compared to each other and with available experimental data. It is seen that the annular flow model typically predicts higher pressure drop compared to the homogeneous model. Finally, the model has also been extended to study the effects of non-uniform heat input along the flow direction. The results show that the non-uniform power map can have a very strong effect on the overall fluid dynamics and heat transfer

  14. 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)

  15. Experimental investigation on critical heat flux and transition boiling of water flow under increased pressure

    International Nuclear Information System (INIS)

    In connection with reactor safety problems (LOCA) a measuring technique has been developed which enables, within the parameter range of medium pressure (0.11 MPa - 1.20 MPa) and low mass flow densities (10 kg/m2s - 500 kg/m2s), exact experimental investigations of critical heat flux and transition boiling of water under quasi-stationary conditions. The system consists of a vertical, temperature-controlled short test section with water flowing upwards inside; an experimental loop controllable to a large extent; a quick automatic data acquisition, and numeric evaluation procedures. Quasi-stationary measured boiling curves, from nucleate boiling to film boiling (circa 450deg C), demonstrate the importance of pressure, mass flow density, and inlet subcooling, the boiling pressure being the most important parameter. The linear course of the boiling curves during transition boiling is remarkable. A frequently suspected hysteresis of the boiling curve could not be detected. The influence of surface effects (contact angle) clearly decreases with increasing pressure. For the empirical correlation of the measured data by means of indices, a statement was chosen that normalizes the heat flux density of transition boiling to the maximum heat flux density at the beginning of the post-CHF range. As a result, the experimental data obtained, and the correlation developed from them, show a better heat transfer in transition boiling than conservatively assumed in general in literature. The temperature-controlled measurements of complete boiling curves supply data for critical heat flow density and the corresponding wall overheating. A comparison with the uncontrolled operation of the test section shows differences of 5-6% only within the range of measurement accuracy of such experiments. (orig.) With 36 figs., 5 tabs

  16. Off-Label Drug Use

    Science.gov (United States)

    ... Your Local Offices Close + - Text Size Off-label Drug Use What is off-label drug use? In the United States new drugs are ... unapproved use of a drug. Is off-label drug use legal? The off-label use of FDA- ...

  17. ON/OFF

    DEFF Research Database (Denmark)

    Jensen, Marianne; Chræmmer, Tanja; Christensen, Trine Søby

    2015-01-01

    Rapporten beskriver projektet On/Off på- og aftagning af kompressionsstrømper, som var et samarbejde mellem Aalborg Kommune, virksomheden Bjørn Nielsen Rehab og Hospitalsartikler A/S samt Lab. X. I Aalborg Kommune indgik to Living Labs i samarbejdet: Hjemmeplejen Område Vest og på Fremtidens...

  18. Off beat: pluralizing rhythm

    NARCIS (Netherlands)

    J.H. Hoogstad; B. Stougaard Pedersen

    2013-01-01

    Off Beat: Pluralizing Rhythm draws attention to rhythm as a tool for analyzing various cultural objects. In fields as diverse as music, culture, nature, and economy, rhythm can be seen as a phenomenon that both connects and divides. It suggests a certain measure with which people, practices, and cul

  19. Nuclear spin-off

    International Nuclear Information System (INIS)

    This booklet gives examples of 'nuclear spin off', from research programmes carried out for the UKAEA, under the following headings; non destructive testing; tribology; environmental protection; flow measurement; material sciences; mechanical engineering; marine services; biochemical technology; electronic instrumentation. (U.K.)

  20. Private Airlines Take Off

    Institute of Scientific and Technical Information of China (English)

    ISABELDING

    2005-01-01

    OKAY Airways, the first wholly private airline in China, took off from the Tianjin Binhai International Airport on March Ⅱ. Carrying a total of 80 passengers, the 189-seat Boeing 737-900 leased from Korea Airlines was bound for Kunming, capital of Yunnan Province, via Changsha,

  1. High-speed infrared thermography for the measurement of microscopic boiling parameters on micro- and nano-structured surfaces

    International Nuclear Information System (INIS)

    Micro- and nano-scale structures on boiling surfaces can enhance nucleate boiling heat transfer coefficient (HTC) and critical heat flux (CHF). A few studies were conducted to explain the enhancements of HTC and CHF using the microscopic boiling parameters. Quantitative measurements of microscopic boiling parameters are needed to understand the physical mechanism of the boiling heat transfer augmentation on structured surfaces. However, there is no existing experimental techniques to conveniently measure the boiling parameters on the structured surfaces because of the small (boiling on micro- and nano-structured surfaces. The visualization results are analyzed to obtain the microscopic boiling parameters. Finally, quantitative microscopic boiling parameters are used to interpret the enhancement of HTC and CHF. In this study, liquid-vapor phase distributions of each surface were clearly visualized by IR thermography during the nucleate boiling phenomena. From the visualization results, following microscopic boiling parameters were quantitatively measured by image processing. - Number density of dry patch, NDP IR thermography technique was demonstrated by nucleate pool boiling experiments with M- and N surfaces. The enhancement of HTC and CHF could be explained by microscopic boiling parameters

  2. Nucleate Pool Boiling of Pure Liquids and Binary Mixtures :Part I—Analytical Model for Boiling Heat Transfer of Pure Liquids on Smooth Tubes

    Institute of Scientific and Technical Information of China (English)

    GuoqingWang; YingkeTan; 等

    1996-01-01

    A mechanism is proposed for nucleate pool boiling heat transfer along with a general model for both pure liquids and binary mixtrues.A combined physical model of bubble growth is also proposed along with a corresponding bubble growth model for pure liquids on smooth tubes.Using the general model and the bubble growth model for pure liquids,an analyticasl model for nucleate pool boiling heat transfer of pure liquids on smooth tubes is developed.

  3. Crisis behaviour of the reactive recoil of a water jets under conditions of explosive boiling

    International Nuclear Information System (INIS)

    One presents the measurement results of the reactive force of boiling up water jet flowing through short channel into the atmosphere depending on superheating degree and at various evaporation mechanisms. The intensive fluctuation evaporation of water (explosive boiling) and presence of a plane perpendicular to the channel axis are shown to result in abrupt reduction of the reactive recoil value

  4. The Gibbs Energy Basis and Construction of Boiling Point Diagrams in Binary Systems

    Science.gov (United States)

    Smith, Norman O.

    2004-01-01

    An illustration of how excess Gibbs energies of the components in binary systems can be used to construct boiling point diagrams is given. The underlying causes of the various types of behavior of the systems in terms of intermolecular forces and the method of calculating the coexisting liquid and vapor compositions in boiling point diagrams with…

  5. Teaching Structure-Property Relationships: Investigating Molecular Structure and Boiling Point

    Science.gov (United States)

    Murphy, Peter M.

    2007-01-01

    A concise, well-organized table of the boiling points of 392 organic compounds has facilitated inquiry-based instruction in multiple scientific principles. Many individual or group learning activities can be derived from the tabulated data of molecular structure and boiling point based on the instructor's education objectives and the students'…

  6. Microwave super-heated boiling of organic liquids: Origin, effect and application

    NARCIS (Netherlands)

    Chemat, F.; Esveld, E.

    2001-01-01

    This paper reports the state of the art of the microwave super-heated boiling phenomenon. When a liquid is heated by microwaves, the temperature increases rapidly to reach a steady temperature while refluxing. It happens that this steady state temperature can be up to 40 K higher than the boiling po

  7. IAEA/IWGFR benchmark tests on sodium boiling noise detection. Part 1

    International Nuclear Information System (INIS)

    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). 2 refs, 21 figs, 2 tabs

  8. Generation of shockwave and vortex structures at the outflow of a boiling water jet

    Science.gov (United States)

    Alekseev, M. V.; Lezhnin, S. I.; Pribaturin, N. A.; Sorokin, A. L.

    2014-12-01

    Results of numerical simulation for shock waves and generation of vortex structures during unsteady outflow of boiling liquid jet are presented. The features of evolution of shock waves and vortex structures formation during unsteady outflow of boiling water are compared with corresponding structures during unsteady gas outflow.

  9. Experimental study on augmentation of nucleate boiling heat transfer on nano porous surfaces

    Energy Technology Data Exchange (ETDEWEB)

    Park, Young Jae; Kim, Hyung Dae [Kyung Hee Univ., Seoul (Korea, Republic of)

    2012-10-15

    Nucleate boiling broadly occurs in thermal hydraulic and safety systems of nuclear power plant (NPP). Heat transfer performance of nucleate boiling is closely related to efficiency and safety of NPPs. Hence, there have been numerous researches to effectively enhance nucleate boiling heat transfer performance. A number of recent studies have reported significant enhancements in nucleate boiling heat transfer coefficient (NBHTC) and critical heat flux (CHF) by fabricating nano/microscale structures on a boiling surface. Wei et al. showed that both NBHTC and CHF can be significantly enhanced with micro pin finned structures. They explained enhancement of NBHTC and CHF that occurred by increase in effective heat transfer area due to micro pin finned structures. Ahn et al. reported 100% enhancement in CHF on a boiling surface with nano/micro hybrid structures. They analyzed CHF enhancement that was caused by improvement of surface wettability on Nano/micro hybrid structures. In this study, an ordered nano porous surface was prepared using anodized aluminum oxide (AAO) technique and nucleate boiling heat transfer performance was examined in a pool with FC 72. Furthermore, the pool boiling result on the nano porous surface was interpreted based on heterogeneous bubble nucleation theory from a cavity.

  10. Prediction of a subcooled boiling flow with advanced two-phase flow models

    International Nuclear Information System (INIS)

    Highlights: ► In this study, advanced two-phase flow models were examined to enhance the prediction capability of subcooled boiling flows for the CFD code. ► They consist of Sγ bubbles size, new wall boiling and two-phase logarithmic wall function models. ► The benchmark calculation confirms that advanced two-phase flow models show good prediction results. - Abstract: Prediction of bubble size which governs interfacial transfer terms between the two phases is of importance for an accurate prediction of the subcooled boiling flow. In the present work, a mechanistic bubbles size model, Sγ was examined to enhance the prediction capability of subcooled boiling flows for the CFD (computational fluid dynamics) code. In addition to this, advanced subcooled boiling models such as new wall boiling and two-phase logarithmic wall function models were also applied for an improvement of energy partitioning and two-phase turbulence models, respectively. The benchmark calculation against the DEBORA subcooled boiling data confirms that the Sγ bubble size model with the two advanced subcooled boiling models shows good prediction results and is applicable to the wide range of flow conditions that are expected in the nominal and postulated accidental conditions of a nuclear power plant.

  11. 78 FR 46378 - La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact...

    Science.gov (United States)

    2013-07-31

    ... COMMISSION La Crosse Boiling Water Reactor, Environmental Assessment and Finding of No Significant Impact... of Title 10 of the Code of Federal Regulations (10 CFR) for the La Crosse Boiling Water Reactor... modifying or adding EP requirements in Section 50.47, Section 50.54, and Appendix E of 10 CFR part 50 (76...

  12. 77 FR 27097 - LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI

    Science.gov (United States)

    2012-05-08

    ... COMMISSION LaCrosse Boiling Water Reactor, Exemption From Certain Requirements, Vernon County, WI AGENCY...) 73.55, for the LaCrosse Boiling Water Reactor (LACBWR). This Environmental Assessment (EA) has been... revised 10 CFR 73.55 through the issuance of a final rule on March 27, 2009 (74 FR 13926). Section...

