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

OpenAIRE

Mirmanto

2013-01-01

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

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

International Nuclear Information System (INIS)

Unal, H.C.

1977-01-01

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

Development of an experimental apparatus for boiling analysis

International Nuclear Information System (INIS)

Castro, A.J.A. de.

1984-04-01

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

An experimental investigation of triggered film boiling destabilisation

International Nuclear Information System (INIS)

Naylor, P.

1985-03-01

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

Analysis of water hammer in control rod drive systems of boiling water reactor nuclear power plants

International Nuclear Information System (INIS)

Safwat, H.H.; Arastu, A.H.; Lau, S.

1983-01-01

The method of characteristics is applied to analyze water hammer in BWR (Boiling Water Reactor) Control Rod Drive (CRD) Systems following fast opening of scram valves. The modelling of the CRD mechanism is presented. Numerical predictions are compared to experimental data. (author)

Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

Energy Technology Data Exchange (ETDEWEB)

S.T. Revankar; W. Zhou; Gavin Henderson

2008-07-08

The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to : 1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, 2) assess the RELAP5 and TRACE computer code against the experimental data, and 3) develop mathematical model and ehat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal- hydraulic codes assessment.

Experimental and Thermalhydraulic Code Assessment of the Transient Behavior of the Passive Condenser System in an Advanced Boiling Water Reactor

International Nuclear Information System (INIS)

S.T. Revankar; W. Zhou; Gavin Henderson

2008-01-01

The main goal of the project was to study analytically and experimentally the condensation heat transfer for the passive condenser system such as GE Economic Simplified Boiling Water Reactor (ESBWR). The effect of noncondensable gas in condenser tube and the reduction of secondary pool water level to the condensation heat transfer coefficient was the main focus in this research. The objectives of this research were to: (1) obtain experimental data on the local and tube averaged condensation heat transfer rates for the PCCS with non-condensable and with change in the secondary pool water, (2) assess the RELAP5 and TRACE computer code against the experimental data, and (3) develop mathematical model and heat transfer correlation for the condensation phenomena for system code application. The project involves experimentation, theoretical model development and verification, and thermal-hydraulic codes assessment

Planned experimental studies on natural-circulation and stability performance of boiling water reactors in four experimental facilities and first results (NACUSP)

Energy Technology Data Exchange (ETDEWEB)

Kruijf, W.J.M. de E-mail: kruijf@iri.tudelft.nl; Ketelaar, K.C.J.; Avakian, G.; Gubernatis, P.; Caruge, D.; Manera, A.; Hagen, T.H.J.J. van der; Yadigaroglu, G.; Dominicus, G.; Rohde, U.; Prasser, H.-M.; Castrillo, F.; Huggenberger, M.; Hennig, D.; Munoz-Cobo, J.L.; Aguirre, C

2003-04-01

Within the 5th Euratom framework programme the NACUSP project focuses on natural-circulation and stability characteristics of Boiling Water Reactors (BWRs). This paper gives an overview of the research to be performed. Moreover, it shows the first results obtained by one of the four experimental facilities involved. Stability boundaries are given for the low-power low-pressure operating range, measured in the CIRCUS facility. The experiments are meant to serve as a future validation database for thermohydraulic system codes to be applied for the design and operation of BWRs.

Planned experimental studies on natural-circulation and stability performance of boiling water reactors in four experimental facilities and first results (NACUSP)

International Nuclear Information System (INIS)

Kruijf, W.J.M. de; Ketelaar, K.C.J.; Avakian, G.; Gubernatis, P.; Caruge, D.; Manera, A.; Hagen, T.H.J.J. van der; Yadigaroglu, G.; Dominicus, G.; Rohde, U.; Prasser, H.-M.; Castrillo, F.; Huggenberger, M.; Hennig, D.; Munoz-Cobo, J.L.; Aguirre, C.

2003-01-01

Within the 5th Euratom framework programme the NACUSP project focuses on natural-circulation and stability characteristics of Boiling Water Reactors (BWRs). This paper gives an overview of the research to be performed. Moreover, it shows the first results obtained by one of the four experimental facilities involved. Stability boundaries are given for the low-power low-pressure operating range, measured in the CIRCUS facility. The experiments are meant to serve as a future validation database for thermohydraulic system codes to be applied for the design and operation of BWRs

An experimental study of forced convective flow boiling CHF in nanofluid

International Nuclear Information System (INIS)

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

2008-01-01

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

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

International Nuclear Information System (INIS)

Yan Mingyu; Qiu Suizheng; Jia Dounan

2005-01-01

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

Potential uses of high gradient magnetic filtration for high-temperature water purification in boiling water reactors

International Nuclear Information System (INIS)

Elliott, H.H.; Holloway, J.H.; Abbott, D.G.

1979-01-01

Studies of various high-temperature filter devices indicate a potentially positive impact for high gradient magnetic filtration on boiling water reactor radiation level reduction. Test results on in-plant water composition and impurity crystallography are presented for several typical boiling water reactors (BWRs) on plant streams where high-temperature filtration may be particularly beneficial. An experimental model on the removal of red iron oxide (hematite) from simulated reactor water with a high gradient magnetic filter is presented, as well as the scale-up parameters used to predict the filtration efficiency on various high temperature, in-plant streams. Numerical examples are given to illustrate the crud removal potential of high gradient magnetic filters installed at alternative stream locations under typical, steady-state, plant operating conditions

Intelligent information data base of flow boiling characteristics in once-through steam generator for integrated type marine water reactor

International Nuclear Information System (INIS)

Inasaka, Fujio; Nariai, Hideki

1998-01-01

Valuable experimental knowledge with flow boiling characteristics of the helical-coil type once-through steam generator was converted into an intelligent information data base program. The program was created as a windows application using the Visual Basic. Main functions of the program are as follows: (1) steady state flow boiling analysis of any helical-coil type once-through steam generator, (2) analysis and comparison with the experimental data, (3) reference and graph display of the steady state experimental data, (4) reference of the flow instability experimental data and display of the instability threshold correlated by each parameter, (5) summary of the experimental apparatus. (6) menu bar such as a help and print. In the steady state analysis, the region lengths of subcooled boiling, saturated boiling, and super-heating, and the temperature and pressure distributions etc. for secondary water calculated. Steady state analysis results agreed well with the experimental data, with the exception of the pressure drop at high mass velocity. The program will be useful for the design of not only the future integrated type marine water reactor but also the small sized water reactor with helical-coil type steam generator

Nuclear boiling heat transfer and critical heat flux in titanium dioxide-water nanofluids

International Nuclear Information System (INIS)

Okawa, Tomio; Takamura, Masahiro; Kamiya, Takahito

2011-01-01

Nucleate boiling heat transfer was experimentally studied for saturated pool boiling of water-based nanofluids. Since significant nanoparticle deposition on the heated surface was observed after the nucleate boiling in nanofluids, measurement of CHF was also carried out using the nanoparticle deposited heated surface; pure water was used in the CHF measurement. In the present work, the heated surface was a 20 mm diameter cupper surface, and titanium-dioxide was selected as the material of nanoparticles. Experiments were performed for upward- and downward-facing surfaces. Although the CHFs for the downward-facing surface were generally lower than those for the upward-facing surface, the CHFs for the nanoparticle deposited surface were about 1.9 times greater than those for the bare surface in both the configurations. The CHF improvement corresponded well to the reduction of the surface contact angle. During the nucleate boiling in nanofluids, the boiling heat transfer showed peculiar behavior; it was first deteriorated, then improved, and finally approached to an equilibrium state. This observation indicated that the present nanofluid had competing effects to deteriorate and improve the nucleate boiling heat transfer. It was assumed that the wettability and the roughness of the heated surface were influenced by the deposited nanoparticles to cause complex variation of the number of active nucleation sites. During the nucleate boiling of pure water using the downward-facing surface, a sudden increase in the wall temperature was observed stochastically probably due to the accumulation of bubbles beneath the heated surface. Such behavior was not observed when the pure water was replaced by the nanofluid. (author)

Experimental study and modelling of transient boiling

International Nuclear Information System (INIS)

Baudin, Nicolas

2015-01-01

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

Study on water boiling noises in a large volume

International Nuclear Information System (INIS)

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

1977-01-01

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

Experimental study on transient boiling heat transfer

International Nuclear Information System (INIS)

Visentini, R.

2012-01-01

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

Experimental Study on Boiling Crisis in Pool Boiling

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-12-01

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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...

Predictors of Drinking Water Boiling and Bottled Water Consumption in Rural China: A Hierarchical Modeling Approach.

Science.gov (United States)

Cohen, Alasdair; Zhang, Qi; Luo, Qing; Tao, Yong; Colford, John M; Ray, Isha

2017-06-20

Approximately two billion people drink unsafe water. Boiling is the most commonly used household water treatment (HWT) method globally and in China. HWT can make water safer, but sustained adoption is rare and bottled water consumption is growing. To successfully promote HWT, an understanding of associated socioeconomic factors is critical. We collected survey data and water samples from 450 rural households in Guangxi Province, China. Covariates were grouped into blocks to hierarchically construct modified Poisson models and estimate risk ratios (RR) associated with boiling methods, bottled water, and untreated water. Female-headed households were most likely to boil (RR = 1.36, p water, or use electric kettles if they boiled. Our findings show that boiling is not an undifferentiated practice, but one with different methods of varying effectiveness, environmental impact, and adoption across socioeconomic strata. Our results can inform programs to promote safer and more efficient boiling using electric kettles, and suggest that if rural China's economy continues to grow then bottled water use will increase.

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Experimental study on forced convection boiling heat transfer on molten alloy

International Nuclear Information System (INIS)

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

1999-01-01

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

Heat transfer enhancement on nucleate boiling

International Nuclear Information System (INIS)

Zhuang, M.; Guibai, L.

1990-01-01

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

Cavitational boiling of liquids

International Nuclear Information System (INIS)

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

1986-01-01

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

Saturated flow boiling heat transfer in water-heated vertical annulus

International Nuclear Information System (INIS)

Sun Licheng; Yan Changqi; Sun Zhonning

2005-01-01

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

Experimental study on boiling heat transfer to an impinging jet on a hot block

International Nuclear Information System (INIS)

Kamata, Choko

1997-01-01

Previous studies on boiling heat transfer by impinging jets are mainly concerned with the impinging point by using small heat transfer surfaces of about 20 mm. An experimental study of the boiling heat transfer to an impinging water jet on a massive hot block is made. The upward heating surface is made of copper. Its diameter and nozzle diameter are 80 mm and 2.2 mm, respectively. The velocity of the impinging jet was varied from 0.6 to 2.1 m/s. Saturated water normally impinged on the heating surface, flowed radially, and subsequently dispersed into the atmosphere. The present study clarifies that heat transfer characteristics vary with the temperature of heat transfer surface, and also with the distance from the impinging point. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

Water jet intrusion into hot melt concomitant with direct-contact boiling of water

Energy Technology Data Exchange (ETDEWEB)

Sibamoto, Yasuteru [Japan Atomic Energy Research Inst., Tokai Research Establishment, Tokai, Ibaraki (Japan)

2005-08-01

Boiling of water poured on surface of high-temperature melt (molten metal or metal oxide) provides an efficient means for heat exchange or cooling of melt. The heat transfer surface area can be extended by forcing water into melt. Objectives of the present study are to elucidate key factors of the thermal and hydrodynamic interactions for the water jet injection into melt (Coolant Injection mode). Proposed applications include in in-vessel heat exchangers for liquid metal reactor and emergency measures for cooling of molten core debris in severe accidents of light water reactor. Water penetration into melt may occurs also as a result of fuel-coolant interaction (FCI) in modes other than CI, it is anticipated that the present study contributes to understand the fundamental mechanism of the FCI process. The previous works have been limited on understanding the melt-water interaction phenomena in the water-injection mode because of difficulty in experimental measurement where boiling occurs in opaque invisible hot melt unlike the melt-injection mode. We conducted visualization and measurement of melt-water-vapor multiphase flow phenomena by using a high-frame-rate neutron radiography technique and newly-developed probes. Although limited knowledge, however, has been gained even such an approach, the experimental data were analyzed deeply by comparing with the knowledge obtained from relevant matters. As a result, we succeeded in revealing several key phenomena and validity in the conditions under which stable heat transfer is established. Moreover, a non-intrusive technique for measurement of the velocity and pressure fields adjacent to a moving free surface is developed. The technique is based on the measurement of fluid surface profile, which is useful for elucidation of flow mechanism accompanied by a free surface like the present phenomena. (author)

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

Directory of Open Access Journals (Sweden)

Misyura Sergey Ya.

2018-01-01

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

An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

International Nuclear Information System (INIS)

Boing, L.E.; Henley, D.R.; Manion, W.J.; Gordon, J.W.

1989-12-01

Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs

An evaluation of alternative reactor vessel cutting technologies for the experimental boiling water reactor at Argonne National Laboratory

Energy Technology Data Exchange (ETDEWEB)

Boing, L.E.; Henley, D.R. (Argonne National Lab., IL (USA)); Manion, W.J.; Gordon, J.W. (Nuclear Energy Services, Inc., Danbury, CT (USA))

1989-12-01

Metal cutting techniques that can be used to segment the reactor pressure vessel of the Experimental Boiling Water Reactor (EBWR) at Argonne National Laboratory (ANL) have been evaluated by Nuclear Energy Services. Twelve cutting technologies are described in terms of their ability to perform the required task, their performance characteristics, environmental and radiological impacts, and cost and schedule considerations. Specific recommendations regarding which technology should ultimately be used by ANL are included. The selection of a cutting method was the responsibility of the decommissioning staff at ANL, who included a relative weighting of the parameters described in this document in their evaluation process. 73 refs., 26 figs., 69 tabs.

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Enhanced Natural Convection in a Metal Layer Cooled by Boiling Water

International Nuclear Information System (INIS)

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

2004-01-01

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

Dynamic model for a boiling water reactor

International Nuclear Information System (INIS)

Muscettola, M.

1963-07-01

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

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

International Nuclear Information System (INIS)

Bechta, S.V.; Vitol, S.A.; Krushinov, E.V.

1999-01-01

Experimental results are presented on the interaction between corium melt and water supplied onto its surface. The tests were conducted on the Rasplav-2' experimental facility. Induction melting in a cold crucible was used to produce the melt. The following data have been obtained: heat transfer at water boiling on the melt surface, aerosol release, structure of the post-interaction solidified corium. The corium melt had the following composition, mass %: 60%UO 2 - 16%ZrO 2 - 15%Fe 2 O 3 - 6%Cr 2 O 3 -3%Ni 2 O 3 . The melt surface temperature was 1650-1700degC. (author)

An overview of the U.S. Department of Energy Experimental Boiling Water Reactor Decontamination and Decommissioning Project

International Nuclear Information System (INIS)

Murphie, W.E.; Mckernan, M.L.

1991-01-01

This paper provides an overview of the U.S. Department of Energy's (DOE) Experimental Boiling Water Reactor (EBWR) Decontamination and Decommissioning (D and D) Project. Physical decommissioning work started in 1986 and is scheduled for completion in 1994. The project total estimated cost is 14.3 million (1990, U.S.) dollars. The reactor pressure vessel will be removed by segmentation. Another notable project feature is that D and D operations were planned for and carried out with a small work force comprised of four to six D and D laborers, one or two health physics technicians, an engineer, and a project manager. When the D and D work is completed the facility will be recycled for other productive uses. (author)

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

International Nuclear Information System (INIS)

Smith, P.F.

1992-01-01

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

Film boiling heat transfer in liquid helium

International Nuclear Information System (INIS)

Inai, Nobuhiko

1979-01-01

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

Boiling curve in high quality flow boiling

International Nuclear Information System (INIS)

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

1980-01-01

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

Technical report on operating experience with boiling water reactor offgas systems

International Nuclear Information System (INIS)

Lo, R.; Barrett, L.; Grimes, B.; Eisenhut, D.

1978-03-01

Over 100 reactor years of Boiling Water Reactor (BWR) operating experience have been accumulated since the first commercial operation of BWRs. A number of incidents have occurred involving the ''offgas'' of these Boiling Water Reactors. This report describes the generation and processing of ''offgas'' in Boiling Water Reactors, the safety considerations regarding systems processing the ''offgas'', operating experience involving ignitions or explosions of ''offgas'' and possible measures to reduce the likelihood of future ignitions or explosions and to mitigate the consequences of such incidents should they occur

An Experimental Study on the Pool Boiling Heat Transfer on a Square Surface

International Nuclear Information System (INIS)

Kim, Jae Kwang

2000-02-01

An experimental study was carried out to identify the various regimes of natural convective boiling and to determine the Critical Heat Flux (CHF) on a square surface. The basic knowledge on the boiling heat transfer and CHF on the square surface is necessary for various engineering problems, such as the design of compact heat exchangers, cooling of CPU chips, and design of the external cooling mechanism for the reactor during the severe accidents in the nuclear power plants. The heater block made of copper with cartridge heaters in it is submerged in a water tank with windows for visualization. The heater surface has dimension of 70mm x 70mm and the maximum heat flux capacity is about 1.8MW/m 2 . The boiling heat transfer coefficient for the various flow regimes up to CHF has been measured for upward facing surface, vertical surface, and nearly horizontal downward facing surfaces. The temperatures of the heater block are measured by the thermocouples imbedded in the heater block. As the heat flux increases from 100kW/m 2 to 1.0MW/m 2 , the heat-transfer regime changes from the nucleate boiling to the CHF. Near 1.0MW/m 2 , the heat transfer regime suddenly changed from nucleate boiling to film boiling and it resulted in a rapid heat up of the heater block. The various boiling patterns on the vertical surface, upward facing surface, and downward facing surface are observed by a high speed video camera whose frame rate is 1000fps. An explosive vapor generation on the heated surface, whose size and frequency are characterized by the heat flux and inclination angle, is observed

Pool boiling of water on nano-structured micro wires at sub-atmospheric conditions

Science.gov (United States)

Arya, Mahendra; Khandekar, Sameer; Pratap, Dheeraj; Ramakrishna, S. Anantha

2016-09-01

Past decades have seen active research in enhancement of boiling heat transfer by surface modifications. Favorable surface modifications are expected to enhance boiling efficiency. Several interrelated mechanisms such as capillarity, surface energy alteration, wettability, cavity geometry, wetting transitions, geometrical features of surface morphology, etc., are responsible for change in the boiling behavior of modified surfaces. Not much work is available on pool boiling at low pressures on microscale/nanoscale geometries; low pressure boiling is attractive in many applications wherein low operating temperatures are desired for a particular working fluid. In this background, an experimental setup was designed and developed to investigate the pool boiling performance of water on (a) plain aluminum micro wire (99.999 % pure) and, (b) nano-porous alumina structured aluminum micro wire, both having diameter of 250 µm, under sub-atmospheric pressure. Nano-structuring on the plain wire surface was achieved via anodization. Two samples, A and B of anodized wires, differing by the degree of anodization were tested. The heater length scale (wire diameter) was much smaller than the capillary length scale. Pool boiling characteristics of water were investigated at three different sub-atmospheric pressures of 73, 123 and 199 mbar (corresponding to T sat = 40, 50 and 60 °C). First, the boiling characteristics of plain wire were measured. It was noticed that at sub-atmospheric pressures, boiling heat transfer performance for plain wire was quite low due to the increased bubble sizes and low nucleation site density. Subsequently, boiling performance of nano-structured wires (both Sample A and Sample B) was compared with plain wire and it was noted that boiling heat transfer for the former was considerably enhanced as compared to the plain wire. This enhancement is attributed to increased nucleation site density, change in wettability and possibly due to enhanced pore scale

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

International Nuclear Information System (INIS)

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

1989-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-15

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

Experimental investigation of nucleate boiling on heated surfaces under subcooled conditions

International Nuclear Information System (INIS)

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

2011-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Bechta, S.V.; Vitol, S.A.; Krushinov, E.V.; Granovsky, V.S.; Sulatsky, A.A.; Khabensky, V.B. [Sci. Res. Technol. Inst., Leningrad (Russian Federation); Lopukh, D.B.; Petrov, Y.B.; Pechenkov, A.Y. [St. Petersburg Electrotechnical University (SPbEU), Prof. Popov st 5/3, St. Petersburg (Russian Federation)

2000-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Bechta, S.V.; Vitol, S.A.; Krushinov, E.V. [Research Institute of Technology, Sosnovy Bor (NITI) (RU)] [and others

1999-07-01

Experimental results are presented on the interaction between corium melt and water supplied onto its surface. The tests were conducted on the Rasplav-2' experimental facility. Induction melting in a cold crucible was used to produce the melt. The following data have been obtained: heat transfer at water boiling on the melt surface, aerosol release, structure of the post-interaction solidified corium. The corium melt had the following composition, mass %: 60%UO{sub 2}- 16%ZrO{sub 2}- 15%Fe{sub 2}O{sub 3} - 6%Cr{sub 2}O{sub 3}-3%Ni{sub 2}O{sub 3}. The melt surface temperature was 1650-1700degC. (author)

On Stability of Natural-circulation-cooled Boiling Water Reactors during Start-up (Experimental Results)

International Nuclear Information System (INIS)

Manera, A.; Van der Hagen, T.H.J.J.

2002-01-01

The characteristics of flashing-induced instabilities, which are of importance during the start-up phase of natural-circulation Boiling Water Reactors (BWRs), are studied. Experiments at typical start-up conditions (low power and low pressure) are carried out on a steam/water natural circulation loop. The mechanism of flashing-induced instability is analyzed in detail and it is found that non-equilibrium between phases and enthalpy transport plays an important role in the instability process. Pressure and steam volume in the steam dome are found to have a stabilizing effect. The main characteristics of the instabilities have been analyzed. (authors)

Water inventory management in condenser pool of boiling water reactor

International Nuclear Information System (INIS)

Gluntz, D.M.

1996-01-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Boiling in the presence of boron compounds in light water reactors

International Nuclear Information System (INIS)

Nakath, Richard

2014-01-01

The scope of the thesis on boiling in the presence of boron compounds in light water reactors was to study the effects of the boron compound addition on the heat removal from the fuel elements. For an effective cooling of the fuel elements in case of boiling processes a high heat transfer coefficient is of importance. Up to now experimental studies were not performed under reactor specific conditions, for instance with respect to the geometry of the flow conditions, high temperature and pressure levels were not represented. Therefore the experiments in the frame of the thesis were using reactor specific parameters. The test facility SECA (study into the effects of coolant additives) was designed and constructed. The experiments simulated the conditions of normal PWR operation, accidental PWR and accidental BWR conditions.

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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

Science.gov (United States)

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

2015-04-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Coarse mesh finite element method for boiling water reactor physics analysis

International Nuclear Information System (INIS)

Ellison, P.G.

1983-01-01

A coarse mesh method is formulated for the solution of Boiling Water Reactor physics problems using two group diffusion theory. No fuel assembly cross-section homogenization is required; water gaps, control blades and fuel pins of varying enrichments are treated explicitly. The method combines constrained finite element discretization with infinite lattice super cell trial functions to obtain coarse mesh solutions for which the only approximations are along the boundaries between fuel assemblies. The method is applied to bench mark Boiling Water Reactor problems to obtain both the eigenvalue and detailed flux distributions. The solutions to these problems indicate the method is useful in predicting detailed power distributions and eigenvalues for Boiling Water Reactor physics problems

Method of providing extended life expectancy for components of boiling water reactors

International Nuclear Information System (INIS)

Niedrach, L.W.

