Device to position selectively a tool carried by a vehicle moving on the perforated plate of a tube bundle

International Nuclear Information System (INIS)

Bernardin, M.

1985-01-01

The aim of the invention is an examination device for a tube bundle of an apparatus such as, but not restrictively, a steam generator, situated in a dangerous zone, e.g. radioactive and designed to be introduced into the water box of the said and placed against the perforated plate of the tube bundle by an operator working outside of the said apparatus and able to operate whatever the vertical or horizontal position of the tube plate. The device has a selectively positionable tool - carrying vehicle comprising pistons positioning fingers extendable into the tubes and mounted on extendable supports perpendicular to the pistons and to each other, and an articulated telescopic arm fixed at one end to a rotary mounting on the vehicle and at the other end to an access opening in the vessel containing the tube plate, to hold the vehicle against the plate [fr

Investigation on flow patterns and transition characteristics in a tube-bundle channel

International Nuclear Information System (INIS)

Xiang Wenyuan; Lu Yonghong; Zhao Guisheng

2012-01-01

Tube-bundle channels have been widely used in condenser-evaporator and other industrial heat-exchange equipment. The characteristics of two-phase flow patterns and their transitions for refrigerant R-113 through a vertical tube-bundle channel are experimentally investigated using high-speed camera. Experiments show that there are four main flow patterns in the tube-bundle channel, which are bubbly flow, bubbly-churn flow, churn flow and annular flow. And in the same cross-section of tube- bundle channels, it is shown that there might be different flow patterns in different sub-channels. The flow pattern transitions exhibit unsynchronized in different sub-channels. On the basis of experimental research, the flow pattern map is drawn and analyses are made on the comparison of differences between boiling flow patterns in a circular tube and those in a tube-bundle channel. (authors)

Characteristics of CANDU fuel bundles that caused pressure tube fretting at the bundle midplane

Energy Technology Data Exchange (ETDEWEB)

Dennier, D; Manzer, A M [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Koehn, E [Ontario Hydro, Toronto, ON (Canada)

1996-12-31

Detailed measurements on new bundles, and those that caused fretting during in- and out-reactor tests, have given insight into the factors responsible for fretting at the midplane of the inlet bundle. Bottom fuel elements that were attached near radial endplate spokes and had inboard bearing pads in the rolled joint cavity produced a significant portion of the observed fret marks. These elements are influenced by several driving forces that deflect the centre bearing pads towards the pressure tube surface. The evidence suggests that slight changes in bundle design may be possible to reduce pressure tube fretting. (author). 4 refs., 3 tabs., 8 figs.

Fuel cell integral bundle assembly including ceramic open end seal and vertical and horizontal thermal expansion control

Science.gov (United States)

Zafred, Paolo R [Murrysville, PA; Gillett, James E [Greensburg, PA

2012-04-24

A plurality of integral bundle assemblies contain a top portion with an inlet fuel plenum and a bottom portion containing a base support, the base supports a dense, ceramic air exhaust manifold having four supporting legs, the manifold is below and connects to air feed tubes located in a recuperator zone, the air feed tubes passing into the center of inverted, tubular, elongated, hollow electrically connected solid oxide fuel cells having an open end above a combustion zone into which the air feed tubes pass and a closed end near the inlet fuel plenum, where the open end of the fuel cells rest upon and within a separate combination ceramic seal and bundle support contained in a ceramic support casting, where at least one flexible cushion ceramic band seal located between the recuperator and fuel cells protects and controls horizontal thermal expansion, and where the fuel cells operate in the fuel cell mode and where the base support and bottom ceramic air exhaust manifolds carry from 85% to all of the weight of the generator.

Support and tool displacement device for the attachment of a tube bundle on a tubular plate of a steam generator

International Nuclear Information System (INIS)

Morisot, M.; Werle, R.; Michaud, J.P.

1983-01-01

The steam generator is being assembled, disposed with its axis horizontal and its tubular plate vertical; the device described in this patent, allows to automatize the preparation stages of the tubular plate and the attachment of the bundle, to shorten the construction of the steam generator and to remove drudgeries done by hand on the tubular plate or the tubes of the bundle. The invention can be applied to the construction of PWR steam generators [fr

Effect of power variations across a fuel bundle and within a fuel element on fuel centerline temperature in PHWR bundles in uncrept and crept pressure tubes

International Nuclear Information System (INIS)

Onder, E.N.; Roubtsov, D.; Rao, Y.F.; Wilhelm, B.

2017-01-01

Highlights: • Pressure tube creep effect on fuel pin power and temperatures was investigated. • Noticeable effects were observed for 5.1% crept pressure tube. • Bundle eccentricity effect on power variations was insignificant for uncrept channels. • Difference of 112 °C was observed between top & bottom elements in 5.1% crept channel. • Not discernible fission gas release was expected with temperature difference of 112 °C. - Abstract: The neutron flux and fission power profiles through a fuel bundle and across a fuel element are important aspects of nuclear fuel analysis in multi-scale/multi-physics modelling of Pressurized Heavy Water Reactors (PHWRs) with advanced fuel bundles. Fuel channels in many existing PHWRs are horizontal. With ageing, pressure tubes creep and fuel bundles in these pressure tubes are eccentrically located, which results in an asymmetric coolant flow distribution between the top and bottom of the fuel bundles. The diametral change of the pressure tube due to creep is not constant along the fuel channel; it reaches a maximum in the vicinity of the maximum neutron flux location. The cross-sectional asymmetric positioning of fuel bundles in a crept pressure tube contributes to an asymmetric power distribution within a ring of fuel elements. Modern reactor physics lattice codes (such as WIMS-AECL) are capable of predicting the details of power distribution from basic principles. Thermalhydraulics subchannel codes (such as ASSERT-PV) use models to describe inhomogeneous power distribution within and across fuel elements (e.g., flux tilt model, different powers in different ring elements, or radial power profiles). In this work, physics and thermalhydraulics codes are applied to quantify the effect of eccentricity of a fuel bundle on power variations across it and within a fuel element, and ultimately on the fuel temperature distribution and fuel centerline temperature, which is one of the indicators of fuel performance under normal

Experimental study of thermal–hydraulic performance of cam-shaped tube bundle with staggered arrangement

International Nuclear Information System (INIS)

Bayat, Hamidreza; Lavasani, Arash Mirabdolah; Maarefdoost, Taher

2014-01-01

Highlights: • Thermal–hydraulic performance of a non-circular tube bundle has been investigated experimentally. • Tubes were mounted in staggered arrangement with two longitudinal pitch ratios 1.5 and 2. • Drag coefficient and Nusselt number of tubes in second row was measured. • Friction factor of this tube bundle is lower than circular tube bundle. • Thermal–hydraulic performance of this tube bundle is greater than circular tube bundle. - Abstract: Flow and heat transfer from cam-shaped tube bank in staggered arrangement is studied experimentally. Tubes were located in test section of an open loop wind tunnel with two longitudinal pitch ratios 1.5 and 2. Reynolds number varies in range of 27,000 ⩽ Re D ⩽ 42,500 and tubes surface temperature is between 78 and 85 °C. Results show that both drag coefficient and Nusselt number depends on position of tube in tube bank and Reynolds number. Tubes in the first column have maximum value of drag coefficient, while its Nusselt number is minimum compared to other tubes in tube bank. Moreover, pressure drop from this tube bank is about 92–93% lower than circular tube bank and as a result thermal–hydraulic performance of this tube bank is about 6 times greater than circular tube bank

Modeling fluid forces and response of a tube bundle in cross-flow induced vibrations

International Nuclear Information System (INIS)

Khushnood, Shahab; Khan, Zaffar M.; Malik, M. Afzaal; Koreshi, Zafarullah; Khan, Mahmood Anwar

2003-01-01

Flow induced vibrations occur in process heat exchangers, condensers, boilers and nuclear steam generators. Under certain flow conditions and fluid velocities, the fluid forces result in tube vibrations and possible damage of tube, tube sheet or baffle due to fretting and fatigue. Prediction of these forces is an important consideration. The characteristics of vibration depend greatly on the fluid dynamic forces and structure of the tube bundle. It is undesirable for the tube bundles to vibrate excessively under normal operating conditions because tubes wear and eventual leakage can occur leading to costly shutdowns. In this paper modeling of fluid forces and vibration response of a tube in a heat exchanger bundle has been carried out. Experimental validation has been performed on an existing refinery heat exchanger tube bundle. The target tube has been instrumented with an accelerometer and strain gages. The bundle has been studied for pulse, sinusoidal and random excitations. Natural frequencies and damping of the tubes have also been computed. Experimental fluid forces and response shows a reasonable agreement with the predictions. (author)

Experimental study on heat transfer with condensation of vapors of pure nitrogen tetroxide with nitrogen oxide additions on a bundle of horizontal tubes

International Nuclear Information System (INIS)

Batishcheva, T.M.; Derov, B.T.; Kolykhan, L.I.; Pulyaev, V.F.

1977-01-01

The results of an experimental investigation of heat transfer during condensation of pure N 2 O 4 vapours and with NO admixtures on the outside surface of a bundle of horizontal tubes are considered. The tests with pure N 2 O 4 have been performed at pressures between 0.3-1.0 MPa in the range of thermal loads 22-121 kW/m 2 , temperature heads of 5-33 grades with complete condensation and evaporation. The content of admixtures boiling at high temperatures do not exceed 0.8%. A concentration of noncondensing nitrogen oxide in a gas phase have changed in the range of 3-27%. It is shown, that a concentration of noncondensible NO doesn't result in a considerable decrease of the heat transfer intensity as well as in the case of condensation of vapour-liquid mixtures. The generalized criterion relations are presented

Fuel bundle to pressure tube fretting in Bruce and Darlington

Energy Technology Data Exchange (ETDEWEB)

Norsworthy, A G; Ditschun, A [Atomic Energy of Canada Ltd., Mississauga, ON (Canada)

1996-12-31

As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. (author). 4 refs., 2 tabs., 13 figs.

Fuel bundle to pressure tube fretting in Bruce and Darlington

International Nuclear Information System (INIS)

Norsworthy, A.G.; Ditschun, A.

1995-01-01

As the fuel channel elongates due to creep, the fuel string moves relative to the inlet until the fuel pads at the inboard end eventually separate from the spacer sleeve, and the fuel resides on the burnish mark of the pressure tube. The bundle is then supported in a fashion which contributes to increased levels of vibration. Those pads which (due to geometric variation) have contact loads with the pressure tube within a certain range, vibrate, and cause significant fretting on the burnish mark, and further along at the midplane of the bundle. Inspection of the pressure tubes in Bruce A, Bruce B, and Darlington has revealed fret damage up to 0.55 mm at the burnish mark and slightly lower than this at the inlet bundle midplane. To date, all fret marks have been dealt with successfully without the need for tube replacement, but a program of work has been initiated to understand the mechanism and reduce the fretting. Such understanding is necessary to guide future design changes to the fuel bundle, to guide future inspection programs, to guide maintenance programs, and for longer term strategic planning. This paper discusses how the understanding of fretting has evolved and outlines a current hypothesis for the mechanism of fretting. The role of bundle geometry, excitation forces, and reactor conditions are reviewed, along with options under consideration to mitigate damage. (author). 4 refs., 2 tabs., 13 figs

Application of tube critical heat flux tables to annuli and rod bundles

International Nuclear Information System (INIS)

Ulrych, G.

1985-01-01

The purpose of this paper is to show that tables for the critical heat flux (CHF) in tubes have a much wider range of applicability than only to tubes. With the proper choice of a characteristic length replacing the tube diameter as a parameter the validity of the tables can be expanded to more complex geometries. The paper describes how the tables must be applied to annuli or rod bundles. The data base for comparisons is mainly taken from the open literature. For rod bundles the proposed methodology was checked for very different geometries including rod bundles from very tight hexagonal to extremely open square bundle arrays. It is concluded that the tables give reasonable results for a wide range of hydraulic diameters

Fuel rod bundles proposed for advanced pressure tube nuclear reactors

International Nuclear Information System (INIS)

Prodea, Iosif; Catana, Alexandru

2010-01-01

The paper aims to be a general presentation for fuel bundles to be used in Advanced Pressure Tube Nuclear Reactors (APTNR). The characteristics of such a nuclear reactor resemble those of known advanced pressure tube nuclear reactors like: Advanced CANDU Reactor (ACR TM -1000, pertaining to AECL) and Indian Advanced Heavy Water Reactor (AHWR). We have also developed a fuel bundle proposal which will be referred as ASEU-43 (Advanced Slightly Enriched Uranium with 43 rods). The ASEU-43 main design along with a few neutronic and thermalhydraulic characteristics are presented in the paper versus similar ones from INR Pitesti SEU-43 and CANDU-37 standard fuel bundles. General remarks regarding the advantages of each fuel bundle and their suitability to be burned in an APTNR reactor are also revealed. (authors)

Heat transfer in tube bundles of heat exchangers with flow baffles induced forced mixing

International Nuclear Information System (INIS)

AbuRomia, M.M.; Chu, A.W.; Cho, S.M.

1976-01-01

Thermal analysis of shell-and-tube heat exchangers is being investigated through geometric modeling of the unit configuration in addition to considering the heat transfer processes taking place within the tube bundle. The governing equations that characterize the heat transfer from the shell side fluid to the tube side fluid across the heat transfer tubewalls are indicated. The equations account for the heat transfer due to molecular conduction, turbulent thermal diffusion, and forced fluid mixing among various shell side fluid channels. The analysis, though general in principle, is being applied to the Clinch River Breeder Reactor Plant-Intermediate Heat Exchanger, which utilizes flow baffles appropriately designed for induced forced fluid mixing in the tube bundle. The results of the analysis are presented in terms of the fluid and tube wall temperature distributions of a non-baffled and baffled tube bundle geometry. The former case yields axial flow in the main bundle region while the latter is associated with axial/cross flow in the bundle. The radial components of the axial/cross flow yield the necessary fluid mixing that results in reducing the thermal unbalance among the heat transfer to the allowable limits. The effect of flow maldistribution, present on the tube or shell sides of the heat exchangers, in altering the temperature field of tube bundles is also noted

Damping in heat exchanger tube bundles. A review

International Nuclear Information System (INIS)

Iqbal, Qamar; Khushnood, Shahab; Ghalban, Ali Roheim El; Sheikh, Nadeem Ahmed; Malik, Muhammad Afzaal; Arastu, Asif

2007-01-01

Damping is a major concern in the design and operation of tube bundles with loosely supported tubes in baffles for process shell and tube heat exchangers and steam generators which are used in nuclear, process and power generation industries. System damping has a strong influence on the amplitude of vibration. Damping depends upon the mechanical properties of the tube material, geometry of intermediate supports and the physical properties of shell-side fluid. Type of tube motion, number of supports, tube frequency, vibration amplitude, tube mass or diameter, side loads, support thickness, higher modes, shell-side temperature etc., affect damping in tube bundles. The importance of damping is further highlighted due to current trend of larger exchangers with increased shell-side velocities in modern units. Various damping mechanisms have been identified (Friction damping, Viscous damping, Squeeze film damping, Support damping. Two-Phase damping, and very recent-Thermal damping), which affect the performance of process exchangers and steam generators with respect to flow induced vibration design, including standard design guidelines. Damping in two-phase flow is very complex and highly void fraction, and flow-regime dependent. The current paper focuses on the various known damping mechanisms subjected to both single and two-phase cross-flow in process heat exchangers and steam generators and formulates the design guidelines for safer design. (author)

Generalization of results on experimental study on resistance of ribbed tube bundles

International Nuclear Information System (INIS)

Lokshin, V.A.; Fomina, V.N.

1978-01-01

On a wide experimental basis a new technique of calculating aerodynamic resistance of chess and passage tube bundles is worked out using inner diameter of carrier tube and ribbing coefficient as geometric parameters in formulae. New calculated formulae are based on a wider experimental material. Their structure is more simple. They are in good agreement with formulae for smooth tube bundles in the chess arrangement

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

International Nuclear Information System (INIS)

Royen Van, E.

2011-11-01

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

Tube bundle vibrations due to cross flow under the influence of turbulence

Energy Technology Data Exchange (ETDEWEB)

Popp, K.; Romberg, O. [Institute of Mechanics, University of Hannover (Germany)

1998-10-01

Tube bundles are often used in heat exchangers and chemical reactors. Besides of large heat transfer capacities and small pressure drops in the apparatus a safe design against vibration damages is demanded. For many years extensive investigations concerning the dynamical behaviour of tube bundles subjected to cross-flow have been carried out in the wind tunnel of the Institute of Mechanics at the University of Hannover. In the last years the investigations were concentrated on the experimental investigations of different flow excitation mechanisms in a fully flexible bundle as well as in a bundle with one single flexibly mounted tube in an otherwise fixed array with variable geometry and changing equilibrium position. The aim of the studies was the determination of the stability boundaries, i.e. the critical reduced fluid velocity depending on the reduced damping coefficient in a wide parameter region. Theoretical investigations of the stability behaviour on the basis of an one dimensional flow model as well as experimental investigations of the influence of turbulence on the stability boundaries have been carried out. Here, for certain tube bundle configurations an increased turbulence has a stabilizing effect and leads to a shift of the stability boundaries to higher velocities. The change of the turbulence was realised by using turbulence grids at the inlet of the bundles or thin Prandtl-tripwires at the tube surfaces. Flow visualization studies at the original experimental set-up under relevant Reynolds numbers give an impression of the flow pattern. At this time an investigation of the exciting fluid forces is carried out using a flexibly mounted pressure test tube. A survey about some recent investigations is given. (orig.)

Process and device for locating a defective tube, particularly in the tube bundle of a steam generator

International Nuclear Information System (INIS)

Denis, Jean.

1977-01-01

A process is described for locating a defective tube, particularly in the tube bundle of a steam generator of the reversed U tube kind with the ends connected to a tube plate, marking with the bottom of the generator casing a space separated into two adjacent collectors, respectively for the inlet and outlet of a primary fluid flowing inside the tubes of the bundle, these being externally washed by a secondary vaporizing fluid. In this process a television camera that can be inserted into the casing is used. This process consists in transmitting to a display system outside the generator an image of the tube plate in each collector by means of a directional television camera and then to place over this image a luminous marker to locate the end or the faulty tube [fr

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

International Nuclear Information System (INIS)

Agostini, F.