  13. High conversion pressurized water reactor with boiling channels

    Energy Technology Data Exchange (ETDEWEB)

    Margulis, M., E-mail: maratm@post.bgu.ac.il [The Unit of Nuclear Engineering, Ben Gurion University of the Negev, POB 653, Beer Sheva 84105 (Israel); Shwageraus, E., E-mail: es607@cam.ac.uk [Department of Engineering, University of Cambridge, CB2 1PZ Cambridge (United Kingdom)

    2015-10-15

    Highlights: • Conceptual design of partially boiling PWR core was proposed and studied. • Self-sustainable Th–{sup 233}U fuel cycle was utilized in this study. • Seed-blanket fuel assembly lattice optimization was performed. • A coupled Monte Carlo, fuel depletion and thermal-hydraulics studies were carried out. • Thermal–hydraulic analysis assured that the design matches imposed safety constraints. - Abstract: Parametric studies have been performed on a seed-blanket Th–{sup 233}U fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts suggested substantial reduction in the core power density is needed in order to operate under nominal PWR system conditions. Boiling flow regime in the seed region allows more heat to be removed for a given coolant mass flow rate, which in turn, may potentially allow increasing the power density of the core. In addition, reduced moderation improves the breeding performance. A two-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to be 104 W/cm{sup 3}, created a map of potentially feasible designs. It was found that several options have the potential to achieve end of life fissile inventory ratio above unity, which implies potential feasibility of a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. Finally, a preliminary three-dimensional coupled neutronic and thermal–hydraulic analysis for a single seed-blanket fuel assembly was performed. The results indicate that axial void distribution changes drastically with burnup. Therefore

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

    International Nuclear Information System (INIS)

    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 CaSO4 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/m2 and CaSO4 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

  15. High conversion pressurized water reactor with boiling channels

    International Nuclear Information System (INIS)

    Highlights: • Conceptual design of partially boiling PWR core was proposed and studied. • Self-sustainable Th–233U fuel cycle was utilized in this study. • Seed-blanket fuel assembly lattice optimization was performed. • A coupled Monte Carlo, fuel depletion and thermal-hydraulics studies were carried out. • Thermal–hydraulic analysis assured that the design matches imposed safety constraints. - Abstract: Parametric studies have been performed on a seed-blanket Th–233U fuel configuration in a pressurized water reactor (PWR) with boiling channels to achieve high conversion ratio. Previous studies on seed-blanket concepts suggested substantial reduction in the core power density is needed in order to operate under nominal PWR system conditions. Boiling flow regime in the seed region allows more heat to be removed for a given coolant mass flow rate, which in turn, may potentially allow increasing the power density of the core. In addition, reduced moderation improves the breeding performance. A two-dimensional design optimization study was carried out with BOXER and SERPENT codes in order to determine the most attractive fuel assembly configuration that would ensure breeding. Effects of various parameters, such as void fraction, blanket fuel form, number of seed pins and their dimensions, on the conversion ratio were examined. The obtained results, for which the power density was set to be 104 W/cm3, created a map of potentially feasible designs. It was found that several options have the potential to achieve end of life fissile inventory ratio above unity, which implies potential feasibility of a self-sustainable Thorium fuel cycle in PWRs without significant reduction in the core power density. Finally, a preliminary three-dimensional coupled neutronic and thermal–hydraulic analysis for a single seed-blanket fuel assembly was performed. The results indicate that axial void distribution changes drastically with burnup. Therefore, some means of

  16. Early detection of coolant boiling in research reactors with MTR-type fuel

    Energy Technology Data Exchange (ETDEWEB)

    Kozma, R.; Turkcan, E.; Verhoef, J.P. [Netherlands Energy Research Foundation (ECN), Petten (Netherlands)

    1992-10-01

    A reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that the noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University of Delft, The Netherlands. (author). 8 refs., 11 figs.

  17. Early detection of coolant boiling in research reactors with MTR-type fuel

    International Nuclear Information System (INIS)

    In this paper, a reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University Delft, The Netherlands

  18. Early detection of coolant boiling in research reactors with MTR-type fuel

    International Nuclear Information System (INIS)

    A reactor core monitoring system having the function of early detection of boiling in the coolant channels of research reactors with MTR-type fuel is introduced. The system is based on the on-line analysis of signals of various ex-core and in-core neutron detectors. Early detection of coolant boiling cannot be accomplished by the evaluation of the DC components of these detectors in a number of practically important cases of boiling anomaly. It is shown that the noise component of the available neutron detector signals can be used for the detection of boiling in these cases. Experiments have been carried out at a boiling setup in the research reactor HOR of the Interfaculty Reactor Institute, Technical University of Delft, The Netherlands. (author). 8 refs., 11 figs

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

    International Nuclear Information System (INIS)

    This paper presents results of ex-vessel boiling experiments performed in the CYBL (cylindrical boiling) facility, a reactor-scale facility which has a tank-within-a-tank design simulating the reactor vessel and the reactor cavity. Experiments with uniform and edge-peaked heat flux distributions up to 20 W/cm2 across the vessel bottom are performed. Boiling outside the reactor vessel is found to be subcooled nucleate boiling. The subcooling is mainly due to the gravity head which results from flooding the sides of the reactor vessel. The boiling process exhibits a cyclic pattern with four distinct phases: direct liquid/solid contact, bubble nucleation and growth, coalescence, and vapor mass dispersion (ejection). The results suggest that the flooded cavity in a passive PWR like the AP-600 should be capable of cooling the RPV in the central region of the lower head. 9 figs., 1 tab., 14 refs

  20. Simulation of boiling flow experiments close to CHF with the NEPTUNE-CFD code

    International Nuclear Information System (INIS)

    A three-dimensional two-fluid code NEPTUNECFD has been validated against the ASU (Arizona State University) [1] and DEBORA [2, 3] boiling flow experiments. Nucleate boiling processes in the subcooled flow boiling regime have been studied on ASU experiments. Within this scope a new wall function is implemented in the NEPTUNECFD V1.0.6 code to improve the prediction of flow parameters in the boiling boundary layer. The capability of the code to predict boiling flow regime close to critical heat flux (CHF) conditions has been assessed on selected DEBORA experiments. It was shown that the code is able to predict wall temperature excursion and a sharp void fraction increase near the heated wall, which are characteristic phenomena for CHF conditions. (author)

  1. In-air PIXE for analyzing heavy metals in water boiled in pans

    International Nuclear Information System (INIS)

    The release rates of heavy metals from pans were measured for boiling water as well as for an acidic solution prior to an investigation on the release or sorption of trace elements due to cooking of food by boiling. The boiled samples were condensed and analyzed by means of in-air PIXE. The release of heavy metals was measured for five kinds of pans. For all pans the release rates were considerably more increased by boiling of a 5% solution of acetic acid. Furthermore it was found that by using the alumina coated aluminum pan (alumina pan) the respective release rates of Fe, Cu and Zn were all less than 50 μg per 100 cm2 of the pan surface dipped in the solution, and that monitoring of the contents of aluminum in the boiled solution enabled the estimation of the contribution of metal elements from the pan wall. (orig.)

  2. Pool boiling heat transfer on heterogeneous wetting surface with hydrophobic dots

    International Nuclear Information System (INIS)

    The boiling heat transfer mechanism of pool boiling is fundamental phenomena for understanding the phase change nature. Of many surface characteristics, the effects of wettability of heating surface is focused on as the dominant parameter for bubble dynamics and boiling heat transfer. In this study, highly controlled heating surfaces via MEMs technique were used for understanding the boiling heat transfer of heterogeneous wetting surfaces mixed by hydrophobic dots and a hydrophilic substrate. The diameter of hydrophobic dots and area ratio of phobic dots to heating area were regulated. The range of phobic dot diameter and area ratio were 50∼1000μm and 18.33∼54.3%, respectively. The performance of boiling heat transfer of each surface were evaluated by comparing with a wholly hydrophilic surface. It will contribute to understand the mechanism and criterion of enhanced heating surface condition by modified surface treatment procedure

  3. A fractal study for nucleate pool boiling heat transfer of nanofluids

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    In this paper, a fractal model for nucleate pool boiling heat transfer of nanofluids is developed based on the fractal distribution of nanoparticles and nucleation sites on boiling surfaces. The model shows the dependences of the heat flux on nanoparticle size and the nanoparticle volume fraction of the suspension, the fractal dimension of the nanoparticle and nucleation site, temperature of nanofluids and properties of fluids. The fractal model predictions show that the natural convection stage continues relatively longer in the case of nanofluids. The addition of nanoparticles causes a decrease of the pool nucleate boiling heat transfer. The nucleate pool boiling heat transfer coefficient is decreased by increasing particle concentration. An excellent agreement between the proposed model predictions and experimental data is found. The validity of the fractal model for nucleate pool boiling heat transfer is thus verified.

  4. Acoustic measurements of the boiling stability tests on THORS sodium loop

    International Nuclear Information System (INIS)

    Acoustic data of boiling stability tests on the THORS (Thermal-Hydraulic Out-of-Reactor Safety) facility were obtained using three sodium-immersible high temperature microphones. The data was analyzed in both the time and frequency domains and provides the following information: (1) the acoustic signal due to sodium boiling was clearly observed; (2) the signal level and the repetition rate of boiling pulses are directly proportional to the applied heat flux; (3) a typical boiling pulse consists of a high frequency signal due mainly to the bubble collapses and a low frequency void oscillation; (4) the frequency spectra of the boiling and background pulses can be mostly assigned to various acoustic resonance frequencies of the THORS loop

  5. Modelling of CRUD growth phenomena on PWR fuel rods under nucleate boiling conditions

    International Nuclear Information System (INIS)

    PWR primary circuit materials undergo general corrosion leading to a release of metallic element release and subsequent process of particle deposition and ion precipitation on the primary circuit surfaces. The species accumulated on fuel rods are activated by neutron flux. Consequently, crud erosion and dissolution induce primary coolant contamination. In French PWRs, 58Co volume activity is generally low and almost constant (< 30 MBq.m-3) throughout an ordinary operating cycle. In some specific cases, a significant increase in volume activity is observed after the middle of a cycle (100-1000 MBq.m-3 for 58Co) when conditions for nucleate boiling are locally reached in certain fuel assemblies. Indeed, it is well known that nucleate boiling intensifies the deposition process. The thickness of the crud layer can reach some micrometers in non-boiling areas, whereas it can reach 100 micrometers in boiling areas. Crud growth in boiling conditions can be related to three phenomena: bubble growth induces deposition process (called boiling deposition), bubbles induce concentration increase at crud-coolant interface (called enrichment and modelled by the enrichment factor, the ratio between the wall concentration and the bulk concentration) and vaporisation induces concentration increase inside the crud. A literature review on the modelling of these phenomena and on the crud structure in nucleate boiling conditions has been carried out. The OSCAR [1] calculation code developed by the CEA to predict surface and volume activities in a single phase PWR primary circuit was chosen as a basis for present study. Ability to describe local nucleate boiling conditions was added to this code leading to realistic modelling of subsequent volume activity increase. In this article, we present the results obtained using a modified version of the OSCAR PC V1.2 calculation code including: - A double phase thermal-hydraulic module, - A model of boiling crud growth, able to calculate inner

  6. Flow boiling of water on nanocoated surfaces in a microchannel

    CERN Document Server

    Phan, Hai Trieu; Marty, Philippe; Colasson, Stéphane; Gavillet, Jérôme

    2010-01-01

    Experiments were performed to study the effects of surface wettability on flow boiling of water at atmospheric pressure. The test channel is a single rectangular channel 0.5 mm high, 5 mm wide and 180 mm long. The mass flux was set at 100 kg/m2 s and the base heat flux varied from 30 to 80 kW/m2. Water enters the test channel under subcooled conditions. The samples are silicone oxide (SiOx), titanium (Ti), diamond-like carbon (DLC) and carbon-doped silicon oxide (SiOC) surfaces with static contact angles of 26{\\deg}, 49{\\deg}, 63{\\deg} and 103{\\deg}, respectively. The results show significant impacts of surface wettability on heat transfer coefficient.