1992-01-01

This patent describes a containment for a boiling water nuclear reactor, a stainless steel containment, the containment having a deposit of a metal of the platinum group of metals on the surfaces thereof exposed to high temperature, high pressure water of the boiling water nuclear reactor

Cycle studies: material balance estimation in the domain of pressurized water and boiling water reactors. Experimental qualification

International Nuclear Information System (INIS)

Chabert, Christine

1994-01-01

This study is concerned with the physics of the fuel cycle the aim being to develop and make recommendations concerning schemes for calculating the neutronics of light water reactor fuel cycles. A preliminary study carried out using the old fuel cycle calculation scheme APOLLO1- KAFKA and the library SERMA79 has shown that for the compositions of totally dissolved assemblies from Pressurized Water Reactors (type 17*17) and also for the first time, for Boiling Water Reactor assemblies (type 8*8), the differences between calculation and measurement are large and must be reduced. The integration of the APOLLO2 neutronics code into the fuel cycle calculation scheme improves the results because it can model the situation more precisely. A comparison between APOLLO1 and APOLLO2 using the same options, demonstrated the consistency of the two methods for PWR and BWR geometries. Following this comparison, we developed an optimised scheme for PWR applications using the library CEA86 and the code APOLLO2. Depending on whether the information required is the detailed distribution of the composition of the irradiated fuel or the average composition (estimation of the total material balance of the fuel assembly), the physics options recommended are different. We show that the use of APOLLO2 and the library CEA86 improves the results and especially the estimation of the Pu 239 content. Concerning the Boiling Water Reactor, we have highlighted the need to treat several axial sections of the fuel assembly (variation of the void-fraction, heterogeneity of composition). A scheme using Sn transport theory, permits one to obtain a better coherence between the consumption of U 235 , the production of plutonium and burnup, and a better estimation of the material balance. (author) [fr

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

Science.gov (United States)

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

2017-12-01

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

Experimental investigation of boiling-water nuclear-reactor parallel-channel effects during a postulated loss-of-coolant accident

International Nuclear Information System (INIS)

Conlon, W.M.; Lahey, R.T. Jr.

1982-12-01

This report describes an experimental study of the influence of parallel channel effects (PCE) on the distribution of emergency core spray cooling water in a Boiling Water Nuclear Reactor (BWR) following a postulated design basis loss of coolant accident (LCA). The experiments were conducted in a scaled test section in which the reactor coolant was simulated by Freon-114 at conditions similar to those postulated to occur in the reactor vessel shortly after a LOCA. A BWR/4 was simulated by a (PCE) test section which contained three parallel heated channels to simulate fuel assemblies; a core bypass channel, and a jet pump channel. The test section also inlcuded scaled regions to simulate the lower and upper plena, downcomer, and steam separation regions of a BWR. A series of nine transient experiments were conducted, in which the lower plenum vaporization rate and heater rod power were varied while the core spray flow rate was held constant to simulate that of a BWR/4. During these experiments the flow distribution and heat transfer phenomena were observed and measured

Operating experience of natural circulation core cooling in boiling water reactors

International Nuclear Information System (INIS)

Kullberg, C.; Jones, K.; Heath, C.

1993-01-01

General Electric (GE) has proposed an advanced boiling water reactor, the Simplified Boiling Water Reactor (SBWR), which will utilize passive, gravity-driven safety systems for emergency core coolant injection. The SBWR design includes no recirculation loops or recirculation pumps. Therefore the SBWR will operate in a natural circulation (NC) mode at full power conditions. This design poses some concerns relative to stability during startup, shutdown, and at power conditions. As a consequence, the NRC has directed personnel at several national labs to help investigate SBWR stability issues. This paper will focus on some of the preliminary findings made at the INEL. Because of the broad range of stability issues this paper will mainly focus on potential geysering instabilities during startup. The two NC designs examined in detail are the US Humboldt Bay Unit 3 BWR-1 plant and Dodewaard plant in the Netherlands. The objective of this paper will be to review operating experience of these two plants and evaluate their relevance to planned SBWR operational procedures. For completeness, experimental work with early natural circulation GE test facilities will also be briefly discussed

Boiling water system of nuclear power plants (BWR)

International Nuclear Information System (INIS)

Martias Nurdin

1975-01-01

About 85% of the world electric generators are light water reactors. It shows that LWR is technologically and economically competitive with other generators. The Boiling Water Reactor (BWR) is one of the two systems in the LWR group. The techniques of BWR operation in several countries, especially low and moderate power BWR, are presented. The discussion is made in relation with the interconnection problems of electric installation in developing countries, including Indonesia, where the total electric energy installation is low. The high reliability and great flexibility of the operation of a boiling water reactor for a sufficiently long period are also presented. Component standardization for BWR system is discussed to get a better technological and economical performance for further development. (author)

Mark I 1/5-scale boiling water reactor pressure suppresion experiment quick-look report

International Nuclear Information System (INIS)

Lai, W.; Collins, E.K.

1977-01-01

This report is intended as a ''quick-look'' report summarizing the experimental results obtained from pressure suppression experiment numbers 2.1, 2.2, and 2.3 that were performed on the Lawrence Livermore Laboratory's 1/5-scale boiling water reactor (BWR) Mark I pressure suppression experimental facility on April 26, 1977. A brief description of the general nature of the tests and a summary of the actual tests that were performed are given

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

International Nuclear Information System (INIS)

Johannsen, K.; Meinen, W.

1984-01-01

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

Status report on the Experimental Boiling Water Reactor (EBWR) Decontamination and Decommissioning (D ampersand D) Project

International Nuclear Information System (INIS)

Sears, L.; Garlock, G.; Mencarelli, R.; Fellhauer, C.

1994-01-01

ALARON Corporation is under contract, to Argonne National Laboratory - East (ANL-E), to complete the decontamination and decommissioning of the Experimental Boiling Water Reactor (EBWR). The project, begun, in 1986 by ANL-E personnel, is projected to be completed by the end of 1994. The final phase of work was awarded to ALARON in December 1993 with the scope of work including the disassembly and removal of all remaining reactor internals, the reactor vessel, the lead bio-shield, the core liner, and the activated portion of the concrete bio-shield. This paper discusses the work undertaken beginning in January 1994 and continuing through July 1994. During this period the required pre-mobilization documentation was prepared and approved, mobilization was completed, and the reactor internals, reactor vessel, lead bio-shield and core liner were removed. The paper will compare the planned schedule to the actual schedule, discuss problems encountered, review volume reduction techniques and health and safety issues including radiological aspects of the project

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

International Nuclear Information System (INIS)

Yu Zhenwan; Jia Dounan; Li Linjiao; Mu Quanhou

1989-01-01

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

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

OpenAIRE

Yu, Jian; Koyama, Shigeru; Momoki, Satoru

1995-01-01

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

Boiling of water in flow restricted areas modeled by colloidal silica deposits

International Nuclear Information System (INIS)

Peixinho, Jorge; Lefevre, Gregory; Coudert, Francois-Xavier; Hurisse, Olivier

2012-09-01

Understanding the effects of particle deposits on evaporation and boiling of water represents an important issue for EDF because it causes a severe reduction in efficiency particularly in steam generators of pressurized water reactor. These deposits are made of oxide metallic particles and the deposition process depends on multiple factors. Here we mimic deposits using a simple system made of hydrophilic silica particles. The present study reports experiments on evaporation or boiling of water confined in the pores of colloidal mono-dispersed silica micro-sphere deposits. The boiling of water confined in the pores of the colloidal crystal is studied using optical microscopy, scanning electron microscopy, nitrogen adsorption, water adsorption through infrared attenuated total reflectance spectroscopy, differential scanning calorimetry and thermal gravimetric analysis. By comparison of the results with silica deposits and alumina membranes with cylindrical pores, our study shows that the morphology of the pores contributes to the evaporation and boiling of water. The measurements suggest that particle resuspension and crust formation take place during drying at elevated temperature and are responsible for cracks formation within the deposit film. (authors)

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

Absence of genotoxic activity from milk and water boiled in microwave oven in somatic cells from Drosophila melanogaster

International Nuclear Information System (INIS)

Dias, Cristina das Dores.

2003-01-01

This paper reports an experiment for evaluation of the possible genotoxic effects of food prepared in a microwave oven, through the mutation test and somatic recombination, in wings of Drosophila melanogaster. Two crossing have been performed: a standard cross-ST and a high bioactivation cross - HB resulting in marked trans -heterozygote descendents (MH) and balanced heterozygotes (BH). The 72 hours larvas were fed with water and milk boiled both in the microwave oven and in the traditional way. The MH individual wings were analyzed, where the spots can be induced either by mutation or mitotic recombination. The experiment presented negative results related to the genotoxic effects of the water and milk boiled using the microwave oven, in MH descendents of both crossing. Therefore, under these experimental conditions, genotoxic activity were not presented by milk and water boiled in the microwave oven. However, an extensive study using different techniques is necessary to investigate the action of the food prepared in the microwave oven on the genetic material

A new correlation for nucleate pool boiling of aqueous mixtures

International Nuclear Information System (INIS)

Thome, J.R.; Shakir, S.

1987-01-01

A new mixture boiling correlation was developed for nucleate pool boiling of aqueous mixtures on plain, smooth tubes. The semi-empirical correlation models the rise in the local bubble point temperature in a mixture caused by the preferential evaporation of the more volatile component during bubble growth. This rise varies from zero at low heat fluxes (where only single-phase natural convection is present) up to nearly the entire boiling range at the peak heat flux (where latent heat transport is dominant). The boiling range, which is the temperature difference between the dew point and bubble point of a mixture, is used to characterize phase equilibrium effects. An exponential term models the rise in the local bubble point temperature as a function of heat flux. The correlation was compared against binary mixture boiling data for ethanol-water, methanol-water, n-propanol-water, and acetone-water. The majority of the data was predicted to within 20%. Further experimental research is currently underway to obtain multicomponent boiling data for aqueous mixtures with up to five components and for wider boiling ranges

Vapor bubble behavior in subcooled flow boiling in annuli heated by water

International Nuclear Information System (INIS)

Licheng Sun; Zhongning Sun; Changqi Yan

2005-01-01

Full text of publication follows: This paper describes experimental and theoretical work conducted on vapor bubble behavior in subcooled flow boiling at atmospheric pressure. The test section is mainly consisted of two concentrically installed circular tubes, the outside tube is made of quartz and therefore all test courses can be visualized. Water is forced to flow through annuli with gap sizes of 3 mm and 5 mm, and is heated by high temperature water in the inner tube. The main objective is to visually study the bubble behavior of subcooled flow boiling water in the condition of surface heated by water. The results show that bubbles depart from wall directly or slide a certain distance before departure, this is same as that heated by electricity. There exists a bubble layer near the wall, most bubbles move and disappear in the layer after departure, the bubble sliding behavior is not very obvious in 5 mm annulus, however, we found that most bubbles in 3 mm annulus will slide a long distance before departure and their growth courses are different from usual experimental results. The bubbles are not always growing, but shrinking a little quickly after growing for some time, and then the course will repeat for some times till they depart from wall or disappeared, the collision and coalescence of bubbles is very common and makes the bubbles depart from wall more easily in 3 mm annulus. At last, the forces on bubbles growing and detaching in flow along the wall are analyzed to comprehend these phenomena more accurately. (authors)

Numerical and Experimental Study of Mechanisms Involved in Boiling Histotripsy.

Science.gov (United States)

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

2017-12-01

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

Nuclear fuel performance in boiling water reactors

International Nuclear Information System (INIS)

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

1981-01-01

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

A stability identification system for boiling water nuclear reactors

International Nuclear Information System (INIS)

Belblidia, L.A.; Chevrier, A.

1994-01-01

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

Instrumenting a pressure suppression experiment for a MK I boiling water reactor: another measurements engineering challenge

International Nuclear Information System (INIS)

Shay, W.M.; Brough, W.G.; Miller, T.B.

1977-01-01

A scale test facility of a pressure suppression system from a boiling water reactor was instrumented with seven types of transducers to obtain high-accuracy experimental data during a hypothetical loss-of-coolant accident. The instrumentation verified the analysis of the dynamic loading of the pressure suppression system

Critical heat flux of subcooled flow boiling in a narrow tube

International Nuclear Information System (INIS)

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

1986-01-01

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

Investigation of water films on fuel rods in boiling water reactors using neutron tomography

Energy Technology Data Exchange (ETDEWEB)

Lanthen, Jonas

2006-09-15

In a boiling water reactor, thin films of liquid water around the fuel rods play a very important role in cooling the fuel, and evaporation of the film can lead to fuel damage. If the thickness of the water film could be measured accurately the reactor operation could be both safer and more economical. In this thesis, the possibility to use neutron tomography, to study thin water films on fuel rods in an experimental nuclear fuel set-up, has been investigated. The main tool for this has been a computer simulation software. The simulations have shown that very thin water films, down to around 20 pm, can be seen on fuel rods in an experimental set-up using neutron tomography. The spatial resolution needed to obtain this result is around 300 pm. A suitable detector system for this kind of experiment would be plastic fiber scintillators combined with a CCD camera. As a neutron source it would be possible to use a D-D neutron generator, which generates neutrons with energies of 2.5 MeV. Using a neutron generator with a high enough neutron yield and a detector with high enough detection efficiency, a neutron tomography to measure thin water films should take no longer than 25 - 30 minutes.

Investigation of water films on fuel rods in boiling water reactors using neutron tomography

International Nuclear Information System (INIS)

Lanthen, Jonas

2006-09-01

In a boiling water reactor, thin films of liquid water around the fuel rods play a very important role in cooling the fuel, and evaporation of the film can lead to fuel damage. If the thickness of the water film could be measured accurately the reactor operation could be both safer and more economical. In this thesis, the possibility to use neutron tomography, to study thin water films on fuel rods in an experimental nuclear fuel set-up, has been investigated. The main tool for this has been a computer simulation software. The simulations have shown that very thin water films, down to around 20 pm, can be seen on fuel rods in an experimental set-up using neutron tomography. The spatial resolution needed to obtain this result is around 300 pm. A suitable detector system for this kind of experiment would be plastic fiber scintillators combined with a CCD camera. As a neutron source it would be possible to use a D-D neutron generator, which generates neutrons with energies of 2.5 MeV. Using a neutron generator with a high enough neutron yield and a detector with high enough detection efficiency, a neutron tomography to measure thin water films should take no longer than 25 - 30 minutes

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

Energy Technology Data Exchange (ETDEWEB)

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

2008-05-15

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

Experimental investigation and mechanistic modelling of dilute bubbly bulk boiling

International Nuclear Information System (INIS)

Kutnjak, Josip

2013-01-01

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

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

Development of advanced boiling water reactor for medium capacity

International Nuclear Information System (INIS)

Kazuo Hisajima; Yutaka Asanuma

2005-01-01

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

Two-phase flow boiling pressure drop in small channels

International Nuclear Information System (INIS)

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

2016-01-01

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

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

DEFF Research Database (Denmark)

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

2016-01-01

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

An experimental study on pool boiling characteristics of carbon nano tube (CNT) and fullerene (C-60) nanofluids

International Nuclear Information System (INIS)

Ai, Melani

2009-02-01

In recent years, it was found that pool boiling critical heat flux (CHF) increases in nanofluids. The CHF conditions are important for safe and economic design of many heat transfer units including nuclear reactor. In this study, our objective is to evaluate the impact of Carbone Nano Tubes (Singlewalled CNTs and Multiwalled CNTs) and Fullerene (C-60) nanofluids at different particle concentration on pool boiling critical heat flux experimentally at saturated conditions. Multiwalled CNT and fullerene (C-60) added in the pure water at three volume concentrations (0.01%, 0.001%, and 0.0001%). Singlewalled CNT nanoparticles added in the pure water at two volume concentrations (0.0005%, and 0.0001%). For the dispersion of nanoparticles in pure water, several treatments were performed. Multiwalled CNTs and Fullerene (C-60) prepared using acid treatment, meanwhile two treatment are using for Singlewalled CNTs: (1)Singlewalled CNTs prepared using polymer treatment, (2)Singlewalled CNTs prepared using pre polymerization of micelle treatment. The zeta potential of CNTs and Fullerene nanofluids were in the range of 13-71 mV. The zeta potential of nanofluids was constant for more than one month. It concludes that the treatment has been succeeded produces water dispersible CNTs and Fullerene nanofluids with good stability. The critical heat flux (CHFs) of the solution is enhanced greatly for all nanofluids. Enhanced (∼167.9%) CHF was observed for solutions with Multiwalled CNT nanoparticles with concentration 0.01 vol%. Enhanced (∼109.4%) CHF was observed for solutions with Singlewalled CNT nanoparticles with concentration 0.0005 vol%. Enhanced (∼108.9%) CHF was observed for solutions with Fullerene nanoparticles with concentration 0.01 vol%. The pool boiling Heat Transfer Coefficient (HTCs) of the CNTs nanofluids are lower than those of pure water in the entire nucleate boiling regime. On the other hand, the pool boiling HTCs of Fullerene nanofluids are higher than

Experimental investigation of tube length effect on nucleate pool boiling heat transfer

International Nuclear Information System (INIS)

Kang, Myeong-Gie

1998-01-01

The effect of a vertically installed tube length on the nucleate pool boiling heat transfer coefficient under atmospheric pressure has been empirically obtained using various combination of major parameters for application to advanced light water reactor design. The experimental data for q'' versus ΔT test are counted as 1,063 points and can cover the extent of D = 9.7 ∼ 25.4 mm, ε = 15.1 ∼ 60.9 nm, H = 5.25 ∼ 30.93, and q'' ≤ 160 kW/m 2 . The experimental results show that a shorter tube is more efficient to increase heat transfer rate due to smaller bubble slug formation on the tube surface. The effect of tube length is greatly observed before H(= L/D) gets 50. After that, the heat flux decreases linearly with H increase. To quantify tube length effect, a new empirical correlation has been developed based on the experimental data bank for pool boiling heat transfer and some parametric studies have been done using the newly developed empirical correlation to broaden its applicability. The newly developed empirical correlation has the form of q'' 0.019ε 0.570 ΔT 4.676 /(D 1.238 H 0.072 ) and can predict the experimental data within ± 20% bound. Heat transfer characteristics can be changed with tube length variation and the transition point is H ∼ 50. Before the transition point, bubble coalescence is active and heat transfer rate gets rapidly decreased with increasing tube length. After that, heat transfer gets somewhat slowly decreased since bubble coalescence effect gets nearly equilibrium with liquid agitation effect

Some observations on boiling heat transfer with surface oscillation

International Nuclear Information System (INIS)

Miyashita, H.

1992-01-01

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

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.

Boiling water reactor life extension monitoring

International Nuclear Information System (INIS)

Stancavage, P.

1991-01-01

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

Pool film boiling heat transfer, 5

International Nuclear Information System (INIS)

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

1981-01-01

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

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

International Nuclear Information System (INIS)

Monde, M.; Katto, Y.

1978-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2006-10-20

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

Standard Technical Specifications for General Electric Boiling Water Reactors (BWR/5)

International Nuclear Information System (INIS)

Bottimore, R.R.

1980-12-01

The Standard Technical Specifications for General Electric Boiling Water Reactors (GE-STS) is a generic document prepared by the US NRC for use in the licensing process of current General Electric Boiling Water Reactors. The GE-STS sets forth the limits, operating conditions, and other requirements applicable to nuclear reactor facility operation as set forth by Section 50.36 of 10 CFR Part 50 for the protection of the health and safety of the public. The document is revised periodically to reflect current licensing requirements

CIRCUS and DESIRE: Experimental facilities for research on natural-circulation-cooled boiling water reactors

International Nuclear Information System (INIS)

Kruijf, W.J.M. de; Haden, T.H.J.J. van der; Zboray, R.; Manera, A.; Mudde, R.F.

2002-01-01

At the Delft University of Technology two thermohydraulic test facilities are being used to study the characteristics of Boiling Water Reactors (BWRs) with natural circulation core cooling. The focus of the research is on the stability characteristics of the system. DESIRE is a test facility with freon-12 as scaling fluid in which one fuel bundle of a natural-circulation BWR is simulated. The neutronic feedback can be simulated artificially. DESIRE is used to study the stability of the system at nominal and beyond nominal conditions. CIRCUS is a full-height facility with water, consisting of four parallel fuel channels and four parallel bypass channels with a common riser or with parallel riser sections. It is used to study the start-up characteristics of a natural-circulation BWR at low pressures and low power. In this paper a description of both facilities is given and the research items are presented. (author)

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

Energy Technology Data Exchange (ETDEWEB)

Kwang-Won, Lee; Sang-Yong, Lee

1995-09-01

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

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

International Nuclear Information System (INIS)

Morin, Henri; Verdier, Jacques

1964-10-01

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

An experimental study on micro-scale flow boiling heat transfer

International Nuclear Information System (INIS)

Tibirica, Cristiano Bigonha; Ribatski, Gherhardt

2009-01-01

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

An experimental study on micro-scale flow boiling heat transfer

Energy Technology Data Exchange (ETDEWEB)

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

2009-07-01

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

Hydrogen water chemistry for boiling water reactors

International Nuclear Information System (INIS)

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

1985-01-01

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

Possibilities of crack mouth opening displacement (CMOD) measurement under boiling and pressurized water reactor conditions

International Nuclear Information System (INIS)

Ehling, W.

1984-01-01

Fracture mechanics investigations carried out so far in laboratory conditions cover only part of the material stresses, as effects which occur in nuclear powerstations, in particular, such as corrosion and radioactive radiation are largely left out of account. Therefore experiments including these effects were recently carried out in autoclaves, test rigs simulating reactors (HRD experimental plant) and in experimental reactors. An important parameter of experimental fracture mechanics is the measurement of crack opening displacement (COD). The crack opening is measured with socalled clip gauges (transmitters based on strain gauges, which convert mechanical deformation of springs into electrical signals) on standard samples in the laboratory. It was therefore sensible to use these high temperature strain gauges (HTD) for the development of a measuring system for travel for pressurized water and boiling water reactor conditions. (orig.) [de

Radioactive waste management practices with KWU-boiling water reactors

International Nuclear Information System (INIS)

Queiser, H.

1976-01-01

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

Experimental simulation of the water cooling of corium spread over the floor of a BWR containment

Energy Technology Data Exchange (ETDEWEB)

Morage, F.; Lahey, R.T. Jr.; Podowski, M.Z. [Rensselaer Polytechnic Institute, Troy, NY (United States)

1995-09-01

This paper is concerned with an experimental investigation of the cooling effect of water collected on the surface of corium released onto the floor of a BWR drywell. In the present experiments, the actual reactor materials were replaced by simulant materials. Specifically, the results are shown for Freon-11 film boiling over liquid Wood`s metal spread above a solid porous surface through which argon gas was injected. An analysis of the obtained experimental data revealed that the actual film boiling heat transfer between a molten pool of corium and the water above the pool should be more efficient than predicted by using standard correlations for boiling over solid surfaces. This effect will be further augmented by the gas released due to the ablation of concrete floor beneath the corium and percolating towards its upper surface and into through the water layer above.

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

International Nuclear Information System (INIS)

Ricque, Roger; Siboul, Roger

1969-01-01

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

A review of boiling water reactor water chemistry: Science, technology, and performance

International Nuclear Information System (INIS)

Fox, M.J.