2008-07-01

The complexity of the two-phase flow in a tube bundle presents important problems in the design and understanding of the physical phenomena taking place. The working conditions of an evaporator depend largely on the dynamics of the two-phase flow that in turn influence the heat exchange and the pressure drop of the system. A characterization of the flow dynamics, and possibly the identification of the flow pattern in the tube bundle, is thus expected to lead to a better understanding of the phenomena and to reveal on the mechanisms governing the tube bundle. Therefore, the present study aims at providing further insights into two-phase bundle flow through a new visualization system able to provide for the first time a view of the flow in the core of a tube bundle. In addition, the measurement of the light attenuation of a laser beam through the two-phase flow and measurement of the high frequency pressure fluctuations with a piezo-electric pressure transducer are used to characterize the flow. The design and the validation of this new instrumentation also provided a method for the detection of dry-out in tube bundles. This was achieved by a laser attenuation technique, flow visualization, and estimation of the power spectrum of the pressure fluctuation. The current investigation includes results for two different refrigerants, R134a and R236fa, three saturations temperatures T sat = 5, 10 and 15 °C, mass velocities ranging from 4 to 40 kg/sm² in adiabatic and diabatic conditions (several heat fluxes). Measurement of the local heat transfer coefficient and two-phase frictional pressure drop were obtained and utilized to improve the current prediction methods. The heat transfer and pressure drop data were supported by extensive characterization of the two-phase flow, which was to improve the understanding of the two-phase flow occurring in tube bundles. (author)

Heat transfer analysis and effects of feeding tubes arrangement, falling film behavior and backsplash on ice formation around horizontal tubes bundles

International Nuclear Information System (INIS)

Sait, Hani Hussain

2013-01-01

Highlights: • Ice shape around the tubes. • Effects of accumulation of ice around the tubes. • Effects of parallel and series tubes arrangements. • Effects of ice accumulated around the tube surfaces. • Effects of backsplash on ice formation. - Abstract: Excessive electrical load has recently get a lot of attention from electric companies specially in hot countries like Saudi Arabia, where air-conditioning load represents about 75% from the total electrical load. Energy storage by freezing is one of the methods that used to tackle this issue. Ice is formed around horizontal cold tubes that are subjected to falling film of water. Ice quantity is measured, photographed and studied. In this studied the coolant inside the tubes flows in series tube arrangement. The results are compared with previous study in which parallel arrangement was used. In addition the falling film behavior and the resulted backsplash are also investigated. A mathematical model to predict ice formation around the tube is proposed. Comparison of the results of the model with that of the experiments showed that the agreement between the two is acceptable. The results also show a quite reasonable quantity of ice is formed in a short time and the series arrangement is more efficient than parallel one. The falling film shapes and its backsplash has also affected the ice formation

Adjustment of pipe flow explicit friction factor equations for application to tube bundles

International Nuclear Information System (INIS)

Wiltz, Christopher L.; Bowen, Mike D.; Von Olnhausen, Wayne A.

2005-01-01

Full text of publication follows: The accurate determination of single phase friction losses or friction pressure drop in tube bundles is essential in the thermal-hydraulic analyses of components such as nuclear fuel assemblies, heat exchangers and steam generators. Such friction losses are normally calculated using a friction factor, f, along with the experimental observation that the friction pressure drop in a pipe is proportional to the dynamic pressure (1/2 ρV 2 ) of the flow: ΔP = 1/2 ρV 2 (fL/D). In this equation L is the pipe or tube bundle length and D is the hydraulic diameter of the pipe or tube bundle. The friction factor is normally calculated using one of a number of explicit friction factor equations. A significant amount of work has been accomplished in developing explicit friction factor equations. These explicit equations range from approximations, which were developed for ease of numerical evaluation, to those which are mathematically complex but yield very good fits to the test data. These explicit friction factor equations are based on a large experimental data base, nearly all of which comes from pipe flow geometry information, and have been historically applied to tube bundles. This paper presents an adjustment method which may be applied to various explicit friction factor equations developed for pipe flow to accurately predict the friction factor for tube bundles. The characteristic of the adjustment is based on experimental friction pressure loss data obtained by Framatome ANP through flow testing of a nuclear fuel assembly (tube bundle) at its Richland Test Facility (RTF). Through adjustment of previously developed explicit friction factor equations for pipe flow, the vast amount of historical development and experimentation in the area of single phase pipe flow friction loss may be incorporated into the evaluation of single phase friction losses within tube bundles. Comparisons of the application of one or more of the previously

Numerical simulation of flow-induced vibrations in tube bundles

International Nuclear Information System (INIS)

Elisabeth Longatte; Zaky Bendjeddou; Mhamed Souli

2005-01-01

Full text of publication follows: In many industrial components mechanical structures like rod cluster control assembly, fuel assembly and heat exchanger tube bundles are submitted to complex flows causing possible vibrations and damage. Fluid forces are usually split into two parts: structure motion independent forces and fluid-elastic forces coupled with tube motion and responsible for possible dynamic instability development leading to possible short term failures through high amplitude vibrations. Most classical fluid force identification methods rely on structure response experimental measurements associated with convenient data processes. Owing to recent improvements in Computational Fluid Dynamics (C.F.D.), numerical fluid force identification is now practicable in the presence of industrial configurations. The present paper is devoted to numerical simulation of flow-induced vibrations of tube bundles submitted to single-phase cross flows by using C.F.D. codes. Direct Numerical Simulation (D.N.S.), Arbitrary Lagrange Euler formulation (A.L.E.) and code coupling process are involved to predict fluid forces responsible for tube bundle vibrations in the presence of fluid structure and fluid-elastic coupling effects. In the presence of strong multi-physics coupling, simulation of flow-induced vibrations requires a fluid structure code coupling process. The methodology consists in solving in the same time thermohydraulics and mechanics problems by using an A.L.E. formulation for the fluid computation. The purpose is to take into account coupling between flow and structure motions in order to be able to capture coupling effects. From a numerical point of view, there are three steps in the computation: the fluid problem is solved on the computational domain; fluid forces acting on the moving tube are estimated; finally they are introduced in the structure solver providing the tube displacement that is used to actualize the fluid computational domain. Specific

Condensation Analysis of Steam/Air Mixtures in Horizontal Tubes

International Nuclear Information System (INIS)

Lee, Kwon Yeong; Bae, Sung Won; Kim, Moo Hwan

2008-01-01

Perhaps the most common flow configuration in which a convective condensation occurs is a flow in a horizontal circular tube. This configuration is encountered in air-conditioning and refrigeration condensers as well as condensers in Rankine power cycles. Although a convective condensation is also sometimes contrived to occur in a co-current vertical downward flow, a horizontal flow is often preferred because the flow can be repeatedly passed through the heat exchanger core in a serpentine fashion without trapping liquid or vapor in the return bends. Many researchers have investigated a in-tube condensation for horizontal heat exchangers. However, almost all of them obtained tube section-averaged data without a noncondensable gas. Recently, Wu and Vierow have experimentally studied the condensation of steam in a horizontal heat exchanger with air present. In order to measure the condenser tube inner surface temperatures and to calculate the local heat fluxes, they developed an innovative thermocouple design that allowed for nonintrusive measurements. Here we developed a theoretical model using the heat and mass analogy to analyze a steam condensation with a noncondensable gas in horizontal tubes

Assessment of fluid-to-fluid modelling of critical heat flux in horizontal 37-element bundle flows

International Nuclear Information System (INIS)

Yang, S.K.

2006-01-01

Fluid-to-fluid modelling laws of critical heat flux (CHF) available in the literature were reviewed. The applicability of the fluid-to-fluid modelling laws was assessed using available data ranging from low to high mass fluxes in horizontal 37-element bundles simulating a CANDU fuel string. Correlations consisting of dimensionless similarity groups were derived using modelling fluid data (Freon-12) to predict water CHF data in horizontal 37-element bundles with uniform and non-uniform axial-heat flux distribution (AFD). The results showed that at mass fluxes higher than ∼4,000 kg/m 2 s (water equivalent value), the vertical fluid-to-fluid modelling laws of Ahmad (1973) and Katto (1979) predict water CHF in horizontal 37-element bundles with non-uniform AFD with average errors of 1.4% and 3.0% and RMS errors of 5.9% and 6.1%, respectively. The Francois and Berthoud (2003) fluid-to-fluid modelling law predicts CHF in non-uniformly heated 37-element bundles in the horizontal orientation with an average error of 0.6% and an RMS error of 10.4% over the available range of 2,000 to 6,200 kg/m 2 s. (author)

Set-up for steam generator tube bundle washing after explosion expanding the tubes

International Nuclear Information System (INIS)

Osipov, S.I.; Kal'nin, A.Ya.; Mazanenko, M.F.

1985-01-01

Set-up for steam generator tube bundle washing after the explosion expanding of tubes is described. Washing is accomplished by distillate. Steam is added to distillate for heating, and compersed air for preventing hydraulic shock. The set-up is equiped by control equipment. Set-up performances are presented. Time for one steam generator washing constitutes 8-12 h. High economic efficiency is realized due to the set-up introduction

Comparing studies for an optimization of steam-heated tube bundle heat exchangers

International Nuclear Information System (INIS)

Horn, M.

1975-01-01

The problems of designing an apparatus are to be shown by the example of the steam-heated tube bundle heat exchanger, and optimizations are to be carried through by relevant examples. From the results of the optimization, a set of apparatus types is to be derived where the dimensions of the shell and the heat pipes as well as the length of the tube bundle are to be determined by as few data as possible. (orig./TK) [de

Experimental investigation of water sprayed finned heat exchanger tube bundles

International Nuclear Information System (INIS)

Sommer, A.

1987-07-01

Experimental investigations have been made to study the performance of two finned tube-bundle heat exchangers (FORGO type) when wetted by water sprays. The heat exchangers are designed to cool water in a dry cooling tower. The test-elements had a frontal area of 1 m 2 . The water sprays were created by 20 nozzles, 200 mm in front of the heat exchangers. Air velocities at the inlet of the coolers were in the range 0,8 m/s to 12 m/s and initial temperature differences ITD reached 45 degrees C. The test facility was designed to determine the combined latent and sensible heat fluxes in the wetted heat exchanger, the airside pressure drop and the air humidity and temperature at the exchanger inlet and outlet, and to measure the weight of the water wetting the cooler's surface. The sprayed test elements were investigated in different positions, but most of the experiments were carried out in the position with the fins horizontal

An eddy viscosity model for flow in a tube bundle

International Nuclear Information System (INIS)

Soussan, D.; Grandotto, M.

1998-01-01

The work described in this paper is part of the development of GENEPI a 3-dimensional finite element code, designed for the thermalhydraulic analysis of steam generators. It focuses on the implementation of two-phase flow turbulence-induced viscosity in a tube bundle. The GENEPI code, as other industrial codes, uses the eddy viscosity concept introduced by Boussinesq for single phase flow. The concept assumes that the turbulent momentum transfer is similar to the viscous shear stresses. Eddy viscosity formulation is reasonably well known for single phase flows, especially in simple geometries (i.e., in smooth tube, around a single body, or behind a row of bars/tubes), but there exists very little information on it for two-phase flows. An analogy between single and two-phases is used to set up a model for eddy viscosity. The eddy viscosity model examined in this paper is used for a tube bundle geometry and, therefore, is extended to include anisotropy to the classic model. Each of the main flow directions (cross flow inline, cross flow staggered, and parallel flows) gives rise to a specific eddy viscosity formula. The results from a parametric study indicate that the eddy viscosity in the staggered flow is roughly 1.5 times as large as that for the inline cross flow, 60 times as large as that for the parallel flow, and 105 as large as that for the molecular viscosity. Then, the different terms are combined with each other to result in a global eddy viscosity model for a steam generator tube bundle flow. (author)

CHF prediction in rod bundles using round tube data

Energy Technology Data Exchange (ETDEWEB)

Souza, Wallen F.; Veloso, Maria A.F.; Pereira, Cláubia; Costa, Antonella L., E-mail: wallenfds@yahoo.com.br, E-mail: mdora@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear

2017-07-01

The present work concerns the use of 1995 CHF table for uniformly heated round tubes, developed jointly by Canadian and Russian researchers, for the prediction of critical heat fluxes in rod bundles geometries. Comparisons between measured and calculated critical heat fluxes indicate that this table could be applied to rod bundles provided that a suitable correction factor is employed. The tolerance limits associated with the departure from nucleate boiling ratio (DNBR) are evaluated by using statistical analysis. (author)

Lattice Boltzmann simulation of flow across a staggered tube bundle array

Energy Technology Data Exchange (ETDEWEB)

Tiftikçi, A.; Kocar, C., E-mail: ckocar@hacettepe.edu.tr

2016-04-15

Highlights: • Large eddy simulation of the cross-flow in a staggered tube bundle array in 3D was made. • LBM and FVM are used separately as numerical solvers and the results of each method compared with experimental data. • Effect of lattice model is studied for tube bundle flow. • Filter size effects, mesh size effects are studied for VLES turbulence model. - Abstract: The decision on the magnitude of the grid size is a crucial problem in large eddy simulations. Finer mesh requires excessive memory and causes long simulation time. Large eddy simulation model becomes inefficient when the extent of the flow geometry to be simulated with the lattice-Boltzmann method is large. Thus, in this study, it is proposed to investigate the capabilities of three turbulence models, namely, very large eddy simulation, Van Driest and Smagorinsky–Lilly. As a test case, a staggered tube bundle flow experiment is used for the validation and comparison purposes. Sensitivity analyses (including mesh and filter size) have been made. Furthermore, the effect of lattice model is investigated and it is showed that the D3Q27 and D3Q19 models do not differ significantly in lattice-Boltzmann method for this type of flow. The results of turbulence model comparisons for staggered tube bundle flow showed that very large eddy simulation is superior at low resolution. This paper might be considered as a good validation of the lattice-Boltzmann method. In turbulent flow conditions, the code successfully captures the velocity and stress profiles even if the flow is quite complicated.

Numerical analysis on the condensation heat transfer and pressure drop characteristics of the horizontal tubes of modular shell and tube-bundle heat exchanger

International Nuclear Information System (INIS)

Ko, Seung Hwan; Park, Hyung Gyu; Kim, Charn Jung; Park, Byung Kyu

2001-01-01

A numerical analysis of the heat and mass transfer and pressure drop characteristics in modular shell and tube bundle heat exchanger was carried out. Finite concept method based on FVM and κ-ε turbulent model were used for this analysis. Condensation heat transfer enhanced total heat transfer rate 4∼8% higher than that of dry heat exchanger. With increasing humid air inlet velocity, temperature and relative humidity, and with decreasing heat exchanger aspect ratio and cooling water velocity, total heat and mass transfer rate could be increased. Cooling water inlet velocity had little effect on total heat transfer

Improving the thermodynamic efficiency of steam turbine condensers with partial tube replacement and an advanced tube bundle design

International Nuclear Information System (INIS)

Drosdziok, A.; Zorner, W.

1989-01-01

Many different problems have been experienced with power plant condensers all over the world. It has become apparent that plant availability and cost-effectiveness are significantly influenced by the thermodynamic design of the condensers and the materials selected. This paper reports that by refitting older condensers in operating plants it has proven possible to improve thermodynamic efficiency by changing the tube bundle design. In conjunction with the replacement of the cooper-bearing tubing in these condensers, which became necessary because of the introduction of high AVT (All Volatile Treatment) conditioning in the secondary circuit, it has generally been possible to fulfil the requirements imposed on the condensers without a deterioration of plant efficiency. By experience, best results have been obtained by replacing the condenser bundle with an advanced tube bundle design. Apart from solving all problems, this further improves the thermodynamic efficiency of the condensers. In nuclear power plants constructed by the Siemens KWU Group the condensers are tailored to present-day requirements

Experiment and numerical simulation of bubbly two-phase flow across horizontal and inclined rod bundles

International Nuclear Information System (INIS)

Serizawa, A.; Huda, K.; Yamada, Y.; Kataoka, I.

1997-01-01

Experimental and numerical analyses were carried out on vertically upward air-water bubbly two-phase flow behavior in both horizontal and inclined rod bundles with either in-line or staggered array. The inclination angle of the rod bundle varied from 0 to 60 with respect to the horizontal. The measured phase distribution indicated non-uniform characteristics, particularly in the direction of the rod axis when the rods were inclined. The mechanisms for this non-uniform phase distribution is supposed to be due to: (1) Bubble segregation phenomenon which depends on the bubble size and shape: (2) bubble entrainment by the large scale secondary flow induced by the pressure gradient in the horizontal direction which crosses the rod bundle; (3) effects of bubble entrapment by vortices generated in the wake behind the rods which travel upward along the rod axis; and (4) effect of bubble entrainment by local flows sliding up along the front surface of the rods. The liquid velocity and turbulence distributions were also measured and discussed. In these speculations, the mechanisms for bubble bouncing at the curved rod surface and turbulence production induced by a bubble were discussed, based on visual observations. Finally, the bubble behaviors in vertically upward bubbly two-phase flow across horizontal rod bundle were analyzed based on a particle tracking method (one-way coupling). The predicted bubble trajectories clearly indicated the bubble entrapment by vortices in the wake region. (orig.)