  7. Channel-type nuclear reactor with a boiling coolant

    International Nuclear Information System (INIS)

    The invention is aimed at increasing the channel-type reactor safety, in particular, RBMK-type reactors, during accidents resulting in the coolant circulation discontinuation. The reactor core is assembled of vertial technological channels connected in parallel between distributing group collectors and drum-separator. Each technological channel contains a high pressure tube, a fuel assembly with fuel elements and a storage vessel located above the fuel assembly which is filled with water at saturation temperature in the normal operation regime. After dehydration of channels in the course of accident the boiling water from storage vessel is ejected into them. So the device described allows one to reduce the fuel element can temperature in the course of accidents connected with the coolant circulation discontinuation and so to increase the plant safety level

  8. Radial nodalization effects on BWR [boiling water reactor] stability calculations

    International Nuclear Information System (INIS)

    Computer simulations have shown that stability calculations in boiling water reactors (BWRs) are very sensitive to a number of input parameters and modeling assumptions. In particular, the number of thermohydraulic regions (i.e., channels) used in the calculation can affect the results of decay ratio calculations by as much as 30%. This paper presents the background theory behind the observed effects of radial nodalization in BWR stability calculations. The theory of how a radial power distribution can be simulated in time or frequency domain codes by using ''representative'' regions is developed. The approximations involved in this method of solution are reviewed, and some examples of the effect of radial nodalization are presented based on LAPUR code solutions. 2 refs., 4 figs., 2 tabs

  9. A stochastic study of noise in boiling reactors

    International Nuclear Information System (INIS)

    A stochastic point model is considered of a boiling reactor, involving the population of neutrons, the population of delayed neutron precursors, fuel temperature, and the number of bubbles in the coolant as random variables. Whereas the first two variables are related to capture, fission and delayed neutron processes, the other two take into account heat transfer between fuel and coolant and changes in coolant density. The variations in the fuel temperature and the coolant density generate reactivity feedbacks which affects the neutron power spectral density; analysis of the shape of this spectral density is expected to give information on the value of reactor parameters such as, for example, the void coefficient. (U.K.)

  10. Dynamic simulation of a boiling water nuclear reactor

    International Nuclear Information System (INIS)

    For the application of modern control theory, specifically optimal control, to the boiling water reactor, it is necessary to have a linear model that is validated. The nonlinear model of the BWR derived on the basis of physical laws and empirical relations is linearized around an operating point and the model if verified against experimental results by simulating various tests such as the pressure transient test, change in power to recirculating pump etc. The transport delay occurring in the model is approximated by various representations and the results are compared with the exact delay representation. Validation such as discussed in the paper forms the basis for devising appropriate control strategies in the presence of disturbances. (author)

  11. Analytical simulation of boiling water reactor pressure suppression pool swell

    International Nuclear Information System (INIS)

    In a loss-of-coolant accident, the pressure suppression pool of a boiling water reactor swells as a steam/air mixture is expelled from the drywell into the pool and large gas bubbles are formed beneath the surface. Many tests have been performed to quantify pool swell loads, but analytical methods have been limited in their ability to provide accurate loading estimates. With advancement of numerical methods, it is now feasible to numerically simulate the pool swell process. A finite difference solution algorithm is used to solve the transient imcompressible equations for the liquid flow field. Boundary conditions at the fluid-gas interface are determined using a simplified gas flow model. The program is used to simulate several pool swell tests: comparison of the simulation with test data shows good agreement

  12. Analytical simulation of boiling water reactor pressure suppression pool swell

    Energy Technology Data Exchange (ETDEWEB)

    Widener, S.K.

    1986-01-01

    In a loss-of-coolant accident, the pressure suppression pool of a boiling water reactor swells as a steam/air mixture is expelled from the drywell into the pool and large gas bubbles are formed beneath the surface. Many tests have been performed to quantify pool swell loads, but analytical methods have been limited in their ability to provide accurate loading estimates. With advancement of numerical methods, it is now feasible to numerically simulate the pool swell process. A finite difference solution algorithm is used to solve the transient imcompressible equations for the liquid flow field. Boundary conditions at the fluid-gas interface are determined using a simplified gas flow model. The program is used to simulate several pool swell tests: comparison of the simulation with test data shows good agreement.

  13. Theoretical prediction method of subcooled flow boiling CHF

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-12-31

    A theoretical critical heat flux (CHF ) model, based on lateral bubble coalescence on the heated wall, is proposed to predict the subcooled flow boiling CHF in a uniformly heated vertical tube. The model is based on the concept that a single layer of bubbles contacted to the heated wall prevents a bulk liquid from reaching the wall at near CHF condition. Comparisons between the model predictions and experimental data result in satisfactory agreement within less than 9.73% root-mean-square error by the appropriate choice of the critical void fraction in the bubbly layer. The present model shows comparable performance with the CHF look-up table of Groeneveld et al.. 28 refs., 11 figs., 1 tab. (Author)

  14. 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)

  15. Feasibility study on the thorium fueled boiling water breeder reactor

    International Nuclear Information System (INIS)

    The feasibility of (Th,U)O 2 fueled, boiling water breeder reactor based on conventional BWR technology has been studied. In order to determine the potential use of water cooled thorium reactor as a competitive breeder, this study evaluated criticality, breeding and void reactivity coefficient in response to changes made in MFR and fissile enrichments. The result of the study shows that while using light water as moderator, low moderator to fuel volume ratio (MFR=0.5), it was possible to breed fissile fuel in negative void reactivity condition. However the burnup value was lower than the value of the current LWR. On the other hand, heavy water cooled reactor shows relatively wider feasible breeding region, which lead into possibility of designing a core having better neutronic and economic performance than light water with negative void reactivity coefficient. (authors)

  16. Effects of Outflow Area on Pool Boiling in Vertical Annulus

    International Nuclear Information System (INIS)

    To identify the effects of an outflow area on pool boiling heat transfer in a vertical annulus, three different flow recreates were studied experimentally. For the test, a heated tube of smooth stainless steel and water at atmospheric pressure were used. Both annuli with open and closed bottoms were considered. To validate the effects of the outflow area on the heat transfer, the results of the annulus with the reactors were compared with the data for the plain annulus without the reactors. The reduction of the outflow area ultimately results in a decrease in the heat transfer. As the outflow area is very small, a slight increase in heat transfer is also observed. The major cause of this tendency is explained as the difference in the intensity of liquid agitation cause by the movement of coalesced bubbles. It is identified that the convective flow, pulsating flow, and evaporative mechanism are considered as the important mechanisms

  17. Pool boiling of nanoparticle-modified surface with interlaced wettability

    KAUST Repository

    Hsu, Chin-Chi

    2012-01-01

    This study investigated the pool boiling heat transfer under heating surfaces with various interlaced wettability. Nano-silica particles were used as the coating element to vary the interlaced wettability of the surface. The experimental results revealed that when the wettability of a surface is uniform, the critical heat flux increases with the more wettable surface; however, when the wettability of a surface is modified interlacedly, regardless of whether the modified region becomes more hydrophilic or hydrophobic, the critical heat flux is consistently higher than that of the isotropic surface. In addition, this study observed that critical heat flux was higher when the contact angle difference between the plain surface and the modified region was smaller. © 2012 Hsu et al.

  18. Boiling Heat Transfer on Superhydrophilic, Superhydrophobic, and Superbiphilic Surfaces

    CERN Document Server

    Betz, Amy Rachel; Kim, Chang-Jin 'CJ'; Attinger, Daniel

    2012-01-01

    With recent advances in micro- and nanofabrication, superhydrophilic and superhydrophobic surfaces have been developed. The statics and dynamics of fluids on these surfaces have been well characterized. However, few investigations have been made into the potential of these surfaces to control and enhance other transport phenomena. In this article, we characterize pool boiling on surfaces with wettabilities varied from superhydrophobic to superhydrophilic, and provide nucleation measurements. The most interesting result of our measurements is that the largest heat transfer coefficients are reached not on surfaces with spatially uniform wettability, but on biphilic surfaces, which juxtapose hydrophilic and hydrophobic regions. We develop an analytical model that describes how biphilic surfaces effectively manage the vapor and liquid transport, delaying critical heat flux and maximizing the heat transfer coefficient. Finally, we manufacture and test the first superbiphilic surfaces (juxtaposing superhydrophobic ...

  19. Operational margin monitoring system for boiling water reactor power plants

    International Nuclear Information System (INIS)

    This paper reports on an on-line operational margin monitoring system which has been developed for boiling water reactor power plants to improve safety, reliability, and quality of reactor operation. The system consists of a steady-state core status prediction module, a transient analysis module, a stability analysis module, and an evaluation and guidance module. This system quantitatively evaluates the thermal margin during abnormal transients as well as the stability margin, which cannot be evaluated by direct monitoring of the plant parameters, either for the current operational state or for a predicted operating state that may be brought about by the intended operation. This system also gives operator guidance as to appropriate or alternate operations when the operating state has or will become marginless

  20. Resolution of US regulatory issues involving boiling water reactor stability

    International Nuclear Information System (INIS)

    The U.S. Nuclear Regulatory Commission (NRC) and the Boiling Water Reactor Owners Group (BWROG) have been reexamining BWR instability characteristics and consequences since the March 1988 instability event at LaSalle Unit 2. The NRC and BWROG concluded that existing reactor protection systems do not prevent violation of the critical power ratio (CPR) safety limits caused by large asymmetric oscillations. The studies are also examining the need to modify the automatic and operator actions previously developed for response to an anticipated transient without scram (ATWS) event because of oscillation effects not fully considered in previous studies. This paper presents the current status of these studies and an assessment of actions needed to resolve the issue. (author)

  1. Multi-dimensional nodal analysis of boiling water reactor stability

    International Nuclear Information System (INIS)

    A computer program, NUFREQ-3D, was developed for boiling water reactor stability analysis. The code, which incorporates sophisticated thermal-hydraulic model coupled with a space dependent nodal neutronic model, is able to evaluate the system stabilities in terms of state variables such as inlet flow rate, power density, and system pressure. The detailed full 3-D representation was developed for more accurate stability analysis by using the sparse matrix techniques and by a channel grouping procedure. Results of modeling a representative operating BWR system show that spatial coupling has a significant effect on the prediction of stability margins. Comparisons of calculated transfer functions with the measured data also reveal that the code generally predict well the trends of system transfer functions

  2. Improvements in boiling water reactor designs and safety

    International Nuclear Information System (INIS)

    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 discussed in this paper. They include: design simplification; improved safety and reliability; reduced construction, fuel and operating costs; improved maneuverability; and reduced occupational exposure and radwaste. Key features of the ABWR are internal recirculation pumps; fine-motion, electro-hydraulic control rod drives; digital control and instrumentation; multiplexed, fiber optic cabling network; 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 turbine/generator with 52 last stage buckets; and advanced radwaste technology

  3. Behavior of bubble in subcooled boiling with forced convection, 2

    International Nuclear Information System (INIS)

    The objective of this research is to observe the bubble behavior in subcooled boiling with forced convection and to investigate the conditions of heaving test with a heater which initiates isolated bubbles and with high speed motion photography. The observation was made at three inlet subcooling of 15, 30 and 45 K with pressure of 0.3 MPa, mass flux of 1000 kg/m2·s, heat flux of 35 kW/m2. At inlet subcooling of 15 K, bubble velocity is nearly constant in the field of visions for upstream and downstream area. But at the higher subcooling, bubble velocity varies greatly and bubble collapse in the camera field. (author)

  4. Flow Structures Around Micro-bubbles During Subcooled Nucleate Boiling

    Institute of Scientific and Technical Information of China (English)

    WANG Hao; PENG Xiao-Feng; David M. Christopher; WANG Bu-Xuan

    2005-01-01

    The flow structures were investigated around micro bubbles on extremely thin wires during subcooled nucleate boiling. Jet flows emanating from the bubbles were observed visually with the fluid field measurement using high-speed photography and a PIV system. The jet flows induced a strong pumping effect around a bubble. The multi-jet structure was further observed experimentally, indicating the evolution of flow structure around micro bubbles. Numerical simulations explore that the jet flows were induced by a strong Marangoni effect due to high temperature gradients near the wire. The bubble interface with multi-jet structure has abnormal temperature distribution such that the coolest parts were observed at two sides of a bubble extending into the subcooled bulk liquid rather than at the top. Evaporation and condensation on the bubble interface play important roles not only in controlling the intensity of the jet flow, but also in bringing out the multi-jet structure.