1989-02-01

Boiling water reactor (BWR) water chemistry (science, technology, and performance) has been reviewed with an emphasis on the relationships between BWR water quality and corrosion fuel performance, and radiation buildup. A comparison of Nuclear Regulatory Commission (NRC) Regulatory Guide 1.56, the Boiling Water Reactor Owners Group (BWROG) Water Chemistry Guidelines, and Plant Technical Specifications showed that the BWROG Guidelines are more stringent than the NRC Regulatory Guide, which is almost identical to Plant Technical Specifications. Plant performance with respect to BWR water chemistry has shown dramatic improvements in recent years. Up until 1979 BWRs experienced an average of 3.0 water chemistry incidents per reactor-year. Since 1979 the water chemistry technical specifications have been violated an average of only 0.2 times per reactor-year, with the most recent data from 1986-1987 showing only 0.05 violations per reactor-year. The data clearly demonstrate the industry-wide commitment to improving water quality in BWRs. In addition to improving water quality, domestic BWRs are beginning to switch to hydrogen water chemistry (HWC), a remedy for intergranular stress corrosion cracking. Three domestic BWRs are presently operating on HWC, and fourteen more have either performed HWC mini tests or are in various stages of HWC implementation. This report includes a detailed review of HWC science and technology as well as areas in which further research on BWR chemistry may be needed. 43 refs., 30 figs., 8 tabs

DIRECT-CYCLE, BOILING-WATER NUCLEAR REACTOR

Science.gov (United States)

Harrer, J.M.; Fromm, L.W. Jr.; Kolba, V.M.

1962-08-14

A direct-cycle boiling-water nuclear reactor is described that employs a closed vessel and a plurality of fuel assemblies, each comprising an outer tube closed at its lower end, an inner tube, fuel rods in the space between the tubes and within the inner tube. A body of water lying within the pressure vessel and outside the fuel assemblies is converted to saturated steam, which enters each fuel assembly at the top and is converted to superheated steam in the fuel assembly while it is passing therethrough first downward through the space between the inner and outer tubes of the fuel assembly and then upward through the inner tube. (AEC)

On the determination of boiling water reactor characteristics by noise analysis

International Nuclear Information System (INIS)

Kleiss, J.

1983-01-01

In boiling water reactors the main noise source is the boiling process in the core and the most important variable is the neutron flux, thus the effect of the steam bubbles on the neutron flux is studied in detail. An experiment has been performed in a small subcritical reactor to measure the response of a neutron detector to the passage of a single air bubble. A mathematical model for the description of the response was tested and the results agree very well with the experiment. Noise measurements in the Dodewaard boiling water reactor are discussed. The construction of a twin self-powered neutron detector, developed to perform steam velocity measurements in the core is described. The low-frequency part of the neutron noise characteristics is considered. The transfer functions exhibit a good agreement with ones obtained by independent means: control rod step experiments and model calculations. (Auth.)

Boiling Suppression in Convective Flow

International Nuclear Information System (INIS)

Aounallah, Y.

2004-01-01

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

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

International Nuclear Information System (INIS)

Ishida, Naoyuki; Utsuno, Hideaki; Kasahara, Fumio

2003-01-01

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

Modelling of Control Bars in Calculations of Boiling Water Reactors

International Nuclear Information System (INIS)

Khlaifi, A.; Buiron, L.

2004-01-01

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

Sloshing of water in torus pressure-suppression pool of boiling water reactors under earthquake ground motions

International Nuclear Information System (INIS)

Aslam, M.; Godden, W.G.; Scalise, D.T.

1978-08-01

This report presents an analytical and experimental investigation into the sloshing of water in torus tanks under horizontal earthquake ground motions. This study was motivated because of the use of torus tanks for pressure-suppression pools in Boiling Water Reactors. Such a pressure-suppression pool would typically have 80 ft and 140 ft inside and outside diameters, a 30 ft diameter section, and a water depth of 15 ft. A general finite element analysis was developed for all axisymmetric tanks and a computer program was written to obtain time-history plots of sloshing displacements of water and dynamic pressures. Tests were carried out on a 1/60th scale model under sinusoidal as well as simulated earthquake ground motions. Tests and analytical results regarding natural frequencies, surface water displacements, and dynamic pressures were compared and a good agreement was found within the range of displacements studied. The computer program gave satisfactory results within a maximum range of sloshing displacements in the full-size prototype of 30 in. which is greater than the value obtained under the full intensity of the El Centro earthquake (N-S component 1940). The range of linear behavior was studied experimentally by subjecting the torus model to increasing intensities of the El Centro earthquake

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

International Nuclear Information System (INIS)

1987-11-01

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

Decontamination and decommissioning of the Experimental Boiling Water Reactor (EBWR): Project final report, Argonne National Laboratory

International Nuclear Information System (INIS)

Fellhauer, C.R.; Boing, L.E.; Aldana, J.

1997-03-01

The Final Report for the Decontamination and Decommissioning (D ampersand D) of the Argonne National Laboratory - East (ANL-E) Experimental Boiling Water Reactor (EBWR) facility contains the descriptions and evaluations of the activities and the results of the EBWR D ampersand D project. It provides the following information: (1) An overall description of the ANL-E site and EBWR facility. (2) The history of the EBWR facility. (3) A description of the D ampersand D activities conducted during the EBWR project. (4) A summary of the final status of the facility, including the final and confirmation surveys. (5) A summary of the final cost, schedule, and personnel exposure associated with the project, including a summary of the total waste generated. This project report covers the entire EBWR D ampersand D project, from the initiation of Phase I activities to final project closeout. After the confirmation survey, the EBWR facility was released as a open-quotes Radiologically Controlled Area,close quotes noting residual elevated activity remains in inaccessible areas. However, exposure levels in accessible areas are at background levels. Personnel working in accessible areas do not need Radiation Work Permits, radiation monitors, or other radiological controls. Planned use for the containment structure is as an interim transuranic waste storage facility (after conversion)

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

International Nuclear Information System (INIS)

Ha, Sang Jun; No, Hee Cheon

1997-01-01

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

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

International Nuclear Information System (INIS)

Ohtake, Hiroyasu; Wada, Noriyoshi; Koizumi, Yasuo

2002-01-01

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

Physical quality of Simental Ongole crossbred silverside meat at various boiling times

Science.gov (United States)

Riyanto, J.; Cahyadi, M.; Guntari, W. S.

2018-03-01

This study aims to determine the physical quality of silverside beef meat at various boiling times. Samples that have been used are the back thigh or silverside meat. Treatment of boiling meat included TR (meat without boiled), R15 (boiled 15 minutes), and R30 (boiled for 30 minutes). The experimental design using Completely Randomized Design with 3 replications. Each replication was done in triple physical quality test. Determination of physical quality was performed at the Livestock Industry and Processing Laboratory at Sebelas Maret University Surakarta and the Meat Technology Laboratory at the Faculty of Animal Husbandry of Gadjah Mada University. The result of variance analysis showed that boiling affect cooking loss (P≥0.05) and but did not affect (P≤0,05) pH, water holding capacity and meat tenderness. The conclusions of the study showed that boiling for 15 minutes and 30 minutes decreased the cooking loss of Simental Ongole Crossbred silverside meat. Meat physical quality of pH, water holding capacity and the value of tenderness is not affected by boiling for 15 and 30 minutes.

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

International Nuclear Information System (INIS)

Tomita, M.; Haruyama, Y.; Saito, M.

1993-01-01

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 cm 2 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.)

Dry patch formed boiling and burnout in potassium pool boiling

International Nuclear Information System (INIS)

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

1986-01-01

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

Basic philosophy of the safety design of the Toshiba boiling water reactor

International Nuclear Information System (INIS)

Sato, T.

1992-01-01

This paper discusses the safety design of the Toshiba Boiling Water Reactor (TOSBWR) which was created ∼8 years ago. The design concept is intermediate between conventional boiling water reactors (BWRs) and the advanced BWR (ABWR). It utilizes internal pumps and fine motion control rod drive, but the emergency core cooling system (ECCS) configuration is different from both conventional BWRs and the ABWR. The plant output is 1350 MW (electric). The design is based on two important philosophies: the positive cost reduction philosophy and the constant risk philosophy

Boiling water reactor fuel bundle

International Nuclear Information System (INIS)

Weitzberg, A.

1986-01-01

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

The flooding phenomenon and its connection with dry-out in boiling particle beds

International Nuclear Information System (INIS)

Macbeth, R.V.; Anderson, A.R.

1986-03-01

Experimental studies of boiling in particle beds representing reactor core debris have been restricted to very small beds compared with those that may be found in a reactor. The use of air and water to simulate some of the many features of boiling in a particle bed has given results that were inconclusive. The work reported here is that carried out at Winfrith to extend the dryout data to larger particle diameters, and to provide new experimental data which removes earlier doubts, and makes the air-water analogue position much clearer. (U.K.)

Modeling and measurement of boiling point elevation during water vaporization from aqueous urea for SCR applications

International Nuclear Information System (INIS)

Dan, Ho Jin; Lee, Joon Sik

2016-01-01

Understanding of water vaporization is the first step to anticipate the conversion process of urea into ammonia in the exhaust stream. As aqueous urea is a mixture and the urea in the mixture acts as a non-volatile solute, its colligative properties should be considered during water vaporization. The elevation of boiling point for urea water solution is measured with respect to urea mole fraction. With the boiling-point elevation relation, a model for water vaporization is proposed underlining the correction of the heat of vaporization of water in the urea water mixture due to the enthalpy of urea dissolution in water. The model is verified by the experiments of water vaporization as well. Finally, the water vaporization model is applied to the water vaporization of aqueous urea droplets. It is shown that urea decomposition can begin before water evaporation finishes due to the boiling-point elevation

Modeling and measurement of boiling point elevation during water vaporization from aqueous urea for SCR applications

Energy Technology Data Exchange (ETDEWEB)

Dan, Ho Jin; Lee, Joon Sik [Seoul National University, Seoul (Korea, Republic of)

2016-03-15

Understanding of water vaporization is the first step to anticipate the conversion process of urea into ammonia in the exhaust stream. As aqueous urea is a mixture and the urea in the mixture acts as a non-volatile solute, its colligative properties should be considered during water vaporization. The elevation of boiling point for urea water solution is measured with respect to urea mole fraction. With the boiling-point elevation relation, a model for water vaporization is proposed underlining the correction of the heat of vaporization of water in the urea water mixture due to the enthalpy of urea dissolution in water. The model is verified by the experiments of water vaporization as well. Finally, the water vaporization model is applied to the water vaporization of aqueous urea droplets. It is shown that urea decomposition can begin before water evaporation finishes due to the boiling-point elevation.

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

Science.gov (United States)

2013-10-24

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

Development of surface wettability characteristics for enhancing pool boiling heat transfer

International Nuclear Information System (INIS)

Kim, Moo Hwan; Jo, Hang Jin

2010-05-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

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

International Nuclear Information System (INIS)

Zhang, Weizhong; Mishima, Kaichiro; Hibiki, Takashi

2007-01-01

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

An Experimental study of Fullerene (C60) Nano-fluids on Pool Boiling Conditions

International Nuclear Information System (INIS)

Melani, Ai; Shin, Byoong Su; Chang, Soon Heung

2009-01-01

Critical heat flux (CHF) is directly related to the performance of the system since CHF limits the heat transfer of a heat transfer system. Significant enhancement of CHF allows reliable operation of equipment with more margins to operational limit and more economic cost saving. The previous results show that the nano-fluids significantly enhanced pool boiling CHF compared to pure water. It was supposed that CHF enhancement was due to increased thermal conductivity of fluids, change of bubble shape and behavior, and nano-particle coating of the boiling surface. The previous researches also show that mainly the pool boiling experiment was employed metal particles. Fullerene (C 60 ) is a novel carbon allotrope that was first discovered in 1985 by a winner noble 'Sir Harold W.Kroto, Richard E. Smalley and Robert F.Curl Jr'. In this study we report the first CHF experiment in pool boiling conditions using Fullerene (C 60 ) nanofluids

Nonlinear dynamics of boiling water reactors

International Nuclear Information System (INIS)

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

1983-01-01

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

The myth of the boiling point.

Science.gov (United States)

Chang, Hasok

2008-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-12-31

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

Boiling in porous media

International Nuclear Information System (INIS)

1998-01-01

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

Photographic and video techniques used in the 1/5-scale Mark I boiling water reactor pressure suppression experiment

International Nuclear Information System (INIS)

Dixon, D.; Lord, D.

1978-01-01

The report provides a description of the techniques and equipment used for the photographic and video recordings of the air test series conducted on the 1/5 scale Mark I boiling water reactor (BWR) pressure suppression experimental facility at Lawrence Livermore Laboratory (LLL) between March 4, 1977, and May 12, 1977. Lighting and water filtering are discussed in the photographic system section and are also applicable to the video system. The appendices contain information from the photographic and video camera logs

An Experimental Study of Pressure Gradients for Flow of Boiling Water in a Vertical Round Duct. (Part 2)

Energy Technology Data Exchange (ETDEWEB)

Becker, Kurt M; Hernborg, Gunnar; Bode, Manfred

1962-03-15

The present report contains the results of the second phase of an experimental investigation concerning frictional pressure gradients for the flow of boiling water in vertical channels. The test section for this phase consisted of an electric heated stainless steel tube of 3120 mm length and 7.76 mm inner diameter. Data were obtained for pressures between 6 and 41 ata, steam qualities between 0 and 70 per cent, flow rates between 0.025 and 0.210 Kg/sec and surface heat flux between 30 and 91 W/cm. The results are in excellent agreement with our earlier data for flow in a 9.93 mm inner diameter ducts which were presented in report AE-69. From the measurements we conclude that in the range investigated the non dimensional pressure gradient ratio, {phi}{sup 2} is independent of mass flow rate, inlet sub-cooling and surface heat flux. On the basis of the measured pressure gradients, the following empirical equation has been established for engineering use, {phi}{sup 2} = 1 + 2400 (x/p){sup 0.96} This equation correlates our data (more than 1000 points) with a discrepancy of less than {+-} 15 per cent.

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.

Experimental determination of residual stress by neutron diffraction in a boiling water reactor core shroud

International Nuclear Information System (INIS)

Payzant, A.; Spooner, S.; Zhu, Xiaojing; Hubbard, C.R.

1996-01-01

Residual strains in a 51 mm (2-inch) thick 304L stainless steel plate have been measured by neutron diffraction and interpreted in terms of residual stress. The plate, measuring (300 mm) in area, was removed from a 6m (20-ft.) diameter unirradiated boiling water reactor core shroud, and included a multiple-pass horizontal weld which joined two of the cylindrical shells which comprise the core shroud. Residual stress mapping was undertaken in the heat affected zone, concentrating on the outside half of the plate thickness. Variations in residual stresses with location appeared consistent with trends expected from finite element calculations, considering that a large fraction of the residual hoop stress was released upon removal of the plate from the core shroud cylinder

Small-scale experimental study of vaporization flux of liquid nitrogen released on water.

Science.gov (United States)

Gopalaswami, Nirupama; Olewski, Tomasz; Véchot, Luc N; Mannan, M Sam

2015-10-30

A small-scale experimental study was conducted using liquid nitrogen to investigate the convective heat transfer behavior of cryogenic liquids released on water. The experiment was performed by spilling five different amounts of liquid nitrogen at different release rates and initial water temperatures. The vaporization mass fluxes of liquid nitrogen were determined directly from the mass loss measured during the experiment. A variation of initial vaporization fluxes and a subsequent shift in heat transfer mechanism were observed with changes in initial water temperature. The initial vaporization fluxes were directly dependent on the liquid nitrogen spill rate. The heat flux from water to liquid nitrogen determined from experimental data was validated with two theoretical correlations for convective boiling. It was also observed from validation with correlations that liquid nitrogen was found to be predominantly in the film boiling regime. The substantial results provide a suitable procedure for predicting the heat flux from water to cryogenic liquids that is required for source term modeling. Copyright © 2015 Elsevier B.V. All rights reserved.

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

International Nuclear Information System (INIS)

Gocmanac, M.; Luxat, J.C.

2012-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-15

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

Characterization of the TIP4P-Ew water model: vapor pressure and boiling point.

Science.gov (United States)

Horn, Hans W; Swope, William C; Pitera, Jed W

2005-11-15

The liquid-vapor-phase equilibrium properties of the previously developed TIP4P-Ew water model have been studied using thermodynamic integration free-energy simulation techniques in the temperature range of 274-400 K. We stress that free-energy results from simulations need to be corrected in order to be compared to the experiment. This is due to the fact that the thermodynamic end states accessible through simulations correspond to fictitious substances (classical rigid liquids and classical rigid ideal gases) while experiments operate on real substances (liquids and real gases, with quantum effects). After applying analytical corrections the vapor pressure curve obtained from simulated free-energy changes is in excellent agreement with the experimental vapor pressure curve. The boiling point of TIP4P-Ew water under ambient pressure is found to be at 370.3+/-1.9 K, about 7 K higher than the boiling point of TIP4P water (363.7+/-5.1 K; from simulations that employ finite range treatment of electrostatic and Lennard-Jones interactions). This is in contrast to the approximately +15 K by which the temperature of the density maximum and the melting temperature of TIP4P-Ew are shifted relative to TIP4P, indicating that the temperature range over which the liquid phase of TIP4P-Ew is stable is narrower than that of TIP4P and resembles more that of real water. The quality of the vapor pressure results highlights the success of TIP4P-Ew in describing the energetic and entropic aspects of intermolecular interactions in liquid water.

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

International Nuclear Information System (INIS)

Yasuo Koizumi; Hiroyasu Ohtake; Tomoyuki Suzuki

2002-01-01

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

Review of boiling water reactor small break loss of coolant accidents

International Nuclear Information System (INIS)

Gururaj, P.M.; Dua, S.S.; Rao, A.S.

1981-01-01

This paper presents a review of the analytical and the experimental work performed by the General Electric Company to determine the performance of boiling water reactors (BWR) following postulated small break accidents (SBA). This review paper addresses the following issues: (1) the response of the BWR following small loss of inventory events; (2) methods of analysis and their justification; (3) necessity, if any, of operator action and the length of time available in which such action can be performed; and (4) operator interface following the SBA event. The results from these SBA studies for different BWR product lines show that even with the multiple system failures assumed, the BWR can successfully withstand an SBA. For a typical BWR/6, it takes the failure of 13 water delivery pumps to cause any significant core heatup. The only operator actions determined to be necessary are simple ones and ample time is available to the operator to perform these actions, if needed

Numerical simulation of progressive inlet orifices in boiling water reactor fuel

International Nuclear Information System (INIS)

Lundgren, Sara

2004-07-01

This thesis was carried out at Forsmark Nuclear Power Plant. The power plant in Forsmark consists of three boiling water reactors (BWR) which produce about 17% of Swedish electricity. In a BWR the nuclear reactions are used to boil water inside the reactor vessel. The water works both as a coolant and as a moderator and the resulting steam is used directly to run the turbines. A problem when running a BWR at low flow conditions is the density wave oscillations that might occur to the water flow inside the fuel assemblies. These oscillations arise due to the connection between power and flow rate in a heated channel with two-phase flow. In order to improve the stability performance of the channel an orifice plate is placed at the inlet of each fuel assembly. Today these orifice plates have sharp edges and a constant resistance coefficient. Experimental work has been done with progressive orifices, the edge of which is half-oval in shape. The advantage of progressive orifices is the lower pressure losses with an increase of the Reynolds number, a similar phenomenon that appears in external flow around curved bodies. Since there are high costs associated with experimental generation of high- temperature and high-pressure data, it is of some interest to be able to reproduce and generate data using Computational Fluid Dynamics (CFD). This work deals with the possibility to use the CFD-code Fluent to do numerical simulations of the flow through progressive orifices. The following conclusions may be drawn from the numerical results: All simulations using Reynolds-Averaged Navier-Stokes (RANS) turbulence models, two-dimensional and three-dimensional, capture an abrupt decrease of the resistance coefficient at higher Reynolds numbers. Two-equation models seem to under-predict the critical Reynolds number. The five-equation Reynolds Stress Model (RSM) gives a critical Reynolds number of the same order of magnitude of that measured in experiments. No major differences have

Critical heat flux in subcooled and low quality boiling

International Nuclear Information System (INIS)

Maroti, L.

1976-06-01

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

Electrochemical measurements and modeling predictions in boiling water reactors under various operating conditions

International Nuclear Information System (INIS)

Indig, M.E.

1991-01-01

One important issue for providing life extension to operating boiling water nuclear reactors (BWRs) is the control of stress corrosion cracking in all sections of the primary coolant circuit. This paper links experimental and theoretical methods that provide understanding and measurements of the critical parameter, the electrochemical potential (ECP), and its application to determining crack growth rate among and within the family of BWRs. Measurement of in-core ECP required the development of a new family of radiation-resistant sensors. With these sensors, ECPs were measured in the core and piping of two operating BWRs. Concurrent crack growth measurements were used to benchmark a crack growth prediction algorithm with measured ECPs

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

International Nuclear Information System (INIS)

Tsuiki, Makoto; Sakurada, Koichi; Yoshida, Hiroyuki.

1988-01-01

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)

An Analysis of Burnout Conditions for Flow of Boiling Water in Vertical Round Ducts

Energy Technology Data Exchange (ETDEWEB)

Becker, Kurt M; Persson, P

1963-06-15

A method of predicting the burnout conditions for flow of boiling water in vertical round ducts is presented. The analysis predicts that the burnout conditions are independent of the L/d-ratio and the inlet temperature, and that the burnout steam quality decreases with increasing surface heat flux and increasing mass velocity. It was also found that the burnout steam quality at low pressures increases with the pressure and reaches a maximum at approximately 70 kg/cm, and thereafter decreases with a further increase of the pressure. The theoretical result compares very well with experimental data from different sources.

An Analysis of Burnout Conditions for Flow of Boiling Water in Vertical Round Ducts

International Nuclear Information System (INIS)

Becker, Kurt M.; Persson, P.

1963-06-01

A method of predicting the burnout conditions for flow of boiling water in vertical round ducts is presented. The analysis predicts that the burnout conditions are independent of the L/d-ratio and the inlet temperature, and that the burnout steam quality decreases with increasing surface heat flux and increasing mass velocity. It was also found that the burnout steam quality at low pressures increases with the pressure and reaches a maximum at approximately 70 kg/cm, and thereafter decreases with a further increase of the pressure. The theoretical result compares very well with experimental data from different sources

Effect of ionite decomposition products on the reactor coolant pH in a boiling-water reactor

International Nuclear Information System (INIS)

Bredikhin, V.Ya.; Moskvin, L.N.