Influence of structure improvement of guide tubes and bundles in pressurized water reactor (PWR) on drop of control rods

International Nuclear Information System (INIS)

Shen Xiuzhong; Yu Pingan; Yang Guanyue

1996-01-01

In order to alleviate the cross hydraulic load on control rod guide tubes and bundles, some protective sleeves are added to those near the upper plenum outlet nozzles (4 symmetric bundles: 02-26, 03-25, 11-29, 12-28). In a 1/4 scale transparent model of the PWR upper plenum of Qinshan Nuclear Power Station, water was chosen as the fluid and hydraulic experiments with improved control rod guide tubes and bundles were carried out. The results were carefully compared with those of the experiments with unimproved control rod guide tubes and bundles. It is concluded that adding protective sleeves to the control rod guide tubes and bundles near the outlet nozzles will help to lighten the hydraulic load on them and make certain of the free movement and rapid dropping of control rods in the tubes and bundles in emergency by order

Fluid-Elastic Instability Tests on Parallel Triangular Tube Bundles with Different Mass Ratio Values under Increasing and Decreasing Flow Velocities

Directory of Open Access Journals (Sweden)

Xu Zhang

2016-01-01

Full Text Available To study the effects of increasing and decreasing flow velocities on the fluid-elastic instability of tube bundles, the responses of an elastically mounted tube in a rigid parallel triangular tube bundle with a pitch-to-diameter ratio of 1.67 were tested in a water tunnel subjected to crossflow. Aluminum and stainless steel tubes were tested, respectively. In the in-line and transverse directions, the amplitudes, power spectrum density functions, response frequencies, added mass coefficients, and other results were obtained and compared. Results show that the nonlinear hysteresis phenomenon occurred in both tube bundle vibrations. When the flow velocity is decreasing, the tubes which have been in the state of fluid-elastic instability can keep on this state for a certain flow velocity range. During this process, the response frequencies of the tubes will decrease. Furthermore, the response frequencies of the aluminum tube can decrease much more than those of the stainless steel tube. The fluid-elastic instability constants fitted for these experiments were obtained from experimental data. A deeper insight into the fluid-elastic instability of tube bundles was also obtained by synthesizing the results. This study is beneficial for designing and operating equipment with tube bundles inside, as well as for further research on the fluid-elastic instability of tube bundles.

Falling film flow, heat transfer and breakdown on horizontal tubes

International Nuclear Information System (INIS)

Rogers, J.T.

1980-11-01

Knowledge of falling film flow and heat transfer characteristics on horizontal tubes is required in the assessment of certain CANDU reactor accident sequences for those CANDU reactors which use moderator dump as one of the shut-down mechanisms. In these reactors, subsequent cooling of the calandria tubes is provided by falling films produced by sprays. This report describes studies of falling film flow and heat transfer characteristics on horizontal tubes. Analyses using integral methods are given for laminar and turbulent flow, ignoring and accounting for momentum effects in the film. Preliminary experiments on film flow stability on horizontal tubes are described and various mechanisms of film breakdown are examined. The work described in this report shows that in LOCA with indefinitely delayed ECI in the NPD or Douglas Point (at 70 percent power) reactors, the falling films on the calandria tubes will not be disrupted by any of the mechanisms considered, provided that the pressure tubes do not sag onto the calandria tubes. However, should the pressure tubes sag onto the calandria tubes, film disruption will probably occur

Radial holding device of the tube bundle casing and of the tube support plates by elastic stops

International Nuclear Information System (INIS)

Comic, G.

1995-01-01

Each stop comprises a first piece fixed on the tube bundle casing and contacting the inner face of the pressure casing by the intermediary of screws. A second piece abutting the tube support plate and constraining it is housed to form a piston in a cavity of the first piece. 5 figs

Calculation of vapour bubble growth on the lower generatrix of horizontal tubes

International Nuclear Information System (INIS)

Chajka, V.D.

1987-01-01

The known models of vapour bubble growth are compared with experimental data. Cinematographic study of vapour formation during water boiling was carried out with elements of horizontal tubes of copper 10, 16, 24, 34 and 70 mm in diameter under the pressure of 100 kPa and specific thermal loadings of 20 and 40 kW/m 2 . According to the experimental data the main volume of vapour phase is occupied by vapour bubbles from the lower part of the horizontal tube. Five stages of vapour bubble growth on the lower generatrix of the horizontal tube: nucleation, growth to the point of breaking off from nucleate centre, the breaking off from the nucleate centre, the tube surface flowing around during floating up, the breaking off from the tube surface, were singled out. The shape of vapour volume varied during the whole period of the bubble growth and it was mainly determined by the horizontal tube diameter. The change of vapour bubble radius in time is the function of the horizontal tube diameter. Comparison of the experimental data with the known models of vapour bubble growth has shown, that every stage of vapour bubble growth on the lower generatrix of the tube is determined by the complex of thermal and hydrodynamic conditions, the effect of which depends on the horizontal tube diameter

Effects of Angle of Rotation on Pool Boiling Heat Transfer of V-shape Tube Bundle

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

The most important facility for the systems is a passive heat exchanger that transfers core decay heat to the cold water in a water storage tank under atmospheric pressure. Since the space for the installation of the heat exchanger is usually limited, developing more efficient heat exchangers is important. In general, pool boiling is generated on the surface of the heat exchanging tube. The major design parameter of the heat exchanger is a tube arrangement. The upper tube is affected by the lower tube and the enhancement of the heat transfer on the upper tube is estimated by the bundle effect. Since heat transfer is related to the conditions of a tube surface, bundle geometries, and a liquid type, lots of studies have been carried out for the combinations of those parameters. An experimental study was performed to investigate the effects of the angle of rotation on pool boiling heat transfer of a V-shape tube bundle. For the test, two smooth stainless steel tubes of 19 mm outside diameter and the water at atmospheric pressure were used. The enhancement of the heat transfer is clearly observed when the angle becomes to 90° where the upper tube has the maximum region of influence by the lower tube. The convective flow and liquid agitation enhance heat transfer while the coalescence of the bubbles deteriorates heat transfer.

A thermal mixing model of crossflow in tube bundles for use with the porous body approximation

International Nuclear Information System (INIS)

Ashcroft, J.; Kaminski, D.A.

1996-06-01

Diffusive thermal mixing in a heated tube bundle with a cooling fluid in crossflow was analyzed numerically. From the results of detailed two-dimensional models, which calculated the diffusion of heat downstream of one heated tube in an otherwise adiabatic flow field, a diffusion model appropriate for use with the porous body method was developed. The model accounts for both molecular and turbulent diffusion of heat by determining the effective thermal conductivity in the porous region. The model was developed for triangular shaped staggered tube bundles with pitch to diameter ratios between 1.10 and 2.00 and for Reynolds numbers between 1,000 and 20,000. The tubes are treated as nonconducting. Air and water were considered as working fluids. The effective thermal conductivity was found to be linearly dependent on the tube Reynolds number and fluid Prandtl number, and dependent on the bundle geometry. The porous body thermal mixing model was then compared against numerical models for flows with multiple heated tubes with very good agreement

Horizontal beam tubes in FRM-II

International Nuclear Information System (INIS)

Coors, D.; Vanvor, D.

2001-01-01

The new research reactor in Garching FRM-II is equipped with 10 leak tight horizontal beam tubes (BT1 - BT10), each of them consisting of a beam tube structure taking an insert with neutron channels. The design of all beam tube structures is similar whereas the inserts are adapted to the special requirements of the using of each beam tube. Inside the reflector tank the beam tube structures are shaped by the inner cones which are made of Al-alloy with circular and rectangular cross sections. They are located in the region of maximum neutron flux (exception BT10), they are directly connected to the flanges of the reflector tank, their lengths are about 1.5 m (exception BT10) and their axes are directed tagentially to the core centre thus contributing to a low γ-noise at the experiments. (orig.)

The various phenomena encountered in tube-bundles in cross-flow

International Nuclear Information System (INIS)

Gibert, R.J.

1975-01-01

The various vibrational phenomena induced on tube bundles in a cross flow are classified. The research program is concerned with mechanical phenomena observed on mock-ups with tube row structures. It is intended for specifying the coefficients controlling the appearance of two different phenomena: the first one entailing a change in the vortex shedding and consequently the mechanical source, the other one entailing a frequency spread of vibrations (floating instability). The research is to improve heat exchanger performance and cost [fr

Three-dimensional Effects of Turburlent Flow in an In-Line Tube Bundle

DEFF Research Database (Denmark)

Meyer, Knud Erik

1998-01-01

Velocities have been measured with laser Doppler anemometry between tubes in cross-flow in a small in-line tube bundle with longitudinal to transverse pitches of 1.5Dx1.8D and a Reynolds number based on mean velocity in minimum flow section of Re=30000. At most locations a single recirculation zone...... is found behind each tube. However, the direction of circulation changes sign along the tube with a period of about 2~tube diameters. Three different patterns of such recirculation zones have been observed. Each pattern is very stable and does not change under undisturbed flow conditions....

Single-Phase Crossflow Mixing in a Vertical Tube Bundle Geometry : An Experimental Study

NARCIS (Netherlands)

Mahmood, A.

2011-01-01

The vertical rod/tube bundle geometry has a wide variety of industrial applications. Typical examples are the core of light water nuclear reactors (LWR) and vertical tube steam generators. In the core of a LWR, primarily coolant flows upward but their also exist a flow in lateral direction, called

Composite Coiled Tubing for Extended Reach in Horizontal Oil Wells

DEFF Research Database (Denmark)

Costache, Andrei; Berggreen, Christian

2017-01-01

Conventional steel coiled tubing cannot reach along the entire length of very long horizontal oil wells. A lighter and more buoyant coiled tube is made possible using composite materials. The high stiffness to weight ratio of fiber reinforced polymers, coupled with a lower coefficient of friction......, has the potential of greatly extending the reach in horizontal oil wells. This study shows how to design composite coiled tubing and gives a comprehensive discussion about the most influential parameters. Several solutions, using glass-fiber and carbon are considered. Finite element models are used...

Condensate subcooling near tube exit during horizontal in-tube condensation

International Nuclear Information System (INIS)

Hashizume, K.; Abe, N.; Ozeki, T.

1992-01-01

In-tube condensation is encountered in various applications for heat exchangers, such as domestic air-conditioning equipment, industrial air-cooled condensers, and moisture separator reheaters (MSRs) for nuclear power pants. Numerous research work has been conducted to predict the condensation heat transfer coefficient, and we have now enough information for thermal design of heat exchangers with horizontal in-tube condensation. Most of the research is analytical and/or experimental work in the annular or stratified flow regime, or experimental work on bulk condensation, i.e., from saturated vapor to complete condensation. On the other hand, there exist few data about the heat transfer phenomena in the very lower-quality region near the tube exit. The purpose of this paper is to clarify the condensation heat transfer phenomena near the tube exit experimentally and analytically, and to predict the degree of condensate subcooling

Contribution at the vibrations study of tube bundles in a transversal flow

International Nuclear Information System (INIS)

Antunes, J.

1986-03-01

The steam generators tubes bundles attended vibratory risks under flow. In this work we present the experimental and theoretical analysis which shows the necessary to approach this problem with taking into account the non-linear dynamic interaction between tubes and supports. An entirety of experiences put in clearness the importance of little clearance between the tubes and their supports. Methods for numerical simulation of the tubes vibratory response are proposed. Parametric analysis are presented, which permit to find simple laws concerning the influence of system parameters on its vibratory behaviour. This work is completed by analytical study of two unstable oscillators [fr

A new accelerator tube and column for a horizontal 8 MV tandem

International Nuclear Information System (INIS)

Sundquist, M.L.; Rathmell, R.D.; Raatz, J.E.

1990-01-01

A horizontal 8 MV tandem is being installed in an existing tank at Kyoto University in Japan. This NEC Model 8UDH is the largest horizontal Pelletron constructed to date. The terminal is charged by two Pelletron chains. The acceleration tube is a metal and ceramic construction made into tube sections with a length of 30 cm each. This tube design adds 27% more live ceramic than in the standard NEC tube design, which had heated apertures in 5 cm long shorted regions every 20 cm. The column structure and tube design are reviewed. (orig.)

Dismantling Experiment of Mock-up Tube Bundle of Steam Generator

International Nuclear Information System (INIS)

Kim, Sung Kyun; Lee, Kune Woo

2010-01-01

A SG (steam generator) is one of the biggest decommissioning components in nuclear power plants and one has been replaced 2∼6 times during the whole operation of a nuclear power plant. The old SG should be decommissioned for the purpose of the volume reduction of radioactive waste. Among the components of SG, the tube bundle is one of the most difficult items to be dismantled due to the fact that it is very hard to cut since it is made of Inconel 600 which has high resistance of corrosion and abrasion. Moreover, All cutting process should be performed by remotely since radioactive contamination of the internal surface of SG tubes is very high (about 150,000∼300,000 Bq/cm 2 ). Therefore, it is necessary to choose the appropriate cutting methods by the pros and cons analysis for candidate dismantling technologies and to do experiment study for the validation. In this study, the results of cutting experiment for a mock-up bundle by using band saw cutting method are described herein

Fluid-structure interaction in tube bundles: homogenization methods, physical analysis

International Nuclear Information System (INIS)

Broc, D.; Sigrist, J.F.

2009-01-01

It is well known that the movements of a structure may be strongly influenced by fluid. This topic, called 'Fluid Structure Interaction' is important in many industrial applications. Tube bundles immersed in fluid are found in many cases, especially in nuclear industry: (core reactors, steam generators,...). The fluid leads to 'inertial effects' (with a decrease of the vibration frequencies) and 'dissipative effects' (with higher damping). The paper first presents the methods used for the simulation of the dynamic behaviour of tube bundles immersed in a fluid, with industrial examples. The methods used are based on the Euler equations for the fluid (perfect fluid), which allow to take into account the inertial effects. It is possible to take into account dissipative effects also, by using a Rayleigh damping. The conclusion focuses on improvements of the methods, in order to take into account with more accuracy the influence of the fluid, mainly the dissipative effects, which may be very important, especially in the case of a global fluid flow. (authors)

Pressure distribution over tube surfaces of tube bundle subjected to two phase cross flow

International Nuclear Information System (INIS)

Sim, Woo Gun

2013-01-01

Two phase vapor liquid flows exist in many shell and tube heat exchangers such as condensers, evaporators and nuclear steam generators. To understand the fluid dynamic forces acting on a structure subjected to a two phase flow, it is essential to obtain detailed information about the characteristics of a two phase flow. The characteristics of a two phase flow and the flow parameters were introduced, and then, an experiment was performed to evaluate the pressure loss in the tube bundles and the fluid dynamic force acting on the cylinder owing to the pressure distribution. A two phase flow was pre mixed at the entrance of the test section, and the experiments were undertaken using a normal triangular array of cylinders subjected to a two phase cross flow. The pressure loss along the flow direction in the tube bundles was measured to calculate the two phase friction multiplier, and the multiplier was compared with the analytical value. Furthermore, the circular distributions of the pressure on the cylinders were measured. Based on the distribution and the fundamental theory of two phase flow, the effects of the void fraction and mass flux per unit area on the pressure coefficient and the drag coefficient were evaluated. The drag coefficient was calculated by integrating the measured pressure coefficient and the drag coefficient were evaluated. The drag coefficient was calculated by integrating the measured pressure on the tube by a numerical method. It was found that for low mass fluxes, the measured two phase friction multipliers agree well with the analytical results, and good agreement for the effect of the void fraction on the drag coefficients, as calculated by the measured pressure distributions, is shown qualitatively, as compared to the existing experimental results

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-10-15

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

Heat transfer performance during in-tube condensation in horizontal smooth, micro-fin and herringbone tubes

OpenAIRE

2008-01-01

M.Ing. An experimental investigation was conducted into the heat transfer characteristics of horizontal smooth, micro-fin and herringbone tubes during in-tube condensation. The study focused on the heat transfer coefficients of refrigerants R-22, R-134a and R-407C inside the three tubes. The herringbone tube results were compared to the smooth and micro-fin tube results. The average increase in the heat transfer coefficient when compared to the smooth tube was found to be as high as 322% w...

Numerical Analysis of Turbulent Flow around Tube Bundle by Applying CAD Best Practice Guideline

International Nuclear Information System (INIS)

Lee, Gong Hee; Bang, Young Seok; Woo, Sweng Woong; Cheng, Ae Ju

2013-01-01

In this study, the numerical analysis of a turbulent flow around both a staggered and an incline tube bundle was conducted using Annoys Cfx V. 13, a commercial CAD software. The flow was assumed to be steady, incompressible, and isothermal. According to the CAD Best Practice Guideline, the sensitivity study for grid size, accuracy of the discretization scheme for convection term, and turbulence model was conducted, and its result was compared with the experimental data to estimate the applicability of the CAD Best Practice Guideline. It was concluded that the CAD Best Practice Guideline did not always guarantee an improvement in the prediction performance of the commercial CAD software in the field of tube bundle flow

Free vibration analysis of a steam generator tube bundle with and without lateral support

International Nuclear Information System (INIS)

King, D.M.

1979-04-01

The vibrational modes and frequency characteristics of a pressurized water reactor (PWR) steam generator tube bundle assembly with and without lateral support in a fluid environment are analyzed. The idealized half-model was constructed using the SAP-IV finite element code. Free vibration analyses were performed for an in-air case and a submerged in-water case, each with different constraint conditions at steam generator tube bundle assembly support plates 10 and 11. These constraint conditions included having both support plates free, having both support plates fixed, and having support plate 11 free while support plate 10 was fixed. It was found that as the support plate constraints were removed, the frequency range for each case increased significantly

Two-phase flow field simulation of horizontal steam generators

Energy Technology Data Exchange (ETDEWEB)

Rabiee, Ataollah; Kamalinia, Amir Hossein; Hadad, Kamal [School of Mechanical Engineering, Shiraz University, Shiraz (Iran, Islamic Republic of)

2017-02-15

The analysis of steam generators as an interface between primary and secondary circuits in light water nuclear power plants is crucial in terms of safety and design issues. VVER-1000 nuclear power plants use horizontal steam generators which demand a detailed thermal hydraulics investigation in order to predict their behavior during normal and transient operational conditions. Two phase flow field simulation on adjacent tube bundles is important in obtaining logical numerical results. However, the complexity of the tube bundles, due to geometry and arrangement, makes it complicated. Employment of porous media is suggested to simplify numerical modeling. This study presents the use of porous media to simulate the tube bundles within a general-purpose computational fluid dynamics code. Solved governing equations are generalized phase continuity, momentum, and energy equations. Boundary conditions, as one of the main challenges in this numerical analysis, are optimized. The model has been verified and tuned by simple two-dimensional geometry. It is shown that the obtained vapor volume fraction near the cold and hot collectors predict the experimental results more accurately than in previous studies.