  5. Boiling water reactor operator training and qualification in Japan

    International Nuclear Information System (INIS)

    Nuclear power plant operators in Japan are individuals employed by each electric power company. A recruit goes through his company's training; afterwards, he is given a qualification rating and is assigned to practical duty. The only formal qualification authorized by the Japanese government is the full-fledged shift supervisor. Other classifications such as assistant shift supervisor, shift foreman, reactor operator, and subreactor operator are all designated and appointed by each company's in-house regulations. As a part of the training system, power companies that require the use of a full-scope simulator in their training programs utilize the boiling water reactor (BWR) and pressurized water reactor operator training centers. Both were set up independently of the power companies. A synopsis of the BWR Operator Training Center Corp. (BTC ) and its training systems, features, performance evaluation, curriculum improvement, and related items is presented

  6. Boiling heat transfer during rapid quenching using standard nanofluids

    International Nuclear Information System (INIS)

    Rapid quenching experiments were conducted using standard nanofluid (Si 0.01, 0.001 vol%) under atmospheric temperature and pressure. Standard nanofluid prepared by the PLAL technique, which has the narrow band size distribution and good dispersibility in base fluid. Two types of a platinum wire(bare and deposited with nanoparticles) as a heat source and temperature sensor were rapidly dipped in two type of fluid(base water, i.e. de-ionized water and Si nanofluid). During quench, the cooling rates were recorded and then recalculated to complete the boiling curve. Higher cooling rate was observed in the case of using a wire deposited with nanopartiles, while no significant difference between base fluid and nanofluid with a virgin wire was observed

  7. Nucleate boiling pressure drop in an annulus: Book 3

    International Nuclear Information System (INIS)

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D2O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ''power tilt'' or asymmetric heating of the inner and outer annulus walls. The test facility uses H2O rather than D2O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists of data plots and summary files of temperature measurements

  8. Nucleate boiling pressure drop in an annulus: Book 6

    Energy Technology Data Exchange (ETDEWEB)

    1992-11-01

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists of a summary of temperature measurements to include recorded minima, maxima, averages and standard deviations.

  9. Nucleate boiling pressure drop in an annulus: Book 7

    Energy Technology Data Exchange (ETDEWEB)

    1992-11-01

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists solely of tables of temperature measurements; minima, maxima, averages and standard deviations being measured.

  10. Nucleate boiling pressure drop in an annulus: Book 4

    Energy Technology Data Exchange (ETDEWEB)

    Block, J.A.; Crowley, C.; Dolan, F.X.; Sam, R.G.; Stoedefalke, B.H.

    1992-11-01

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists of data plots and summary files of temperature measurements.

  11. Nucleate boiling pressure drop in an annulus: Book 8

    Energy Technology Data Exchange (ETDEWEB)

    1992-11-01

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists of tables of temperature measurements.

  12. Nucleate boiling pressure drop in an annulus: Book 3

    Energy Technology Data Exchange (ETDEWEB)

    Block, J.A.; Crowley, C.; Dolan, F.X.; Sam, R.G.; Stoedefalke, B.H.

    1992-11-01

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D{sub 2}O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ``power tilt`` or asymmetric heating of the inner and outer annulus walls. The test facility uses H{sub 2}O rather than D{sub 2}O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists of data plots and summary files of temperature measurements.

  13. Boiling water reactor stability analysis in the time domain

    International Nuclear Information System (INIS)

    Boiling water nuclear reactors may experience density wave instabilities. These instabilities cause the density, and consequently the mass flow rate, to oscillate in the shrouded fuel bundles. This effect causes the nuclear power generation to oscillate due to the tight coupling of flow to power, especially under gravity-driven circulation. In order to predict the amplitude of the power oscillation, a time domain transient analysis tool may be employed. The modeling tool must have sufficient hydrodynamic detail to model natural circulation in two-phase flow as well as the coupled nuclear feedback. TRAC/BF1 is a modeling code with such capabilities. A dynamic system model has been developed for a typical boiling water reactor. Using this tool it has been demonstrated that density waxes may be modeled in this fashion and that their resultant hydrodynamic and nuclear behavior correspond well to simple theory. Several cases have been analyzed using this model, the goal being to determine the coupling between the channel hydrodynamics and the nuclear power. From that study it has been concluded that two-phase friction controls the extent of the oscillation and that the existing conventional methodologies of implementing two-phase friction into analysis codes of this type can lead to significant deviation in results from case to case. It has also been determined that higher dimensional nuclear feedback models reduce the extent of the oscillation. It has also been confirmed from a nonlinear dynamic standpoint that the birth of this oscillation may be described as a Hopf Bifurcation

  14. Nucleate boiling pressure drop in an annulus: Book 2

    International Nuclear Information System (INIS)

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D2O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ''power tilt'' or asymmetric heating of the inner and outer annulus walls. The test facility uses H2O rather than D2O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. Nineteen test series and a total of 178 tests were performed. Testing addressed the effects of: Heat flux; pressure; helium gas; power tilt; ribs; asymmetric heat flux

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

    International Nuclear Information System (INIS)

    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

  16. Nucleate boiling pressure drop in an annulus: Book 4

    International Nuclear Information System (INIS)

    The application of the work described in this report is the production reactors at the Savannah River Site, and the context is nuclear reactor safety. The Loss of Coolant Accident (LOCA) scenario considered involves a double-ended break of a primary coolant pipe in the reactor. During a LOCA, the flow through portions of the reactor may reverse direction or be greatly reduced, depending upon the location of the break. The reduced flow rate of coolant (D2O) through the fuel assembly channels of the reactor -- downflow in this situation -- can lead to boiling and to the potential for flow instabilities which may cause some of the fuel assembly channels to overheat and melt. That situation is to be avoided. The experimental approach is to provide a test annulus which simulates geometry, materials, and flow conditions in a Mark-22 fuel assembly (Coolant Channel 3) to the extent possible. The annulus has a full-scale geometry, and in fat uses SRL dummy hardware for the inner annulus wall in the ribbed geometry. The materials aluminum. The annulus is uniformly heated in the axial direction, but the circumferential heat flux can be varied to provide ''power tilt'' or asymmetric heating of the inner and outer annulus walls. The test facility uses H2O rather than D2O, but it includes the effects of dissolved helium gas present in the reactor. The key analysis approaches are: To compare the minima in the measured demand curves with analytical criteria, in particular the Saha-Zuber (1974) model; and to compare the pressure and temperature as a function of length in the annulus with an integral model for flow boiling in a heated channel. This document consists of data plots and summary files of temperature measurements

  17. Concrete peeling off device

    International Nuclear Information System (INIS)

    The present invention concerns a device for peeling off activated concretes in processing for discarding a reactor of a nuclear reactor facility. The device comprises a gyrotron for generation microwaves, an irradiator for irradiating output microwaves, a reflection mirror for reflecting and converging the microwaves and irradiating them to a material to be irradiated and a first rotating means for rotating the irradiator and the reflection mirror in parallel with the axis of the gyrotron while maintaining the positional relation between the irradiator and the reflection mirror. When the position of the microwaves irradiated on concrete walls are moved in a circumferential direction and the central axes of the rotational axis and the material to be irradiated are aligned, then the intensity of the irradiation of the microwaves at each of the irradiation points can be maintained constant without changing the focal distance of the reflected microwaves thereby enabling to peel off concretes efficiently. If operation conditions are controlled based on information such as temperature at the periphery of the microwave irradiation positions, the shape and the color of the material to be irradiated and the distance to the material to be irradiated, a concrete peeling off device of high reliability can be obtained. (N.H.)

  18. Model for calculation of the critical heat fluxes when boiling in a swirling subcooled flow under nonuniform heating

    International Nuclear Information System (INIS)

    Mechanism of heat transfer crisis under boiling in a subcooled flow differs essentially from the crisis classical concepts. The paper describes a physical model of heat transfer crisis under boiling in a highly subcooled twisted flow at the tube perimeter heterogeneous heating. Ratio for the critical heat flow at boiling in a subcooled twisted flow is obtained

  19. 78 FR 35990 - All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos...

    Science.gov (United States)

    2013-06-14

    ... COMMISSION All Operating Boiling-Water Reactor Licensees With Mark I And Mark II Containments; Docket Nos... Licensees operate boiling-water reactors (BWRs) with Mark I and Mark II containment designs. II. The events... Boiling Water Reactors with Mark I and Mark II Containments'' (November 26, 2012). Option 2 in...

  20. 75 FR 26967 - Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water...

    Science.gov (United States)

    2010-05-13

    ... Areas Subject to a Boil-Water Advisory; Availability AGENCY: Food and Drug Administration, HHS. ACTION... entitled ``Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water Advisory.'' This guidance is intended to advise food manufacturers that once a boil-water advisory has...

  1. Steady State Film Boiling Heat Transfer Simulated With Trace V4.160

    International Nuclear Information System (INIS)

    This paper presents the results of the assessment and analysis of TRACE v4.160 heat transfer predictions in the post-CHF (critical heat flux) region and discusses the possibilities to improve the TRACE v4.160 code predictions in the film boiling heat transfer when applying different film boiling correlations. For this purpose, the TRACE v4.160-calculated film boiling heat flux and the resulting maximum inner wall temperatures during film boiling in single tubes were compared with experimental data obtained at the Royal Institute of Technology (KTH) in Stockholm, Sweden. The experimental database included measurements for pressures ranging from 30 to 200 bar and coolant mass fluxes from 500 to 3000 kg/m2s. It was found that TRACE v4.160 does not produce correct predictions of the film boiling heat flux, and consequently of the maximum inner wall temperature in the test section, under the wide range of conditions documented in the KTH experiments. In particular, it was found that the standard TRACE v4.160 under-predicts the film boiling heat transfer coefficient at low pressure-low mass flux and high pressure-high mass flux conditions. For most of the rest of the investigated range of parameters, TRACE v4.160 over-predicts the film boiling heat transfer coefficient, which can lead to non-conservative predictions in applications to nuclear power plant analyses. Since no satisfactory agreement with the experimental database was obtained with the standard TRACE v4.160 film boiling heat transfer correlations, we have added seven film boiling correlations to TRACE v4.160 in order to investigate the possibility to improve the code predictions for the conditions similar to the KTH tests. The film boiling correlations were selected among the most commonly used film boiling correlations found in the open literature, namely Groeneveld 5.7, Bishop (2 correlations), Tong, Konkov, Miropolskii and Groeneveld-Delorme correlations. The only correlation among the investigated, which

  2. Modeling the Rapid Boil-Off of a Cryogenic Liquid When Injected into a Low Pressure Cavity

    Science.gov (United States)

    Lira, Eric

    2016-01-01

    Many launch vehicle cryogenic applications require the modeling of injecting a cryogenic liquid into a low pressure cavity. The difficulty of such analyses lies in accurately predicting the heat transfer coefficient between the cold liquid and a warm wall in a low pressure environment. The heat transfer coefficient and the behavior of the liquid is highly dependent on the mass flow rate into the cavity, the cavity wall temperature and the cavity volume. Testing was performed to correlate the modeling performed using Thermal Desktop and Sinda Fluint Thermal and Fluids Analysis Software. This presentation shall describe a methodology to model the cryogenic process using Sinda Fluint, a description of the cryogenic test set up, a description of the test procedure and how the model was correlated to match the test results.