1982-01-01

The effect of products resulting from thermal radiolysis of ionites on water-chemical regime of NPP with RBMK is considered basing on investigations conducted in a boiling type experimental reactor. Data are presented on dynamics of changes in the specific electric conductivity and pH of the coolant following destruction of ion exchange groups and ionite matrix under the effect of reactor radiation. The authors draw a conclusion that radiation destruction of ionito fine disperse suspension or high-molecular soluble compounds in the reactor are, probably, one of the main reasons for variations in pH values of the coolant at NPP in non-correction water chemical regime

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

International Nuclear Information System (INIS)

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

1979-01-01

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

Advanced boiling water reactor

International Nuclear Information System (INIS)

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

1999-01-01

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

Outline of advanced boiling water reactor

International Nuclear Information System (INIS)

Yoshio Matsuo

1987-01-01

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

An Experimental Study of Pressure Gradients for Flow of Boiling Water in Vertical Round Ducts (Part 4)

Energy Technology Data Exchange (ETDEWEB)

Becker, Kurt M; Hernborg, Gunnar; Bode, Manfred

1962-07-01

The present report contains the experimental results from the fourth and last phase of an investigation concerning frictional pressure gradients for flow of boiling water in vertical channels. The test section for this phase consisted of an electric heated stainless steel tube of 3120 mm length and 12.99 mm inner diameter. Data were obtained for pressures between 6 and 10 ata, steam qualities between 0 and 0.70, mass flow rates between 0.04 and 0.164 kg/sec. Only one value of 65 W/cm{sup 2} were used for the surface heat flux. The results are in excellent agreement with our earlier data for flow in 9. 93, 7. 76 and 3. 94 mm inner diameter ducts previously presented, and our conclusions given in those reports have been verified. On the basis of the measured pressure gradients, the following empirical equation has been established for engineering use. {chi}{sup 2} = 1 + 2600*(x/p){sup 0.96} This equation correlates our data within an accuracy of {+-} 15 per cent. Considering the data from all four ducts investigated, we have found that the following equation correlates the data with a discrepancy less than {+-} 20 per cent: {chi}{sup 2} = 1 + 2500*(x/p){sup 0.96} and we conclude that for engineering purposes, the effect of diameter is of no significance.

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

Science.gov (United States)

Demarly, Etienne; Baglietto, Emilio

2017-11-01

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

Spectral shift rod for the boiling water reactor

International Nuclear Information System (INIS)

Yokomizo, O.; Kashiwai, S.; Nishida, K.; Orii, A.; Yamashita, J.; Mochida, T.

1993-01-01

A Boiling Water Reactor core concept has been proposed using a new fuel component called spectral shift rod (SSR). The SSR is a new type of water rod in which a water level is formed during core operation. The water level can be controlled by the core recirculation flow rate. By using SSRs, the reactor can be operated with all control rods withdrawn through the operation cycle as well as that a much larger natural uranium saving is possible due to spectral shift operation than in current BWRs. The steady state and transient characteristics of the SSRs have been examined by experiments and analyses to certify the feasibility. In a reference design, a four times larger spectral shift width as for the current BWR has been obtained. (orig.)

An Experimental Study of Pressure Gradients for Flow of Boiling Water in a Vertical Round Duct. (Part 3)

Energy Technology Data Exchange (ETDEWEB)

Becker, Kurt M; Hernborg, Gunnar; Bode, Manfred

1962-07-01

The present report contains the results of the third phase of an experimental investigation concerning frictional pressure gradients for flow of boiling water in vertical channels. The test section for this phase consisted of an electric heated stainless steel tube of 3120 mm length and 3.94 mm inner diameter. Data were obtained for pressures between 8 and 41 ata, steam qualities between 0 and 58 %, flow rates between 0.0075 and 0.048 kg/sec and surface heat flux between 20 and 83 W/cm. The results are in excellent agreement with our earlier data for flow in 9.93 and 7.76 mm inner diameter ducts which were presented in reports AE-69 and AE-70. The present measurements substantiate our earlier conclusion that the non dimensional pressure gradient ratio, {psi}{sup 2} , is, in the range investigated, independent of mass flow rate, inlet subcooling and surface heat flux. On the basis of the measured pressure gradients, the following empirical equation has been established for engineering use: {psi}{sup 2} = 1 + 2400(x/p){sup 0.96} This equation correlates our data (about 800 points) with a discrepancy less than {+-} 15 per cent and is identical with the corresponding equation obtained from measurements with the 7.76 mm duct.

Decontamination flange film characterization for a boiling water reactor under hydrogen water chemistry

International Nuclear Information System (INIS)

Baston, V.F.; Garbauskas, M.F.; Bozeman, J.

1996-01-01

Stainless steel artifacts removed from a boiling water reactor class 4 plant that operated under hydrogen water chemistry and experienced a difficult decontamination were submitted for oxide film characterization. The results reported for the corrosion film composition and structure are consistent with existing theoretical concepts for stainless steel corrosion, spinel structure site preferences (octahedral or tetrahedral) for transition metal ions, and potential-pH diagrams. The observed zinc effects on film stability and lower cobalt incorporation are also consistent with these theoretical concepts

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

International Nuclear Information System (INIS)

Ohtake, Hiroyasu; Wada, Noriyoshi; Koizumi, Yasuo

2003-01-01

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

SBWR: A simplified boiling water reactor

International Nuclear Information System (INIS)

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

1987-01-01

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

CHF enhancement in pool boiling of nanofluid : effect of nanoparticle-coating on heating surface

International Nuclear Information System (INIS)

Kim, Hyung Dae; Kim, Moo Hwan

2005-01-01

Recently researches to enhance CHF using the nanofluid, a new kind of heat transfer fluid in which nano-particles are uniformly and stably dispersed, were attempted. You showed that nanofluid, containing only 0.005 g/l of alumina nanoparticle, make the dramatic increase (∼200%) in CHF in pool boiling at the pressure of 2.89 psia (Tsat=60 .deg. C). They concluded that the abnormal CHF enhancement of nanofluid cannot be explained with any existing models of CHF. Vassallo performed the experimental studies on pool boiling heat transfer in water-SiO 2 nanofluid under atmospheric pressure. They showed a remarkable increase in CHF for nanofluid and also found that the stable film boiling at temperatures close to the melting point of the boiling surface are achievable with the nanofluid. After the experiments, they observed that the formation of the thin silica coating on the wire heater was occurred. This paper focuses on the experimental study of the effect of nanoparticle-coating on CHF enhancement in pool boiling of nanofluid. In this regard, pool boiling CHF values are measured and compared (a) from bare heater immersed in nanofluid and (b) from nanoparticle-coated heater, which is generated by deposition of suspended nanoparticles during pool boiling of nanofluid, immersed in pure water, and (c) from nanoparticle-coated heater immersed in nanofluid. And the microstructure of each heating surface is investigated from photography taken using SEM

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

The analogy between the bubbling of air into water and nucleate boiling at saturation temperature

International Nuclear Information System (INIS)

Wallis, G.B.

1960-01-01

This paper presents a case for the separate consideration of the hydrodynamic and thermal aspects of nucleate boiling. It is shown how boiling phenomena may be simulated in detail by the use of porous media to introduce air bubbles into water. Points of similarity and equivalence are described and analysed. (author)

A real-time BWR [boiling water reactor] stability measurement system

International Nuclear Information System (INIS)

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

1987-01-01

This paper describes the characteristics of a portable, real-time system used for nonperturbational measurements of stability in boiling water reactors. The algorithm used in this system estimates the closed-loop asymptotic decay ratio using only the naturally occurring neutron noise and it is based on the univariate autoregressive methodology

The mechanism of heat transfer in transition boiling

International Nuclear Information System (INIS)

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

1989-01-01

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

Boiling and burnout phenomena under transient heat input, 1

International Nuclear Information System (INIS)

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

1976-01-01

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

CFD simulation of subcooled flow boiling at low pressure

International Nuclear Information System (INIS)

Koncar, B.; Mavko, B.

2001-01-01

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

Converting high boiling hydrocarbons

Energy Technology Data Exchange (ETDEWEB)

Terrisse, H; DuFour, L

1929-02-12

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

Calculation of steam content in a draught section of a tank-type boiling water cooled reactor

International Nuclear Information System (INIS)

Panajotov, D.P.; Gorburov, V.I.

1989-01-01

Structural and hydrodynamic features of a two-phase flow in a draught section of a tank-type boiling water cooled reactor are considered. A calculated model of the steady flow and methods for determining steam content and phase rate profiles under the maximum steam content at the section axis and at some distance from it are proposed. Steam content distribution by height quantitatively agrees with experimental data for the VK-50 reactor. Calculation technique allows one to obtain steam content and phase rate profiles at the section outlet

Vent clearing during a simulated loss-of-coolant accident in Mark I boiling-water-reactor pressure-suppression system

International Nuclear Information System (INIS)

Pitts, J.H.; McCauley, E.W.

1978-01-01

The response of the pressure-suspension containment system of Mark I boiling-water reactors to a loss-of-coolant accident (LOCA) is being studied. This response is a design basis for light-water nuclear reactors. Part of the study is being carried out on a 1 / 5 -scale experimental facility that models the pressure-suppression containment system of the Peach Bottom 2 nuclear power plant. The test series reported here focused on the initial or air-clearing phase of a hypothetical LOCA. Measured forces, measured pressures, and the hydrodynamic phenomena (observed with high-speed cameras) show a logical interrelationship

Mark I 1/5-scale boiling water reactor pressure suppression experiment. Quick-look report for test numbers 1.0(a) and 1.0(b) performed on March 4 and 8, 1977

International Nuclear Information System (INIS)

McCauley, E.W.; Pitts, J.H.

1977-01-01

The experimental results obtained from pressure suppression experiment numbers 1.0(a) and 1.0(b) that were performed on the Lawrence Livermore Laboratory's 1 / 5 -scale boiling water reactor (BWR) Mark I pressure suppression experimental facility are summarized

An experimental study of flow boiling chf with porous surface coatings and surfactant solutions

International Nuclear Information System (INIS)

Sarwar, Mohammad Sohail

2007-02-01

The boiling crisis or critical heat flux (CHF) phenomenon is an enormously studied topic of the boiling heat transfer. The great interest in the CHF is due to practical motives, since it is desirable to design an equipment (heat exchanger or boiler, etc) to operate at as high a heat flux as possible with optimum heat transfer rates but without the risk of physical burnout. This study consists of two parts of flow boiling CHF experiment: with porous surface coated tubes and by using surfactant solutions as working fluid. In first part, the effect of micro- and nano-porous inside surface coated vertical tubes on the CHF was determined for flow boiling of water in vertical round tubes at atmospheric pressure. CHF was measured for a smooth and three different coated tubes, at mass fluxes of 100∼300 kg/m 2 s and two inlet subcooling temperatures (50 .deg. C and 75 .deg. C). Greater CHF enhancement was found with microporous coatings. Al 2 O 3 microporous coatings with particle size <10 μm and coating thickness of 50 μm showed the best CHF enhancement. The maximum increase in the CHF was about 25% for microporous Al 2 O 3 . A wettability test was performed to study the physical mechanism of increase of CHF with microporous coated surfaces and contact angle was measured for smooth and coated surfaces. Pressure drop measurements were also performed across the coated tubes using the DP-cell apparatus. In second part, surfactant effect on the CHF was determined for water flow boiling at atmospheric pressure in a closed loop filled with solution of tri-sodium phosphate (TSP, Na 3 PO 4 ·12H 2 O). The TSP is usually added to the containment sump water to adjust pH level during accident in nuclear power plants. The CHF was measured for four different surfactant solutions of water in vertical tubes, at different mass fluxes (100 ∼ 500 kg/m 2 s) and two inlet subcooling temperatures (50 .deg. C and 75 .deg. C). Surfactant solutions in the range of 0.05%∼0.2% at low mass

Thermal analysis and design of a passive reflux condenser for the simplified boiling water reactor

International Nuclear Information System (INIS)

Bijlani, C.; Patti, F.; Prasad, V.

1993-01-01

At present, the advanced light water reactors (ALWRS) in the United States are being designed to remove reactor decay heat for a period of 72 h following a postulated loss-of-coolant accident (LOCA). The water in the pools external to the containment is evaporated or boiled off to remove the decay heat. It is presumed that the water in the pools can be replenished within 72 h through operator actions or outside assistance. Some countries in Europe require that the plant be designed to remove the reactor decay heat for a much longer duration than 72 h without external assistance. This paper presents an analysis and design of a passive heat exchanger called a reflux condenser (RC), which was considered for an ALWR-the 600-MW(electric) simplified boiling water reactor. The RC is required to condense the steam formed when the water in the pool in which the passive containment cooling system (PCCS) is immersed boils following a LOCA. The RCs are nuclear non-safety related. This paper presents steady-state performance of an RC at various outdoor air dry-bulb temperatures under still air conditions

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

International Nuclear Information System (INIS)

Kim, M.C.

1981-01-01

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

Power distribution monitoring system in the boiling water cooled reactor core

International Nuclear Information System (INIS)

Leshchenko, Yu.I.; Sadulin, V.P.; Semidotskij, I.I.

1987-01-01

Consideration is being given to the system of physical power distribution monitoring, used during several years in the VK-50 tank type boiling water cooled reactor. Experiments were conducted to measure the ratios of detector prompt and activation currents, coefficients of detector relative sensitivity with respect to neutrons and effective cross sections of 103 Rh interaction with thermal and epithermal neutrons. Mobile self-powered detectors (SPD) with rhodium emitters are used as the power distribution detectors in the considered system. All detectors move simultaneously with constant rate in channels, located in fuel assembly central tubes, when conducting the measurements. It is concluded on the basis of analyzing the obtained data, that investigated system with calibrated SPD enables to monitor the absolute power distribution in fuel assemblies under conditions of boiling water cooled reactor and is independent of thermal engineering measurements conducted by in core instruments

Burnout Conditions for Flow of Boiling Water in Vertical Rod Clusters

Energy Technology Data Exchange (ETDEWEB)

Becker, Kurt M

1962-05-15

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

Burnout Conditions for Flow of Boiling Water in Vertical Rod Clusters

International Nuclear Information System (INIS)

Becker, Kurt M.

1962-05-01

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

Recent computer applications in boiling water reactor power plants

International Nuclear Information System (INIS)

Hiraga, Shoji; Joge, Toshio; Kiyokawa, Kazuhiro; Kato, Kanji; Nigawara, Seiitsu

1976-01-01

Process computers in boiling water reactor power plants have won the position of important equipments for the calculation of the core and plant performances and for data logging. Their application technique is growing larger and larger every year. Here, two systems are introduced; plant diagnostic system and computerized control panel. The plant diagnostic system consists of the part processing the signals from a plant, the operation part mainly composed of a computer to diagnose the operating conditions of each system component using input signal, and the result display (CRT or typewriter). The concept on the indications on control panels in nuclear power plants is changing from ''Plant parameters and to be indicated on panel meters as much as possible'' to ''Only the data required for operation are to be indicated.'' Thus the computerized control panel is attracting attention, in which the process computer for processing the operating information and CRT display are introduced. The experimental model of that panel comprises and operator's console and a chief watchmen's console. Its functions are dialogic data access and the automatic selection of preferential information. (Wakatsuki, Y.)

Power distribution effects on boiling water reactor stability

International Nuclear Information System (INIS)

Damiano, B.; March-Leuba, J.

1989-01-01

The work presented in this paper deals with the effects of spatial power distributions on the stability of boiling water reactors (BWRs). It is shown that a conservative power distribution exists for which the stability is minimal. These results are relevant because they imply that bounding stability calculations are possible and, thus, a worst-possible scenario may be defined for a particular BWR geometry. These bounding calculations may, then, be used to determine the maximum expected limit-cycle peak powers

SWR 1000: The new boiling water reactor power plant concept

International Nuclear Information System (INIS)

Brettschuh, W.

1999-01-01

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)

Technical Basis for Water Chemistry Control of IGSCC in Boiling Water Reactors

Science.gov (United States)

Gordon, Barry; Garcia, Susan

Boiling water reactors (BWRs) operate with very high purity water. However, even the utilization of near theoretical conductivity water cannot prevent intergranular stress corrosion cracking (IGSCC) of sensitized stainless steel, wrought nickel alloys and nickel weld metals under oxygenated conditions. IGSCC can be further accelerated by the presence of certain impurities dissolved in the coolant. The goal of this paper is to present the technical basis for controlling various impurities under both oxygenated, i.e., normal water chemistry (NWC) and deoxygenated, i.e., hydrogen water chemistry (HWC) conditions for mitigation of IGSCC. More specifically, the effects of typical BWR ionic impurities (e.g., sulfate, chloride, nitrate, borate, phosphate, etc.) on IGSCC propensities in both NWC and HWC environments will be discussed. The technical basis for zinc addition to the BWR coolant will also provided along with an in-plant example of the most severe water chemistry transient to date.

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

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

International Nuclear Information System (INIS)

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

1980-06-01

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

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.

Subcooled boiling heat transfer on a finned surface

International Nuclear Information System (INIS)

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

1992-01-01

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

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

International Nuclear Information System (INIS)

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

1991-04-01

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

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...

Steady-state nucleate pool boiling mechanism at low heat fluxes

International Nuclear Information System (INIS)

Bastos, L.E.G.

1979-01-01

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

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

International Nuclear Information System (INIS)

Jansen, Franz

1966-01-01

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

Loss of coolant accident at boiling water reactors

International Nuclear Information System (INIS)

Ramirez G, R.

1975-01-01

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)

TRAC-BD1: transient reactor analysis code for boiling-water systems

International Nuclear Information System (INIS)

Spore, J.W.; Weaver, W.L.; Shumway, R.W.; Giles, M.M.; Phillips, R.E.; Mohr, C.M.; Singer, G.L.; Aguilar, F.; Fischer, S.R.

1981-01-01

The Boiling Water Reactor (BWR) version of the Transient Reactor Analysis Code (TRAC) is being developed at the Idaho National Engineering Laboratory (INEL) to provide an advanced best-estimate predictive capability for the analysis of postulated accidents in BWRs. The TRAC-BD1 program provides the Loss of Coolant Accident (LOCA) analysis capability for BWRs and for many BWR related thermal hydraulic experimental facilities. This code features a three-dimensional treatment of the BWR pressure vessel; a detailed model of a BWR fuel bundle including multirod, multibundle, radiation heat transfer, leakage path modeling capability, flow-regime-dependent constitutive equation treatment, reflood tracking capability for both falling films and bottom flood quench fronts, and consistent treatment of the entire accident sequence. The BWR component models in TRAC-BD1 are described and comparisons with data presented. Application of the code to a BWR6 LOCA is also presented

The interpretation of neutron noise in boiling water reactors

International Nuclear Information System (INIS)

John, T.M.; Singh, O.P.

1985-01-01

Some qualitative results of neutron noise in a boiling water reactor (BWR) are reported. By using one-group theory, it has been shown that the neutron flux fluctuations caused by a distributed source in space, representative of the coolant boiling noise in BWRs, can be considered as made up of two components: The first one, having a global character, is a quickly varying function of frequency and follows the fundamental mode solution in space; the second, called nonglobal (local), follows the spatial variation of noise-source intensity distribution and is independent of frequency for ω γΣ, this component decreases with increasing frequency. The formulation indicates that the global component is quite sensitive to the neutron multiplication factor of the system and, for the local component, the medium behaves like a nonmultiplying one. The global effect is dominant at lower frequencies in a critical system, and the local effect is dominant at higher fre quencies

Characteristics of liquid and boiling sodium flows in heating pin bundles

International Nuclear Information System (INIS)

Menant, Bernard

1976-01-01

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

Experimental investigations of heat transfer during sodium boiling in fuel assembly model in justification of advanced fast reactor safety

International Nuclear Information System (INIS)

Khafizov, R.R.; Poplavskij, V.M.; Rachkov, V.I.; Sorokin, A.P.; Ashurko, Yu.M.; Volkov, A.V.; Ivanov, E.F.; Privezentsev, V.V.

2015-01-01

The experimental facility is built up and investigation of heat exchange during sodium boiling in simulated fast reactor core assembly in conditions of natural and forced circulation with sodium plenum and upper end shield model are conducted. It is shown that in the presence of sodium plenum there is possibility to provide long-term cooling of fuel assembly when heat flux density on the surface of fuel element simulator up to 140 and 170 kW/m 2 in conditions of natural and forced circulation, respectively. The obtained data is used for improving calculational model of sodium boiling process in fuel assembly and calculational code COREMELT verification. It is pointed out that heat transfer coefficients in the case of liquid metal boiling in fuel assemblies are slightly over the ones in the case of liquid metals boiling in pipes and pool boiling [ru

Water chemistry in boiling water reactors - A Leibstadt-specific overview

International Nuclear Information System (INIS)

Sarott, F.-A.

2005-01-01

The boiling water reactor (BWR) consists of two main water circuits: the water-steam cycle and the main cooling water system. In the introduction, the goals and tasks of the BWR plant chemistry are described. The most important objectives are the prevention of system degradation by corrosion and the minimisation of radiation fields. Then a short description of the BWR operation principle, including the water steam cycle, the transport of various impurities by the steam, removing impurities from the condensate, the reactor water clean-up system, the balance of plant and the main cooling water system, is given. Subsequently, the focus is set on the water-steam cycle chemistry. In order to fulfil the somewhat contradictory requirements, the chemical parameters must be well balanced. This is achieved by the water chemistry control method called 'normal water chemistry'. Other additional methods are used for the solution to different problems. The 'zinc addition method' is applied to reduce high radiation levels around the recirculation loops. The 'hydrogen water chemistry method' and the 'noble metal chemical addition method' are used to protect the reactor core components and piping made of stainless steel against stress corrosion cracking. This phenomenon has been observed for about 40 years and is partly due to the strong oxidising conditions in the BWR water. Both mitigation methods are used by the majority of the BWR plants all over the world (including the two Swiss NPPs Muehleberg and Leibstadt). (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-12-15

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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

Numerical analysis on pool boiling using user defined function

Energy Technology Data Exchange (ETDEWEB)

Ryu, Sung Uk; Jeon, Byong Guk; Kim, Seok; Euh, Dong-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

PAFS (passive auxiliary feedwater system) adopted in the APR+ (Advanced Power Reactor Plus) of Korea is one such application. When PAFS is activated with an actuation signal, steam from the steam generator passes through heat exchanger tubes submerged in a water tank of the PAFS. Outside these heat exchanger tubes, nucleate boiling phenomena appears. In the present work, a numerical study is reported on three-dimensional transient state pool boiling of water having an immersed heat source. The velocity vector fields during the decrease in the water level are numerically investigated in a pool, and the accuracy of the results is checked by comparing the experimental results conducted using the PIV techniques by Kim et al. These numerical results can be used as basic research data for an analysis and prediction of the natural circulation phenomena in the cooling tank of the passive safety system in a nuclear power plant.

Prestressed concrete pressure vessels for boiling water reactors

International Nuclear Information System (INIS)

Menon, S.

1979-12-01

Following a general description of the Scandinavian cooperative project on prestressed concrete pressure vessels for boiling water reactors, detailed discussion is given in four appendices of the following aspects: the verification programme of tests and studies, the development and testing of a liner venting system, a preliminary safety philosophy and comparative assessment of cold and hot liners. Vessel failure probability is briefly discussed and some figures presented. The pressure gradients in the vessel wall resulting from various stipulated linear cracks, with a liner venting system are presented graphically. (JIW)

Studies on validation possibilities for computational codes for criticality and burnup calculations of boiling water reactor fuel; Untersuchungen zu Validierungsmoeglichkeiten von Rechencodes fuer Kritikalitaets- und Abbrandrechnungen von Siedewasserreaktor-Brennstoff

Energy Technology Data Exchange (ETDEWEB)

Behler, Matthais; Hannstein, Volker; Kilger, Robert; Sommer, Fabian; Stuke, Maik

2017-06-15

The Application of the method of Burn-up Credit on Boiling Water Reactor fuel is much more complex than in the case of Pressurized Water Reactors due to the increased heterogeneity and complexity of the fuel assemblies. Strongly varying enrichments, complex fuel assembly geometries, partial length fuel rods, and strong axial variations of the moderator density make the verification of conservative irradiation conditions difficult. In this Report, it was investigated whether it is possible to take into account the burn-up in criticality analyses for systems with irradiated Boiling Water Reactor fuel on the basis of freely available experimental data and by additionally applying stochastic methods. In order to achieve this goal, existing methods for stochastic analysis were adapted and further developed in order to being applicable to the specific conditions needed in Boiling Water Reactor analysis. The aim was to gain first insight whether a workable scheme for using burn-up credit in Boiling Water Reactor applications can be derived. Due to the fact that the different relevant quantities, like e.g. moderator density and the axial power profile, are strongly correlated, the GRS-tool SUnCISTT for Monte-Carlo uncertainty quantification was used in the analysis. This tool was coupled to a simplified, consistent model for the irradiation conditions. In contrast to conventional methods, this approach allows to simultaneously analyze all involved effects.