Unsteady Model for Transverse Fluid Elastic Instability of Heat Exchange Tube Bundle

Directory of Open Access Journals (Sweden)

Jun Liu

2014-01-01

Full Text Available From the viewpoint of practical application, based on the unsteady analytical model for transverse fluid elastic instability of tube array proposed by Yetisir and the linear attenuation function introduced by Li Ming, a new explicit model based on nonsteady state “streamtube” hypothesis is proposed and solved using complex number method. In the model, numerical integral is avoided and inappropriate aspects in Li Ming model are modified. Using the model, the fluid elastic instability analysis of a single flexible tube is made. The stability graphs for four typical types of tube array are plotted and contrasted with experimental results. It is found that the current explicit model is effective in the analysis of transverse fluid elastic instability of tube bundle.

Investigation and application of reduced-order methods for flows study in heat exchanger tube bundles

International Nuclear Information System (INIS)

Pomarede, M.

2012-01-01

The objective of this thesis is to study the ability of model reduction for investigations of flow-induced vibrations in heat exchangers tube bundle systems.These mechanisms are a cause of major concern because heat exchangers are key elements of nuclear power plants and on-board stoke-holds.In a first part, we give a recall on heat exchangers functioning and on vibratory problems to which they are prone. Then, complete calculations leaded with the CFD numerical code Code-Saturne are carried out, first for the flow around a single circular cylinder (fixed then elastically mounted) and then for the case of a tube bundle system submitted to cross-flow. Reduced-order method POD is applied to the flow resolution with fixed structures. The obtained results show the efficiency of this technique for such configurations, using stabilization methods for the dynamical system resolution in the tube-bundle case. Multiphase-POD, which is a method enabling the adaptation of POD to fluid-structure interactions, is applied. Large displacements of a single cylinder elastically mounted under cross-flow, corresponding to the lock-in phenomenon,are well reproduced with this reduction technique. In the same way, large displacements of a confined moving tube in a bundle are shown to be faithfully reconstructed.Finally, the use of model reduction is extended to parametric studies. First, we propose to use the method which consists in projecting Navier-Stokes equations for several values of the Reynolds number on to a unique POD basis. The results obtained confirm the fact that POD predictability is limited to a range of parameter values. Then, a basis interpolation method, constructed using Grassmann manifolds and allowing the construction of a POD basis from other pre-calculated basis, is applied to basic cases. (author)

CANFLEX fuel bundle impact test

International Nuclear Information System (INIS)

Chang, Seok Kyu; Chung, C. H.; Park, J. S.; Hong, S. D.; Kim, B. D.

1997-08-01

This document outlines the test results for the impact test of the CANFLEX fuel bundle. Impact test is performed to determine and verify the amount of general bundle shape distortion and defect of the pressure tube that may occur during refuelling. The test specification requires that the fuel bundles and the pressure tube retain their integrities after the impact test under the conservative conditions (10 stationary bundles with 31kg/s flow rate) considering the pressure tube creep. The refuelling simulator operating with pneumatic force and simulated shield plug were fabricated and the velocity/displacement transducer and the high speed camera were also used in this test. The characteristics of the moving bundle (velocity, displacement, impacting force) were measured and analyzed with the impact sensor and the high speed camera system. The important test procedures and measurement results were discussed as follows. 1) Test bundle measurements and the pressure tube inspections 2) Simulated shield plug, outlet flange installation and bundle loading 3) refuelling simulator, inlet flange installation and sensors, high speed camera installation 4) Perform the impact test with operating the refuelling simulator and measure the dynamic characteristics 5) Inspections of the fuel bundles and the pressure tube. (author). 8 refs., 23 tabs., 13 figs

RELAP5 analysis of reflux condensation behavior in heat transfer tube bundle of a steam generator

International Nuclear Information System (INIS)

Minami, Noritoshi; Chikusa, Toshiaki; Nagae, Takashi; Murase, Michio

2007-01-01

In case of loss of the residual heat removal system and other alternative cooling methods under mid-loop operation during shutdown of the pressurized water reactor plant, reflux condensation in the steam generator (SG) may be an effective heat removal mechanism. In reflux condensation experiments 7.2c with injection of nitrogen gas using the BETHSY facility in France, which is a scale model of a pressurized water reactor plant, 34 heat transfer tubes were divided into two kinds of flow patterns, which were steam forward flow and nitrogen reverse flow. In this study, we simulated the BETHSY experiments using the transient analysis code RELAP5. Modifying calculation equations for interfacial friction force and wall friction force between the inlet plenum and heat transfer tubes, nitrogen reverse flow was successfully simulated. In calculations with alteration of the flow area ratio to two flow channels for the heat transfer tube bundle, the number of active tubes with the maximum nitrogen recirculation flow rate agreed rather well with the observed number of active tubes. In calculations with three flow channels for the heat transfer tube bundle, the average number of active tubes in several calculations with different flow area ratios of the three flow channels predicted the number of active tubes well. (author)

Theoretical and experimental investigations of CHF in round tubes and rod bundles

International Nuclear Information System (INIS)

Hwang, Dae Hyun

1994-02-01

A knowledge of the condition leading to critical heat flux (CHF) is of great importance in the design of nuclear reactors. Although many efforts have been devoted to the subject of CHF during the last few decades, information on the burnout phenomenon at low velocity condition is very limited. Furthermore, in most cases, the applicable range of a bundle CHF correlation is restricted to a narrow region mainly due to the limitation of the CHF data base used in the correlation development. In view of these points, theoretical and experimental investigations are performed in this study for round tubes and rod bundles. A CHF prediction model for low velocity conditions is proposed throughout the assessment of CHF data from various sources with mass velocities less than 500 kg/m 2 s. The CHF data base is classified into seven groups with respect to the flow pattern characteristics at CHF conditions. CHF data for each group is analyzed by several CHF prediction models including; the flooding correlations, the flow regime transition criteria, the complete evaporation model, and the empirical correlations. At zero inlet flow or extremely low mass velocity conditions, the flooding correlation can be used for predicting CHF employing appropriate constant. In the slug or churn-turbulent flow regime, CHF seems to occur at the annular flow transition conditions. When CHF occurs at the annular flow region, the empirical correlation such as AECL CHF lookup table gives accurate predictions except for the ranges where density-wave instability is expected. A phenomenological model for the prediction of dryout locations under flooding-limited CHF condition is developed based on the liquid film dryout model and the two-phase mixture level theory. The mass and energy conservation equations are applicable to the liquid film considering no entrainment of liquid droplets from the film region. The variation of the two-phase mixture level after the onset of flooding is calculated based on

Effects of inter-tube coupling on the electro-optical properties of silicon carbide nanotube bundles studied by density functional theory

Science.gov (United States)

Behzad, Somayeh

2015-09-01

The electronic and optical properties of bundled armchair and zigzag silicon carbide nanotubes (SiCNTs) are investigated by using density functional theory. The effects of inter-tube coupling on the electronic dispersions of SiCNT bundles are demonstrated. It was found that the band structure of (6, 0) SiCNT bundle shows metallic feature. The calculated dielectric functions of the armchair and zigzag bundles are similar to that of the isolated tubes, except for the appearance of broadened peaks, small shifts of peak positions about 0.1 eV and increasing of peak intensities. For (6, 0) SiCNT with smaller radius, by considering interband and interaband transitions, the band structure coupling causes an extra peak at low energies.

Mechanical and chemical cleaning of the tubes bundles of the moisture separator reheaters (GSS) of Nuclear power plants

International Nuclear Information System (INIS)

Guerra, Patrice; Ruiz, Jose T.; Ureta, Roman; Carreres, Cristina; Virginie, Le-Guerroue

2012-09-01

The cleaning operation concerns the 'GSS' system (GSS stands for moisture separator reheaters, MSR) which are classified as 'watch quality guarantee', not classified as safety facility and subjected to Pressure Equipment regulations. The follow-up of the operational GSS (steel carbon) of EDF nuclear power plants CP0 group reveals a clog rate due to a relevant magnetite deposits that could result in equipment damage, loss of availability and loss of plant productivity. The pressure drop between inlet and outlet of the heating steam is close to maximum design criterion. The service consisted in designing, developing, qualifying and carrying out a process which removes clog from the inside of GSS U-tubes bundle located in the vapor circuit and which respects the equipment integrity and ensures the process harmlessness. This cleaning has to enable the complete removal of deposits and oxides (magnetite) in order to recover a passage diameter and a surface finish equivalent to the origin, thus avoiding the replacement of the GSS and obtaining a considerable reduction of costs. To do so, LAINSA and SOLARCA designed, developed, qualified and operated on 14 GSS bundles, by carrying out the following operations: - Cartography of the GSS tubes bundles clogging state; - Pre-Mechanical cleaning to un-block the sealed tubes and release the inside tubes passing; - Isolation of the bundle and check of leaks of the system; - Chemical cleaning with the efficiency and harmlessness parameters follow-up: - Acid Phase by means of weak organic acids to eliminate all the deposits; - Passivation phase; - Final Rinsing respecting the customer criteria; - Drying; - Waste management and waste treatment. The implementation of this operation enables the elimination of the whole deposits (magnetite) and oxides located inside the GSS tube bundle and thus to recover a passage diameter inside the tubes, and a pressure drop close to a new system and therefore to enables the

Measurement of gas-liquid two-phase flow around horizontal tube bundle using SF6-water. Simulating high-pressure high-temperature gas-liquid two-phase flow of PWR/SG secondary coolant side at normal pressure

International Nuclear Information System (INIS)

Ishikawa, Atsushi; Imai, Ryoj; Tanaka, Takahiro

2014-01-01

In order to improve prediction accuracy of analysis code used for design and development of industrial products, technology had been developed to create and evaluate constitutive equation incorporated in analysis code. The experimental facility for PWR/SG U tubes part was manufactured to measure local void fraction and gas-liquid interfacial velocity with forming gas-liquid upward two-phase flow simulating high-pressure high-temperature secondary coolant (water-steam) rising vertically around horizontal tube bundle. The experimental facility could reproduce flow field having gas-liquid density ratio equivalent to real system with no heating using SF6 (Sulfur Hexafluoride) gas at normal temperature and pressure less than 1 MPa, because gas-liquid density ratio, surface tension and gas-liquid viscosity ratio were important parameters to determine state of gas-liquid two-phase flow and gas-liquid density ratio was most influential. Void fraction was measured by two different methods of bi-optical probe and conductivity type probe. Test results of gas-liquid interfacial velocity vs. apparent velocity were in good agreement with existing empirical equation within 10% error, which could confirm integrity of experimental facility and appropriateness of measuring method so as to set up original constitutive equation in the future. (T. Tanaka)

Application of field synergy principle for optimization fluid flow and convective heat transfer in a tube bundle of a pre-heater

International Nuclear Information System (INIS)

Hamid, Mohammed O.A.; Zhang, Bo; Yang, Luopeng

2014-01-01

The big problems facing solar-assisted MED (multiple-effect distillation) desalination unit are the low efficiency and bulky heat exchangers, which worsen its systematic economic feasibility. In an attempt to develop heat transfer technologies with high energy efficiency, a mathematical study is established, and optimization analysis using FSP (field synergy principle) is proposed to support meaning of heat transfer enhancement of a pre-heater in a solar-assisted MED desalination unit. Numerical simulations are performed on fluid flow and heat transfer characteristics in a circular and elliptical tube bundle. The numerical results are analyzed using the concept of synergy angle and synergy number as an indication of synergy between velocity vector and temperature gradient fields. Heat transfer in elliptical tube bundle is enhanced significantly with increasing initial velocity of the feed seawater and field synergy number and decreasing of synergy angle. Under the same operating conditions of the two designs, the total average synergy angle is 78.97° and 66.31° in circular and elliptical tube bundle, respectively. Optimization of the pre-heater by FSP shows that in case of elliptical tube bundle design, the average synergy number and heat transfer rate are increased by 22.68% and 35.98% respectively. - Highlights: • FSP (field synergy principle) is used to investigate heat transfer enhancement. • Numerical simulations are performed in circular and elliptical tubes pre-heater. • Numerical results are analyzed using concept of synergy angle and synergy number. • Optimization of elliptical tube bundle by FSP has better performance

The Vibration Analysis of Tube Bundles Induced by Fluid Elastic Excitation in Shell Side of Heat Exchanger

Science.gov (United States)

Bao, Minle; Wang, Lu; Li, Wenyao; Gao, Tianze

2017-09-01

Fluid elastic excitation in shell side of heat exchanger was deduced theoretically in this paper. Model foundation was completed by using Pro / Engineer software. The finite element model was constructed and imported into the FLUENT module. The flow field simulation adopted the dynamic mesh model, RNG k-ε model and no-slip boundary conditions. Analysing different positions vibration of tube bundles by selecting three regions in shell side of heat exchanger. The results show that heat exchanger tube bundles at the inlet of the shell side are more likely to be failure due to fluid induced vibration.

Helically coiled tube heat exchanger

International Nuclear Information System (INIS)

Harris, A.M.

1981-01-01

In a heat exchanger such as a steam generator for a nuclear reactor, two or more bundles of helically coiled tubes are arranged in series with the tubes in each bundle integrally continuing through the tube bundles arranged in series therewith. Pitch values for the tubing in any pair of tube bundles, taken transverse to the path of the reactor coolant flow about the tubes, are selected as a ratio of two unequal integers to permit efficient operation of each tube bundle while maintaining the various tube bundles of the heat exchanger within a compact envelope. Preferably, the helix angle and tube pitch parallel to the path of coolant flow are constant for all tubes in a single bundle so that the tubes are of approximately the same length within each bundle

Measurement of unsteady flow forces in inline and staggered tube bundles with fixed and vibrating tubes

International Nuclear Information System (INIS)

Michel, A.; Heinecke, E.; Decken, C.B. von der.

1986-01-01

Unsteady flow forces arising in heat exchangers with cross-flow may lead to serious vibrations of the tubes. These vibrations can destroy the tubes in the end supports or in the baffles, which would require expensive repairs. The flow forces reach unexpectedly by high values if the vibration of the tube intensifies these forces. To clear up this coupling mechanism the flow forces and the vibration amplitude were measured simultaneously in a staggered and in an inline tube bundle. Considering the tube as a one-mass oscillator excited by the flow force, the main parameters can be derived, i.e. dynamic pressure, reduced mass, eigenfrequency and damping. These parameters form a dimensionless model number describing the coherence of the vibration amplitude and the force coefficient. The validity of this number has been confirmed by varying the test conditions. With the aid of this model number, the expected force coefficient can be calculated and then using a finite-element program information can be obtained about mechanical tensions and the lifetime of the heat exchanger tubes. With this model number the results of other authors, who measured the vibration amplitude only, could be confirmed in good agreement. The experiments were carried out in air with Reynolds numbers 10 4 5 . (orig.) [de

Heat transfer in a membrane assisted fluidised bed with immersed horizontal tubes

NARCIS (Netherlands)

Deshmukh, S.A.R.K.; Volkers, S.; van Sint Annaland, M.; Kuipers, J.A.M.

2004-01-01

The effect of gas permeation through horizontally immersed membrane tubes on the heat transfer characteristics in a membrane assisted fluidised bed was investigated experimentally. Local time-averaged heat transfer coefficients from copper tubes arranged in a staggered formation with the membrane

Heat transfer in a membrane assisted fluidized bed with immersed horizontal tubes

NARCIS (Netherlands)

Deshmukh, S.A.R.K.; Volkers, Sander; van Sint Annaland, M.; Kuipers, J.A.M.

2005-01-01

The effect of gas permeation through horizontally immersed membrane tubes on the heat transfer characteristics in a membrane assisted fluidized bed operated in the bubbling fluidization regime was investigated experimentally. Local time-averaged heat transfer coefficients from copper tubes arranged

An experimental study of aerosol penetration through horizontal tubes and strom-type loops

International Nuclear Information System (INIS)

Wong, F.S.; McFarland, A.R.; Anand, N.K.

1996-01-01

Because some designers of aerosol transport systems use the assumption that aerosol penetration through a system is maximized of the flow Reynolds number is 2,800, we have conducted tests to determine if such an assumption is appropriate. Although we do not believe that optimal performance of an aerosol sample transport system can be presented solely in terms of the Reynolds number, we have presented our results in terms of that parameter to compare our work with the results of an earlier study. Two types of experiments were performed. First, the penetration of liquid aerosol particles through horizontal tubes was experimentally investigated for a range of design and operational conditions. For a particle size of 10 μm aerodynamic diameter, the maximum penetration through a 6.7 mm diameter tube was associated with a Reynolds number of approximately 2,000; the maximum penetration through a tube of 15.9 mm occurred at a Reynolds number of about 3,000; and the maximum penetration through a 26.7 mm diameter tube occurred at about 4,000. It was also experimentally demonstrated that for a fixed flow rate through a horizontal tube, there is an optimum tube diameter for which the aerosol penetration is a maximum. An early study dealing with aerosol particle penetration through a 16.8 mm inside diameter loop of tubing (two vertical tubes, two horizontal tubes and three 90 degrees bends) suggested there was a fixed Reynolds number for optimal aerosol penetration independent of particle size. Those experiments were repeated here and the agreement with those tests is excellent. 16 refs., 8 figs., 3 tabs

Simulation of turbulent flow over staggered tube bundles using multi-relaxation time lattice Boltzmann method

International Nuclear Information System (INIS)

Park, Jong Woon; Choi, Hyun Gyung

2014-01-01

A turbulent fluid flow over staggered tube bundles is of great interest in many engineering fields including nuclear fuel rods, heat exchangers and especially a gas cooled reactor lower plenum. Computational methods have evolved for the simulation of such flow for decades and lattice Boltzmann method (LBM) is one of the attractive methods due to its sound physical basis and ease of computerization including parallelization. In this study to find computational performance of the LBM in turbulent flows over staggered tubes, a fluid flow analysis code employing multi-relaxation time lattice Boltzmann method (MRT-LBM) is developed based on a 2-dimensional D2Q9 lattice model and classical sub-grid eddy viscosity model of Smagorinsky. As a first step, fundamental performance MRT-LBM is investigated against a standard problem of a flow past a cylinder at low Reynolds number in terms of drag forces. As a major step, benchmarking of the MRT-LBM is performed over a turbulent flow through staggered tube bundles at Reynolds number of 18,000. For a flow past a single cylinder, the accuracy is validated against existing experimental data and previous computations in terms of drag forces on the cylinder. Mainly, the MRT-LBM computation for a flow through staggered tube bundles is performed and compared with experimental data and general purpose computational fluid dynamic (CFD) analyses with standard k-ω turbulence and large eddy simulation (LES) equipped with turbulence closures of Smagrinsky-Lilly and wall-adapting local eddy-viscosity (WALE) model. The agreement between the experimental and the computational results from the present MRT-LBM is found to be reasonably acceptable and even comparable to the LES whereas the computational efficiency is superior. (orig.)