  3. Energy consumption, destruction of exergy and boil off during the process of liquefaction, transport and regasification of liquefied natural gas

    Energy Technology Data Exchange (ETDEWEB)

    Stradioto, Diogo Angelo; Schneider, Paulo Smith [Dept. of Mechanical Engineering. Universidade Federal do Rio Grande do Sul, Porto Alegre (Brazil)], e-mail: pss@mecanica.ufrgs.br

    2010-07-01

    A supply chain of Liquefied Natural Gas (LNG) is composed by several processes like extraction, purification, liquefaction, storage, transport, regasification and distribution. In all these stages, processes need of energy. The main objective of this work is to quantify the energy consumption, mass loss and exergy destruction occurred throughout the chain. Results show that the process of liquefaction is the largest consumer of energy. Storage and transport by ship are responsible for the bigger mass losses and regasification is the process of larger destruction of exergy. A case study is performed considering a stream of pure methane at the input of a liquefaction plant, and evaluates energy along the chain, ending up at the distribution of NG after its regasification. (author)

  4. Development of the boil off lithium vapor source (BOLVAPS) for a PBFA-II lithium ion source

    International Nuclear Information System (INIS)

    A thin layer of dense lithium plasma may be used for the ion source in intense ion beam diodes such as the one being developed for inertial confinement fusion experiments with the PBFA-ll accelerator. Production of a thin vapor layer might be the first step in providing the plasma. Production of a lithium vapor layer by rapid evaporation and its ionization have been described. In these experiments, lithium was evaporated, by rapid (several μs) ohmic heating, from a 1 μm thick layer of LiAg alloy deposited on a 25 μm thick Ta foil. The vapor density was limited to ≅ 1 x 10/sup 16/ cm/sup -3/ in these experiments by termination of the heating pulse which arced through the evolving vapor. The thickness of the layer (a few mm) was determined by the heating pulse duration. A somewhat denser and thinner plasma layer is desired for the PBFA-II applied-B diode. To use this technique in a diode, the heater and lithium layer must both be thin-films in a low inductance configuration. The authors are now performing an experiment using a submicrosecond duration fast heating pulse to test lithium evaporation from a coaxial structure composed of a molybdenum core, an alumina-titania insulator overcoated with a 10 μm thick tungsten heater layer, and 1 μm thick LiAg lithium reservoir. The geometry of the electrical circuit is topologically the same as one required for the PBFA-II diode. They are also designing a BOLVAPS source for use with the PBFA-II diode. Progress in the fast-pulsed experiment and the design for PBFA-II are presented

  5. Validation of RETRAN-03 by simulating a Peach Bottom turbine trip and boil-off at the FIST facility

    International Nuclear Information System (INIS)

    The RETRAN-03 computer code was validated by simulating two tests that were performed at the Full Integral Simulation Test (FIST) facility. RETRAN-03 results of a Peach Bottom turbine trip (4PTT1) and failure to maintain water level at decay power (T1QUV) are compared with the FIST test data. Sensitivity to various model nodalizations and RETRAN-03 slip options were studied by comparing results of FIST test T1QUV. Core uncovery time is sensitive to the upper downcomer and upper plenum nodalization. The RETRAN-03 analysis of FIST test 4PTT1 was compared to a previous TRAC-BWR analysis of the test

  6. A look-up table for fully developed film-boiling heat transfer

    International Nuclear Information System (INIS)

    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

  7. An assessment of boiling as a method of household water treatment in South India.

    Science.gov (United States)

    Juran, Luke; MacDonald, Morgan C

    2014-12-01

    This article scrutinizes the boiling of water in Tamil Nadu and Puducherry, India. Boiling, as it is commonly practiced, improves water quality, but its full potential is not being realized. Thus, the objective is to refine the method in practice, promote acceptability, and foster the scalability of boiling and household water treatment (HWT) writ large. The study is based on bacteriological samples from 300 households and 80 public standposts, 14 focus group discussions (FGDs), and 74 household interviews. Collectively, the data fashion both an empirical and ethnographic understanding of boiling. The rate and efficacy of boiling, barriers to and caveats of its adoption, and recommendations for augmenting its practice are detailed. While boiling is scientifically proven to eliminate bacteria, data demonstrate that pragmatics inhibit their total destruction. Furthermore, data and the literature indicate that a range of cultural, economic, and ancillary health factors challenge the uptake of boiling. Fieldwork and resultant knowledge arrive at strategies for overcoming these impediments. The article concludes with recommendations for selecting, introducing, and scaling up HWT mechanisms. A place-based approach that can be sustained over the long-term is espoused, and prolonged exposure by the interveners coupled with meaningful participation of the target population is essential. PMID:25473989

  8. Experiments on HFE-7100 pool boiling at atmospheric pressure in horizontal narrow spaces

    Energy Technology Data Exchange (ETDEWEB)

    Guglielmini, G.; Misale, M.; Priarone, A. [Universita degli Studi di Genova (Italy). DIPTEM - Sezione di Termoenergetica e Condizionamento Ambientale

    2009-07-01

    Experiments were performed to examine the pool boiling heat transfer and critical heat flux on a smooth copper circular surface, confined by a face-to-face parallel unheated surface, by changing the gap between the surfaces and the unheated surface diameter. Pool boiling data at atmospheric pressure were obtained for saturated HFE-7100. The gap values investigated, between the boiling surface and the adiabatic one, were s 0.5, 1.0, 2.0, 3.5 mm. To confine the boiling surface, two different Plexiglas plates were used: the former characterised by a diameter D = 60 mm, large as the overall test section support, the latter characterised by a diameter D = 30 mm, large to cover only copper boiling surface (d = 30 mm). For each configuration, boiling curves were obtained up to the thermal crisis. For both different types of confinement, it was observed that the boiling curves match at low wall superheat, except for s = 0.5 mm, 1 mm. However, at high wall superheat, a drastic reduction in heat transfer as well as CHF appears decreasing the channel width s; for all gap sizes, this reduction is less pronounced for the smaller confinement wall (D = 30 mm). Instead, at low wall superheat for gap of 0.5 and 1.0 mm, the heat transfer coefficient is higher for diameter disc of 60 mm. CHF data were also compared with a literature correlation (Misale and al., 2009). (author)

  9. Effects of rolling motion on thermal–hydraulic characteristics of boiling flow in rectangular narrow channel

    International Nuclear Information System (INIS)

    Highlights: • Pressure drop fluctuation is enhanced with increasing rolling amplitude and period. • The phase difference between flow rate and pressure drop fluctuation is 1/4 period. • Amplitude of boiling heat transfer coefficient increases with increasing heat flux. • The curve of boiling heat transfer coefficient fluctuations is close to sine curve. - Abstract: Experimental investigations on thermal–hydraulic characteristics of boiling flow in a rectangular narrow channel under rolling motion conditions are carried out. This experiment is designed to elucidate the phenomena of boiling flow under rolling motion and to give the corresponding rational explanations. The results show that the amplitudes of fluctuations of pressure drop, flow rate, fluid and wall temperatures, and boiling heat transfer coefficient increase with the increasing of rolling amplitude and rolling period. The phase difference of flow rate fluctuation and pressure drop fluctuation is 1/4 period, and the saturated water temperature fluctuations of the test section delay 2–3 s behind the pressure drop fluctuations. The time average boiling heat transfer coefficients of the rectangular narrow channel under rolling motion are equal to those under static conditions. The amplitude of boiling heat transfer coefficient of test section increases with increasing heat flux and flow rate, while decreases with increasing system pressure

  10. Experimental investigation on partial pool boiling heat transfer in pure liquids

    Directory of Open Access Journals (Sweden)

    Fazel Seyed Ali Alavi

    2016-01-01

    Full Text Available Saturated partial pool boiling heat transfer has been experimentally investigated on a horizontal rod heater. The boiling liquids are including water and ethanol. The heating section is made by various materials including SS316, copper, aluminum and brass. Experiments have been performed at several degrees of surface roughness ranging between 30 and 360 micrometer average vertical deviation. The measurements are including boiling heat transfer coefficient, bubble departing diameter and frequency and also nucleation site density. The data have been compared to major existing correlations. It is shown that experimental data do not match with major correlations at the entire range of experiments with acceptable accuracy. In this article, the boiling heat transfer area has been divided in two complementary areas, the induced forced convection area and the boiling affected area. Based on two dimensionless groups, including Eötvös and Roshko numbers, a semi-empirical model have been proposed to predict the boiling heat transfer coefficient. It is shown that the proposed model provides improved performance in prediction of the boiling heat transfer coefficient in comparison with to existing correlations.

  11. Semi-empirical modeling of pool boiling heat transfer in binary mixtures

    International Nuclear Information System (INIS)

    Highlights: • The boiling heat transfer coefficient of mixtures are less than those of ideal. • Evaporation of the volatile component increases the V–L interfacial temperature. • The transition q/A from free convection to boiling is about 20 kW per square meter. -- Abstract: Pool boiling heat transfer has been investigated for various binary mixtures, including acetone/isopropanol, water/acetone, water/methanol, water/ethanol, water/isopropanol, water/monoethanolamine, water/diethanolamine and water/triethyleneglycol as test solutions. Many correlations have been developed to predict the pool boiling heat transfer coefficient in mixtures in the past few decades, however the predicted values are not confirming. In addition, the application of many existing correlations requires some individual adjusting parameters that may be not available for every system. In this investigation, a new set of experimental data are presented. These data have been compared to major existing correlations. It is observed that the pool boiling heat transfer coefficients in mixtures are less than the ideal boiling heat transfer coefficient. A new semi-empirical model has been proposed based on the mass transfer resistance to predict the boiling heat transfer coefficient with satisfactory accuracy. The new model does not include any tuning parameter and is applicable to any given binary system. The performance of the proposed model is superior to most existing correlations

  12. Cooling of hot bubbles by surface texture during the boiling crisis

    Science.gov (United States)

    Dhillon, Navdeep; Buongiorno, Jacopo; Varanasi, Kripa

    2015-11-01

    We report the existence of maxima in critical heat flux (CHF) enhancement for pool boiling on textured hydrophilic surfaces and reveal the interaction mechanism between bubbles and surface texture that governs the boiling crisis phenomenon. Boiling is a process of fundamental importance in many engineering and industrial applications but the maximum heat flux that can be absorbed by the boiling liquid (or CHF) is limited by the boiling crisis. Enhancing the CHF of industrial boilers by surface texturing can lead to substantial energy savings and reduction in greenhouse gas emissions on a global scale. However, the fundamental mechanisms behind this enhancement are not well understood, with some previous studies indicating that CHF should increase monotonically with increasing texture density. However, using pool boiling experiments on a parametrically designed set of plain and nano-textured micropillar surfaces, we show that there is an optimum intermediate texture density that maximizes CHF and further that the length scale of this texture is of fundamental significance. Using imbibition experiments and high-speed optical and infrared imaging, we reveal the fundamental mechanisms governing the CHF enhancement maxima in boiling crisis. We acknowledge funding from the Chevron corporation.

  13. Boiling heat transfer phenomenon base on the event of loca and severe accident

    International Nuclear Information System (INIS)

    Research and development base on TMI-2 NPP accident mostly directed to vessel and core performance. The majority of research was conducted which aimed on boiling heat transfer phenomenon, begin by loss of coolant accident (LOCA) until severe accident, in which core meltdown. Study on boiling heat transfer has been done by simulation on core bottom re-flooding process and a narrow gap cooling. The results of experimental research which was conducted by BATAN concerning LOCA and severe accident are giving a clearly picture, in how boiling heat transfer phenomenon was occurs during sequent of nuclear reactors accident, especially TMI-2 accident. The mapping of heat transfer base on transient temperature data was created in boiling curve form which was shown the differences of heat flux in three boiling regimes, both in pool boiling and flow boiling. The experimental simulation of LOCA shown that the CHF value (67.31 kW/m2 ) is small than the CHF value of severe accident (262 kW/m2). (author)

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

    International Nuclear Information System (INIS)

    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/(m2·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)

  15. Nucleate pool-boiling heat transfer - II. Assessment of prediction methods

    International Nuclear Information System (INIS)

    Part I of this paper has identified all significant boiling surface parameters affecting nucleate pool-boiling heat transfer and has investigated their parametric trends, thus providing a measure of the state of the art in this area. This part of the paper examines the existing prediction methods for the heat transfer coefficient (HTC) under this boiling regime. Six heat transfer pool-boiling correlations that are well known in the literature have been selected and their prediction accuracy has been assessed against available and well-documented experimental databases. These databases provide HTCs obtained: (i) under pool-boiling conditions of fluids such as water, ethanol, R-113, and n-heptane; and (ii) on the following large-size horizontal surfaces: thick plates (made of copper, aluminum, brass, and stainless steel), and a horizontal circular disk (plated with a thin layer of polished chromium). For completeness, the microgeometry characteristics of several boiling surfaces are included here, even though they are not fully utilized in the present analysis. The surface microgeometry has been characterized by 14 roughness parameters measured with a laser profilometer. The analysis concludes that within the investigated ranges of boiling conditions, working fluids and boiling surfaces, the Rohsenow and Pioro nucleate pool-boiling correlations are the most accurate among those assessed. The Rohsenow and Pioro correlations use constants and powers for non-dimensional numbers that correspond to a specific surface-fluid combination, as opposed to the other correlations that use fixed values regardless of the surface-fluid combination. (author)