Influence of the heater material on the critical heat load at boiling of liquids on surfaces with different sizes

Science.gov (United States)

Anokhina, E. V.

2010-05-01

Data on critical heat loads q cr for the saturated and unsaturated pool boiling of water and ethanol under atmospheric pressure are reported. It is found experimentally that the critical heat load does not necessarily coincide with the heat load causing burnout of the heater, which should be taken into account. The absolute values of q cr for the boiling of water and ethanol on copper surfaces 65, 80, 100, 120, and 200 μm in diameter; tungsten surface 100 μm in diameter; and nichrome surface 100 μm in diameter are obtained experimentally.

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

International Nuclear Information System (INIS)

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

2013-01-01

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

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

International Nuclear Information System (INIS)

Ha, Sang Jun

1998-02-01

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

Comparison of boiling and chlorination on the quality of stored drinking water and childhood diarrhoea in Indonesian households.

Science.gov (United States)

Fagerli, K; Trivedi, K K; Sodha, S V; Blanton, E; Ati, A; Nguyen, T; Delea, K C; Ainslie, R; Figueroa, M E; Kim, S; Quick, R

2017-11-01

We compared the impact of a commercial chlorination product (brand name Air RahMat) in stored drinking water to traditional boiling practices in Indonesia. We conducted a baseline survey of all households with children 1000 MPN/100 ml (RR 1·86, 95% CI 1·09-3·19) in stored water than in households without detectable E. coli. Although results suggested that Air RahMat water treatment was associated with lower E. coli contamination and diarrhoeal rates among children water treatment by boiling, Air RahMat use remained low.

A model for oxidizing species concentrations in boiling water reactors

International Nuclear Information System (INIS)

Sun, B.; Chexal, B.; Pathania, R.; Chun, J.; Ballinger, R.; Abdollahian, D.

1993-01-01

To evaluate and control the intergranular stress corrosion cracking of boiling water reactor (BWR) vessel internal components requires knowledge of the concentration of oxidizing species that affects the electrochemical potentials in various regions of a BWR. In a BWR flow circuit, as water flows through the radiation field, the radiolysis process and chemical reactions lead to the production of species such as oxygen, hydrogen, and hydrogen peroxide. Since chemistry measurements are difficult inside BWRs, analytical tools have been developed by Ruiz and Lin, Ibe and Uchida and Chun and Ballinger for estimating the concentration of species that provide the necessary input for water chemistry control and material protection

Experimental analysis of refrigerants flow boiling inside small sized microfin tubes

Science.gov (United States)

Diani, Andrea; Rossetto, Luisa

2017-07-01

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

Calculation of limit cycle amplitudes in commercial boiling water reactors

International Nuclear Information System (INIS)

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

1984-01-01

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)

New pool boiling data for water with copper-foam metal at sub-atmospheric pressures: Experiments and correlation

Energy Technology Data Exchange (ETDEWEB)

Choon, Ng Kim; Chakraborty, Anutosh; Aye, Sai Maung; Xiaolin, Wang [Department of Mechanical Engineering, National University of Singapore, 10 Kent Ridge Crescent, Singapore 119260 (Singapore)

2006-08-15

Over the past decades, pool boiling heat transfer of water has been investigated extensively by many scientists and researchers at system pressures varying from atmospheric to near critical pressure. However, at sub-atmospheric pressures conditions there is a dearth of data, particularly when the vapour pressures are less than 10kPa. The authors have conducted a detailed study of pool boiling of water in an evaporator where its system pressure was about 1.8kPa. The heat flux for pool boiling was derived from an uniform radiant heaters up to 5W/cm{sup 2} (or a total heating rate of 125W within an area of 25cm{sup 2}), a region that is of interest for the cooling of CPUs. (author)

Boiling water reactor simulator. Workshop material

International Nuclear Information System (INIS)

2003-01-01

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

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

International Nuclear Information System (INIS)

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

2017-01-01

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

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

... DEPARTMENT OF HEALTH AND HUMAN SERVICES Food and Drug Administration [Docket No. FDA-2010-D-0236] Guidance for Industry: Use of Water by Food Manufacturers in Areas Subject to a Boil-Water Advisory; Availability AGENCY: Food and Drug Administration, HHS. ACTION: Notice. SUMMARY: The Food and Drug...

Aging assessment of Residual Heat Removal systems in Boiling Water Reactors

International Nuclear Information System (INIS)

Lofaro, R.J.; Aggarwal, S.

1992-01-01

The effects of aging on Residual Heat Removal systems in Boiling Water Reactors have been studied as part of the Nuclear Plant Aging Research Program. The aging phenomena has been characterized by analyzing operating experience from various national data bases. In addition, actual plant data was obtained to supplement and validate the data base findings

An experimental and theoretical analysis of void fraction dynamics in a boiling channel

International Nuclear Information System (INIS)

Romberg, T.M.

1977-01-01

This paper describes an experimental and theoretical investigation of the void fraction dynamics at the exit of a test boiling channel which is operated near the 'instability threshold power' (the power level at which coolant flow instabilities occur). Dynamic measurements of the perturbations in channel inlet flow-rate, power input and exit void fraction are analysed using multivariate spectral analysis. The resulting experimental cross-spectral density functions between flow-rate/exit void fraction and power input/exit void fraction agree favourably with those calculated by a linearised hydrodynamic model in the frequency domain. (Author)

Effect of water impurities on stress corrosion cracking in a boiling water reactor

International Nuclear Information System (INIS)

Ljungbery, L.G.; Cubicciotti, D

1985-01-01

A series of stress corrosion tests, including corrosion potential and water chemistry measurements, has been performed in the Swedish Ringhals-1 boiling water reactor. Tests have been run under reactor start-up and reactor power operation with normal reactor water conditions and with alternate water chemistry in which hydrogen is added to the feedwater to suppress stress corrosion cracking. During one alternate water chemistry test, there was significant intergranular corrosion cracking of sensitized stainless specimens. It is shown that nitrate and sulfate, arising from an accidental resin intrusion, are likely causes. Nitrate increases the oxidizing power of the water, and sulfate enhances cracking under oxidizing conditions. During another test under start-up conditions, enhanced transgranular stress corrosion cracking in low alloy steels and possibly initiation of cracking in a nickel base alloy was observed as a result of resin intrusion into the reactor water. The intrusion produced acid and sulfate, which are believed to enhance hydrogen cracking conditions

Liquid-solid contact measurements using a surface thermocouple temperature probe in atmospheric pool boiling water

International Nuclear Information System (INIS)

Lee, L.Y.W.; Chen, J.C.; Nelson, R.A.

1984-01-01

Objective was to apply the technique of using a microthermocouple flush-mounted at the boiling surface for the measurement of the local-surface-temperature history in film and transition boiling on high temperature surfaces. From this measurement direct liquid-solid contact in film and transition boiling regimes was observed. In pool boiling of saturated, distilled, deionized water on an aluminum-coated copper surface, the time-averaged, local-liquid-contact fraction increased with decreasing surface superheat. Average contact duration increased monotonically with decreasing surface superheat, while frequency of liquid contact reached a maximum of approx. 50 contacts/s at a surface superheat of approx. 100 K and decreased gradually to 30 contacts/s near the critical heat flux. The liquid-solid contact duration distribution was dominated by short contacts 4 ms at low surface superheats, passing through a relatively flat contact duration distribution at about 80 0 K. Results of this paper indicate that liquid-solid contacts may be the dominant mechanism for energy transfer in the transition boiling process

Experimental study of the effect of the reduced graphene oxide films on nucleate boiling performances of inclined surfaces

International Nuclear Information System (INIS)

Kim, Ji Hoon; Kong, Byeong Tak; Kim, Ji Min

2016-01-01

For the enhancing the CHF, surface coating techniques are available. Yang et al. performed small scale boiling experiments for the vessel lower head, which was coated by aluminum/copper micro particles. Recently, graphene has received much attention for applications in thermal engineering due to its large thermal conductivity. Ahn et al. used a silicon dioxide substrate, which was coated graphene films, as a heating surface during pool boiling experiments. The graphene films inhibited the formation of hot spots, increasing the CHF. For applying novel material 'Graphene' in nuclear industry, here we investigated the effects of graphene film coatings on boiling performances. The experimental pool boiling facility, copying the geometry of lower head of reactor, was designed for verifying orientation effects. The effects of graphene films coating on varied inclined heater surfaces were investigated. The CHF values were increased at every case, but the increased amounts were decreased for downward heater surfaces. At the downward-facing region, however, coating the RGO films would change the CHF mechanisms and boiling heat transfer performances. Generally, RGO films, made by colloidal fabrication, has defects on each flakes.

Correlation between the solubility of aromatic hydrocarbons in water and micellar solutions, with their normal boiling points

International Nuclear Information System (INIS)

Almgren, M.; Grieser, F.; Powell, J.R.; Thomas, J.K.

1979-01-01

A linear correlation between the logarithm of the solubility in water of aromatic hydrocarbons and their normal boiling points is shown. Similarly, the logarithm of the distribution ratio of aromatic hydrocarbons in aqueous micellar solution is shown to be linearly related to the boiling points of the hydrocarbons. 2 figures, 2 tables

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-04-01

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

Acceleration of two-phase flow by boiling, 1

International Nuclear Information System (INIS)

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

1975-01-01

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

Enhancement of pool boiling heat transfer in water using sintered copper microporous coatings

Energy Technology Data Exchange (ETDEWEB)

Jun, Seong Chul; KIm, Jin Sub; You, Seung M. [Dept. of Mechanical Engineering, The University of Texas at Dallas, Richardson (United States); Son, Dong Gun; KIm, Hwan Yeol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-08-15

Pool boiling heat transfer of water saturated at atmospheric pressure was investigated experimentally on Cu surfaces with high-temperature, thermally-conductive, microporous coatings (HTCMC). The coatings were created by sintering Cu powders on Cu surfaces in a nitrogen gas environment. A parametric study of the effects of particle size and coating thickness was conducted using three average particle sizes (APSs) of 10 μm, 25 μm, and 67 μm and various coating thicknesses. It was found that nucleate boiling heat transfer (NBHT) and critical heat flux (CHF) were enhanced significantly for sintered microporous coatings. This is believed to have resulted from the random porous structures that appear to include reentrant type cavities. The maximum NBHT coefficient was measured to be approximately 400 kW/m2k with APS 67 μm and 296 μm coating thicknesses. This value is approximately eight times higher than that of a plain Cu surface. The maximum CHF observed was 2.1 MW/m2 at APS 67 μm and 428 μm coating thicknesses, which is approximately double the CHF of a plain Cu surface. The enhancement of NBHT and CHF appeared to increase as the particle size increased in the tested range. However, two larger particle sizes (25 μm and 67 μm) showed a similar level of enhancement.

Study of the initiation of subcooled boiling during power transients

International Nuclear Information System (INIS)

VanVleet, R.J.

1985-01-01

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

Technical support to the Nuclear Regulatory Commission for the boiling water reactor blowdown heat transfer program

International Nuclear Information System (INIS)

Rice, R.E.

1976-09-01

Results are presented of studies conducted by Aerojet Nuclear Company (ANC) in FY 1975 to support the Nuclear Regulatory Commission (NRC) on the boiling water reactor blowdown heat transfer (BWR-BDHT) program. The support provided by ANC is that of an independent assessor of the program to ensure that the data obtained are adequate for verification of analytical models used for predicting reactor response to a postulated loss-of-coolant accident. The support included reviews of program plans, objectives, measurements, and actual data. Additional activity included analysis of experimental system performance and evaluation of the RELAP4 computer code as applied to the experiments

U.S. experience with hydrogen water chemistry in boiling water reactors

International Nuclear Information System (INIS)

Cowan, R.L.; Head, R.A.; Indig, M.E.; Ruiz, C.P.; Simpson, J.L.

1988-01-01

Hydrogen water chemistry in boiling water reactors is currently being adopted by many utilities in the U.S., with eleven units having completed preimplementation test programs, four units operating permanently with hydrogen water chemistry, and six other units in the process of installing permanent equipment. Intergranular stress corrosion cracking protection is required for the recirculation piping system and other regions of the BWR systems. The present paper explores progress in predicting and monitoring hydrogen water chemistry response in these areas. Testing has shown that impurities can play an important role in hydrogen water chemistry. Evaluation of their effects are also performed. Both computer modeling and in plant measurements show that each plant will respond uniquely to feedwater hydrogen addition. Thus, each plant has its own unique hydrogen requirement for recirculation system protecion. Furthermore, the modeling, and plant measurements show that different regions of the BWR respond differently to hydrogen injection. Thus, to insure protection of components other than the recirculation systems may require more (or less) hydrogen demand than indicated by the recirculation system measurements. In addition, impurities such as copper can play a significant role in establishing hydrogen demand. (Nogami, K.)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

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

International Nuclear Information System (INIS)

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

1999-01-01

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

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

International Nuclear Information System (INIS)

Horiike, Hiroshi; Kuriyama, Masaaki; Morita, Hiroaki

1982-01-01

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

Boiling heat transfer on single phosphor bronze and copper mesh microstructures

Directory of Open Access Journals (Sweden)

Orman Łukasz J.

2014-03-01

Full Text Available The paper presents experimental results of boiling heat transfer of distilled water and ethyl alcohol on surfaces covered with single layers of wire mesh structures made of phosphor bronze and copper. For each material two kinds of structures have been considered (higher and lower in order to determine the impact of the height of the structure on boiling heat transfer. The wire diameter of the copper meshes was 0,25 mm and 0,32 mm, while of the bronze meshes: 0,20 mm and 0,25 mm. The structures had the same mesh aperture (distance between the wires – 0,50 mm for copper and 0,40 for bronze but different wire diameter and, consequently, different height of the layers. The tests have been performed under ambient pressure in the pool boiling mode. The obtained results indicate a visible impact of the layer height on the boiling heat transfer performance of the analysed microstructures.

Feasibility study on the thorium fueled boiling water breeder reactor

International Nuclear Information System (INIS)

PetrusTakaki, N.

2012-01-01

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)

Boiling point of volatile liquids at various pressures

Directory of Open Access Journals (Sweden)

Luisa Maria Valencia

2017-07-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-04-01

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Experimental study of the effect of the reduced graphene oxide films on nucleate boiling performances of inclined surfaces

Energy Technology Data Exchange (ETDEWEB)

Kim, Ji Hoon; Kong, Byeong Tak [Incheon National University, Incheon (Korea, Republic of); Kim, Ji Min [POSTECH, Pohang (Korea, Republic of); and others

2016-05-15

For the enhancing the CHF, surface coating techniques are available. Yang et al. performed small scale boiling experiments for the vessel lower head, which was coated by aluminum/copper micro particles. Recently, graphene has received much attention for applications in thermal engineering due to its large thermal conductivity. Ahn et al. used a silicon dioxide substrate, which was coated graphene films, as a heating surface during pool boiling experiments. The graphene films inhibited the formation of hot spots, increasing the CHF. For applying novel material 'Graphene' in nuclear industry, here we investigated the effects of graphene film coatings on boiling performances. The experimental pool boiling facility, copying the geometry of lower head of reactor, was designed for verifying orientation effects. The effects of graphene films coating on varied inclined heater surfaces were investigated. The CHF values were increased at every case, but the increased amounts were decreased for downward heater surfaces. At the downward-facing region, however, coating the RGO films would change the CHF mechanisms and boiling heat transfer performances. Generally, RGO films, made by colloidal fabrication, has defects on each flakes.

Outline of the advanced boiling water reactor (ABWR)

International Nuclear Information System (INIS)

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

1986-01-01

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

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

Directory of Open Access Journals (Sweden)

Sarafraz M.M.

2012-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

THE PREDICTION OF VOID VOLUME IN SUBCOOLED NUCLEATE POOL BOILING

Energy Technology Data Exchange (ETDEWEB)

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

1963-11-15

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

Domestic and overseas development of advanced boiling water reactors

International Nuclear Information System (INIS)

Hatazawa, Mamoru; Fuchino, Satoshi; Nakada, Kotaro

2010-01-01

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

Fault tolerant digital control systems for boiling water reactors

International Nuclear Information System (INIS)

Chakraborty, S.; Cash, N.R.

1986-01-01

In a Boiling Water Reactor nuclear power plant, the power generation control function is divided into several systems, each system controlling only a part of the total plant. Presently, each system is controlled by conventional analog or digital logic circuits with little interaction for coordinated control. The advent of microprocessors has allowed the development of distributed fault-tolerant digital controls. The objective is to replace these conventional controls with fault-tolerant digital controls connected together with digital communication links to form a fully integrated nuclear power plant control system

Production induced boiling and cold water entry in the Cerro Prieto geothermal reservoir indicated by chemical and physical measurements

Energy Technology Data Exchange (ETDEWEB)

Grant, M.A. (DSIR, Wellington, New Zealand); Truesdell, A.H.; Manon, A.

1981-01-01

Chemical and physical data suggest that the relatively shallow western part of the Cerro Prieto reservoir is bounded below by low permeability rocks, and above and at the sides by an interface with cooler water. There is no continuous permeability barrier around or immediately above the reservoir. Permeability within the reservoir is dominantly intergranular. Mixture with cooler water rather than boiling is the dominant cooling process in the natural state, and production causes displacement of hot water by cooler water, not by vapor. Local boiling occurs near most wells in response to pressure decreases, but no general vapor zone has formed.

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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

Science.gov (United States)

Merte, Herman, Jr.

1988-01-01

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

Heat-transfer correlations for natural convection boiling

International Nuclear Information System (INIS)

Stephan, K.; Abdelsalam, M.

1980-01-01

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

Controllability studies for an advanced CANDU boiling light water reactor

International Nuclear Information System (INIS)

Lepp, R.M.; Hinds, H.W.

1976-12-01

Bulk controllability studies carried out as part of a conceptual design study of a 1200 MWe CANDU boiling-light-water reactor fuelled with U 235 - or Pu-enriched uranium oxide are outlined. The concept, the various models developed for its simulation on a hybrid computer and the perturbations used to test system controllability, are described. The results show that this concept will have better bulk controllability than similar CANDU-BLW reactors fuelled with natural uranium. (author)

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

International Nuclear Information System (INIS)

Malmazet, Erik de

2009-01-01

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

Elimination of 137Cs from trefoil (leaf and stem), ''Mitsuba'', cryptotaenia japonica hassk, boiled in a distilled and salted waters

International Nuclear Information System (INIS)

Motegi, Misako; Miyake, Sadaaki; Ohsawa, Takashi; Nakazawa, Kiyoaki; Izumo, Yoshiro

1999-01-01

Elimination of 137 Cs from highly accumulated trefoil (leaf and stem) through boiling in distilled and salted water were investigated in relation to study the effect of cooking and processing on biochemical states of radionuclides (RI) contaminating in foods. 137 Cs was hardly eliminated from the trefoil immersed in a distilled water at room temperature (about 15degC) during 10 min. 137 Cs was considerably eliminated from the trefoil when boiled in a distilled water, 0.3-3.0% salt concentration of the water and soy sauce: about 40-60% (after 2 min), 70-85% (5 min) and 80-90% (10 min), respectively. Elimination of 137 Cs in the soy sauce (e.g. 77.0±2.9%, at 1% salt concentration after 10 min) was restrictive comparing to that in the salt water (93.4±2.3%). These results are expected to contribute to evaluate the radiation exposure to man when a boiled trefoil contaminating with 137 Cs was ingested. (author)

Gamma heated subassembly for sodium boiling experiments

International Nuclear Information System (INIS)

Artus, S.C.

1975-01-01

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

Gamma heated subassembly for sodium boiling experiments

Energy Technology Data Exchange (ETDEWEB)

Artus, S.C.

1975-01-01

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

The Nuclear option for U.S. electrical generating capacity additions utilizing boiling water reactor technology

International Nuclear Information System (INIS)

Garrity, T.F.; Wilkins, D.R.

1993-01-01

The technology status of the Advanced Boiling Water (ABWR) and Simplified Boiling Water (SBWR) reactors are presented along with an analysis of the economic potential of advanced nuclear power generation systems based on BWR technology to meet the projected domestic electrical generating capacity need through 2005. The forecasted capacity needs are determined for each domestic North American Electric Reliability Council (NERC) region. Extensive data sets detailing each NERC region's specific generation and load characteristics, and capital and fuel cost parameters are utilized in the economic analysis of the optimal generation additions to meet this need by use of an expansion planning model. In addition to a reference case, several sensitivity cases are performed with regard to capital costs and fuel price escalation

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-01-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

International Nuclear Information System (INIS)

Gavilian-Moreno, Carlos; Espinosa-Paredes, Gilberto

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-04-15

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

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

Directory of Open Access Journals (Sweden)

Carlos J. Gavilán-Moreno

2016-04-01

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

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)

Nucleate boiling heat transfer

International Nuclear Information System (INIS)

Saiz Jabardo, J.M.

2009-01-01

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

Enhancement of pool boiling heat transfer coefficients using carbon nanotubes

International Nuclear Information System (INIS)

Park, Ki Jung; Jung, Dong Soo

2007-01-01

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

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

International Nuclear Information System (INIS)

Becker, Kurt M.

1962-01-01

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

To the analysis of reactor noise in boiling water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1972-01-01

The paper contains some basic thoughts on the problem of neutron flux oscillations in power reactors. The advantages of self-powered detectors and their function are explained. In addition, noise measurements of the boiling water reactors at Lingen and Holden are described, and the possibilities of an employment of vanadium detectors for the analysis of reactor noise are discussed. The final pages of the paper contain a complete list of the author's publications in the field of reactor noise analysis. (RW/AK) [de

Boiling water reactor stability analysis in the time domain

International Nuclear Information System (INIS)

Borkowski, J.A.

1991-01-01

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

Natural Circulation with Boiling

Energy Technology Data Exchange (ETDEWEB)

Mathisen, R P

1967-09-15

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

Stainless steels in boiling water reactors. Corrosion problems and possible solutions

International Nuclear Information System (INIS)

Combrade, P.; Desestret, A.; Leroy, F.; Coriou, H.

1977-01-01

In boiling water reactors, the heat-carrying water may have an up to 0.1 or even 0.2 ppm oxygen content, which can make it highly agressive at operating temperature for stainless steels subject to high physical stresses. Several metallurgical solutions can be considered, and in particular the use of stainless steels having a mixed austenitic-ferritic structure or of standard austenitic steels (18.10 or 18.10 Mo, such as AISI 304 and 316) with carefully controlled carbon and alloy element contents. The behavior of these steels during prolonged tests in water at 288 0 C with a 30 and even 100 ppm oxygen content turned out to be quite satisfactory [fr

Pool-Boiling Heat-Transfer Enhancement on Cylindrical Surfaces with Hybrid Wettable Patterns.