Simulation of turbulent flow over staggered tube bundles using multi-relaxation time lattice Boltzmann method

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Woon; Choi, Hyun Gyung [Dongguk Univ., Gyeongju (Korea, Republic of). Nuclear and Energy Engineering Dept.

2014-02-15

A turbulent fluid flow over staggered tube bundles is of great interest in many engineering fields including nuclear fuel rods, heat exchangers and especially a gas cooled reactor lower plenum. Computational methods have evolved for the simulation of such flow for decades and lattice Boltzmann method (LBM) is one of the attractive methods due to its sound physical basis and ease of computerization including parallelization. In this study to find computational performance of the LBM in turbulent flows over staggered tubes, a fluid flow analysis code employing multi-relaxation time lattice Boltzmann method (MRT-LBM) is developed based on a 2-dimensional D2Q9 lattice model and classical sub-grid eddy viscosity model of Smagorinsky. As a first step, fundamental performance MRT-LBM is investigated against a standard problem of a flow past a cylinder at low Reynolds number in terms of drag forces. As a major step, benchmarking of the MRT-LBM is performed over a turbulent flow through staggered tube bundles at Reynolds number of 18,000. For a flow past a single cylinder, the accuracy is validated against existing experimental data and previous computations in terms of drag forces on the cylinder. Mainly, the MRT-LBM computation for a flow through staggered tube bundles is performed and compared with experimental data and general purpose computational fluid dynamic (CFD) analyses with standard k-ω turbulence and large eddy simulation (LES) equipped with turbulence closures of Smagrinsky-Lilly and wall-adapting local eddy-viscosity (WALE) model. The agreement between the experimental and the computational results from the present MRT-LBM is found to be reasonably acceptable and even comparable to the LES whereas the computational efficiency is superior. (orig.)

Fluid-Elastic Instability of U-Tube Bundle in Air-Water Two-Phase Flow

International Nuclear Information System (INIS)

Chu, In Cheol; Lee, Chang Hee; Yun, Young Jung; Chung, Heung June

2007-03-01

Using steam generator U-tube flow-induced vibration test facility, the flow-induced vibration characteristics of U-tube in row 34-44 and line 71-77 were investigated. Air and water at room temperature and near atmospheric pressure were used as working fluids. In the present experiments, followings were evaluated under two-phase cross-flow condition: the fundamental vibration responses and the critical gap velocity for a fluid-elastic instability of U-tubes, the damping ratio and hydrodynamic mass of U-tubes. In addition, the fluid-elastic instability factor, K, was preliminary assessed using Connors' relation. In the case of the U-tubes which are not supported by partial egg-crate in OPR100 steam generator, it has been found that the vibration displacement of those U-tubes are highly possible to exceed the design limit even by a turbulent excitation mechanism. The damping ratio of U-tubes measured in the present experiments was significantly higher than the OPR1000 steam generator design value. The fluid-elastic instability factor of U-tube bundle obtained in the present experiments were preliminary evaluated to be mostly in the range of 6.5-10.5

Process and device for detecting and localizing leaks in a tube bundle heat exchanger when it is stopped

International Nuclear Information System (INIS)

Germain, J.L.; Jeanneteau, E.; Loisy, F.

1986-01-01

The device can be used to detect the tubes presenting leaks in a tube bundle exchanger of a light water reactor. This device comprises a feeding point to fill the secondary part of the exchanger, in which the tubes are immersed, with a pressure mixture of vector gas (air) and helium. It has also a feeding point to establish in the tube a sweeping air flow. An analysis apparatus, such as a spectrograph, measures the helium content of air at the outlet of each tube [fr

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 thermal-hydraulic performance of PCCS with horizontal tube heat exchangers: single U-tube test

International Nuclear Information System (INIS)

Nakamura, Hideo; Anoda, Yoshinari; Arai, Kenji; Kurita, Tomohisa

2000-01-01

JAERI and JAPC started a cooperative study to verify performance of a PCCS (Passive Containment Cooling System) using horizontal heat exchanger for next-generation BWR in 1998. A test facility with a horizontal single U-tube was constructed in JAERI in 1999 to investigate fundamental condensation behavior under influences of non-condensable gas. Preliminary pre-test analyses were performed using RELAP5/ MOD3.2.1.2 code to expect the experimental outcomes by incorporating a correlation for condensation degradation because of non-condensable gas by Ueno et al. for better prediction. Preliminary results from both experiments (shakedown) and pre-test analyses indicated that the PCCS using horizontal U-tube heat exchanger is promising. Steam generated under assumed severe accident conditions; steam generation rate approx. = 1% core power, non-condensable gas concentration of 1% and simulated containment vessel pressure of 0.7 MPa, was totally condensed with a small differential pressure across inlet and outlet plenum. Experimental data will be accumulated to develop models and correlations for a better prediction of responses of the PCCS using horizontal heat exchanger during postulated severe accidents. (author)

Convective heat transfer in foams under laminar flow in pipes and tube bundles.

Science.gov (United States)

Attia, Joseph A; McKinley, Ian M; Moreno-Magana, David; Pilon, Laurent

2012-12-01

The present study reports experimental data and scaling analysis for forced convection of foams and microfoams in laminar flow in circular and rectangular tubes as well as in tube bundles. Foams and microfoams are pseudoplastic (shear thinning) two-phase fluids consisting of tightly packed bubbles with diameters ranging from tens of microns to a few millimeters. They have found applications in separation processes, soil remediation, oil recovery, water treatment, food processes, as well as in fire fighting and in heat exchangers. First, aqueous solutions of surfactant Tween 20 with different concentrations were used to generate microfoams with various porosity, bubble size distribution, and rheological behavior. These different microfoams were flowed in uniformly heated circular tubes of different diameter instrumented with thermocouples. A wide range of heat fluxes and flow rates were explored. Experimental data were compared with analytical and semi-empirical expressions derived and validated for single-phase power-law fluids. These correlations were extended to two-phase foams by defining the Reynolds number based on the effective viscosity and density of microfoams. However, the local Nusselt and Prandtl numbers were defined based on the specific heat and thermal conductivity of water. Indeed, the heated wall was continuously in contact with a film of water controlling convective heat transfer to the microfoams. Overall, good agreement between experimental results and model predictions was obtained for all experimental conditions considered. Finally, the same approach was shown to be also valid for experimental data reported in the literature for laminar forced convection of microfoams in rectangular minichannels and of macrofoams across aligned and staggered tube bundles with constant wall heat flux.

Random excitation forces in tube bundles subjected to two-phase cross-flow

International Nuclear Information System (INIS)

Taylor, C.E.; Pettigrew, M.J.; Currie, I.G.

1996-01-01

Data from two experimental programs have been analyzed to determine the characteristics of the random excitation forces associated with two-phase cross-flow in tube bundles. Large-scale air-water flow loops in France and Canada were used to generate the data. Tests were carried out on cantilevered, clamped-pinned, and clamped-clamped tubes in normal-square, parallel-triangular, and normal-triangular configurations. Either strain gages or force transducers were used to measure the vibration response of a centrally located tube as the tue array was subjected to a wide range of void fractions and flow rates. Power spectra were analyzed to determine the effect of parameters such as tube diameter, frequency, flow rate, void fraction, and flow regime on the random excitation forces. Normalized expressions for the excitation force power spectra were found to be flow-regime dependent. In the churn flow regime, flow rate and void fraction had very little effect on the magnitude of the excitation forces. In the bubble-plug flow regime, the excitation forces increased rapidly with flow rate and void fraction

Flooding in a loop with a vertical and a horizontal tube connected by an elbow

International Nuclear Information System (INIS)

Yan Changqi

1994-01-01

The experimental research of flooding and flow-reverse in a test loop which a vertical and a horizontal tube connected by an elbow is introduced. According to the experimental results, the effects of the elbow on flooding and flow-reverse is analyzed. The experimental results is compared with the results obtained in vertical tubes. The effect of horizontal tube length and hysteresis in de-flooding are analyzed. Dimensionless parameters was used in data process. The correlations for predicting the flooding point, de-flooding point, completed carry up and flow reverse points are given

Theoretical study of evaporation heat transfer in horizontal microfin tubes: stratified flow model

Energy Technology Data Exchange (ETDEWEB)

Honda, H; Wang, Y S [Kyushu Univ., Inst. for Materials Chemistry and Engineering, Kasuga, Fukuoka (Japan)

2004-08-01

The stratified flow model of evaporation heat transfer in helically grooved, horizontal microfin tubes has been developed. The profile of stratified liquid was determined by a theoretical model previously developed for condensation in horizontal microfin tubes. For the region above the stratified liquid, the meniscus profile in the groove between adjacent fins was determined by a force balance between the gravity and surface tension forces. The thin film evaporation model was applied to predict heat transfer in the thin film region of the meniscus. Heat transfer through the stratified liquid was estimated by using an empirical correlation proposed by Mori et al. The theoretical predictions of the circumferential average heat transfer coefficient were compared with available experimental data for four tubes and three refrigerants. A good agreement was obtained for the region of Fr{sub 0}<2.5 as long as partial dry out of tube surface did not occur. (Author)

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...

Establishment of Measurement Techniques for Sliding Bubble on a Horizontal Tube

International Nuclear Information System (INIS)

Kim, Yu-Na Kim; Park, Goon-Cherl; Cho, Hyoung-Kyu

2015-01-01

The mechanistic wall boiling model includes many parameters relevant with bubble behaviors, such as the bubble departure diameter, bubble lift-off diameter, bubble waiting time, etc. Although there have been a large number of studies investigating bubble behavior, the subjects of observation are almost bubbles on a plane or vertical tube. Since the bubble motion is highly influenced by the directions of gravitational force and the heating surfaces, it is expected that the bubble behavior on a horizontal tube is largely different from those on the other geometry. The heat exchanger of APR+ has horizontal U-tube configuration installed in a water pool, of which diameter is 50mm. The study aims to establish measurement techniques for sliding bubbles on a horizontal tube. The measurement parameters include the diameter, interface area, volume, and velocity of the bubble. Additionally, in order to analyze the force acting on the bubble, liquid velocity measurement method was proposed. This paper presents the procedure of the measurement; the phase separation technique, 3-D reconstruction technique, and velocity measurement techniques. For visualization of the sliding bubble behavior, bubble and liquid velocity measurement methods were established which use two high speed cameras and a continuous LASER for the PTV and PIV. Three steps for the bubble shape and velocity measurement (the phase separation, 3-D reconstruction, and velocity calculation), were successfully set up and verified. A PIV technique which uses two different time duration for two regions where the velocity difference is huge was proposed and tested. Using these methods, various information regarding a sliding bubble can be obtained such as bubble and liquid velocities, shape, volume, surface area etc

Calculation of heat transfer in transversely stream-lined tube bundles with chess arrangement

International Nuclear Information System (INIS)

Migaj, V.K.

1978-01-01

A semiempirical theory of heat transfer in transversely stream-lined chess-board tube bundles has been developed. The theory is based on a single cylinder model and involves external flow parameter evaluation on the basis of the solidification principle of a vortex zone. The effect of turbulence is estimated according to experimental results. The method is extended to both average and local heat transfer coefficients. Comparison with experiment shows satisfactory agreement

Sea water desalination by horizontal tubes evaporator

International Nuclear Information System (INIS)

Mohammadi, H.K.; Mohit, M.

1986-01-01

Desalinated water supplies are one of the problems of the nuclear power plants located by the seas. This paper explains saline water desalination by a Horizontal Tube Evaporator (HTE) and compares it with flash evaporation. A thermo compressor research project using HTE method has been designed, constructed, and operated at the Esfahan Nuclear Technology Center ENTC. The poject's ultimate goal is to obtain empirical formulae based on data gathered during operation of the unit and its subsequent development towards design and construction of desalination plants on an industrial scale

Apparatus for inspecting and repairing a pressurized-water reactor's steam generator heat exchanger tubes

International Nuclear Information System (INIS)

Mueller, O.; Roettger, H.; Kasti, H.; Hagen, H.G.

1976-01-01

Described is an apparatus provided for use with a pressurized-water reactor' steam generator having a manifold chamber enclosing the bottom side of a horizontal tube sheet having holes therethrough in which are mounted the tubes of a heat exchanger tube bundle. The manifold chamber has a manhole giving access to the tube's bottom side to permit internal inspection or repair of the tubes by registration of an end of a flexible guide conduit with the tube sheet holes and through which a flexible carrier can be guided for insertion via these holes in the tube sheet and through the tubes extending from the tube sheet's other side

Research for rolling effects on flow pattern of gas-water flow in horizontal tubes

International Nuclear Information System (INIS)

Luan Feng; Yan Changqi

2007-01-01

The flow pattern transition of two-phase flow is caused by the inertial force resulted from rolling and incline of horizontal tubes under rolling state. an experimental study on the flow patterns of gas-water flow was carried out in horizontal tubes under rolling state, which rolling period is 15 second and rolling angle is 10 degrees, and a pattern flow picture is shown. It was found that there are two flow patterns in one rolling period under some gas flux and water flux. (authors)

Fluidelastic instability of a tube bundle preferentially flexible in the flow direction to simulate u-bend in-plane vibration

International Nuclear Information System (INIS)

Pettigrew, M.; Violette, R.; Mureithi, N.

2006-01-01

Almost all the available data about fluidelastic instability of heat exchanger tube bundles concerns tubes that are axisymetrically flexible. In those cases, the instability is found to be mostly in the direction transverse to the flow. Thus, the direction parallel to the flow has raised less concern in terms of bundle stability. However, the flat bar supports used in steam generators for preventing U-tubes vibration may not be as effective in the in-plane direction as in the out-of-plane direction. The possibility that fluidelastic instability can develop in the flow direction must then be assessed. In the present work, tests were done to study the fluidelastic instability of a cluster of seven tubes much more flexible in the flow direction than in the lift direction. The array configuration is rotated triangular with a pitch to diameter ratio of 1.5. The array was subjected to two-phase (air-water) cross flow. Well-defined fluidelastic instabilities were observed albeit at somewhat higher flow velocities than for axisymetrically flexible tubes. This so far unknown phenomenon may be of concern if some supports become ineffective in the in-plane direction. (author)

Experimental determination of the local temperature distribution in the cladding tubes of a sodium-cooled pin bundle caused by grid spacers

International Nuclear Information System (INIS)

Moeller, R.; Tschoeke, H.

1980-01-01

The cladding tubes of reactor core elements are highly stressed structural elements. Their careful design includes the following: (a) the mathematical determination of the maximum cladding tube temperatures; (b) the determination of the maximum permissible fatigue strengths and creep strains of the materials; and (c) the safety distance between the nominal cladding tube hot spots and the permissible extreme cladding tube temperature. The maximum cladding tube temperatures occur on the top edge of the core and, due to radial power gradients, in the wrapper-wall region of a pin bundle. If grid spacers are now used for fixing the pins as in the SNR fuel elements, a careful check must be made of whether and to what degree temperature peaks in the region of the supports have an influence on the cladding tube design. Initial experimental investigations on a sodium-cooled pin bundle model of the SNR-300 fuel element were carried out to throw light on these special problems. This is reported in the following together with the results so far obtained. (U.K.)

The evaluation of validity of the RELAP5/Mod3 flow regime map for horizontal small diameter tubes at low pressure

Energy Technology Data Exchange (ETDEWEB)

Agafonova, N. [St. Petersburg State Technical Univ. (Russian Federation); Banati, J. [Lappeenranta Univ. of Technology (Finland)

1997-12-31

RELAP5/MOD3 code was developed for Western type power water reactors with vertical steam generators. Thus, this code should be validated also for WWER design with horizontal steam generators. In application for horizontal steam generators the situation with two-phase flow inside small diameter tubes is possible when the first circuit pressure drops in accident below the pressure level in the boiling water. It is known that computer codes have not always modelled correctly the two-phase flow inside horizontal tubes at low pressures (less than 4-6 MPa). It may be the result of erroneous prediction of the flow regime. Correct prediction of the flow regime is especially important for the fully or partly stratified flow in horizontal tubes. The aim of this study is the attempt of verification of the flow regime map, which is used in the RELAP5/MOD3 computer code for two-phase flow in horizontal small diameter tubes. `Small diameter tube` means according RELAP5/MOD3 that the inner diameter of the tube is less (or equal) than 0.018 m. The inner tube diameter in horizontal steam generators is equal 0.013 m. (orig.). 19 refs.

The evaluation of validity of the RELAP5/Mod3 flow regime map for horizontal small diameter tubes at low pressure

Energy Technology Data Exchange (ETDEWEB)

Agafonova, N [St. Petersburg State Technical Univ. (Russian Federation); Banati, J [Lappeenranta Univ. of Technology (Finland)

1998-12-31

RELAP5/MOD3 code was developed for Western type power water reactors with vertical steam generators. Thus, this code should be validated also for WWER design with horizontal steam generators. In application for horizontal steam generators the situation with two-phase flow inside small diameter tubes is possible when the first circuit pressure drops in accident below the pressure level in the boiling water. It is known that computer codes have not always modelled correctly the two-phase flow inside horizontal tubes at low pressures (less than 4-6 MPa). It may be the result of erroneous prediction of the flow regime. Correct prediction of the flow regime is especially important for the fully or partly stratified flow in horizontal tubes. The aim of this study is the attempt of verification of the flow regime map, which is used in the RELAP5/MOD3 computer code for two-phase flow in horizontal small diameter tubes. `Small diameter tube` means according RELAP5/MOD3 that the inner diameter of the tube is less (or equal) than 0.018 m. The inner tube diameter in horizontal steam generators is equal 0.013 m. (orig.). 19 refs.