  16. On-site staffing requirements for a simplified boiling water reactor (SBWR)

    International Nuclear Information System (INIS)

    In 1992 the total generating costs were estimated by EPRI for a baseload, nth-of-a-kind advanced reactor with the following cost distribution: capital cost 62%, operation and maintenance (O and M) cost 20%, fuel cost 16%, and decommissioning cost 2%. Thus the O and M cost is a significant component of the total cost of electricity, second only to the capital cost. The O and M cost in turn can be split into: cost for on-site staff, maintenance materials, supplies and expenses, off-site technical support, regulatory fees, insurance premiums and administration. The costs for on-site staff is about 30% of the total O and M cost. In 1992, the US Council for Energy Awareness (USCEA) estimated the on-site staffing for a typical 600 MWe advanced reactor to be about 330 with 25 (full time equivalent, FTE) contractors. This estimate was reevaluated by EPRI, and the staffing was modified based on a reengineering of the organizational structure that eliminated unnecessary layers of vertical management. As a result of this review, the on-site staffing was decreased to 259 with 25 (FTE) contractors, for a total of 284 people. The Dodewaard power plant (GKN) in The Netherlands is a 60 MWe facility with a natural circulating reactor. Since the 600 MWe Simplified Boiling Water Reactor (SBWR), an advanced reactor, also utilizes a natural circulating reactor with other passive safety features it was desired to extrapolate the GKN staffing to the SBWR. Also, some of the European O and M practices that utilize fewer skilled labor are reflected. This paper provides the results of the comparison between the EPRI recommendations and the staffing based on GKN experience

  17. The installation of PEANO at the Halden Boiling Water Reactor: first test and results

    International Nuclear Information System (INIS)

    After extensive testing of PEANO with data from process simulators, the next step was to set-up an installation in a real process, where the signal validation is performed online. For this purpose the Halden Boiling Water Reactor was used. One implication is that recorded process data from past operation would be used for the training of the system. This type of data is corrupted with errors, faults, process noise or even previous sensor problems, which need to be removed before the data can be used. The pre-processing was therefore a very import step during this installation. A 15 minute average of 29 process signals, spread out over the primary, secondary and tertiary loops, was used. At the end of the design process a fuzzy-neural network resulted containing 5 clusters, that has been trained with over 20.000 patterns. To establish the TCP/IP connection to the process computer and receiving the process data in real-time, some extra software was developed. With this installation it has been shown that it is possible to have the PEANO Server and PEANO Client (monitoring unit) running on one machine (e.g. in the control room), while additional monitoring units are connected from a remote location (e.g. main office building). The first results show that the installation of PEANO is capable of performing its validation task properly, even during transients. Both start-up and shutdown situations can be handled without any problems. In situations where incoming patterns represent unknown process situations that have not been encountered during the training, the 'I don't know' answer was given. To test the ability to detect a sensor failure off-line tests have been run, where sensor faults and drifts were added (author) (ml)

  18. Experimental investigation of the formation of critical boiling nuclei in superheated sodium

    International Nuclear Information System (INIS)

    This work deals with the problem of the growth of boiling nuclei up to critical size in superheated sodium. Contrary to previous investigations the sodium was superheated to a certain degree without initiating the onset of boiling. The lifetime of the superheated state - defined as the waiting time - and thereby the time dependence of the growth of boiling nuclei was measured. The experiments were performed with stagnant sodium in an almost gas free working apparatus. It was not aspired to simulate real reactor conditions. Sodium temperature and superheat were up to 9600C and 3600C respectively. The purity of the sodium was varied by cold trap purification. (orig./TK)

  19. Criticality in the slowed-down boiling crisis at zero gravity

    OpenAIRE

    Charignon, Thomas; Lloveras Muntané, Pol Marcel; CHATAIN, Denis; Truskinovsky, Lev; Vives, Eduard; Beysens, Daniel; Nikolayev, Vadim

    2015-01-01

    Boiling crisis is a transition between nucleate and film boiling. It occurs at a threshold value of the heat flux from the heater called CHF (critical heat flux). Usually, boiling crisis studies are hindered by the high CHF and short transition duration (below 1 ms). Here we report on experiments in hydrogen near its liquid-vapor critical point, in which the CHF is low and the dynamics slow enough to be resolved. As under such conditions the surface tension is very small, the experiments are ...

  20. Pervaporation investigation of recovery of volatile compounds from brown crab boiling juice.

    Science.gov (United States)

    Martínez, Rodrigo; Sanz, M Teresa; Beltrán, Sagrario

    2014-10-01

    Pervaporation has been used to obtain aroma concentrates from brown crab boiling juice. The boiling juice and the obtained permeate have been analysed by Headspace Solid Phase Dynamic Extraction Gas Chromatography/Mass Spectrometry. The effect of feed temperature on the pervaporation performance of the membrane has been analysed. The permeate aroma profile, at 25 ℃ and 40 ℃, was different from that of the boiling juice. Enrichment factors for some of the volatile compounds were much lower than those obtained in model aqueous dilute solutions. Pervaporation performance can be significantly improved by modifying the permeant circuit to include two condensation stages. PMID:23897977

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

    International Nuclear Information System (INIS)

    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

  2. Diffusive and radiative effects on vaporization times of drops in film boiling

    Science.gov (United States)

    Baumeister, K. J.; Choessow, G. J.

    1972-01-01

    Diffusive and radiative effects are incorporated into an analysis for the vaporization time of drops in film boiling. The momentum, energy, and continuity equations are solved with some appropriate simplifications so as to obtain a simple closed form solution for the overall film boiling heat transfer coefficient. Next, a theoretical expression for the droplet vaporization time is developed and compared to the measured vaporization times of water droplets vaporizing into air, argon, nitrogen, and helium. The agreement between experiment and theory is good. Under the helium blanket, the diffusive evaporative component is significant in comparison to the film boiling component.

  3. Experimental Study on Convective Boiling Heat Transfer in Vertical Narrow Gap Annular Tube

    Institute of Scientific and Technical Information of China (English)

    Li Bin; He Anding; Wang Yueshe; Zhou Fangde

    2001-01-01

    Experiments are conducted to investigate the characteristics of single-phase forced-flow convection and boiling heat transfer of R113 flowing through annular tube with gap of 1, 1.5 and 2.5 mm, and also the visualization test are carried out to get two-phase flow regime. The data show that the Nusselt numbers for the narrow-gap are higher than those predicted by traditional large channel correlation and boiling heat transfer is enhanced. Based on the data obtained in this investigation, correlations for single-phase, forced convection and flow boiling in annular tube of different gap size has been developed.

  4. Comprehensive Evaluation and Prediction of Enhancement of Boiling Heat Transfer with Additives

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    A model of evaluation and prediction of enhancement of boiling heat transfer with additives has been propoeed according to fuzzy fundamentals. Correlative appraisement of boiling heat transfer augmentation was done with the model based on 39 additives which were tested by the authors and other researchers. The results show that the evaluation of 35 additives is consistent with experiments, which means that the accuracy of the model is 89.7 percent. In addition, the prediction of the ability of boiling heat transfer enhancement with sodium oleate,polyethylene glycol and Tween-40 is also in good agreement with correspondent experiments.

  5. Early detection of nucleate boiling and spectral analysis of acoustical noise

    International Nuclear Information System (INIS)

    The development of a reliable detection technique for the onset of boiling has been further pursued. Besides the already studied tube geometry, a more realistic annular set up has been used where a fuel pin model, electrically heated, is placed. Using accelerometers on the pin, on the structure and on specific instrumentation cables the onset of boiling was clearly monitored by the emergence of a typical resonance frequency. The influence of pressure and heat power was analysed in detail. Furthermore, a glass model has been constructed in order to better correlate the observed measurement with the boiling parameters, as bubble radius, frequency and collapse mode

  6. Pressure measurements in boiling particle beds with water at 1 bar

    International Nuclear Information System (INIS)

    Pressures have been measured at the top and bottom of uniformly heated beds of uniform spherical particles with water boiling at atmospheric pressure. Particle sizes used vary from 0.22 to 5 mm diameter and bed heights from 50 to 150 mm. The pressures have been recorded at power levels up to dry-out. The results show how much liquid remains in a boiling bed at different power levels and how the liquid/vapour phase pressure losses vary. The results give a valuable insight into the working of a boiling bed. (author)

  7. Hydrodynamics and heat transfer in a sodium boiling flow. Application to LMFBR safety analyses

    International Nuclear Information System (INIS)

    Experimental and theoretical results in the field of sodium boiling flows are presented. Application to LMFBR safety analyses are presented. The emphasis is mainly focused on thermohydraulic consequencies of sodium boiling. One shows how general knowledge of two phase flow applies to sodium boiling; what is particular to liquid metals; what is specific to subassembly geometry; what happens in steady state; what happens in transient regime. The analysis is based on experimental evidence. Simplified analysis are proposed. Code interpretation are given (NATREX, MANDRIN, BACCHUS)

  8. Boiled coffee does not increase serum cholesterol in gerbils and hamsters.

    OpenAIRE

    Mensink, R.P.; Zock, P. L.; Katan, M B; A. C. Beynen

    1992-01-01

    In contrast to drip filter coffee, boiled coffee increases the serum cholesterol level in man. To identify the substance(s) responsible for this effect, it is necessary to find an animal model sensitive to boiled coffee. In this study, three groups of 20 male gerbils and three groups of six male hamsters were fed a control diet or a control diet supplemented with either freeze-dried boiled coffee or freeze-dried filtered coffee. At the end of the 5-week feeding period serum cholesterol levels...

  9. Kandlikar third number map for flow boiling in micro-channels and micro-gravity

    Directory of Open Access Journals (Sweden)

    Awad M.M.

    2015-01-01

    Full Text Available As an extension of the recent work of Baldassari and Marengo (Baldassari C., Marengo M., Flow Boiling in Microchannels and Microgravity, Progress in Energy and Combustion Science 39 (2013 1, pp. 1-36, this note presents Kandlikar third number (K3 map for flow boiling in microchannels and microgravity. Using several data points available in the literature, Kandlikar third number (K3 map was plotted versus the hydraulic diameter (dh as the characteristic dimension for flow boiling in microchannels and microgravity. The ranges of the Kandlikar third number (K3, calculated using the hydraulic diameter (dh, are presented.

  10. Bubble departure in pool and flow boiling systems: A review and latest developments

    International Nuclear Information System (INIS)

    Many of the vapor bubble departure diameter correlations for pool and flow boiling which have been proposed in the open literature are reviewed. In addition, the recent unified bubble detachment model for pool and flow boiling proposed by Zeng et al. (1992a, 1992b) is discussed. It is demonstrated that the unified model, which requires the vapor bubble growth rate as an input, is the only one which satisfactorily predicts vapor bubble departure diameters over the entire range of boiling conditions for which bubble detachment data exist

  11. An investigation into the influence of different parameters on the onset of boiling in minichannels

    OpenAIRE

    Piasecka Magdalena

    2013-01-01

    The paper presents experimental studies on boiling heat transfer in rectangular minichannels. The investigations focus on the transition from single phase forced convection to nucleate boiling, i.e., in the zone of boiling incipience. The experiment has been carried out with FC-72, R-123 and R-11 at the Reynolds number below 4700, corresponding to mass flow rate range 95-710 kg/(m s). The main part of the test section is a minichannel of pre-set depth from 0.7 to 2 mm and width (20, 40 and 60...