Science.gov (United States)

Kumar C S, Sujith; Chang, Yao Wen; Chen, Ping-Hei

2017-04-10

In this study, pool-boiling heat-transfer experiments were performed to investigate the effect of the number of interlines and the orientation of the hybrid wettable pattern. Hybrid wettable patterns were produced by coating superhydrophilic SiO2 on a masked, hydrophobic, cylindrical copper surface. Using de-ionized (DI) water as the working fluid, pool-boiling heat-transfer studies were conducted on the different surface-treated copper cylinders of a 25-mm diameter and a 40-mm length. The experimental results showed that the number of interlines and the orientation of the hybrid wettable pattern influenced the wall superheat and the HTC. By increasing the number of interlines, the HTC was enhanced when compared to the plain surface. Images obtained from the charge-coupled device (CCD) camera indicated that more bubbles formed on the interlines as compared to other parts. The hybrid wettable pattern with the lowermost section being hydrophobic gave the best heat-transfer coefficient (HTC). The experimental results indicated that the bubble dynamics of the surface is an important factor that determines the nucleate boiling.

The challenge of improving boiling: lessons learned from a randomized controlled trial of water pasteurization and safe storage in Peru.

Science.gov (United States)

Heitzinger, K; Rocha, C A; Quick, R E; Montano, S M; Tilley, D H; Mock, C N; Carrasco, A J; Cabrera, R M; Hawes, S E

2016-07-01

Boiling is the most common method of household water treatment in developing countries; however, it is not always effectively practised. We conducted a randomized controlled trial among 210 households to assess the effectiveness of water pasteurization and safe-storage interventions in reducing Escherichia coli contamination of household drinking water in a water-boiling population in rural Peru. Households were randomized to receive either a safe-storage container or a safe-storage container plus water pasteurization indicator or to a control group. During a 13-week follow-up period, households that received a safe-storage container and water pasteurization indicator did not have a significantly different prevalence of stored drinking-water contamination relative to the control group [prevalence ratio (PR) 1·18, 95% confidence interval (CI) 0·92-1·52]. Similarly, receipt of a safe-storage container alone had no effect on prevalence of contamination (PR 1·02, 95% CI 0·79-1·31). Although use of water pasteurization indicators and locally available storage containers did not increase the safety of household drinking water in this study, future research could illuminate factors that facilitate the effective use of these interventions to improve water quality and reduce the risk of waterborne disease in populations that boil drinking water.

Experiences in stability testing of boiling water reactors

International Nuclear Information System (INIS)

March-Leuba, J.; Otaduy, P.J.

1986-01-01

The purpose of this paper is to summarize experiences with boiling water reactor (BWR) stability testing using noise analysis techniques. These techniques have been studied over an extended period of time, but it has been only recently that they have been well established and generally accepted. This paper contains first a review of the problem of BWR neutronic stability, focusing on its physical causes and its effects on reactor operation. The paper also describes the main techniques used to quantify, from noise measurements, the reactor's stability in terms of a decay ratio. Finally, the main results and experiences obtained from the stability tests performed at the Dresden and the Browns Ferry reactors using noise analysis techniques are summarized

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

Digital control application for the advanced boiling water reactor

International Nuclear Information System (INIS)

Fennern, L.E.; Pearson, T.; Wills, H.D.; Swire Rhodes, L.; Pearson, R.L.

1986-01-01

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

Boiling induced mixed convection in cooling loops

International Nuclear Information System (INIS)

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

2000-01-01

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

A study of forced convective subcooled flow boiling

International Nuclear Information System (INIS)

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

1979-01-01

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

Critical superheats upon boiling of dissociating liquids

International Nuclear Information System (INIS)

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

1985-01-01

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

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

International Nuclear Information System (INIS)

Ladisch, R.

1980-06-01

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

Interfacial area transport of subcooled boiling flow in a vertical annulus

Energy Technology Data Exchange (ETDEWEB)

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

2014-03-15

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

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

International Nuclear Information System (INIS)

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

1992-01-01

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

The Effect of Different Boiling and Filtering Devices on the Concentration of Disinfection By-Products in Tap Water

Directory of Open Access Journals (Sweden)

Glòria Carrasco-Turigas

2013-01-01

Full Text Available Disinfection by-products (DBPs are ubiquitous contaminants in tap drinking water with the potential to produce adverse health effects. Filtering and boiling tap water can lead to changes in the DBP concentrations and modify the exposure through ingestion. Changes in the concentration of 4 individual trihalomethanes (THM4 (chloroform (TCM, bromodichloromethane (BDCM, dibromochloromethane (DBCM, and bromoform (TBM, MX, and bromate were tested when boiling and filtering high bromine-containing tap water from Barcelona. For filtering, we used a pitcher-type filter and a household reverse osmosis filter; for boiling, an electric kettle, a saucepan, and a microwave were used. Samples were taken before and after each treatment to determine the change in the DBP concentration. pH, conductivity, and free/total chlorine were also measured. A large decrease of THM4 (from 48% to 97% and MX concentrations was observed for all experiments. Bromine-containing trihalomethanes were mostly eliminated when filtering while chloroform when boiling. There was a large decrease in the concentration of bromate with reverse osmosis, but there was a little effect in the other experiments. These findings suggest that the exposure to THM4 and MX through ingestion is reduced when using these household appliances, while the decrease of bromate is device dependent. This needs to be considered in the exposure assessment of the epidemiological studies.

Elimination of {sup 137}Cs from trefoil (leaf and stem), ``Mitsuba``, cryptotaenia japonica hassk, boiled in a distilled and salted waters

Energy Technology Data Exchange (ETDEWEB)

Motegi, Misako; Miyake, Sadaaki; Ohsawa, Takashi; Nakazawa, Kiyoaki [Saitama Inst. of Public Health (Japan); Izumo, Yoshiro

1999-07-01

Elimination of {sup 137}Cs from highly accumulated trefoil (leaf and stem) through boiling in distilled and salted water were investigated in relation to study the effect of cooking and processing on biochemical states of radionuclides (RI) contaminating in foods. {sup 137}Cs was hardly eliminated from the trefoil immersed in a distilled water at room temperature (about 15degC) during 10 min. {sup 137}Cs was considerably eliminated from the trefoil when boiled in a distilled water, 0.3-3.0% salt concentration of the water and soy sauce: about 40-60% (after 2 min), 70-85% (5 min) and 80-90% (10 min), respectively. Elimination of {sup 137}Cs in the soy sauce (e.g. 77.0{+-}2.9%, at 1% salt concentration after 10 min) was restrictive comparing to that in the salt water (93.4{+-}2.3%). These results are expected to contribute to evaluate the radiation exposure to man when a boiled trefoil contaminating with {sup 137}Cs was ingested. (author)

Absence of genotoxic activity from milk and water boiled in microwave oven in somatic cells from Drosophila melanogaster; Ausencia da atividade genotoxica do leite e agua, fervidos com microondas, em celulas somaticas de Drosophila melanogaster

Energy Technology Data Exchange (ETDEWEB)

Dias, Cristina das Dores. E-mail: crisddias@yahoo.com.br

2003-07-01

This paper reports an experiment for evaluation of the possible genotoxic effects of food prepared in a microwave oven, through the mutation test and somatic recombination, in wings of Drosophila melanogaster. Two crossing have been performed: a standard cross-ST and a high bioactivation cross - HB resulting in marked trans -heterozygote descendents (MH) and balanced heterozygotes (BH). The 72 hours larvas were fed with water and milk boiled both in the microwave oven and in the traditional way. The MH individual wings were analyzed, where the spots can be induced either by mutation or mitotic recombination. The experiment presented negative results related to the genotoxic effects of the water and milk boiled using the microwave oven, in MH descendents of both crossing. Therefore, under these experimental conditions, genotoxic activity were not presented by milk and water boiled in the microwave oven. However, an extensive study using different techniques is necessary to investigate the action of the food prepared in the microwave oven on the genetic material.

Experimental study on CHF characteristics of water-TiO2 nano-fluids

International Nuclear Information System (INIS)

Kim, Hyung Dae; Kim Moo Hwan; Kim, Jeong Bae

2006-01-01

CHF characteristics of nano-fluids were investigated with different volumetric concentrations of TiO 2 nanoparticles. Pool boiling experiments indicated that the application of nano-fluids, instead of pure water, as a cooling liquid significantly increased the CHF. SEM (Scanning Electron Microscope) observations subsequent to the pool boiling experiments revealed that nanoparticles were coated on the heating surface during pool boiling of nano-fluids. In order to investigate the roles of nanoparticles in CHF enhancement of nano-fluids, pool boiling experiments were performed using (a) a nanoparticle-coated heater, prepared by pool boiling of nano-fluids, immersed in pure water and (b) a nanoparticle-coated heater immersed in nano-fluids. The results demonstrated two different roles of nanoparticles in CHF enhancement using nano-fluids: the effect of nanoparticles coated on the heater surface and the effect of nanoparticles suspended in nano-fluids

Comparison of carbon monoxide levels during heating of ice and water to boiling point with a camping stove.

Science.gov (United States)

Leigh-Smith, Simon; Watt, Ian; McFadyen, Angus; Grant, Stan

2004-01-01

To determine whether using a camping stove to bring a pan of ice to boiling point produces higher carbon monoxide (CO) concentration than would bringing a pan of water to boiling point. The hypothesis was that ice would cause greater CO concentration because of its greater flame-cooling effect and, consequently, more incomplete combustion. This was a randomized, prospective observational study. After an initial pilot study, CO concentration was monitored during 10 trials for each of ice and water. A partially ventilated 200-L cardboard box model was developed and then used inside a chamber at -6 degrees C. Ice temperature and volume, water temperature and volume, pan size, and flame characteristics were all standardized. Temperature of the heated medium was monitored to determine time to boiling point. Carbon monoxide concentration was monitored every 30 seconds for the first 3 minutes, then every minute until the end of each 10-minute trial. There was no significant difference (P > .05) in CO production levels between ice and water. Each achieved a similar mean plateau level of approximately 400 ppm CO concentration with a similar rate of rise. However, significantly higher (P = .014) CO concentration occurred at 4 and 5 minutes when the flame underwent a yellow flare; this occurred only on 3 occasions when ice was the medium. There were no significant differences for CO production between bringing a pan of ice or water to boiling point. In a small number of ice trials, the presence of a yellow flame resulted in high CO concentration. Yellow flares might occur more often with ice or snow melting, but this has not been proven.

Identification of dynamic basins in boiling fluxes

International Nuclear Information System (INIS)

Juanico, L.E.

1997-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Braeuer, H.; Mayinger, F.

1988-10-01

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

An Experimental Study of Pressure Gradients for Flow of Boiling Water in a Vertical Round Duct. (Part 1)

Energy Technology Data Exchange (ETDEWEB)

Becker, Kurt M; Hernborg, Gunnar; Bode, Manfred

1962-03-15

Frictional pressure gradients for flow of boiling water in a vertical channel have been measured in a wide range of variables. The test section consisted of an electrically heated 10 mm inner diameter stainless steel tube of 3120 mm length. Data were obtained for pressures between 6 and 42 ata, steam qualities between 0 and 80 %, flow rates between 0.03 and 0.40 kg/sec and surface heat flux between 24 and 80 W/cm{sup 2}. Preliminary measurements of heat transfer and pressure drop for one phase flow of water showed an excellent agreement with one phase flow theory. Extrapolating our data to 100 % quality, an excellent agreement with one-phase flow theory is also found for this case. The two phase flow results are generally 0 - 40 % higher than the results of Martinelli and Nelson. Extrapolating our data to 137 ata fine agreement is found with the results of Sher and Green. On the basis of the measured pressure gradients, a very simple empirical equation has been established for engineering use. This equation correlates our data (more than 1000 points) with a maximum discrepancy of - 20 % and with an average discrepancy of - 5 %.

Confined boiling of the n-pentane in a horizontal space

International Nuclear Information System (INIS)

Cardoso, E.M.; Passos, J.C.; Stutz, B.

2009-01-01

This paper presents experimental results for saturated nucleated boiling of n-Pentane on a heating surface facing upward, at atmospheric pressure, for different degrees of confinement, s = 0.2, 0.5, 0.7 and s = 13 mm, corresponding to Bond numbers 0.13, 0.32, 0.45 and 8.35. Comparative studies with results from literature, together with analysis of the experimental data allowing the improvement of the experimental apparatus. The results show the enhancement of boiling heat transfer with a decreasing distance s between the heating surface and an unheated surface. The experimental heat transfer coefficients for unconfined boiling, s=13 mm, are compared with three empirical correlations. (author)

Confined boiling of the n-pentane in a horizontal space

Energy Technology Data Exchange (ETDEWEB)

Cardoso, E.M.; Passos, J.C. [Universidade Federal de Santa Catarina (UFSC), Florianopolis, SC (Brazil). Dept. de Engenharia Mecanica. LABSOLAR; Stutz, B. [CNRS, Villeurbanne (France). Institut National des Sciences Appliquees de Lyon. Centre Thermique

2009-07-01

This paper presents experimental results for saturated nucleated boiling of n-Pentane on a heating surface facing upward, at atmospheric pressure, for different degrees of confinement, s = 0.2, 0.5, 0.7 and s = 13 mm, corresponding to Bond numbers 0.13, 0.32, 0.45 and 8.35. Comparative studies with results from literature, together with analysis of the experimental data allowing the improvement of the experimental apparatus. The results show the enhancement of boiling heat transfer with a decreasing distance s between the heating surface and an unheated surface. The experimental heat transfer coefficients for unconfined boiling, s=13 mm, are compared with three empirical correlations. (author)

Overview of activities for the reduction of dose rates in Swiss boiling water reactors

International Nuclear Information System (INIS)

Alder, H.P.; Schenker, E.

1993-01-01

Since March 1990, zinc has been added to the reactor water of the boiling water reactor (BWR) Leibstadt (KKL) and, since January 1991, iron has been added to the BWR Muehleberg (KKM). These changes in reactor water chemistry were accompanied by a comprehensive R+D programme. This paper covers three selected topics: a) the statistical analysis of KKL reactor water data before and after zinc addition; b) the analysis of the KKL reactor water during the 1991 annual shutdown; c) laboratory autoclave tests to clarify the role of water additives on the cobalt deposition on austenitic steel surfaces. (author) 2 figs., 4 tabs

An improved mechanistic critical heat flux model for subcooled flow boiling

Energy Technology Data Exchange (ETDEWEB)

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

1998-12-31

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

An improved mechanistic critical heat flux model for subcooled flow boiling

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

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

Science.gov (United States)

Khoshechin, Mohsen; Salimi, Farhad; Jahangiri, Alireza

2018-04-01

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

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

International Nuclear Information System (INIS)

Geraldo, Issa Cherif; Bose, Tanmoy; Pekpe, Komi Midzodzi; Cassar, Jean-Philippe; Mohanty, A.R.; Paumel, Kévin

2014-01-01

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

Preliminary results from film boiling destabilisation experiments

International Nuclear Information System (INIS)

Naylor, P.

1984-05-01

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

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.

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)

1998-12-31

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.

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

International Nuclear Information System (INIS)

Hyvaerinen, J.; Kouhia, J.

1997-01-01

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

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

Science.gov (United States)

Herman, Cila

1996-01-01

compared to values obtained for the same system without electric fields. Imposing an external electric field holds the promise to improve pool boiling heat transfer in low gravity, since a phase separation force other than gravity is introduced. The goal of our research is to experimentally investigate the potential of EHD and the mechanisms responsible for EHD heat transfer enhancement in boiling in low gravity conditions.

Best estimate approach for the evaluation of critical heat flux phenomenon in the boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

Kaliatka, Tadas; Kaliatka, Algirdas; Uspuras, Eudenijus; Vaisnoras, Mindaugas [Lithuanian Energy Institute, Kaunas (Lithuania); Mochizuki, Hiroyasu; Rooijen, W.F.G. van [Fukui Univ. (Japan). Research Inst. of Nuclear Engineering

2017-05-15

Because of the uncertainties associated with the definition of Critical Heat Flux (CHF), the best estimate approach should be used. In this paper the application of best-estimate approach for the analysis of CHF phenomenon in the boiling water reactors is presented. At first, the nodalization of RBMK-1500, BWR-5 and ABWR fuel assemblies were developed using RELAP5 code. Using developed models the CHF and Critical Heat Flux Ratio (CHFR) for different types of reactors were evaluated. The calculation results of CHF were compared with the well-known experimental data for light water reactors. The uncertainty and sensitivity analysis of ABWR 8 x 8 fuel assembly CHFR calculation result was performed using the GRS (Germany) methodology with the SUSA tool. Finally, the values of Minimum Critical Power Ratio (MCPR) were calculated for RBMK-1500, BWR-5 and ABWR fuel assemblies. The paper demonstrate how, using the results of sensitivity analysis, to receive the MCPR values, which covers all uncertainties and remains best estimated.

Analysis of scrams and forced outages at boiling water reactors

International Nuclear Information System (INIS)

Earle, R.T.; Sullivan, W.P.; Miller, K.R.; Schwegman, W.J.

1980-07-01

This report documents the results of a study of scrams and forced outages at General Electric Boiling Water Reactors (BWRs) operating in the United States. This study was conducted for Sandia Laboratories under a Light Water Reactor Safety Program which it manages for the United States Department of Energy. Operating plant data were used to identify the causes of scrams and forced outages. Causes of scrams and forced outages have been summarized as a function of operating plant and plant age and also ranked according to the number of events per year, outage time per year, and outage time per event. From this ranking, identified potential improvement opportunities were evaluated to determine the associated benefits and impact on plant availability

Investigation of boiling water reactor stability and limit-cycle amplitude

International Nuclear Information System (INIS)

Damiano, B.; March-Leuba, J.A.; Euler, J.A.

1991-01-01

Galerkin's method has been applied to a boiling water reactor (BWR) dynamics model consisting of the point kinetics equations, which describe the neutronics, and a feedback transfer function, which describes the thermal hydraulics. The result is a low-order approximate solution describing BWR behavior during small-amplitude limit-cycle oscillations. The approximate solution has been used to obtain a stability condition, show that the average reactor power must increase during limit-cycle oscillations, and qualitatively determine how changes in transfer function values affect the limit-cycle amplitude. 6 refs., 2 figs., 2 tabs

Safety systems and features of boiling and pressurized water reactors

International Nuclear Information System (INIS)

Khair, H. O. M.

2012-06-01

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)

Boiling-Water Reactor internals aging degradation study

International Nuclear Information System (INIS)

Luk, K.H.

1993-09-01

This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR

Boiling in microchannels: a review of experiment and theory

International Nuclear Information System (INIS)

Thome, John R.

2004-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-07-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Study of film boiling collapse behavior during vapor explosion

International Nuclear Information System (INIS)

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

1996-06-01

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

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

International Nuclear Information System (INIS)

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

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-07-01

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

Experimental study of the structure of vapor phase during boiling of R134a on heat exchange surfaces of heat pump

Science.gov (United States)

Ustinov, D. A.; Sukhikh, A. A.; Sidenkov, D. V.; Ustinov, V. A.

2017-10-01

The heat supply by means of heat pumps is considered now as a rational method of local heating which can lead to economy of primary fuel. At use of low-potential heat, for example, the heat of a ground (5 … 18 °C) or ground waters (8 … 10°C) only small depressing of temperature of these sources (on 3 … 5°C) is possible that demands application of heat exchangers with intensified heatmass transfer surfaces. In thermal laboratory of TOT department the 200 W experimental installation has been developed for research of process of boiling of freon R134a. The principle of action of the installation consists in realisation of reverse thermodynamic cycle and consecutive natural measurement of characteristics of elements of surfaces of heat exchangers of real installations at boiling points of freon from-10°C to +10°C and condensing temperatures from 15°C to 50 °C. The evaporator casing has optical windows for control of process of boiling of freon on ribbed on technology of distorting cut tubes. Temperature measurement in characteristic points of a cycle is provided by copper-constantan thermocouples which by means of ADT are connected to the computer that allows treat results of measurements in a real time mode. The structure of a two-phase flow investigated by means of the optical procedure based on laser technique.

Interaction of the nucleation phenomena at adjacent sites in nucleate boiling

International Nuclear Information System (INIS)

Sultan, M.; Judd, R.L.

1983-01-01

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

Flow dynamics of volume-heated boiling pools

International Nuclear Information System (INIS)

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

1979-01-01

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

CFD for subcooled flow boiling: Simulation of DEBORA experiments

International Nuclear Information System (INIS)

Krepper, Eckhard; Rzehak, Roland

2011-01-01

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

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.

Enhanced Boiling on Micro-Configured Composite Surfaces Under Microgravity Conditions

Science.gov (United States)

Zhang, Nengli; Chai, An-Ti

1999-01-01

bubble detachment manifests itself by a necking process which should not be weakened by reduced gravity. In addition, the composite surfaces introduce no extra pressure drop, no fouling and do not impose significant primary or maintenance costs. All of these suggest that this type of composite is an ideal material for the challenge of accounting for both reliability and economy of the relevant components applied in the ATCSs, the DPSs and other devices in future space missions. The aim of the proposed work is to experimentally investigate high nucleate pool boiling performance on a micro-configured metal-graphite composite surface and to determine the mechanisms of the nucleate boiling heat transfer both experimentally and theoretically. Freon-113 and water will be used as the test liquids to investigate wettability effects on boiling characteristics. The Cu-Gr and Al-Gr composites with various volume fractions of graphite fibers will be tested to obtain the heat transfer characteristic data in the nucleate boiling region and in the CHF regime. In the experiments, the bubble emission and coalescence processes will be recorded by a video camera with a magnifying borescope probe immersed in the working fluid. The temperature profile in the thermal boundary layer on the composite surfaces will be measured by a group of micro thermocouples consisting of four ultra fine micro thermocouples. This instrument was developed and successfully used to measure the temperature profile of evaporating liquid thin layers by the proposers in a study performed at the NASA/Lewis Research Center. A two tier model to explain the nucleate boiling process and the performance enhancement on the composite surfaces has been suggested by the authors. According to the model, the thicknesses of the microlayer and the macrolayer underneath the bubbles and mushrooms, can be estimated by the geometry of the composite surface. The experimental results will be compared to the predictions from the model

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

International Nuclear Information System (INIS)

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

1963-06-01

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

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

International Nuclear Information System (INIS)

Aounallah, Yacine

2008-01-01

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

Boiling temperature measurement for water, methanol, ethanol and their binary mixtures in the presence of a hydrochloric or acetic salt of mono-, di- or tri-ethanolamine at 101.3 kPa

Energy Technology Data Exchange (ETDEWEB)

Wang Junfeng [State Key Lab. of Chem. Resource Eng, College of Chem. Eng., Beijing Univ. of Chem. Tech. Beijing 100029 (China)], E-mail: Licx@mail.buct.edu.cn; Li Xuemei; Meng Hong [College of Chem. Eng.., Beijing Univ. of Chem. Tech. Beijing 100029 (China); Li Chunxi [State Key Lab. of Chem. Resource Eng, College of Chem. Eng., Beijing Univ. of Chem. Tech. Beijing 100029 (China); Wang Zihao [College of Chem. Eng., Beijing Univ. of Chem. Tech. Beijing 100029 (China)

2009-02-15

The boiling temperature at atmospheric pressure were measured for 12 binary systems within the range T = (316 to 379) K and 7 ternary systems using a dual circulation. The systems studied contained water, methanol or ethanol with the following ionic liquids (ILs): monoethanolammonium acetate ([HEMA][Ac]), diethanolammonium acetate ([HDEA][Ac]), triethanolammonium acetate ([HTEA][Ac]) and diethanolammonium chloride ([HDEA]Cl). The experimental VLE results of the IL-containing binary systems were correlated by NRTL equation, and the binary NRTL parameters were used for the prediction of VLE of ternary systems with average absolute deviation of 0.73 K in boiling temperature. The results indicate that [HDEA]Cl can be used as an efficient solvent for the extractive distillation of (ethanol + water) mixture due to its notable salting-out effect, which lower the vapour pressure of water, increase the volatility of ethanol and eliminate the azeotropic phenomenon of the (water + ethanol) mixture at definite IL concentration.