New insights into controlling tube-bundle fouling using alternative amines

International Nuclear Information System (INIS)

Turner, C.W.; Klimas, S.J.; Guzonas, D.A.; Frattini, P.L.; Fruzzetti, K.

2002-01-01

A volatile amine is added to the secondary heat-transport system of a nuclear power plant to reduce the rate of corrosion and corrosion product transport in the feedwater and to protect steam generator (SG) crevices and materials exposed to steam condensate. Volatility and base strength of the amine at the SG operating temperature are two important considerations when choosing the optimum amine (or mixture of amines) for corrosion control in the steam cycle. Atomic Energy of Canada Limited (AECL) and Electric Power Research Institute (EPRI) have been collaborating in an extensive investigation of the effectiveness of amines at controlling the rate of tube-bundle fouling under SG operating conditions. Tests have been performed using a radiotracing technique in a high-temperature fouling loop facility at Chalk River Laboratories operated by AECL. This investigation has provided new insights into the role played by the amine in determining the rate of tube-bundle fouling in the SG. These insights are being used by AECL and EPRI to develop criteria for the selection of an amine that has optimum properties for both corrosion control and deposit control in the secondary heat transport system. The investigation has found that the rate of tube-bundle fouling is strongly dependent upon the surface chemistry of the corrosion products. For example, the fouling rates of fully oxidized iron oxides, such as hematite and lepidocrocite, are at least an order of magnitude greater than the fouling rate of magnetite under identical operating conditions. The difference is related to the sign of the surface charge on the corrosion products at temperature. The choice of amine for pH-control also influences the fouling rate. This was originally thought to be a surface-charge effect as well, but recent tests have suggested that it is related to the role that the amine plays in governing the rate of deposit consolidation on the heat-transfer surface. Amines that promote a high rate of

Tube bundle system: for monitoring of coal mine atmosphere.

Science.gov (United States)

Zipf, R Karl; Marchewka, W; Mohamed, K; Addis, J; Karnack, F

2013-05-01

A tube bundle system (TBS) is a mechanical system for continuously drawing gas samples through tubes from multiple monitoring points located in an underground coal mine. The gas samples are drawn via vacuum pump to the surface and are typically analyzed for oxygen, methane, carbon dioxide and carbon monoxide. Results of the gas analyses are displayed and recorded for further analysis. Trends in the composition of the mine atmosphere, such as increasing methane or carbon monoxide concentration, can be detected early, permitting rapid intervention that prevents problems, such as a potentially explosive atmosphere behind seals, fire or spontaneous combustion. TBS is a well-developed technology and has been used in coal mines around the world for more than 50 years. Most longwall coal mines in Australia deploy a TBS, usually with 30 to 40 monitoring points as part of their atmospheric monitoring. The primary uses of a TBS are detecting spontaneous combustion and maintaining sealed areas inert. The TBS might also provide mine atmosphere gas composition data after a catastrophe occurs in an underground mine, if the sampling tubes are not damaged. TBSs are not an alternative to statutory gas and ventilation airflow monitoring by electronic sensors or people; rather, they are an option to consider in an overall mine atmosphere monitoring strategy. This paper describes the hardware, software and operation of a TBS and presents one example of typical data from a longwall coal mine.

Sheared bioconvection in a horizontal tube

Science.gov (United States)

Croze, O. A.; Ashraf, E. E.; Bees, M. A.

2010-12-01

The recent interest in using microorganisms for biofuels is motivation enough to study bioconvection and cell dispersion in tubes subject to imposed flow. To optimize light and nutrient uptake, many microorganisms swim in directions biased by environmental cues (e.g. phototaxis in algae and chemotaxis in bacteria). Such taxes inevitably lead to accumulations of cells, which, as many microorganisms have a density different to the fluid, can induce hydrodynamic instabilites. The large-scale fluid flow and spectacular patterns that arise are termed bioconvection. However, the extent to which bioconvection is affected or suppressed by an imposed fluid flow and how bioconvection influences the mean flow profile and cell transport are open questions. This experimental study is the first to address these issues by quantifying the patterns due to suspensions of the gravitactic and gyrotactic green biflagellate alga Chlamydomonas in horizontal tubes subject to an imposed flow. With no flow, the dependence of the dominant pattern wavelength at pattern onset on cell concentration is established for three different tube diameters. For small imposed flows, the vertical plumes of cells are observed merely to bow in the direction of flow. For sufficiently high flow rates, the plumes progressively fragment into piecewise linear diagonal plumes, unexpectedly inclined at constant angles and translating at fixed speeds. The pattern wavelength generally grows with flow rate, with transitions at critical rates that depend on concentration. Even at high imposed flow rates, bioconvection is not wholly suppressed and perturbs the flow field.

Fluidic delivery of homogeneous solutions through carbon tube bundles

International Nuclear Information System (INIS)

Srikar, R; Yarin, A L; Megaridis, C M

2009-01-01

A wide array of technological applications requires localized high-rate delivery of dissolved compounds (in particular, biological ones), which can be achieved by forcing the solutions or suspensions of such compounds through nano or microtubes and their bundled assemblies. Using a water-soluble compound, the fluorescent dye Rhodamine 610 chloride, frequently used as a model drug release compound, it is shown that deposit buildup on the inner walls of the delivery channels and its adverse consequences pose a severe challenge to implementing pressure-driven long-term fluidic delivery through nano and microcapillaries, even in the case of such homogeneous solutions. Pressure-driven delivery (3-6 bar) of homogeneous dye solutions through macroscopically-long (∼1 cm) carbon nano and microtubes with inner diameters in the range 100 nm-1 μm and their bundled parallel assemblies is studied experimentally and theoretically. It is shown that the flow delivery gradually shifts from fast convection-dominated (unobstructed) to slow jammed convection, and ultimately to diffusion-limited transport through a porous deposit. The jamming/clogging phenomena appear to be rather generic: they were observed in a wide concentration range for two fluorescent dyes in carbon nano and microtubes, as well as in comparable transparent glass microcapillaries. The aim of the present work is to study the physics of jamming, rather than the chemical reasons for the affinity of dye molecules to the tube walls.

Comparative assessment of condensation models for horizontal tubes

International Nuclear Information System (INIS)

Schaffrath, A.; Kruessenberg, A.K.; Lischke, W.; Gocht, U.; Fjodorow, A.

1999-01-01

The condensation in horizontal tubes plays an important role e.g. for the determination of the operation mode of horizontal steam generators of VVER reactors or passive safety systems for the next generation of nuclear power plants. Two different approaches (HOTKON and KONWAR) for modeling this process have been undertaken by Forschungszentrum Juelich (FZJ) and University for Applied Sciences Zittau/Goerlitz (HTWS) and implemented into the 1D-thermohydraulic code ATHLET, which is developed by the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH for the analysis of anticipated and abnormal transients in light water reactors. Although the improvements of the condensation models are developed for different applications (VVER steam generators - emergency condenser of the SWR1000) with strongly different operation conditions (e.g. the temperature difference over the tube wall in HORUS is up to 30 K and in NOKO up to 250 K, the heat flux density in HORUS is up to 40 kW/m 2 and in NOKO up to 1 GW/m 2 ) both models are now compared and assessed by Forschungszentrum Rossendorf FZR e.V. Therefore, post test calculations of selected HORUS experiments were performed with ATHLET/KONWAR and compared to existing ATHLET and ATHLET/HOTKON calculations of HTWS. It can be seen that the calculations with the extension KONWAR as well as HOTKON improve significantly the agreement between computational and experimental data. (orig.) [de

High temperature technological heat exchangers and steam generators with helical coil assembly tube bundle

International Nuclear Information System (INIS)

Korotaev, O.J.; Mizonov, N.V.; Nikolaevsky, V.B.; Nazarov, E.K.

1990-01-01

Analysis of thermal hydraulics characteristics of nuclear steam generators with different tube bundle arrangements and waste heat boilers for ammonia production units was performed on the basis of operating experience results and research and development data. The present report involves the obtained information. The estimations of steam generator performances and repair-ability are given. The significant temperature profile of the primary and secondary coolant flows are attributed to all steam generator designs. The intermediate mixing is found to be an effective means of temperature profile overcoming. At present the only means to provide an effective mixing in heat exchangers of the following types: straight tubes, field tubes, platen tubes and multibank helical coil tubes (with complicated bend distribution along their length) are section arrangements in series in conjunction with forced and natural mixing in connecting lines. Development of the unificated system from mini helical coil assemblies allows to design and manufacture heat exchangers and steam generators within the wide range of operating conditions without additional expenses on the research and development work

Theoretical modeling of steam condensation in the presence of a noncondensable gas in horizontal tubes

International Nuclear Information System (INIS)

Lee, Kwon-Yeong; Kim, Moo Hwan; Kim, Moo Hwan

2008-01-01

A theoretical model was developed to investigate a steam condensation with a noncondensable gas in a horizontal tube. The heat transfer through the vapor/noncondensable gas mixture boundary layer consists of the sensible heat transfer and the latent heat transfer given up by the condensing vapor, and it must equal that from the condensate film to the tube wall. Therefore, the total heat transfer coefficient is given by the film, condensation and sensible heat transfer coefficients. The film heat transfer coefficients of the upper and lower portions of the tube were calculated separately from Rosson and Meyers (1965) correlation. The heat and mass transfer analogy was used to analyze the steam/noncondensable gas mixture boundary layer. Here, the Nusselt and Sherwood numbers in the gas phase were modified to incorporate the effects of condensate film roughness, suction, and developing flow. The predictions of the theoretical model for the experimental heat transfer coefficients at the top and bottom of the tube were reasonable. The calculated heat transfer coefficients at the top of the tube were higher than those at the bottom of it, as experimental results. As the temperature potential at the top of tube was lower than that at the bottom of it, the heat fluxes at the upper and lower portions of the tube were similar to each other. Generally speaking, however, the model predictions showed a good agreement with experimental data. The new empirical correlation proposed by Lee and Kim (2008) for the vertical tube was applied to the condensation of steam/noncondensable mixture in a horizontal tube. Nusselt theory and Chato correlation were used to calculate the heat transfer coefficients at top and bottom of the horizontal tube, respectively. The predictions of the new empirical correlation were good and very similar with the theoretical model. (author)

Experimental Study about Two-phase Damping Ratio on a Tube Bundle Subjected to Homogeneous Two-phase Flow

Energy Technology Data Exchange (ETDEWEB)

Sim, Woo Gun; Dagdan, Banzragch [Hannam Univ., Daejeon (Korea, Republic of)

2017-03-15

Two-phase cross flow exists in many shell-and-tube heat exchangers such as condensers, evaporators, and nuclear steam generators. The drag force acting on a tube bundle subjected to air/water flow is evaluated experimentally. The cylinders subjected to two-phase flow are arranged in a normal square array. The ratio of pitch to diameter is 1.35, and the diameter of the cylinder is 18 mm. The drag force along the flow direction on the tube bundles is measured to calculate the drag coefficient and the two-phase damping ratio. The two-phase damping ratios, given by the analytical model for a homogeneous two-phase flow, are compared with experimental results. The correlation factor between the frictional pressure drop and the hydraulic drag coefficient is determined from the experimental results. The factor is used to calculate the drag force analytically. It is found that with an increase in the mass flux, the drag force, and the drag coefficients are close to the results given by the homogeneous model. The result shows that the damping ratio can be calculated using the homogeneous model for bubbly flow of sufficiently large mass flux.

A model of the fluid temperature field in a turbulent flow parallel to heated tube bundle

International Nuclear Information System (INIS)

Carvalho Tofani, P. de.

1986-01-01

Basic understanding of thermal-hydraulic phenomena is essential to achieving reactor fuel assembly performance analysis. In this paper, a dimensionless parameter - a normalized fluid temperature - is defined and applied to fluid temperature measurements at particular positions at the exit plane of a bank of nine heated tubes, under different transverse heat flux shapes. This parameter presents an asymptotic trend to equilibrium values, which depend upon considered positions and flux shapes; when increasing the bulk Reynolds Number. Proposed correlations underlie the present approach to predict the fluid temperature field within the tube bundle. (Author) [pt

Experimental and numerical studies of turbulent flow in an in-line tube bundles

Directory of Open Access Journals (Sweden)

Aounalah Mohamed

2012-04-01

Full Text Available In the present paper an experimental and a numerical simulation of the turbulent flow in an in-line tube bundles have been performed. The experiments were carried out using a subsonic wind tunnel. The pressure distributions along the tubes (22 circumferential pressure taping were determined for a variation of the azimuthal angle from 0 to 360deg. The drag and lift forces are measured using the TE 44 balance. The Navier-Stokes equations of the turbulent flow are solved using Reynolds Stress and K-ε, turbulence models (RANS provided by Fluent CFD code. An adapted grid using static pressure, pressure coefficient and velocity gradient, furthermore, a second order upwind scheme were used. The obtained results from the experimental and numerical studies show a satisfactory agreement.

Condensation heat transfer of steam on a single horizontal tube

Science.gov (United States)

Graber, K. A.

1983-06-01

An experimental apparatus was designed, constructed and instrumented in an effort to systematically and carefully study the condensation heat-transfer coefficient on a single, horizontal tube. A smooth, thick-walled copper tube of length 133.5 mm, with an outside diameter of 15.9 mm and an inside diameter of 12.7 mm was instrumented with six wall thermocouples. The temperature rise across the test section was measured accurately using quartz crystal thermometers. The inside heat-transfer coefficient was determined using the Sieder-Tate correlation with leading coefficient of 0.029. Initial steam side data were taken at atmospheric pressure to test the data acquisition/reduction computer programs.

Assessment of CCFL model of RELAP5/MOD3 against simple vertical tubes and rod bundle tests

International Nuclear Information System (INIS)

Cho, Sung Jae; Arne, Nam Sung; Chung, Bub Dong; Kim, Hho Jung

1991-01-01

The CCFL model used in RELAP5/MOD3 version 5m5 has been assessed against simple vertical tubes and rod bundle tests performed at a facility of Korea Atomic Energy Research Institute. The effect of changes in tube diameter and nodalization of tube section were investigated. The roles of interfacial drags on the flooding characteristics are discussed. Difference between the calculation and the experiment are also discussed. A comparison between model assessment results and the test data showed that the calculated value lay well on the experimental flooding curve specified by user, but the pressure jump before onset of flooding was not calculated

CFD analysis of multiphase coolant flow through fuel rod bundles in advanced pressure tube nuclear reactors

International Nuclear Information System (INIS)

Catana, A.; Turcu, I.; Prisecaru, I.; Dupleac, D.; Danila, N.

2010-01-01

The key component of a pressure tube nuclear reactor core is pressure tube filled with a stream of fuel bundles. This feature makes them suitable for CFD thermal-hydraulic analysis. A methodology for CFD analysis applied to pressure tube nuclear reactors is presented in this paper, which is focused on advanced pressure tube nuclear reactors. The complex flow conditions inside pressure tube are analysed by using the Eulerian multiphase model implemented in FLUENT CFD computer code. Fuel rods in these channels are superheated but the liquid is under high pressure, so it is sub-cooled in normal operating conditions on most of pressure tube length. In the second half of pressure tube length, the onset of boiling occurs, so the flow consists of a gas liquid mixture, with the volume of gas increasing along the length of the channel in the direction of the flow. Limited computer resources enforced us to use CFD analysis for segments of pressure tube. Significant local geometries (junctions, spacers) were simulated. Main results of this work are: prediction of main thermal-hydraulic parameters along pressure tube including CHF evaluation through fuel assemblies. (authors)

Flow vibrations and dynamic instability of heat exchanger tube bundles

International Nuclear Information System (INIS)

Granger, S.; Langre, E. de

1995-01-01

This paper presents a review of external-flow-induced vibration of heat exchanger tube bundles. Attention is focused on a dynamic instability, known as ''fluidelastic instability'', which can develop when flow is transverse to the tube axis. The main physical models proposed in the literature are successively reviewed in a critical way. As a consequence, some concepts are clarified, some a priori plausible misinterpretations are rejected and finally, certain basic mechanisms, induced by the flow-structure interaction and responsible for the ultimate onset of fluidelastic instability, are elucidated. Design tools and methods for predictive analysis of industrial cases are then presented. The usual design tool is the ''stability map'', i.e. an empirical correlation which must be interpreted in a conservative way. Of course, when using this approach, the designer must also consider reasonable safety margins. In the area of predictive analysis, the ''unsteady semi-analytical models'' seem to be a promising and efficient methodology. A modern implementation of these ideas mix an original experimental approach for taking fluid dynamic forces into account, together with non-classical numerical methods of mechanical vibration. (authors). 20 refs., 9 figs

Assessment of RELAP5/MOD3.3 condensation models for the tube bundle condensation in the PCCS of ESBWR

International Nuclear Information System (INIS)

Zhou, W.; Wolf, B.; Revankar, S.T.