  12. Nuclear reactor noise investigations on boiling effects in a simulated MTR-type fuel assembly

    International Nuclear Information System (INIS)

    The work includes validation/testing of existing neutron noise methods under well-controlled circumstances, investigation of boiling phenomena in narrow channels, and development of a novel boiling monitoring method. The work has been performed in the NIOBE facility at the HDR. Noise signals of thermocouples in the channel wall are used for velocity profile monitoring. Flow patterns in the boiling coolant are identified by means of analysis of probaof probability density functions and neutron noise spectra. Local noise effects are studied. (DG)

  13. Stability estimation of a large-sized pool boiling superconductor. Dependence on surface orientation and area fraction of surface treatment

    International Nuclear Information System (INIS)

    For the purpose of the stability analysis of a pool boiling superconductor, many studies on the heat transfer of LHe have been conducted. There are a number of variables that affect heat transfer. It is well-known that to variables are surface orientation and surface treatment. Surface orientation of a superconductor is varied by magnet winding. The change is associated with the variation of gravitational force on the surface, thus causing heat transfer characteristics to change. Usually, the surface of a superconductor is treated to improve heat-transfer characteristics; for example, oxidation. During the winding process, the winding machine may strip off the treatment at some locations. The resulting damage may change the heat-transfer characteristics and degrade the stability of the superconducting magnet. In this study, the heat transfer of polished Cu, oxidized Cu and partially oxidized surfaces were measured as a function of orientation. The critical and minimum heat fluxes depend on the area fraction of oxidation. The calculation method for the heat flux of a partially oxidized surface was established. Heat-transfer measurements of a prototype superconductor with polished surfaces were also conducted to change the surface orientation. The heat-transfer characteristics with the prototype superconductor were degraded as compared with those of the Cu surface. It became clear that heat transfer for a stability analysis must be measured using a prototype superconductor. Next, the recovery current for a large-sized pool boiling superconductor, which is a helical coil superconductor, was calculated according to the experimental results of heat transfer. The dependence of the recovery current on surface orientation and the area fraction of surface treatment was estimated. The calculated results were compared with the measured recovery currents of the short sample. The calculated recovery current agreed well with the experimental result. (author)

  14. 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

  15. Some results of experimental investigations of boiling crisis at the low and negative coolant discharge

    International Nuclear Information System (INIS)

    The preliminary analysis of results received at experimental investigations of boiling crisis under conditions of the direct and reverse water flow rate through simplified model of fuel assembly is submitted

  16. Heat Transfer in Nucleate Pool Boiling of Binary and Ternary Refrigerant Mixtures

    Institute of Scientific and Technical Information of China (English)

    赵耀华; 刁彦华; 鹤田隆治; 西川日出男

    2004-01-01

    Heat transfer coefficients in nucleate pool boiling were measured on a horizontal copper surface for refrigerants, HFC-134a, HFC-32, and HFC-125, their binary and ternary mixtures under saturated conditions at 0.9MPa. Compared to pure components, both binary and ternary mixtures showed lower heat transfer coefficients.This deterioration was more pronounced as heat flux was increased. Experimental data were compared with some empirical and semi-empirical correlations available in literature. For binary mixture, the accuracy of the correlations varied considerably with mixtures and the heat flux. Experimental data for HFC-32/134a/125 were also compared with available correlated equation obtained by Thome. For ternary mixture, the boiling range of binary mixture composed by the pure fluids with the lowest and the medium boiling points, and their concentration difference had important effects on boiling heat transfer coefficients.

  17. Critical heat flux of an impinging water jet on a heated surface with boiling

    Energy Technology Data Exchange (ETDEWEB)

    Lee, J.S. [Andong Institute of Informaion Technology, Andong (Korea); Kim, H.D. [Andong National University, Andong (Korea); Choi, K.W. [Incheon University, Incheon (Korea)

    2000-04-01

    The purpose of this paper is to investigate a critical heat flux(CHF) during forced convective subcooled and saturated boiling in free water jet system impinged on a rectangular heated surface. The surface is supplied with subcooled or saturated water through a rectangular jet. Experimental parameters studied are a width of heated surface, a height of supplementary water and a degree of subcooling. Incipient boiling point is observed in the temperature of 6{approx}8 deg.C of superheat of test specimen. CHF depends on jet velocity for various boiling-involved coolant system. CHF also is proportional to the nozzle exit velocity to the power of n, where n is 0.55 and 0.8 for subcooled and saturated boiling, respectively. CHF is enhanced with a higher jet velocity, higher degree of subcooling and smaller width of a heated surface. (author). 18 refs., 13 figs., 1 tab.

  18. Experimental Investigation on Pool Boiling Heat Transfer With Ammonium Dodecyl Sulfate

    Directory of Open Access Journals (Sweden)

    Mr.P. Atcha Rao

    2015-11-01

    Full Text Available We have so many applications related to Pool Boiling. The Pool Boiling is mostly useful in arid areas to produce drinking water from impure water like sea water by distillation process. It is very difficult to distill the only water which having high surface tension. The surface tension is important factor to affect heat transfer enhancement in pool boiling. By reducing the surface tension we can increase the heat transfer rate in pool boiling. From so many years we are using surfactants domestically. It is proven previously by experiments that the addition of little amount of surfactant reduces the surface tension and increase the rate of heat transfer. There are different groups of surfactants. From those I‟m conducting experimentation with anionic surfactant Ammonium Dodecyl Sulfate (ADS, which is most human friendly and three times best soluble than Sodium Dodecyl Sulfate, to test the heat transfer enhancement.

  19. Boiling crisis as inhibition of bubble detachment by the vapor recoil force

    International Nuclear Information System (INIS)

    Boiling crisis is a transition between nucleate and film boiling. In this communication we present a physical model of the boiling crisis based on the vapor recoil effect. Our numerical simulations of the thermally controlled bubble growth at high heat fluxes show how the bubble begins to spread over the heater thus forming a germ for the vapor film. The vapor recoil force not only causes the vapor spreading, it also creates a strong adhesion to the heater that prevents the bubble departure, thus favoring the further bubble spreading. Near the liquid-gas critical point, the bubble growth is very slow and allows the kinetics of the bubble spreading to be observed. Since the surface tension is very small in this regime, only microgravity conditions can preserve a convex bubble shape. Under such conditions, we observed an increase of the apparent contact angle and spreading of the dry spot under the bubble, thus confirming our model of the boiling crisis. (authors)

  20. CFD Simulations and Experimental Verification on Nucleate Pool Boiling of Liquid Nitrogen

    Science.gov (United States)

    Xiaobin, Zhang; Wei, Xiong; Jianye, Chen; Yuchen, Wang; Tang, K.

    To explore the mechanism of nucleate pool boiling of cryogenic fluids, an experimental apparatus was built to conduct a visualization study and verify the CFD boiling model. Apart from the general measurements of the super-heat and heat flux, the influences of super-heat on bubble departure diameters were specially analyzed. Based on the observations, the whole nucleate boiling process from bubble formation to departure from the heated wall can be divided into three stages: low heat flux stage; transitional stage; fully developed nucleate boiling (FDNB) stage. CFD simulations with several existing correlations and the attained values from the experiments for the bubble diameter were finally conducted, and the results fitted well with the present experimental data.

  1. 77 FR 3009 - Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water Reactors

    Science.gov (United States)

    2012-01-20

    ...The U.S. Nuclear Regulatory Commission (NRC) is issuing for public comment a draft NUREG, NUREG-2104, Revision 0, ``Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Advanced Boiling Water...

  2. Heat transfer crises in oxygen nucleate boiling under attenuated gravitation conditions

    International Nuclear Information System (INIS)

    In some physical models of the heat transfer crisis in oxygen nucleate boiling the concepts on boiling micromechanism and its characteristics are used. To test the correctness of these concepts, experimental data are used on critical heat flux densities, departure radii and departure nucleate frequencies during oxygen boiling in the 6x103-7x105 Pa pressure range and relative accelerations (eta=0.01-1). Based on a crisis heat model the formulae are obtained containing different dependence of critical heat flux density on acceleration at high (qsub(cr) approximately etasup(0.4)) and low (qsub(cr) is practically independent of eta) pressures, which agrees with experimental data on oxygen boiling. The data are presented on critical thermal heads; the independence of ΔTsub(cr) of acceleration in the range of the regime parameters investigated is shown

  3. Heat transfer crises in oxygen nucleate boiling under attenuated gravitation conditions

    Energy Technology Data Exchange (ETDEWEB)

    Kirichenko, Yu.A.; Gladchenko, G.M.; Rusanov, K.V. (AN Ukrainskoj SSR, Kharkov. Fiziko-Tekhnicheskij Inst. Nizkikh Temperatur)

    1984-11-01

    In some physical models of the heat transfer crisis in oxygen nucleate boiling the concepts on boiling micromechanism and its characteristics are used. To test the correctness of these concepts, experimental data are used on critical heat flux densities, departure radii and departure nucleate frequencies during oxygen boiling in the 6x10/sup 3/-7x10/sup 5/ Pa pressure range and relative accelerations (eta=0.01-1). Based on a crisis heat model the formulae are obtained containing different dependence of critical heat flux density on acceleration at high (qsub(cr) approximately etasup(0.4)) and low (qsub(cr) is practically independent of eta) pressures, which agrees with experimental data on oxygen boiling. The data are presented on critical thermal heads; the independence of ..delta..Tsub(cr) of acceleration in the range of the regime parameters investigated is shown.

  4. Observation of high heat flux boiling structures in a horizontal pool by a total reflection technique

    International Nuclear Information System (INIS)

    The experiments were carried out for a horizontal pool boiling of saturated water using a transparent ITO heating surface. Details of boiling structure near the heated surface have been clearly observed by applying the total reflection and diagonal view techniques in a synchronized manner. Mechanisms for the bubble coalescence and dry area expansion processes were clearly identified. The base of the large massive bubble was mostly dry with some trapped liquid. The appearance of this large dry area at high heat flux close to CHF was basically resulted from the multiple steps of bubble coalescences which occur while the bubbles are growing, attached to the boiling surface not before they depart from the boiling surface. The thin liquid layer with distributed vapor stems was not observed under the large massive bubble. (author)

  5. Transient CHF enhancement of saturated pool boiling of water using a honeycomb porous media

    International Nuclear Information System (INIS)

    Several studies have been performed to make clear the transient boiling heat transfer during the exponential heat generation which is occurred in reactivity accident of a nuclear reactor. These researches have been focused on the mechanism of the phenomena mainly, not on the enhancement of the transient boiling heat transfer. In a previous study, we proposed a method of CHF enhancement under steady-state conditions using honeycomb porous plate. The CHF was shown experimentally to be enhanced to more than twice that of a plain surface using honeycomb porous plate. The enhancement is considered to result from the capillary supply of liquid onto the heated surface and the release of generated vapor through the channels. In the present paper, enhancement of the transient critical heat flux in pool boiling by the attachment of a honeycomb-structured porous plate on a heated wire is investigated experimentally using water under saturated boiling conditions. (author)

  6. LOGOS. HX: a core simulator for high conversion boiling water reactors

    International Nuclear Information System (INIS)

    A three-dimensional physics simulator 'LOGOS. HX' has been developed for the designing analysis of high conversion boiling water reactor (HCBWR) cores. Its functions, calculational methods, and verification results will briefly be discussed. (author)

  7. Sensory quality and appropriateness of raw and boiled Jerusalem artichoke tubers (Helianthus tuberosus L.)

    DEFF Research Database (Denmark)

    Bach, Vibe; Kidmose, Ulla; Thybo, Anette;

    2013-01-01

    BACKGROUND: The aim of the present study was to investigate the sensory attributes, dry matter and sugar content of five varieties of Jerusalem artichoke tubers and their relation to the appropriateness of the tubers for raw and boiled preparation. RESULTS: Sensory evaluation of raw and boiled...... Jerusalem artichoke tubers was performed by a trained sensory panel and a semi-trained consumer panel of 49 participants, who also evaluated the appropriateness of the tubers for raw and boiled preparation. The appropriateness of raw Jerusalem artichoke tubers was related to Jerusalem artichoke flavour......, green nut flavour, sweetness and colour intensity, whereas the appropriateness of boiled tubers was related to celeriac aroma, sweet aroma, sweetness and colour intensity. In both preparations the variety Dwarf stood out from the others by being the least appropriate tuber. CONCLUSION: A few sensory...

  8. Pool boiling of dielectric liquids on porous graphite and extended copper surfaces

    Science.gov (United States)

    Parker, Jack L.