Boiling temperature measurement for water, methanol, ethanol and their binary mixtures in the presence of a hydrochloric or acetic salt of mono-, di- or tri-ethanolamine at 101.3 kPa

International Nuclear Information System (INIS)

Wang Junfeng; Li Xuemei; Meng Hong; Li Chunxi; Wang Zihao

2009-01-01

The boiling temperature at atmospheric pressure were measured for 12 binary systems within the range T = (316 to 379) K and 7 ternary systems using a dual circulation. The systems studied contained water, methanol or ethanol with the following ionic liquids (ILs): monoethanolammonium acetate ([HEMA][Ac]), diethanolammonium acetate ([HDEA][Ac]), triethanolammonium acetate ([HTEA][Ac]) and diethanolammonium chloride ([HDEA]Cl). The experimental VLE results of the IL-containing binary systems were correlated by NRTL equation, and the binary NRTL parameters were used for the prediction of VLE of ternary systems with average absolute deviation of 0.73 K in boiling temperature. The results indicate that [HDEA]Cl can be used as an efficient solvent for the extractive distillation of (ethanol + water) mixture due to its notable salting-out effect, which lower the vapour pressure of water, increase the volatility of ethanol and eliminate the azeotropic phenomenon of the (water + ethanol) mixture at definite IL concentration

Stability monitoring of a natural-circulation-cooled boiling water reactor

International Nuclear Information System (INIS)

Hagen, T.H.J.J. van der.

1989-01-01

Methods for monitoring the stability of a boiling water reactor (BWR) are discussed. Surveillance of BWR stability is of importance as problems were encountered in several large reactors. Moreover, surveying stability allows plant owners to operate at high power with acceptable stability margins. The results of experiments performed on the Dodewaard BWR (the Netherlands) are reported. This type reactor is cooled by natural circulation, a cooling principle that is also being considered for new reactor designs. The stability of this reactor was studied both with deterministic methods and by noise analysis. Three types of stability are distinguished and were investigated separately: reactor-kinetic stability, thermal-hydraulic stability and total-plant stability. It is shown that the Dodewaard reactor has very large stability margins. A simple yet reliable stability criterion is introduced. It can be derived on-line from thhe noise signal of ex-vessel neutron detectors during normal operation. The sensitivity of neutron detectors to in-core flux perturbations - reflected in the field-of-view of the detector - was calculated in order to insure proper stability surveillance. A novel technique is presented which enables the determination of variations of the in-core coolant velocity by noise correlation. The velocity measured was interpreted on the basis of experiments performed on the air/water flow in a model of a BWR coolant channel. It appeared from this analysis that the velocity measured was much higher than the volume-averaged water and air velocities and the volumetric flux. The applicability of the above-mentioned technique to monitoring of local channel-flow stability was tested. It was observed that stability effects on the coolant velocity are masked by other effects originating from the local flow pattern. Experimental and theoretical studies show a shorter effective fuel time constant in a BWR than was assumed. (author). 118 refs.; 73 figs.; 21 tabs

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

Science.gov (United States)

2012-06-15

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

Benchmark of the CASMO-3G/MICROBURN-B codes for Commonwealth Edison boiling water reactors

International Nuclear Information System (INIS)

Wheeler, J.K.; Pallotta, A.S.

1992-01-01

The Commonwealth Edison Company has performed an extensive benchmark against measured data from three boiling water reactors using the Studsvik lattice physics code CASMO-3G and the Siemens Nuclear Power three-dimensional simulator code MICROBURN-B. The measured data of interest for this benchmark are the hot and cold reactivity, and the core power distributions as measured by the traversing incore probe system and gamma scan data for fuel pins and assemblies. A total of nineteen unit-cycles were evaluated. The database included fuel product lines manufactured by General Electric and Siemens Nuclear Power, wit assemblies containing 7 x 7 to 9 x 9 pin configurations, several water rod designs, various enrichments and gadolina loadings, and axially varying lattice designs throughout the enriched portion of the bundle. The results of the benchmark present evidence that the CASMO-3G/MICROBURN-B code package can adequately model the range of fuel and core types in the benchmark, and the codes are acceptable for performing neutronic analyses of Commonwealth Edison's boiling water reactors

Nucleate Boiling Heat Transfer Studied Under Reduced-Gravity Conditions

Science.gov (United States)

Chao, David F.; Hasan, Mohammad M.

2000-01-01

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

Experimental analysis of R134a flow boiling inside a 5 PPI copper foam

Science.gov (United States)

Diani, A.; Mancin, S.; Rossetto, L.

2014-04-01

Heat dissipation is one of the most important issues for the reliability of electronic equipment. Boiling can be a very efficient heat transfer mechanism when used to face with the electronic technology needs of efficient and compact heat sinks. Recently, cellular structured materials both stochastic and periodic, particularly open cell metal foams, have been proposed as possible enhanced surfaces to lower the junction temperatures at high heat fluxes. Up today, most of the research on metal foams only regards single phase flow, whereas the two phase flow is still almost unexplored. This paper presents an experimental study on the heat transfer of R134a during flow boiling inside a 5 PPI (Pores Per linear Inch) copper foam, which is 5 mm high, 10 mm wide and 200 mm long, and it is brazed on a 10 mm thick copper plate. The experimental measurements were carried out by imposing three different heat fluxes (50, 75, and 100 kW m-2) and by varying the refrigerant mass velocity between 50 and 200 kg m-2 s-1 and the vapour quality from 0.2 to 0.90, at constant saturation temperature (30°C). The effects of the refrigerant mass flow rate, heat flux and vapour quality on the heat transfer coefficient, dry out phenomenon, and pressure drop are studied.

Film boiling heat transfer and vapour film collapse for various geometries

International Nuclear Information System (INIS)

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

2005-01-01

Full text of publication follows: Film boiling heat transfer has application to the safe operation of water-cooled nuclear reactors under fault conditions and it has been studied using nickel-plated copper specimens in transient and steady state experiments. In the transient tests the specimens were held in a water flow; in the steady state investigation a specimen was mounted in an essentially quiescent pool of water. The transient investigation was conducted on two spheres with different diameters, two cylindrical specimens of different lengths in parallel flow, a short cylinder in cross flow and two flat plates with different lengths. The heat transfer coefficient, vapour film thickness (which was estimated from the heat transfer coefficient) and heat flux followed a similar behaviour with changing experimental conditions for all specimens studied. The heat transfer coefficient increased and the vapour film thickness and heat flux decreased as the specimen temperature decreased. As the water subcooling increased the heat transfer coefficient and the heat flux increased while the vapour film thickness decreased. The water velocity was found to have little influence on the film boiling heat transfer results except for the short cylinder in cross flow. The sphere diameter was found to affect the heat transfer results; the heat transfer coefficient and the heat flux were larger, for the larger sphere. No significant effect of the cylinder length on the heat transfer data was observed. However, the heat transfer coefficient was higher (and the average vapour film thinner) for the longer plate than for the shorter plate. Three vapour/liquid interface types were observed namely: 'smooth', 'rippled' and 'turbulent' depending largely on specimen and water temperatures. For all specimens, the maximum heat transfer coefficient, minimum heat flux and minimum film boiling temperature, occurring just before vapour film collapse, were found to increase as the water subcooling

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.

Experimental Investigation of Pool Boiling for Single and Double Heaters Using Printed Circuit Board

International Nuclear Information System (INIS)

Han, Won Seok; Lee, Jae Young

2012-01-01

Over the past several decades, a considerable number of studies have been conducted on boiling heat transfer in pool boiling. Boiling heat transfer is used in a variety of cooling applications, such as heat exchangers, high powered electronics, and nuclear reactors. Nucleate boiling is one of the most efficient heat transfer mechanisms in boiling regime, but it is imperative that the critical heat flux(CHF) should not be exceeded. CHF phenomenon leads to a dramatic rise in wall temperature, decreased heat transfer, and material failure. Although numerous attempts have been made by researchers to demonstrate the CHF, there is little agreement with the CHF mechanism. In recent years, many researchers have been focusing on surface condition using nanoparticles and surface enhancements, such as a micro structure and artificial cavities, due to enhancement of the CHF point. Cooke and Kandlikar used chips etched with microchannels to prove that these structure has the most enhancement effect. They found that the most efficient boiling surface is with a larger channel size and deep etch. The purpose of this paper is to evaluate the heat transfer and CHF of double heaters on printed circuit board(PCB) in pool boiling. In addition, bubble dynamics of nucleate boiling were observed with high speed observation on single and double heaters using PCB heater

Corrosion problems in boiling water reactors and their remedies

International Nuclear Information System (INIS)

Rosborg, B.

1989-01-01

This article briefly presents current corrosion problems in boiling water reactors and their remedies. The problems are different forms of environmentally assisted cracking, and the remedies are divided into material-, environment-, and stress-related remedies. The list of problems comprises: intergranular stress corrosion cracking (IGSCC) in weld-sensitized stainless steel piping; IGSCC in cold-bent stainless steel piping; irradiation-assisted stress corrosion cracking (IASCC) in stainless alloys; IGSCC in high-strength stainless alloys. A prospective corrosion problem, as judged from literature references, and one which relates to plant life, is corrosion fatigue in pressure vessel steel, since the reactor pressure vessel is the most critical component in the BWR pressure boundary as regards plant safety. (author)

Return to nucleate boiling

International Nuclear Information System (INIS)

Shumway, R.W.

1985-01-01

This paper presents a collection of TMIN (temperature of return to nucleate boiling) correlations, evaluates them under several conditions, and compares them with a wide range of data. Purpose is to obtain the best one for use in a water reactor safety computer simulator known as TRAC-B. Return to nucleate boiling can occur in a reactor accident at either high or low pressure and flow rates. Most of the correlations yield unrealistic results under some conditions. A new correlation is proposed which overcomes many of the deficiencies

User preferences and willingness to pay for safe drinking water: Experimental evidence from rural Tanzania.

Science.gov (United States)

Burt, Zachary; Njee, Robert M; Mbatia, Yolanda; Msimbe, Veritas; Brown, Joe; Clasen, Thomas F; Malebo, Hamisi M; Ray, Isha

2017-01-01

Almost half of all deaths from drinking microbiologically unsafe water occur in Sub-Saharan Africa. Household water treatment and safe storage (HWTS) systems, when consistently used, can provide safer drinking water and improve health. Social marketing to increase adoption and use of HWTS depends both on the prices of and preferences for these systems. This study included 556 households from rural Tanzania across two low-income districts with low-quality water sources. Over 9 months in 2012 and 2013, we experimentally evaluated consumer preferences for six "low-cost" HWTS options, including boiling, through an ordinal ranking protocol. We estimated consumers' willingness to pay (WTP) for these options, using a modified auction. We allowed respondents to pay for the durable HWTS systems with cash, chickens or mobile money; a significant minority chose chickens as payment. Overall, our participants favored boiling, the ceramic pot filter and, where water was turbid, PuR™ (a combined flocculant-disinfectant). The revealed WTP for all products was far below retail prices, indicating that significant scale-up may need significant subsidies. Our work will inform programs and policies aimed at scaling up HWTS to improve the health of resource-constrained communities that must rely on poor-quality, and sometimes turbid, drinking water sources. Copyright © 2016 Elsevier Ltd. All rights reserved.

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

International Nuclear Information System (INIS)

Bede, M.

1987-01-01

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

Piping benchmark problems for the General Electric Advanced Boiling Water Reactor

International Nuclear Information System (INIS)

Bezler, P.; DeGrassi, G.; Braverman, J.; Wang, Y.K.

1993-08-01

To satisfy the need for verification of the computer programs and modeling techniques that will be used to perform the final piping analyses for an advanced boiling water reactor standard design, three benchmark problems were developed. The problems are representative piping systems subjected to representative dynamic loads with solutions developed using the methods being proposed for analysis for the advanced reactor standard design. It will be required that the combined license holders demonstrate that their solutions to these problems are in agreement with the benchmark problem set

Does pan diameter influence carbon monoxide levels during heating of water to boiling point with a camping stove?

Science.gov (United States)

Leigh-Smith, Simon; Stevenson, Richard; Watt, Martin; Watt, Ian; McFadyen, Angus; Grant, Stan

2004-01-01

To determine whether pan diameter influences carbon monoxide (CO) concentration during heating of water to boiling point with a camping stove. The hypothesis was that increasing pan diameter increases CO concentration because of greater flame dispersal and a larger flame. This was a randomized, prospective study. A Coleman Dual Fuel 533 stove was used to heat pans of water to boiling point, with CO concentration monitored every 30 seconds for 5 minutes. The stove was inside a partially ventilated 200-L cardboard box model that was inside an environmental chamber at -6 degrees C. Water temperature, water volume, and flame characteristics were all standardized. Ten trials were performed for each of 2 pan diameters (base diameters of 165 mm [small] and 220 mm [large]). There was a significant difference (P = .002) between the pans for CO levels at each measurement interval from 60 seconds onward. These differences were markedly larger after 90 seconds, with a mean difference of 185 ppm (95% CI 115, 276 ppm) for all the results from 120 seconds onwards. This study has shown that there is significantly higher CO production with a large-diameter pan compared with a small-diameter pan. These findings were evident by using a camping stove to heat water to boiling point when a maximum blue flame was present throughout. Thus, in enclosed environments it is recommended that small-diameter pans be used in an attempt to prevent high CO levels.

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

International Nuclear Information System (INIS)

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

1992-01-01

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

Calculation of releases of radioactive materials in gaseous and liquid effluents from boiling water reactors (BWR-GALE Code)

International Nuclear Information System (INIS)

Bangart, R.L.; Bell, L.G.; Boegli, J.S.; Burke, W.C.; Lee, J.Y.; Minns, J.L.; Stoddart, P.G.; Weller, R.A.; Collins, J.T.

1978-12-01

The calculational procedures described in the report reflect current NRC staff practice. The methods described will be used in the evaluation of applications for construction permits and operating licenses docketed after January 1, 1979, until this NUREG is revised as a result of additional staff review. The BWR-GALE (Boiling Water Reactor Gaseous and Liquid Effluents) Code is a computerized mathematical model for calculating the release of radioactive material in gaseous and liquid effluents from boiling water reactors (BWRs). The calculations are based on data generated from operating reactors, field tests, laboratory tests, and plant-specific design considerations incorporated to reduce the quantity of radioactive materials that may be released to the environment

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

Energy Technology Data Exchange (ETDEWEB)

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

1998-12-31

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

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

International Nuclear Information System (INIS)

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

1997-01-01

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

Little low-power boiling never hurt anybody

International Nuclear Information System (INIS)

Dunn, F.E.

1985-01-01

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

An Investigation into Water Chemistry in Primary Coolant Circuit of an Advanced Boiling Water Reactor

International Nuclear Information System (INIS)

Wu, Bing-Jhen; Yeh, Tsung-Kuang; Wang, Mei-Ya; Sheu, Rong-Jiun

2012-09-01

To ensure operation safety, an optimization on the coolant chemistry in the primary coolant circuit of a nuclear reactor is essential no matter what type or generation the reactor belongs to. For a better understanding toward the water chemistry in an advanced boiling water reactor (ABWR), such as the one being constructed in the northern part of Taiwan, and for a safer operation of this ABWR, we conducted a proactive, thorough water chemistry analysis prior to the completion of this reactor in this study. A numerical simulation model for water chemistry analyses in ABWRs has been developed, based upon the core technology we established in the past. This core technology for water chemistry modeling is basically an integration of water radiolysis, thermal-hydraulics, and reactor physics. The model, by the name of DEMACE - ABWR, is an improved version of the original DEMACE model and was used for radiolysis and water chemistry prediction in the Longmen ABWR in Taiwan. Predicted results pertinent to the water chemistry variation and the corrosion behavior of structure materials in the primary coolant circuit of this ABWR under rated-power operation were reported in this paper. (authors)

Research on boiling and two-phase flow

International Nuclear Information System (INIS)

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

1977-01-01

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

One component, volume heated, boiling pool thermohydraulics

International Nuclear Information System (INIS)

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

1993-01-01

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

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

Directory of Open Access Journals (Sweden)

Paulo Medina

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

Experimental Investigation of the Combined Effects of Heat Exchanger Geometries on Nucleate Pool Boiling Heat Transfer in a Scaled IRWST

International Nuclear Information System (INIS)

Kang, Myeong Gie; Chun, Moon Hyun

1996-01-01

In an effort to determine the combined effects of major parameters of heat exchanger tubes on the nucleate pool boiling heat transfer in the scaled in-containment refueling water storage tank (IRWST), a total of 1,966 data for q v ersus ΔT has been obtained using various combinations of tube diameters, surface roughness, and tube orientations. The experimental results show that (1) increased surface roughness enhances heat transfer for both horizontal and vertical tubes, (2) the two heat transfer mechanisms, i.e.,enhanced heat transfer for both horizontal and vertical tubes, (2) the two heat transfer mechanisms, i.e., enhanced heat transfer due to liquid agitation by bubbles generated and reduced heat transfer by the formation of large vapor slugs and bubble coalescence are different in two regions of low heat fluxes (q ≤ 50kW/m 2 ) and high heat fluxes (q > 50kW/m 2 ) depending on the orientation of tubes and the degree of surface roughness, and (3) the heat transfer rate decreases as the tube diameter is increased for both horizontal and vertical tubes, but the effect of tube diameter on the nucleate pool boiling heat transfer for vertical tubes is greater than that for horizontal tubes. Two empirical heat transfer correlations for q , one for horizontal tubes and the other for vertical tubes, are obtained in terms of surface roughness (ε) and tube diameter (D). In addition, a simple empirical correlation for nucleate pool boiling heat transfer coefficient (h b ) is obtained as a function of heat flux (q ) only. 9 figs., 4 tabs., 15 refs. (Author)

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

International Nuclear Information System (INIS)

Ohtake, Hiroyasu; Yamazaki, Satoshi; Koizumi, Yasuo

2004-01-01

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

Physical characteristics of GE [General Electric] BWR [boiling-water reactor] fuel assemblies

International Nuclear Information System (INIS)

Moore, R.S.; Notz, K.J.

1989-06-01

The physical characteristics of fuel assemblies manufactured by the General Electric Company for boiling-water reactors are classified and described. The classification into assembly types is based on the GE reactor product line, the Characteristics Data Base (CDB) assembly class, and the GE fuel design. Thirty production assembly types are identified. Detailed physical data are presented for each assembly type in an appendix. Descriptions of special (nonstandard) fuels are also reported. 52 refs., 1 fig., 6 tabs

The determination of superheated layer thickness for boiling incipience in a vertical thermosiphon reboiler

International Nuclear Information System (INIS)

Shamsuzzoha, M.; Kamil, M.; Alam, S.S.

2003-01-01

The characteristics of the incipient boiling for vertical thermosiphon reboiler were examined in detail. At the onset of boiling, liquid film adjacent to the heating surface, the super-heated layers thickness δ * , must attain a threshold value so that the critical bubble nuclei with radius r c can further grow to the point of detachment. Thus, the value of δ * /r c is of primary importance for the superheat calculation. In the present study a semi-empirical equation was proposed for the incipient point of boiling including the effect of submergence. The results predicted from theoretical analysis are consistent with the experimental data available in the literature. All the data for fluids namely, distilled water, toluene and ethylene glycol having different thermophysical properties were correlated with a unified correlation having mean absolute deviation 12.73%. (author)

Knowledge and abilities catalog for nuclear power plant operators: boiling water reactors

International Nuclear Information System (INIS)

1986-09-01

The Knowledge and Abilities Catalog for Nuclear Power Plant Operators: Boiling-Water Reactors (BWR) (NUREG-1123) provides the basis for the development of content-valid licensing examinations for reactor operators (ROs) and senior reactor operators (SROs). The examinations developed using the BWR Catalog and Examiners' Handbook for Developing Operator Licensing Examinations (NUREG-1121) will cover those topics listed under Title 10, Code of Federal Regulations, Part 55. The BWR Catalog contains approximately 7000 knowledge and ability (K/A) statements for ROs and SROs at boiling water reactors. Each K/A statement has been rated for its importance to the safe operation of the plant in a manner ensuring personnel and public health and safety. The BWR K/A Catalog is organized into five major sections: Plant-wide Generic Knowledge and Ability Statements, Plant Systems grouped by Safety Function, Emergency and Abnormal Plant Evolutions, Components, and Theory. The BWR Catalog represents a modification of the form and content of the K/A Catalog for Nuclear Power Plant Operators: Pressurized Water Reactors (NUREG-1122). First, categories of knowledge and ability statements have been redefined. Second, the scope of the definition of emergency and abnormal plant evolutions has been revised in line with a symptom-based approach. Third, K/As related to the operational applications of theory have been incorporated into the delineations for both plant systems and emergency and abnormal plant evolutions, while K/As pertaining to theory fundamental to plant operation have been delineated in a separate theory section. Finally, the components section has been revised

Study on boiling heat transfer of high temperature liquid sodium

International Nuclear Information System (INIS)

Sakurai, Akira

1978-01-01

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

Age-related degradation of boiling water reactor vessel internals

International Nuclear Information System (INIS)

Ware, A.G.; Shah, V.N.

1992-01-01

Researchers at the Idaho National Engineering Laboratory performed an assessment of the aging of the reactor internals in boiling water reactors (BWRs), and identified the unresolved technical issues related to the degradation of these components. The overall life-limiting mechanism is intergranular stress corrosion cracking (IGSCC). Irradiation-assisted stress corrosion cracking, fatigue, and thermal aging embrittlement are other potential degradation mechanisms. Several failures in BWR internals have been caused by a combination of factors such as environment, high residual or preload stresses, and flow-induced vibration. The ASME Code Section XI in-service inspection requirements are insufficient for detecting aging-related degradation at many locations in reactor internals. Many of the potential locations for IGSCC or fatigue are not accessible for inspection. (orig.)

Burnout Conditions for Flow of Boiling Water in Vertical Rod Clusters

Energy Technology Data Exchange (ETDEWEB)

Becker, Kurt M

1962-07-01

The present report deals with the results of the first phase of an experimental investigation of burnout conditions for flow of boiling water in vertical round ducts. Data were obtained in the following ranges of variables. Pressure 2.4

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

Science.gov (United States)

2011-03-16

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0055] Economic Simplified Boiling Water Reactor Standard Design: GE Hitachi Nuclear Energy; Issuance of Final Design Approval The U.S. Nuclear Regulatory Commission has issued a final design approval (FDA) to GE Hitachi Nuclear Energy (GEH) for the economic...