2011-01-01

The passive containment condenser system (PCCS) in an ESBWR reactor consists of vertical tube bundle submerged in a large pool of water. The condensation model for the PCCS in a thermalhydraulics code RELAP5/MOD3.3 consists of the default Nusselt model and an alternate condensation model from UCB condensation correlation. An assessment of the PCCS condensation model in RELAP5/MOD3.3 was carried out using experiments conducted on a single tube and tube bundle PCCS tests at Purdue University. The experimental conditions were simulated with the default and the alternate condensation models in the REALP5/MOD3.3 beta version of the code. The default model and the UCB model (alternate model) give quite different results on condensation heat transfer for the PCCS. The default model predicts complete condensation well whereas the UCB model predicts the through flow condensation well. Based on this study it was found that none of the models in REALP5 can predict complete condensation as well as the through flow condensation well. (author)

Assessment of RELAP5/MOD3.3 condensation models for the tube bundle condensation in the PCCS of ESBWR

Energy Technology Data Exchange (ETDEWEB)

Zhou, W., E-mail: wenzzhou@cityu.edu.hk [Department of Mechanical and Biomedical Engineering, City University of Hong Kong, Hong Kong (China); Wolf, B. [Purdue University, West Lafayette, IN 47907 (United States); Revankar, S. [Purdue University, West Lafayette, IN 47907 (United States); POSTECH, Pohang (Korea, Republic of)

2013-11-15

The passive containment condenser system (PCCS) in an ESBWR reactor consists of vertical tube bundle submerged in a large pool of water. The condensation model for the PCCS in a thermalhydraulics code RELAP5/MOD3.3 consists of the default Nusselt model and an alternate condensation model from UCB condensation correlation. An assessment of the PCCS condensation model in RELAP5/MOD3.3 was carried out using experiments conducted on a single tube and tube bundle PCCS tests at Purdue University. The experimental conditions were simulated with the default and the alternate condensation models in the REALP5/MOD3.3 beta version of the code. The default model and the UCB model (alternate model) give quite different results on condensation heat transfer for the PCCS. The default model predicts complete condensation well whereas the UCB model predicts the through flow condensation well. Based on this study it was found that none of the models in REALP5 can predict complete condensation as well as the through flow condensation well.

Coulomb friction modelling in numerical simulations of vibration and wear work rate of multispan tube bundles

International Nuclear Information System (INIS)

Antunes, J.; Axisa, F.; Beaufils, B.; Guilbaud, D.

1990-01-01

The working life of heat exchanger multispan tube bundles subjected to flow-induced vibration, is heavily dependent on nonlinear interaction between the loosely supported tubes and their supports. Reliable wear prediction techniques must account for a number of factors controlling impact-sliding tube response, such as tube support gap, contact stiffness, impact damping, Coulomb friction and squeeze film effect at supports. Tube fretting wear potential risk may then be adequately quantified by an equivalent wear work rate. A simple model is presented which accounts for the key aspects of dry friction and is well suited to the efficient explicit numerical integration schemes, specifically through nonlinear model superposition. Extensive parametric two-dimensional simulations, under random vibration induced by flow turbulence, are presented. Also, the effect of permanent tube-support preload, arising from cross flow drag, tube-support misalignment and thermal expansion, is investigated. Results show that frictional forces consistently reduce wear work rates, which decrease for high values of the coefficient of friction. Such reductions may be extremely important for the limiting case when preload and frictional forces are of sufficient magnitude to overcome dynamic forces, preventing tube-support relative motion. (author)

Application of advanced optical probe instrumentation in steam generator tube bundles

International Nuclear Information System (INIS)

Bouchter, J.C.; Gouirand, J.M.; Haquet, J.F.; Ivars, J.F.

1990-01-01

The Department of Energy Transfer (DTE) of the French CEA has been developing for about 15 years optical probe techniques in order to better understand and predict nuclear components dealing with two-phase flows. More recently, in the scope of an International Program, the CEA has made an intensive use of bi-optical probes in order to very precisely investigate the distributions of void fraction and gas velocity in the secondary side of a Steam-generator mock-up operating with Freon 114 (80 degrees C, 9 x 10 5 Pa at nominal conditions). At the present time, the success of this program and the recent progress achieved in the technology of the probe, in particular to withstand higher pressures and temperatures allow us to reasonably think that this device will be soon available for industrial uses. So, this paper deals with the present state of the art of the technique within CEA and in particular it successively describes: what is required of a practical point of view when it comes to perform local measurements within tube bundles and what have been the technical choices to meet these requirements; how the bi-optical probe does operate with an emphasis on the signal processing description; how the whole device accuracy, i.e., the bi-optical probe plus its complete acquisition and signal processing chain, is determined by a calibration procedure comparing first separately then all together the different components to independent numerical and physical reference methods; typical examples of measurements of the emulsion fine structure within tube bundle subchannels as regards with void fraction, gas velocity and bubble granulometries; finally, the recent progress accomplished in terms of, higher reliability, resolution, pressure and temperature resistance

Condensation heat transfer of R22 and R410A in horizontal smooth and microfin tubes

Energy Technology Data Exchange (ETDEWEB)

Kim, Man-Hoe; Shin, Joeng-Seob [Korea Advanced Institute of Science and Technology, Daejeon (Korea). Department of Mechanical Engineering

2005-09-01

An experimental investigation of condensation heat transfer in 9.52 mm O.D. horizontal copper tubes was conducted using R22 and R410A. The test rig had a straight, horizontal test section with an active length of 0.92 m and was cooled by the heat transfer fluid (cold water) circulated in a surrounding annulus. Constant heat flux of 11.0 kW/m{sup 2} was maintained throughout the experiment and refrigerant quality varied from 0.9 to 0.1. The condensation test results at 45 {sup o}C were reported for 40-80 kg/h mass flow rate. The local and average condensation coefficients for seven microfin tubes were presented compared to those for a smooth tube. The average condensation coefficients of R22 and R410A for the microfin tubes were 1.7-3.19 and 1.7-2.94 times larger than those in smooth tube, respectively. (author)

Fluid velocity in outer channels of tube bundle of PGV-1 steam generator prototype

International Nuclear Information System (INIS)

Salgo, C.

1979-01-01

This paper deals with the determination of the fluid velocity and pressure in every channel outside the tube bundle of the prototype model of the steam generator PGV-1, designed jointly by CNEN and NIRA. The results are obtained by the numerical solution of a system of algebraic and differential equations deduced by a mathematical ''channel'' model. Such results agree with the experiments performed on a model of the prototype PGV-1 by Alsthom Technique des Fluides

PACTEL: Experiments on the behaviour of the new horizontal steam generator

Energy Technology Data Exchange (ETDEWEB)

Kouhia, J.; Riikonen, V.; Purhonen, H. [VTT Energy, Lappeenranta (Finland)

1995-12-31

Experiments were performed to study the behaviour of the PACTEL facility, a medium scale integral test loop simulating VVER 440 pressurized water reactors. The study focused on the operation of the new horizontal steam generator model installed in PACTEL. Three experiments were conducted: a small-break test to observe the steam generator behaviour over a range of primary coolant inventories, a hot leg loop seal experiment to study the cyclic behaviour of a loop seal and a loss of secondary side feedwater test to examine the effect of uncovered tubes in the steam generator. A reverse flow was observed in the lower part of the U-tube bundle of the steam generator during natural circulation. The flow reversal point dropped when the tubes uncovered, during secondary inventory reduction. (orig.). 5 refs.

PACTEL: Experiments on the behaviour of the new horizontal steam generator

Energy Technology Data Exchange (ETDEWEB)

Kouhia, J; Riikonen, V; Purhonen, H [VTT Energy, Lappeenranta (Finland)

1996-12-31

Experiments were performed to study the behaviour of the PACTEL facility, a medium scale integral test loop simulating VVER 440 pressurized water reactors. The study focused on the operation of the new horizontal steam generator model installed in PACTEL. Three experiments were conducted: a small-break test to observe the steam generator behaviour over a range of primary coolant inventories, a hot leg loop seal experiment to study the cyclic behaviour of a loop seal and a loss of secondary side feedwater test to examine the effect of uncovered tubes in the steam generator. A reverse flow was observed in the lower part of the U-tube bundle of the steam generator during natural circulation. The flow reversal point dropped when the tubes uncovered, during secondary inventory reduction. (orig.). 5 refs.

Post-Dryout Heat Transfer to a Refrigerant Flowing in Horizontal Evaporator Tubes

Science.gov (United States)

Mori, Hideo; Yoshida, Suguru; Kakimoto, Yasushi; Ohishi, Katsumi; Fukuda, Kenichi

Studies of the post-dryout heat transfer were made based on the experimental data for HFC-134a flowing in horizontal smooth and spiral1y grooved (micro-fin) tubes and the characteristics of the post-dryout heat transfer were c1arified. The heat transfer coefficient at medium and high mass flow rates in the smooth tube was lower than the single-phase heat transfer coefficient of the superheated vapor flow, of which mass flow rate was given on the assumption that the flow was in a thermodynamic equilibrium. A prediction method of post-dryout heat transfer coefficient was developed to reproduce the measurement satisfactorily for the smooth tube. The post dryout heat transfer in the micro-fin tube can be regarded approximately as a superheated vapor single-phase heat transfer.

An evaluation method of critical velocity for self-excited vibration of cross-shaped tube bundle in cross flow

International Nuclear Information System (INIS)

Inada, Fumio; Nishihara, Takashi; Yasuo, Akira; Morita, Ryo

2002-01-01

The applicability of the cross-shaped tube bundle as a lower plenum component of pressure vessel is examined to develop a next generation LWR in Japanese electric utilities. The flow-induced vibration characteristics are not understood well. Methods to evaluate turbulence induced vibration and vortex induced vibration were proposed by CRIEPI. In this study, vibration response is obtained experimentally to propose a method to evaluate self-excited vibration of cross-shaped tube bundle. The self-excited vibration was found to be generated when nondimensional flow velocity was above a critical value. The nondimensional critical velocity of normal configuration is 15% smaller than that of staggered configuration, which means that the nondimensional critical velocity of normal configuration can give conservative evaluation. The result of Reynolds number Re=6.2 x 10 4 agrees well with that of Re=6.8 x 10 5 , in which region, the effect of Reynolds number on the critical velocity is small. (author)

Thermal load non-uniformity estimation for superheater tube bundle damage evaluation

Directory of Open Access Journals (Sweden)

Naď Martin

2018-01-01

Full Text Available Industrial boiler damage is a common phenomenon encountered in boiler operation which usually lasts several decades. Since boiler shutdown may be required because of localized failures, it is crucial to predict the most vulnerable parts. If damage occurs, it is necessary to perform root cause analysis and devise corrective measures (repairs, design modifications, etc.. Boiler tube bundles, such as those in superheaters, preheaters and reheaters, are the most exposed and often the most damaged boiler parts. Both short-term and long-term overheating are common causes of tube failures. In these cases, the design temperatures are exceeded, which often results in decrease of remaining creep life. Advanced models for damage evaluation require temperature history, which is available only in rare cases when it has been measured and recorded for the whole service life. However, in most cases it is necessary to estimate the temperature history from available operation history data (inlet and outlet pressures and temperatures etc.. The task may be very challenging because of the combination of complex flow behaviour in the flue gas domain and heat transfer phenomena. This paper focuses on estimating thermal load non-uniformity on superheater tubes via Computational Fluid Dynamics (CFD simulation of flue gas flow including heat transfer within the domain consisting of a furnace and a part of the first stage of the boiler.

Assessment of ASSERT-PV for prediction of critical heat flux in CANDU bundles

International Nuclear Information System (INIS)

Rao, Y.F.; Cheng, Z.; Waddington, G.M.

2014-01-01

Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for CHF prediction. • CANDU 28-, 37- and 43-element bundle CHF experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of CHF model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against five full-scale CANDU bundle experiments conducted in 1990s and in 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element (CANFLEX) bundles. A total of 15 CHF test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for CANDU reactors. Code predictions of channel dryout power and axial and radial CHF locations were compared against measurements from the SL CHF tests to quantify the code prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, the sensitivity studies evaluated the contribution of each CHF model change or enhancement to the improvement in CHF prediction. Overall, the assessment demonstrated significant improvement in prediction of channel dryout power and axial and radial CHF locations in horizontal fuel channels containing CANDU bundles

A review on critical heat flux in horizontal tubes

International Nuclear Information System (INIS)

Baburajan, P.K.; Gaikwad, Avinash; Prabhu, S.V.

2015-01-01

Coolant channels of PHWR during accident similar to loss of coolant accident (LOCA) may experience different flow transients with low pressure and low flow conditions. In the advanced PHWRs it is desired to have small amount of positive quality at the exit of the coolant channel to increase the thermal efficiency. Investigation on pressure drop and heat transfer coefficient under subcooled boiling condition is important in the design and operation of the PHWRs. Understanding of thermal hydraulic phenomena associated with horizontal flow is also important in the safety and accident management in these reactors. A detailed experimental investigation on the important thermal hydraulic phenomena of horizontal tubes under low pressure and low flow conditions is carried out. The phenomena covered in this work are measurement of diabatic single phase and subcooled boiling pressure drop and local heat transfer coefficients, steady state CHF, effect of upstream flow restrictions on flow transients and CHF, CHF under oscillatory flow and flow decreasing transients. A detailed literature review is carried out on CHF in horizontal channels to take stock of the works being carried out along with current state of the art and to justify the motivation for the experimental study. This paper presents the review of available literature on horizontal CHF with the results of the experimental work. (author)

Effects of Dihedral Angle on Pool Boiling Heat Transfer from Two Tubes in Vertical Alignment

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

One of the major issues in pool boiling heat transfer is a tube arrangement. The upper tube is affected by the lower tube and the enhancement of the heat transfer on the upper tube is estimated by the bundle effect ( h{sub r} ). It is defined as the ratio of the heat transfer coefficient ( h{sub b} ) for an upper tube in a bundle with lower tubes activated to that for the same tube activated alone in the bundle. Since heat transfer is related with the conditions of a tube surface, bundle geometries, and a liquid type, lots of studies have been carried out for the combinations of those parameters. The most effective parameter must be the tube pitch. Many researchers have been investigated its effect on heat transfer enhancement for the tube bundles and the tandem tubes. The heat transfer on the upper tube of the tubes is enhanced compared with the single tube. The upper tube within a tube bundle can significantly increase nucleate boiling heat transfer compared to the lower tubes at moderate heat fluxes. At high heat fluxes these influences disappear and the data merge onto the pool boiling curve of a single tube. It was explained that the major influential factor is the convective effects due to the fluid velocity and the rising bubbles. They used microstructure-R134a or FC-3184 combinations and identified that the increase in the heat flux of the lower tube decreased the superheat ( ∆T{sub sat} ) of the upper tube. The passive condensers adopted in SWR1000 and APR+ has U-type tubes. Those tubes are slightly inclined from the horizontal to prevent the occurrence of the water hammer. Since the pitch between the upper and lower tubes is varying along the tube length, the results for the fixed pitch are not applicable to the analysis of these condensers. Although there are lots of studies introducing results for the effects of inclination angle on pool boiling heat transfer, no results are treating the angle between two tubes. Therefore, the present study is aimed

Horizontal Heat Exchanger Design and Analysis for Passive Heat Removal Systems

Energy Technology Data Exchange (ETDEWEB)

Vierow, Karen

2005-08-29

This report describes a three-year project to investigate the major factors of horizontal heat exchanger performance in passive containment heat removal from a light water reactor following a design basis accident LOCA (Loss of Coolant Accident). The heat exchanger studied in this work may be used in advanced and innovative reactors, in which passive heat removal systems are adopted to improve safety and reliability The application of horizontal tube-bundle condensers to passive containment heat removal is new. In order to show the feasibility of horizontal heat exchangers for passive containment cooling, the following aspects were investigated: 1. the condensation heat transfer characteristics when the incoming fluid contains noncondensable gases 2. the effectiveness of condensate draining in the horizontal orientation 3. the conditions that may lead to unstable condenser operation or highly degraded performance 4. multi-tube behavior with the associated secondary-side effects This project consisted of two experimental investigations and analytical model development for incorporation into industry safety codes such as TRAC and RELAP. A physical understanding of the flow and heat transfer phenomena was obtained and reflected in the analysis models. Two gradute students (one funded by the program) and seven undergraduate students obtained research experience as a part of this program.

Evaluation of droplet deposition in rod bundle

International Nuclear Information System (INIS)

Ji, W.; Gu, C.Y.; Anglart, H.

1997-01-01

Deposition model for droplets in gas droplet two-phase flow in rod bundle is developed in this work using the Lagrangian method. The model is evaluated in a 9-rod bundle geometry. The deposition coefficient in the bundle geometry are compared with that in round tube. The influences of the droplet size and gas mass flow rate on deposition coefficient are investigated. Furthermore, the droplet motion is studied in more detail by dividing the bundle channel into sub-channels. The results show that the overall deposition coefficient in the bundle geometry is close to that in the round tube with the diameter equal to the bundle hydraulic diameter. The calculated deposition coefficient is found to be higher for higher gas mass flux and smaller droplets. The study in the sub-channels show that the ratio between the local deposition coefficient for a sub-channel and the averaged value for the whole bundle is close to a constant value, deviations from the mean value for all the calculated cases being within the range of ±13%. (author)

Assessment of ASSERT-PV for prediction of post-dryout heat transfer in CANDU bundles

International Nuclear Information System (INIS)

Cheng, Z.; Rao, Y.F.; Waddington, G.M.