    This work investigated pool boiling of the dielectric liquids HFE-7100 and FC-72 on plane copper and porous graphite and on copper surfaces with corner pins. The work investigated the effects of surface orientation and liquid subcooling and, for the copper surfaces with corner pins, the effect of surface roughness. In addition, investigations were made studying the heat transfer by natural convection and nucleate boiling, as well as the effects of liquid subcooling (up to 30 K) and surface inclination (0°--upward facing, to 180°--downward facing) on nucleate boiling heat transfer and Critical Heat Flux (CHF). The results are applicable to direct immersion cooling by nucleate boiling of high power computer chips dissipating 50 - 100 W/cm2 while maintaining the junction temperature for the chips below the recommended values (˜85 °C). Pool boiling experiments are performed with degassed HFE-7100 and FC-72 liquids using uniformly heated 10 x 10 mm porous graphite and copper surfaces with corner pins. The measured footprint temperatures and thermal power removed from the surfaces are used to construct the pool boiling curves and determine the critical heat flux and corresponding surface superheat. Results are compared with those obtained on plane copper of same heated footprint area. The obtained CHF values are also compared with those reported in the open literature for plane, micro-porous, and macro-structured surfaces. Digital photographs and video are obtained to help explain and interpret the results. For the first time, natural convection correlations for dielectric liquids on plane, porous, and copper with corner pins developed. These correlations are important to electronic cooling in the stand-by mode when the heat dissipation by the chips is only a few watts. Results show that the power removed by natural convection from surfaces with corner pins is 67% more than from plane Si and Cu surfaces at the same surface superheat. Using porous graphite and copper

  9. Experimental study on a new solar boiling water system with holistic track solar funnel concentrator

    International Nuclear Information System (INIS)

    A new solar boiling water system with conventional vacuum-tube solar collector as primary heater and the holistic solar funnel concentrator as secondary heater had been designed. In this paper, the system was measured out door and its performance was analyzed. The configuration and operation principle of the system are described. Variations of the boiled water yield, the temperature of the stove and the solar irradiance with local time have been measured. Main factors affecting the system performance have been analyzed. The experimental results indicate that the system produced large amount of boiled water. And the performance of the system has been found closely related to the solar radiance. When the solar radiance is above 600 W/m2, the boiled water yield rate of the system has reached 20 kg/h and its total energy efficiency has exceeded 40%.

  10. X ray observations of boiling sodium in a reflux-pool-boiler solar receiver

    Science.gov (United States)

    Moreno, J. B.; Stoker, G. C.; Thompson, K. R.

    1992-01-01

    X ray observations of boiling sodium in a 75-kW sub t reflux-pool-boiler solar receiver operating at up to 800 C were carried out. Both cinematographic and quantitative observations were made. From the cinematography, the pool free surface was observed before and during the start of boiling. During boiling, the free surface rose out of the field of view, and chaotic motion was observed. From the quantitative observations, void fraction in pencil-like probe volumes was inferred, using a linear array of detectors. Useful data were obtained from three of the eight probe volumes. Information from the other volumes was masked by scattered radiation. During boiling, time-averaged void fractions ranged from 0.6 to 0.8. During hot restarts, void fractions near unity occurred and persisted for up to 1/2 second.

  11. Boiling crisis as inhibition of bubble detachment by the vapor recoil force

    CERN Document Server

    Nikolayev, Vadim; Garrabos, Yves

    2016-01-01

    Boiling crisis is a transition between nucleate and film boiling. In this communication we present a physical model of the boiling crisis based on the vapor recoil effect. Our numerical simulations of the thermally controlled bubble growth at high heat fluxes show how the bubble begins to spread over the heater thus forming a germ for the vapor film. The vapor recoil force not only causes the vapor spreading, it also creates a strong adhesion to the heater that prevents the bubble departure, thus favoring the further bubble spreading. Near the liquid-gas critical point, the bubble growth is very slow and allows the kinetics of the bubble spreading to be observed. Since the surface tension is very small in this regime, only microgravity conditions can preserve a convex bubble shape. Under such conditions, we observed an increase of the apparent contact angle and spreading of the dry spot under the bubble, thus confirming our model of the boiling crisis.

  12. Stability analysis on natural circulation boiling water reactors

    International Nuclear Information System (INIS)

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

  13. Practical application of neutron noise analysis at boiling water reactors

    International Nuclear Information System (INIS)

    The present status in the development of neutron noise methods for diagnostic purposes at BWRs is assessed with respect to practical applications. Three items of interest are briefly reviewed. They are concerned with local phenomena found in neutron noise signals at the higher frequency ranges (above several Hertz). The detection of vibrating in-core instrument tubes and the impacting of fuel element boxes were a problem in which neutron noise analysis substantially contributed. The possibility of detecting bypass flow boiling from neutron noise signatures is a recently proposed concept. Most of the research efforts have been applied to the experimental determination of local characteristics of the two-phase flow which dominates the noise sources in a BWR. Steam velocity measurements in fuel bundles by neutron noise techniques and the derivation of semi-empirical data, e.g. void fraction, bundle power and inlet flow rate, and possibly flow pattern recognition are features for practical use. But there are still effects which are not yet completely understood and require further experimental and theoretical investigations. (Auth.)

  14. Multi-cycle boiling water reactor fuel cycle optimization

    Energy Technology Data Exchange (ETDEWEB)

    Ottinger, K.; Maldonado, G.I. [University of Tennessee, 311 Pasqua Engineering Building, Knoxville, TN 37996-2300 (United States)

    2013-07-01

    In this work a new computer code, BWROPT (Boiling Water Reactor Optimization), is presented. BWROPT uses the Parallel Simulated Annealing (PSA) algorithm to solve the out-of-core optimization problem coupled with an in-core optimization that determines the optimum fuel loading pattern. However it uses a Haling power profile for the depletion instead of optimizing the operating strategy. The result of this optimization is the optimum new fuel inventory and the core loading pattern for the first cycle considered in the optimization. Several changes were made to the optimization algorithm with respect to other nuclear fuel cycle optimization codes that use PSA. Instead of using constant sampling probabilities for the solution perturbation types throughout the optimization as is usually done in PSA optimizations the sampling probabilities are varied to get a better solution and/or decrease runtime. The new fuel types available for use can be sorted into an array based on any number of parameters so that each parameter can be incremented or decremented, which allows for more precise fuel type selection compared to random sampling. Also, the results are sorted by the new fuel inventory of the first cycle for ease of comparing alternative solutions. (authors)

  15. Calculation system for physical analysis of boiling water reactors

    International Nuclear Information System (INIS)

    Although Boiling Water Reactors generate a quarter of worldwide nuclear electricity, they have been only little studied in France. A certain interest now shows up for these reactors. So, the aim of the work presented here is to contribute to determine a core calculation methodology with CEA (Commissariat a l'Energie Atomique) codes. Vapour production in the reactor core involves great differences in technological options from pressurised water reactor. We analyse main physical phenomena for BWR and offer solutions taking them into account. BWR fuel assembly heterogeneity causes steep thermal flux gradients. The two dimensional collision probability method with exact boundary conditions makes possible to calculate accurately the flux in BWR fuel assemblies using the APOLLO-2 lattice code but induces a very long calculation time. So, we determine a new methodology based on a two-level flux calculation. Void fraction variations in assemblies involve big spectrum changes that we have to consider in core calculation. We suggest to use a void history parameter to generate cross-sections libraries for core calculation. The core calculation code has also to calculate the depletion of main isotopes concentrations. A core calculation associating neutronics and thermal-hydraulic codes lays stress on points we still have to study out. The most important of them is to take into account the control blade in the different calculation stages. (author)

  16. Aging study of boiling water reactor high pressure injection systems

    International Nuclear Information System (INIS)

    The purpose of high pressure injection systems is to maintain an adequate coolant level in reactor pressure vessels, so that the fuel cladding temperature does not exceed 1,200 degrees C (2,200 degrees F), and to permit plant shutdown during a variety of design basis loss-of-coolant accidents. This report presents the results of a study on aging performed for high pressure injection systems of boiling water reactor plants in the United States. The purpose of the study was to identify and evaluate the effects of aging and the effectiveness of testing and maintenance in detecting and mitigating aging degradation. Guidelines from the United States Nuclear Regulatory Commission's Nuclear Plant Aging Research Program were used in performing the aging study. Review and analysis of the failures reported in databases such as Nuclear Power Experience, Licensee Event Reports, and the Nuclear Plant Reliability Data System, along with plant-specific maintenance records databases, are included in this report to provide the information required to identify aging stressors, failure modes, and failure causes. Several probabilistic risk assessments were reviewed to identify risk-significant components in high pressure injection systems. Testing, maintenance, specific safety issues, and codes and standards are also discussed

  17. Transport Phenomena in Thin Rotating Liquid Films Including: Nucleate Boiling

    Science.gov (United States)

    Faghri, Amir

    2005-01-01

    In this grant, experimental, numerical and analytical studies of heat transfer in a thin liquid film flowing over a rotating disk have been conducted. Heat transfer coefficients were measured experimentally in a rotating disk heat transfer apparatus where the disk was heated from below with electrical resistance heaters. The heat transfer measurements were supplemented by experimental characterization of the liquid film thickness using a novel laser based technique. The heat transfer measurements show that the disk rotation plays an important role on enhancement of heat transfer primarily through the thinning of the liquid film. Experiments covered both momentum and rotation dominated regimes of the flow and heat transfer in this apparatus. Heat transfer measurements have been extended to include evaporation and nucleate boiling and these experiments are continuing in our laboratory. Empirical correlations have also been developed to provide useful information for design of compact high efficiency heat transfer devices. The experimental work has been supplemented by numerical and analytical analyses of the same problem. Both numerical and analytical results have been found to agree reasonably well with the experimental results on liquid film thickness and heat transfer Coefficients/Nusselt numbers. The numerical simulations include the free surface liquid film flow and heat transfer under disk rotation including the conjugate effects. The analytical analysis utilizes an integral boundary layer approach from which

  18. BWR [boiling water reactor] shutdown margin model in SIMULATE-3

    International Nuclear Information System (INIS)

    Boiling water reactor (BWR) technical specifications require that the reactor be kept subcritical (by some prescribed margin) when at room temperature rodded conditions with any one control rod fully withdrawn. The design of an acceptable core loading pattern may require hundreds or thousands of neutronic calculations in order to predict the shutdown margin for each control rod. Direct, full-core, three-dimensional calculations with the SIMULATE-3 two-group advanced nodal code require 3 to 6 CPU min (on a SUN-4 workstation) for each statepoint/control rod that is computed. Such computing and manpower requirements may be burdensome, particularly during the early core design process. These requirements have been significantly reduced by the development of a fast, accurate shutdown margin model in SIMULATE-3. The SIMULATE-3 shutdown margin model achieves a high degree of accuracy and speed without using axial collapsing approximations inherent in many models. The mean difference between SIMULATE-3 one-group and two-group calculations is approximately - 12 pcm with a standard deviation of 35 pcm. The SIMULATE-3 shutdown margin model requires a factor of ∼15 less CPU time than is required for stacked independent two-group SIMULATE-3 calculations

  19. Stability analysis on natural circulation boiling water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Metz, Peter

    1999-05-01

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

  20. Invited talk on ageing management of boiling water reactors (BWRs)

    International Nuclear Information System (INIS)

    A nuclear power plant is built with a certain design life but by managing the operation of the plant with a well designed in-service inspection, repair and replacement programme of the equipment as required we will be able to extend the operation of the plant well beyond it's design life. This is also economically a paying proposition in view of the astronomical cost of construction of a new plant of equivalent capacity. In view of this, there is a growing trend the world over to study the ageing phenomena, especially in respect of nuclear power plant equipment and system which will contribute towards the continued operation of the nuclear power plants beyond their economic life which is fixed mainly to amortize the investments over a period. Tarapur Atomic Power Station (TAPS) which consists of 2 nos. of Boiling Water Reactor (BWRs) with the presently rated capacity of 160 MWe each has been operating for the past 24 years and is completing its 25th year of service by the year 1994 which was considered as its economic life and the plant depreciation as well as fuel supply agreement were based on this period of 25 years. I will be discussing about the available residual life which is much more than the above (25 years) and the studies we have undertaken in respect of the assessment of this residual life. (author). 2 tabs., 6 figs