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-15

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Science.gov (United States)

Chen, Dongsheng; Shi, Yumei

2013-10-01

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

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)

Boiling water reactor liquid radioactive waste processing system

International Nuclear Information System (INIS)

Anon.

1977-01-01

The standard sets forth minimum design, construction and performance requirements with due consideration for operation of the liquid radioactive waste processing system for boiling water reactor plants for routine operation including design basis fuel leakage and design basis occurrences. For the purpose of this standard, the liquid radioactive waste processing system begins at the interfaces with the reactor coolant pressure boundary, at the interface valve(s) in lines from other systems and at those sumps and floor drains provided for liquid waste with the potential of containing radioactive material. The system terminates at the point of controlled discharge to the environment, at the point of interface with the waste solidification system and at the point of recycle back to storage for reuse. The standard does not include the reactor coolant clean-up system, fuel pool clean-up system, sanitary waste system, any nonaqueous liquid system or controlled area storm drains

Film Boiling on Downward Quenching Hemisphere of Varying Sizes

Energy Technology Data Exchange (ETDEWEB)

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

2004-04-01

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

Application of hydrogen water chemistry to moderate corrosive circumstances around the reactor pressure vessel bottom of boiling water reactors

International Nuclear Information System (INIS)

Uchida, Shunsuke; Ibe, Eishi; Nakata, Kiyatomo; Fuse, Motomasa; Ohsumi, Katsumi; Takashima, Yoshie

1995-01-01

Many efforts to preserve the structural integrity of major piping, components, and structures in a boiling water reactor (BWR) primary cooling system have been directed toward avoiding intergranular stress corrosion cracking (IGSCC). Application of hydrogen water chemistry (HWC) to moderate corrosive circumstances is a promising approach to preserve the structural integrity during extended lifetimes of BWRs. The benefits of HWC application are (a) avoiding the occurrence of IGSCC on structural materials around the bottom of the crack growth rate, even if microcracks are present on the structural materials. Several disadvantage caused by HWC are evaluated to develop suitable countermeasures prior to HWC application. The advantages and disadvantages of HWC are quantitatively evaluated base on both BWR plant data and laboratory data shown in unclassified publications. Their trade-offs are discussed, and suitable applications of HWC are described. It is concluded that an optimal amount of Hydrogen injected into the feedwater can moderate corrosive circumstances, in the region to be preserved, without serious disadvantages. The conclusions have been drawn by combining experimental and theoretical results. Experiments in BWR plants -- e.g., direct measurements of electrochemical corrosion potential and crack growth rate at the RPV bottom -- are planned that would collect data to support the theoretical considerations

Coolability of degraded core under reflooding conditions in Nordic boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

Lindholm, I; Pekkarinen, E [VTT Energy, Espoo (Finland); Nilsson, L [Studsvik EcoSafe AB, Nykoeping (Sweden); Sjoevall, H [Teollisuuden Voima Oy, Olkiluoto (Finland)

1995-09-01

Present work is part of the first phase of subproject RAK-2.1 of the new Nordic Co-operative Reactor Safety Program, NKS. The first phase comprises reflooding calculations for the boiling water reactors (BWRs) TVO I/II in Finland and Forsmark 3 in Sweden, as a continuation of earlier severe accident analyses which were made in the SIK-2 project. The objective of the core reflooding studies is to evaluate when and how the core is still coolable with water and what are the probable consequences of water cooling. In the following phase of the RAK-2.1 project, recriticality studies will be performed. Conditions for recriticality might occur if control rods have melted away with the fuel rods intact in a shape that critical conditions can be created in reflooding with insufficiently borated water. Core coolability was investigated for two reference plants, TVO I/II and Forsmark 3. The selected accident cases were anticipated station blackout with or without successful depressurization of reactor coolant system (RCS). The effects of the recovery of emergency core cooling (ECC) were studied by varying the starting time of core reflooding. The start of ECC systems were assigned to reaching a maximum cladding temperature: 1400 K, 1600 K, 1800 K and 2000 K in the core. Cases with coolant injection through the downcomer were studied for TVO I/II and both downcomer injection and core top spray were investigated for Forsmark 3. Calculations with three different computer codes: MAAP 4, MELCOR 1.8.3 and SCDA/RELAP5/MOD 3.1 for the basis for the presented reflooding studies. Presently, and experimental programme on core reflooding phenomena has been started in Kernforschungszentrum Karlsruhe in QUENCH test facility. (EG) 17 refs.

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

International Nuclear Information System (INIS)

Mori, Yasuo; Hijikata, Kunio; Iwata, Shoichiro

1976-01-01

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

Experimental investigation of heat transfer of R134a in pool boiling on stainless steel and aluminum tubes

Science.gov (United States)

Wengler, C.; Addy, J.; Luke, A.

2018-03-01

Due to high energy demand required for chemical processes, refrigeration and process industries the increase of efficiency and performance of thermal systems especially evaporators is indispensable. One of the possibilities to meet this purpose are investigations in enhancement of the heat transfer in nucleate boiling where high heat fluxes at low superheat are transferred. In the present work, the heat transfer in pool boiling is investigated with pure R134a over wide ranges of reduced pressures and heat fluxes. The heating materials of the test tubes are aluminum and stainless steel. The influence of the thermal conductivity on the heat transfer coefficients is analysed by the surface roughness of sandblasted surfaces. The heat transfer coefficient increases with increasing thermal conductivity, surface roughness and reduced pressures. The experimental results show a small degradation of the heat transfer coefficients between the two heating materials aluminum and stainless steel. In correlation with the VDI Heat Atlas, the experimental results are matching well with the predictions but do not accurately consider the stainless steel material reference properties.

Electrochemical potential measurements in boiling water reactors; relation to water chemistry and stress corrosion

International Nuclear Information System (INIS)

Indig, M.E.; Cowan, R.L.

1981-01-01

Electrochemical potential measurements were performed in operating boiling water reactors to determine the range of corrosion potentials that exist from cold standby to full power operation and the relationship of these measurements to reactor water chemistry. Once the corrosion potentials were known, experiments were performed in the laboratory under electrochemical control to determine potentials and equivalent dissolved oxygen concentrations where intergranular stress corrosion cracking (IGSCC) would and would not occur on welded Type-304 stainless steel. At 274 0 C, cracking occurred at potentials that were equivalent to dissolved oxygen concentration > 40 to 50 ppb. With decreasing temperature, IGSCC became more difficult and only severely sensitized stainless steel would crack. Recent in-reactor experiments combined with the previous laboratory data, have shown that injection of small concentrations of hydrogen during reactor operation can cause a significant decrease in corrosion potential which should cause immunity to IGSCC. (author)

Materials behavior in alternate (hydrogen) water chemistry in the Ringhals-1 boiling water reactor

International Nuclear Information System (INIS)

Ljungberg, L.G.; Cubicciotti, D.; Trolle, M.

1986-01-01

In-plant studies on the intergranular stress corrosion cracking (IGSCC) of sensitized austenitic stainless steel (SS) have been performed at the Swedish Ringhals-1 boiling water reactor (BWR). The studies have covered the present [full-temperature (normal)] water chemistry (PWC) and the alternate (primary) water chemistry (AWC) with hydrogen addition. The test techniques applied were constant extension rate testing (CERT) and electrochemical potential (ECP) measurements. The program was covered by extensive environment monitoring. The results verify earlier laboratory studies which show that sensitized austenitic SS is susceptible to IGSCC in PWC, but not in AWC. Other pressure-bearing BWR construction materials are not adversely affected by AWC. The boundary conditions in Ringhals-1 have been established for an AWC, which is defined as an environment that does not produce IGSCC in sensitized SS. The results are compared with a similar program at Dresden-2, and the points of agreement and discordance in the results are discussed. The relevance of ECP measurements for the control of AWC is discussed

Enabling Highly Effective Boiling from Superhydrophobic Surfaces

Science.gov (United States)

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

2018-04-01

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

Impact of boiling conditions on the molecular and sensory profile of a vegetable broth.

Science.gov (United States)

Mougin, Alice; Mauroux, Olivier; Matthey-Doret, Walter; Barcos, Eugenia Maria; Beaud, Fernand; Bousbaine, Ahmed; Viton, Florian; Smarrito-Menozzi, Candice

2015-02-11

Low-pressure cooking has recently been identified as an alternative to ambient and high-pressure cooking to provide food with enhanced organoleptic properties. This work investigates the impact of the cooking process at different pressures on the molecular and sensory profile of a vegetable broth. Experimental results showed similar sensory and chemical profiles of vegetable broths when boiling at 0.93 and 1.5 bar, while an enhancement of sulfur volatile compounds correlated with a greater leek content and savory aroma was observed when boiling at low pressure (80 °C/0.48 bar). Thus, low-pressure cooking would allow preserving the most labile volatiles likely due to the lower water boiling temperature and the reduced level of oxygen. This study evidenced chemical and sensory impact of pressure during cooking and demonstrated that the flavor profile of culinary preparations can be enhanced by applying low-pressure conditions.

Experimental investigation of small diameter two-phase closed thermosyphons charged with water, FC-84, FC-77 and FC-3283

International Nuclear Information System (INIS)

Jouhara, Hussam; Robinson, Anthony J.

2010-01-01

An experimental investigation of the performance of thermosyphons charged with water as well as the dielectric heat transfer liquids FC-84, FC-77 and FC-3283 has been carried out. The copper thermosyphon was 200 mm long with an inner diameter of 6 mm, which can be considered quite small compared with the vast majority of thermosyphons reported in the open literature. The evaporator length was 40 mm and the condenser length was 60 mm which corresponds with what might be expected in compact heat exchangers. With water as the working fluid two fluid loadings were investigated, that being 0.6 ml and 1.8 ml, corresponding to approximately half filled and overfilled evaporator section in order to ensure combined pool boiling and thin film evaporation/boiling and pool boiling only conditions, respectively. For the Fluorinert TM liquids, only the higher fill volume was tested as the aim was to investigate pool boiling opposed to thin film evaporation. Generally, the water-charged thermosyphon evaporator and condenser heat transfer characteristics compared well with available predictive correlations and theories. The thermal performance of the water-charged thermosyphon also outperformed the other three working fluids in both the effective thermal resistance as well as maximum heat transport capabilities. Even so, FC-84, the lowest saturation temperature fluid tested, shows marginal improvement in the heat transfer at low operating temperatures. All of the tested Fluorinert TM liquids offer the advantage of being dielectric fluids, which may be better suited for sensitive electronics cooling applications and were all found to provide adequate thermal performance up to approximately 30-50 W after which liquid entrainment compromised their performance.

Dynamic analysis of the condensate feedwater system in boiling water reactor plants

International Nuclear Information System (INIS)

Tanji, J.; Omori, T.

1982-01-01

The computer code, CONFAC, has been developed for dynamic analysis of the condensate feedwater system in boiling water reactor plants. This code simulates the hydrodynamics in the piping system, the pump dynamics, and the feedwater controller in order to clarify the system transient characteristics in such cases as pump trip incidents. Code verification was performed by comparison between analytical results and actual plant operational data. Satisfactory agreement was obtained. With the code, appropriate pump start/stop interlocks were estimated for preventing pump cavitation in pump trip incidents

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

International Nuclear Information System (INIS)

Aritomi, Masanori; Aoki, Shigebumi; Inoue, Akira

1983-01-01

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

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-05-15

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

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

International Nuclear Information System (INIS)

Bessho, Y.; Uchikawa, S.

1985-01-01

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

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

International Nuclear Information System (INIS)

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

2005-01-01

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

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

OpenAIRE

Schlenker, Markus Thomas

2014-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Schlenker, Markus Thomas

2014-12-19

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

Determining the void fraction in draught sections of a boiling water cooled reactor

International Nuclear Information System (INIS)

Fedulin, V.N.; Barolomej, G.G.; Solodkij, V.A.; Shmelev, V.E.

1987-01-01

Consideration is being given to the problem of improving methods for calculation of the void fraction in large channels of cooling system of the boiling water cooled reactor during two-phase unsteady flow. Investigation of the structure of two-phase flow was conducted in draught section of the VK-50 reactor (diameter D=2 m, height H=3). The method for calculation of the void fraction in channels with H/D ratio close to 1 is suggested

Boiling water reactor modeling capabilities of MMS-02

International Nuclear Information System (INIS)

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

1987-01-01

During the development period for the Modular Modeling System (MMS) library modules, the Boiling Water Reactor (BWR) has been the last major component to be addressed. The BWRX module includes models of the reactor core, reactor vessel, and recirculation loop. A pre-release version was made available for utility use in September 1983. Since that time a number of changes have been incorporated in BWRX to (1) improve running time for most transient events of interest, (2) extend its capability to include certain events of interest in reactor safety analysis, and (3) incorporate a variety of improvements to the module interfaces and user input formats. The purposes of this study were to briefly review the module structure and physical models, to point the differences between the MMS-02 BWRX module and the BWRX version previously available in the TESTREV1 library, to provide guidelines for choosing among the various user options, and to present some representative results

Stability analysis on natural circulation boiling water reactors

International Nuclear Information System (INIS)

Metz, Peter

1999-05-01

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

Recent developments in the modeling of boiling heat transfer mechanisms

International Nuclear Information System (INIS)

Podowski, M.Z.

2009-01-01

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

TRAC-B thermal-hydraulic analysis of the Black Fox boiling water reactor

International Nuclear Information System (INIS)

Martin, R.P.

1993-05-01

Thermal-hydraulic analyses of six hypothetical accident scenarios for the General Electric Black Fox Nuclear Project boiling water reactor were performed using the TRAC-BF1 computer code. This work is sponsored by the US Nuclear Regulatory Commission and is being done in conjunction with future analysis work at the US Nuclear Regulatory Commission Technical Training Center in Chattanooga, Tennessee. These accident scenarios were chosen to assess and benchmark the thermal-hydraulic capabilities of the Black Fox Nuclear Project simulator at the Technical Training Center to model abnormal transient conditions

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

International Nuclear Information System (INIS)

Bergles, A.E.

1977-01-01

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

Corrosion product deposition on fuel element surfaces of a boiling water reactor

International Nuclear Information System (INIS)

Orlov, A.

2011-01-01

solid solutions. These spinels are well-known for their magnetic behaviour. Since non magnetic zinc ferrite (ZnFe 2 O 4 ) may become magnetic when doped with even small amounts of Ni and/or Mn, their occurrence in the deposit layer has been analyzed. The magnetic permeability of zinc ferrite, trevorite (NiFe 2 O 4 ) and jacobsite (MnFe 2 O 4 ) and their solid solutions are estimated by magnetic moment additivity. The theoretical and literature data are compared with the magnetic permeability of the spinel-type fuel deposition layers gained by reactor pool side Eddy current (EC) analyses. The calculated thicknesses and magnetic permeability values of the deposition layers (estimated by MAGNACROX multifrequency EC method) are compared with the values estimated using an “ion magnetic moment additivity” model. The buildup of corrosion product deposits (crud) on the fuel cladding of the boiling water reactor, Kernkraftwerk Leibstadt (KKL) Switzerland has been investigated by using Gibbs Energy Minimization (GEM-Selector code) calculations of thermodynamic equilibrium at in situ temperatures and pressures, and by applying local experimental PIE techniques. Under the KKL water chemistry conditions, Zn addition together with the presence of Ni and Mn induce the formation of (Zn,Ni,Mn)Fe 2 O 4 spinel solid solutions. GEM calculations applied to the boiling zone match with the EPMA findings, indicating that zinc-rich ferritic spinels are formed on KKL fuel cladding mainly at the lower pin elevations under Zn water chemistry conditions. For the prediction of the crud formation on the fuel cladding, Gibbs energy minimization of the element species during water evaporation in the BWR core were modelled. On the base of the literature data, the solid solution model is proposed and used for the modelling. The modelled processes of water evaporation propose that the spinel phase is formed only at large amount of the evaporated water locally at the pin surface. Finally, all the

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

Energy Technology Data Exchange (ETDEWEB)

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

2009-07-01

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

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

International Nuclear Information System (INIS)

Shekriladze, I.G.

2009-01-01

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

Contribution to the boiling curve of sodium

International Nuclear Information System (INIS)

Schins, H.E.J.

1975-01-01

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

A potential of boiling water power reactors with a natural circulation of a coolant

International Nuclear Information System (INIS)

Osmachkin, V.S.; Sokolov, I.N.

1998-01-01

The use of the natural circulation of coolant in the boiling water reactors simplifies a reactor control and facilities the service of the equipment components. The moderated core power loads allows the long fuel burnup, good control ability and large water stock set up the enhancement of safety level. That is considered to be very important for isolated regions or small countries. In the paper a high safety level and effectiveness of BWRs with natural circulation are reviewed. The limitations of flow stability and protection measures are being discussed. Some recent efforts in designing of such reactors are described.(author)

Multiphysics modeling of two-phase film boiling within porous corrosion deposits

Energy Technology Data Exchange (ETDEWEB)

Jin, Miaomiao, E-mail: mmjin@mit.edu; Short, Michael, E-mail: hereiam@mit.edu

2016-07-01

Porous corrosion deposits on nuclear fuel cladding, known as CRUD, can cause multiple operational problems in light water reactors (LWRs). CRUD can cause accelerated corrosion of the fuel cladding, increase radiation fields and hence greater exposure risk to plant workers once activated, and induce a downward axial power shift causing an imbalance in core power distribution. In order to facilitate a better understanding of CRUD's effects, such as localized high cladding surface temperatures related to accelerated corrosion rates, we describe an improved, fully-coupled, multiphysics model to simulate heat transfer, chemical reactions and transport, and two-phase fluid flow within these deposits. Our new model features a reformed assumption of 2D, two-phase film boiling within the CRUD, correcting earlier models' assumptions of single-phase coolant flow with wick boiling under high heat fluxes. This model helps to better explain observed experimental values of the effective CRUD thermal conductivity. Finally, we propose a more complete set of boiling regimes, or a more detailed mechanism, to explain recent CRUD deposition experiments by suggesting the new concept of double dryout specifically in thick porous media with boiling chimneys. - Highlights: • A two-phase model of CRUD's effects on fuel cladding is developed and improved. • This model eliminates the formerly erroneous assumption of wick boiling. • Higher fuel cladding temperatures are predicted when accounting for two-phase flow. • Double-peaks in thermal conductivity vs. heat flux in experiments are explained. • A “double dryout” mechanism in CRUD is proposed based on the model and experiments.

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

International Nuclear Information System (INIS)

Rachedi, M.

1986-01-01

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

Sloshing of water in annular pressure-suppression pool of boiling water reactors under earthquake ground motions

International Nuclear Information System (INIS)

Aslam, M.; Godden, W.G.; Scalise, D.T.

1979-10-01

This report presents an analytical investigation of the sloshing response of water in annular-circular as well as simple-circular tanks under horizontal earthquake ground motions, and the results are verified with tests. This study was motivated because of the use of annular tanks for pressure-suppression pools in Boiling Water Reactors. Such a pressure-suppression pool would typically have 80 ft and 120 ft inside and outside diameters and a water depth of 20 ft. The analysis was based upon potential flow theory and a computer program was written to obtain time-history plots of sloshing displacements of water and the dynamic pressures. Tests were carried out on 1/80th and 1/15th scale models under sinusoidal as well as simulated earthquake ground motions. Tests and analytical results regarding the natural frequencies, surface water displacements, and dynamic pressures were compared and a good agreement was found for relatively small displacements. The computer program gave satisfactory results as long as the maximum water surface displacements were less than 30 in., which is roughly the value obtained under full intensity of El Centro earthquake

Calculation of optimum control rod operation programme for boiling water reactor

International Nuclear Information System (INIS)

Fehr, L.

1978-01-01

Control rod operation programmes are calculated based on a three dimensional Boiling Water Reactor situation model. The position of the control rods at variosu burn-ups is chosen by an optimisation so that the sum of the square deviations of the load density distribution from an optimum distribution ('Haling' distribution) are minimised. Other conditions are remaining critical and observing the thermal limits for central fuel element melting and critical heat surface loading. As an example, an optimum control rod operation programme for the first cycle in Lengen nuclear power station is calculated and is compared with the programme actually used. (orig.) 891 HP [de

Numerical investigation of boiling heat transfer on hydrocarbon mixture refrigerant in vertical rectangular minichannel

OpenAIRE

Huixing Li; Yu Liu

2016-01-01

In order to investigate the characteristics of boiling heat transfer for hydrocarbon mixture refrigerant in plate-fin heat exchanger which is used in the petrochemical industry field, a model was established on boiling heat transfer in vertical rectangular channel. The simulated results were compared with the experimental data from literature. The results show that the deviation between the simulated results and experimental data is within ±15%. Meanwhile, the characteristic of boiling heat t...

High-Speed Imaging of Explosive Droplet Boiling at the Superheat Limit

Science.gov (United States)

Ferris, F. Robert; Hermanson, Jim; Asadollahi, Arash; Esmaeeli, Asghar

2017-11-01

The explosive boiling processes of droplets of diethyl ether (1-2 mm in diameter) at the superheat limit were examined both experimentally and computationally. Experimentally, droplet explosion was studied using a heated bubble column to bring the test droplet to the superheat limit. The droplet fluid was diethyl ether (superheat limit 147 C at 1 bar) with immiscible glycerol employed as the heated host fluid. Tests were carried out at pressures between 0.5 and 4 bar absolute. The pressure rise associated with the explosive boiling event was captured using a piezoelectric quartz pressure transducer with a 1 MHz DAQ system. High-speed imaging of the interfacial behavior during explosive boiling was performed using a Phantom v12.1 camera at a frame rate of up to one million frames per second with the droplets illuminated by diffuse back-lighting. The imaging reveals features of the Rayleigh-Taylor instability at the vapor-liquid interface resulting from the unstable boiling process. Computationally, Direct Numerical Simulations are performed at Southern Illinois University Carbondale to compliment the experimental tests. NSF Award Number 1511152.

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.

Flow boiling heat transfer at low liquid Reynolds number

International Nuclear Information System (INIS)

Weizhong Zhang; Takashi Hibiki; Kaichiro Mishima

2005-01-01

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

Operational margin monitoring system for boiling water reactor power plants

International Nuclear Information System (INIS)

Fukutomi, S.; Takigawa, Y.

1992-01-01

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

BWR [boiling water reactor] core criticality versus water level during an ATWS [anticipated transient without scram] event

International Nuclear Information System (INIS)

Sehgal, B.R.; Peng, C.M.; Maly, J.

1988-01-01

The BWR [boiling water reactor] emergency procedures guidelines recommend management of core water level to reduce the power generated during an anticipated transient without scram (ATWS) event. BWR power level variation has traditionally been calculated in the system codes using a 1-D [one-dimensional] 2-group neutron kinetics model to determine criticality. This methodology used also for calculating criticality of the partially covered BWR cores has, however, never been validated against data. In this paper, the power level versus water level issues in an ATWS severe accident are introduced and the accuracy of the traditional methodology is investigated by comparing with measured data. It is found that the 1-D 2-group treatment is not adequate for accurate predictions of criticality and therefore the system power level for the water level variations that may be encountered in a prototypical ATWS severe accident. It is believed that the current predictions for power level may be too high