2014-01-01

Highlights: • Assessment of the new Canadian subchannel code ASSERT-PV 3.2 for PDO sheath temperature prediction. • CANDU 28-, 37- and 43-element bundle PDO experiments. • Prediction improvement of ASSERT-PV 3.2 over previous code versions. • Sensitivity study of the effect of PDO model options. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The recently released ASSERT-PV 3.2 provides enhanced models for improved predictions of subchannel flow distribution, critical heat flux (CHF), and post-dryout (PDO) heat transfer in horizontal CANDU fuel channels. This paper presents results of an assessment of the new code version against PDO tests performed during five full-size CANDU bundle experiments conducted between 1992 and 2009 by Stern Laboratories (SL), using 28-, 37- and 43-element bundles. A total of 10 PDO test series with varying pressure-tube creep and/or bearing-pad height were analyzed. The SL experiments encompassed the bundle geometries and range of flow conditions for the intended ASSERT-PV applications for existing CANDU reactors. Code predictions of maximum PDO fuel-sheath temperature were compared against measurements from the SL PDO tests to quantify the code's prediction accuracy. The prediction statistics using the recommended model set of ASSERT-PV 3.2 were compared to those from previous code versions. Furthermore, separate-effects sensitivity studies quantified the contribution of each PDO model change or enhancement to the improvement in PDO heat transfer prediction. Overall, the assessment demonstrated significant improvement in prediction of PDO sheath temperature in horizontal fuel channels containing CANDU bundles

Influence of stiffness on CHF for horizontal tubes under LPLF conditions

Energy Technology Data Exchange (ETDEWEB)

Baburajan, P.K. [Nuclear Safety Analysis Division, AERB, Niyamak Bhavan, 400094 (India); Bisht, Govind Singh [Department of Mechanical Engineering, IIT Bombay, 400076 (India); Gaikwad, Avinash J. [Nuclear Safety Analysis Division, AERB, Niyamak Bhavan, 400094 (India); Prabhu, S.V., E-mail: svprabhu@iitb.ac.in [Department of Mechanical Engineering, IIT Bombay, 400076 (India)

2014-10-01

Highlights: • Effect of stiffness on the CHF in horizontal tube under LPLF conditions is studied. • CHF increases with the increase in stiffness. • Correlation for the prediction of CHF as a function of stiffness is developed. • Correlation for mass flux at CHF in terms of stiffness and initial mass flux is given. • RELAP5 is capable of predicting the effect of stiffness on CHF. - Abstract: Studies reported in the past on critical heat flux (CHF) are mostly limited to vertical flow, large channel diameter, high pressure and high mass flux. Since horizontal flow is commonly encountered in boiler tubes, refrigerating equipments and nuclear reactor fuel channels (PHWR), there is a need to understand horizontal flow CHF, generate sufficient experimental database and to develop reliable predictive method. Few studies are reported on the effect of upstream flow restrictions on flow instabilities and CHF. The present work investigates the effect of upstream flow restriction on CHF in horizontal flow at near atmospheric pressure conditions. In the present study, stiffness is defined as the ratio of upstream flow restriction pressure drop to the test section pressure drop. The classification of a flow boiling system as soft or stiff on the basis of quantification of the stiffness is attempted. Experimental data shows an increase in the CHF with the increase in the stiffness for a given initial mass flux. A correlation for the prediction of CHF under various stiffness conditions is developed. A correlation is suggested to predict the mass flux at CHF as a function of stiffness and initial mass flux. Modeling and transient analysis of the stiffness effect on CHF is carried out using the thermal hydraulic system code RELAP5. The predicted phenomena are in agreement with the experimental observations.

Bundling and mergers in energy markets

International Nuclear Information System (INIS)

Granier, Laurent; Podesta, Marion

2010-01-01

Does bundling trigger mergers in energy industries? We observe mergers between firms belonging to various energy markets, for instance between gas and electricity providers. These mergers enable firms to bundle. We consider two horizontally differentiated markets. In this framework, we show that bundling strategies in energy markets create incentives to form multi-market firms in order to supply bi-energy packages. Moreover, we find that this type of merger is detrimental to social welfare. (author)

Investigation of the integrity of u-bend tube bundles subjected to flow-induced vibrations

Energy Technology Data Exchange (ETDEWEB)

Hassan, M. [University of Guelph, Guelph, Ontario (Canada); Riznic, J. [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

2012-07-01

Maintaining the integrity of nuclear steam generator (SG) tubes in CANDU reactors is a major safety issue since they maintain the physical barrier between the primary and secondary coolants. The integrity of these tubes can be compromised due to flow-induced vibrations in the form of fatigue and fretting wear damage. Wear is a result of the tube impacting and sliding against its loose supports, and it becomes more severe as the tube/support clearance increases. The vibration is caused by fluid flow around these tubes through turbulence and fluidelastic instability mechanisms. Supports are installed to stiffen the structure and to ensure safe and stable operation. The U-bend region is the most critical part since it is subjected to high cross flow. Therefore, special attention is paid to properly supporting this region. However, in some situations, tube support plates (TSP) located on the straight part of the tube may deteriorate to the point where extremely large clearances, or even total wastage of the supports, may result. One possible cause for such a situation is corrosion and/or excessive fretting wear. This loss of TSP may affect the rate of wear in the U-bend portion of the tube due to the increased flexibility in this region. The integrity could be seriously breached as result of a potential support loss. This paper addresses the flow-induced vibrations (FIV) aspect, consequences, and suggested remedies for support degradation. This analysis will include fretting wear producing parameters, such as impact force and normal work rate. Turbulence and fluidelastic instability (FEI) are considered to be the main excitation mechanisms. The investigation is conducted through a numerical simulation of the full Ubend tube bundles including modelling the variable flow distribution, flow excitation, impact, and friction at the supports. (author)

Tube structural integrity evaluation of Palo Verde Unit 1 steam generators for axial upper-bundle cracking

International Nuclear Information System (INIS)

Woodman, B.W.; Begley, J.A.; Brown, S.D.; Sweeney, K.; Radspinner, M.; Melton, M.

1995-01-01

The analysis of the issue of upper bundle axial ODSCC as it apples to steam generator tube structural integrity in Unit 1 at the Palo Verde Nuclear generating Station is presented in this study. Based on past inspection results for Units 2 and 3 at Palo Verde, the detection of secondary side stress corrosion cracks in the upper bundle region of Unit 1 may occur at some future date. The following discussion provides a description and analysis of the probability of axial ODSCC in Unit 1 leading to the exceedance of Regulatory Guide 1.121 structural limits. The probabilities of structural limit exceedance are estimated as function of run time using a conservative approach. The chosen approach models the historical development of cracks, crack growth, detection of cracks and subsequent removal from service and the initiation and growth of new cracks during a given cycle of operation. Past performance of all Palo Verde Units as well as the historical performance of other steam generators was considered in the development of cracking statistics for application to Unit 1. Data in the literature and Unit 2 pulled tube examination results were used to construct probability of detection curves for the detection of axial IGSCC/IGA using an MRPC (multi-frequency rotating panake coil) eddy current probe. Crack growth rates were estimated from Unit 2 eddy current inspection data combined with pulled tube examination results and data in the literature. A Monte-Carlo probabilistic model is developed to provide an overall assessment of the risk of Regulatory Guide exceedance during plant operation

Advances in the manufacture of clad tubes and components for PHWR fuel bundle

International Nuclear Information System (INIS)

Saibaba, N.; Jha, S.K.; Chandrasekha, B.; Tonpe, S.; Jayaraj, R.N.

2010-01-01

Fuel bundles for Pressurized Heavy Water Reactors (PHWRs) consists of Uranium di-oxide pellets encapsulated into thin wall Zircaloy clad tubes. Other components such as end caps, bearing pads and spacer pads are the integral elements of the fuel bundle. As the fuel assembly is subjected to severe operating conditions of high temperature and pressure in addition to continual irradiation exposure, all the components are manufactured conforming to stringent specifications with respect to chemical composition, mechanical & metallurgical properties and dimensional tolerances. The integrity of each component is ensured by NDE at different stages of manufacture. The manufacturing route for fuel tubes and components comprise of a combination of thermomechanical processing and each process step has marked effect on the final properties. The fuel tubes are manufactured by processing the extruded blanks in four stage cold pilgering with intermediate annealing and final stress relieving operation. The bar material is produced by hot extrusion followed by multi-pass swaging and intermediate annealing. Spacer pads and bearing pads are manufactured by blanking and coining of Zircaloy sheet which is made by a combination of hot and cold rolling operations. Due to the small size and stringent dimensional requirements of these appendages, selection of production route and optimization of process parameters are important. This paper discusses about various measures taken for improving the recoveries and mechanical and corrosion properties of the tube, sheet and bar materials being manufactured at Nuclear Fuel Complex, Hyderabad For the production of clad tubes, modifications at extrusion stage to reduce the wall thickness variation, introduction of ultrasonic testing of extruded blanks, optimization of cold working and heat treatment parameters at various stages of production etc. were done. The finished bar material is subjected to 100% Ultrasonic and eddy current testing to ensure

Assessment of horizontal in-tube condensation models using MARS code. Part I: Stratified flow condensation

Energy Technology Data Exchange (ETDEWEB)

Jeon, Seong-Su [Department of Engineering Project, FNC Technology Co., Ltd., Bldg. 135-308, Seoul National University, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Department of Nuclear Engineering, Seoul National University, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Hong, Soon-Joon, E-mail: sjhong90@fnctech.com [Department of Engineering Project, FNC Technology Co., Ltd., Bldg. 135-308, Seoul National University, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Park, Ju-Yeop; Seul, Kwang-Won [Korea Institute of Nuclear Safety, 19 Kuseong-dong, Yuseong-gu, Daejon (Korea, Republic of); Park, Goon-Cherl [Department of Nuclear Engineering, Seoul National University, Gwanak-gu, Seoul 151-744 (Korea, Republic of)

2013-01-15

Highlights: Black-Right-Pointing-Pointer This study collected 11 horizontal in-tube condensation models for stratified flow. Black-Right-Pointing-Pointer This study assessed the predictive capability of the models for steam condensation. Black-Right-Pointing-Pointer Purdue-PCCS experiments were simulated using MARS code incorporated with models. Black-Right-Pointing-Pointer Cavallini et al. (2006) model predicts well the data for stratified flow condition. Black-Right-Pointing-Pointer Results of this study can be used to improve condensation model in RELAP5 or MARS. - Abstract: The accurate prediction of the horizontal in-tube condensation heat transfer is a primary concern in the optimum design and safety analysis of horizontal heat exchangers of passive safety systems such as the passive containment cooling system (PCCS), the emergency condenser system (ECS) and the passive auxiliary feed-water system (PAFS). It is essential to analyze and assess the predictive capability of the previous horizontal in-tube condensation models for each flow regime using various experimental data. This study assessed totally 11 condensation models for the stratified flow, one of the main flow regime encountered in the horizontal condenser, with the heat transfer data from the Purdue-PCCS experiment using the multi-dimensional analysis of reactor safety (MARS) code. From the assessments, it was found that the models by Akers and Rosson, Chato, Tandon et al., Sweeney and Chato, and Cavallini et al. (2002) under-predicted the data in the main condensation heat transfer region, on the contrary to this, the models by Rosson and Meyers, Jaster and Kosky, Fujii, Dobson and Chato, and Thome et al. similarly- or over-predicted the data, and especially, Cavallini et al. (2006) model shows good predictive capability for all test conditions. The results of this study can be used importantly to improve the condensation models in thermal hydraulic code, such as RELAP5 or MARS code.

Evaporating heat transfer of R22 and R410A in horizontal smooth and microfin tubes

Energy Technology Data Exchange (ETDEWEB)

Kim, Man-Hoe; Shin, Joeng-Seob [Korea Advanced Institute of Science and Technology, Daejeon (Korea). Department of Mechanical Engineering

2005-09-01

An experimental investigation of evaporating heat transfer in 9.52 mm O.D. horizontal copper tubes was conducted. The refrigerants tested were R22 and the near-azeotropic mixture, R410A. The test rig had a straight, horizontal test section with an active length of 0.92 m and was heated by the heat transfer fluid (hot water) circulated in a surrounding annulus. Constant heat flux of 11.0 kW/m{sup 2} was maintained and refrigerant quality varied from 0.2 to 0.8.. The results were reported for evaporation at 15 {sup o}C in a 0.92 m long test section for 30-60 kg/h mass flow rate. The local and average heat transfer coefficients for seven microfin tubes were presented compared to those for a smooth tube. The average evaporation heat transfer coefficients of R22 and R410A for the microfin tubes were 1.86-3.27 and 1.64-2.99 times higher than those for the smooth tube, respectively. When compared to R22 at the same test conditions, the evaporating heat transfer coefficients for R410A were 97-129% of R22. (author)

Mixed convection heat transfer between a steam/non-condensable gas mixture and an inclined finned tube bundle

Energy Technology Data Exchange (ETDEWEB)

De Cachard, F.; Lompersky, S.; Monauni, G.R. [Paul Scherrer Institute, Villigen (Switzerland). Thermal Hydraulic Lab.

1999-07-01

An experimental and analytical program was performed at PSI (Paul Scherrer Institute) to study the performance of a finned-tube condenser in the presence of non-condensable gases at low gas mass fluxes. The model developed for this application includes mixed convection heat transfer between the vapour/non-condensable mixture and the finned tubes, heat conduction through the fins and tubes, and evaporative heat transfer inside the tubes. On the gas, heat transfer correlations are used, and the condensation rate is calculated using the heat/mass transfer analogy. A combination of various available correlations for forced convection in staggered finned tube bundles is used, together with a correction accounting for superimposed natural convection. For the condensate heat transfer resistance, the beatty and Katz model for gravity driven liquid films on the tubes is used. The fine efficiency is accounted for using classical iterative calculations. Evaporative heat transfer inside the tubes is predicted using the Chen correlation. The finned tube condenser model has been assessed against data obtained at the PSI LINX facility with two test condensers. For the 62 LINX experiments performed, the model predictions are very good, i.e., less then 10% standard deviation between experimental and predicted results.

Visualization of two-phase gas-liquid flow regimes in horizontal and slightly-inclined circular tubes

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Livia Alves [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil); Nuclear Engineering Institute (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)], E-mail: livia@lasme.coppe.ufrj.br; Cunha Filho, Jurandyr; Su, Jian [Coordenacao dos Programas de Pos-Graduacao de Engenharia (COPPE/UFRJ), RJ (Brazil). Nuclear Engineering Program], Emails: cunhafilho@lasme.coppe.ufrj.br, sujian@lasme.coppe.ufrj.br; Faccini, Jose Luiz Horacio [Nuclear Engineering Institute (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)], E-mail: faccini@ien.gov.br

2010-07-01

In this paper a flow visualization study was performed for two-phase gas-liquid flow in horizontal and slightly inclined tubes. The test section consists of a 2.54 cm inner diameter stainless steel circular tube, followed by a transparent acrylic tube with the same inner diameter. The working fluids were air and water, with liquid superficial velocities ranging from 0:11 to 3:28 m/s and gas superficial velocities ranging from 0:27 to 5:48 m/s. Flow visualization was executed for upward flow at 5 deg and 10 deg and downward flow at 2:5 deg, 5 deg and 10 deg, as well as for horizontal flow. The visualization technique consists of a high-speed digital camera that records images at rates of 125 and 250 frames per second of a concurrent air-water mixture through a transparent part of the tube. From the obtained images, the flow regimes were identified (except for annular flow), observing the effect of inclination angles on flow regime transition boundaries. Finally, the experimental results were compared with empirical and theoretical flow pattern maps available in literature. (author)

Numerical simulation, statistical and hybrid turbulence modelling in a tube bundle under crossflow at high Reynolds number in the context of fluid-structure interaction

International Nuclear Information System (INIS)

Marcel, T.

2011-01-01

The prediction of fluid-elastic instabilities that develop in a tube bundle is of major importance for the design of modern heat exchangers in nuclear reactors, to prevent accidents associated with such instabilities. The fluid-elastic instabilities, or flutter, cause material fatigue, shocks between beams and damage to the solid walls. These issues are very complex for scientific applications involving the nuclear industry. This work is a collaboration between EDF, CEA and IMFT. It aims to improve the numerical simulation of the fluid-structure interaction in the tube bundle, in particular in the range of critical parameters contribute to the onset of damping negative system and the fluid-elastic instability. (author) [fr

Heat transfer to immersed horizontal tubes in gas fluidized bed dryers

Energy Technology Data Exchange (ETDEWEB)

Jonassen, Ola

1999-10-01

The main objective of this study was to construct heat pump fluidized bed dryers of the FHT type with improved dewatering capacity for a given size of the dryer. The use of heat exchangers immersed in the fluidized bed drying chambers is an important part of the FHT (Fluidized Bed High Temperature Heat Pump Dryer) concept. A pilot plant FHT dryer was built and successfully tested on fish meal raw material and seaweed. The plant included two fluidized bed drying chambers with immersed heat exchangers. The gain in water vapor of the drying air through the chambers was increased up to four times that of adiabatic drying. The energy saving concept was retained as a SMER ratio of 3.5 to 4.7 was measured in the same tests. Therefore optimization of the immersed heat exchangers was considered the most important single objective for this work. The optimization study of the heat exchangers was confined to the actual operating conditions for the dryers using: (1) Bubbling gas fluidized beds were used, (2) air as the only type of fluidising gas, (3) beds at atmospheric pressure, (4) bed temperatures below 100 {sup o}C, (5) fluidized particles of Geldart classes B and D, (6) horizontal tube banks for the immersed heat exchanger, and the influence of radiation heat transfer was ignored. The heat transfer study was confined to the fluidized bed side of the heat exchanger surface. It was concluded early in this work that the bubbles play a major role in generating the particle circulation inside the bed and hence also in heat transfer. Publications describing the most important bubble induced mechanisms contributing to high rates of heat transfer were found to be limited. Therefore the first part of this study was aimed at establishing a method for locating and measuring the size and rise velocity of bubbles inside the bed. The method established through this work using differential pressure measurements from two static pressure probes was used later in the study of heat transfer

Full-scale fire experiments on vertical horizontal cable trays

International Nuclear Information System (INIS)

Mangs, J.; Keski-Rahkonen, O.

1997-10-01

Two full-scale fire experiments on PVC cables used in nuclear power plants were carried out, one with cables in vertical position and one with cables in horizontal position. The vertical cable bundle, 3 m high, 300 mm wide and 30 mm thick, was attached to a steel cable ladder. The vertical bundle experiment was carried out in nearly free space with three walls near the cable ladder guiding air flow in order to stabilise flames. The horizontal cable experiment was carried out in a small room with five cable bundles attached to steel cable ladders. Three of the 2 m long cable bundles were located in an array, equally spaced above each other near one long side of the room and two correspondingly near the opposite long side. The vertical cable bundle was ignited with a small propane gas burner beneath the lower edge of the bundle. The horizontal cable bundles were ignited with a small propane burner beneath the lowest bundle in an array of three bundles. Rate of heat release by means of oxygen consumption calorimetry, mass change, CO 2 , CO and smoke production rate and gas, wall and cable surface temperatures were measured as a function of time, as well as time to sprinkler operation and failure of test voltage in cables. Additionally, the minimum rate of heat release needed to ignite the bundle was determined. This paper concentrates on describing and recording the experimental set-up and the data obtained. (orig.)

Analysis of flow-induced vibration of heat exchanger and steam generator tube bundles using the AECL computer code PIPEAU-2

International Nuclear Information System (INIS)

Gorman, D.J.

1983-12-01

PIPEAU-2 is a computer code developed at the Chalk River Nuclear Laboratories for the flow-induced vibration analysis of heat exchanger and steam generator tube bundles. It can perform this analysis for straight and 'U' tubes. All the theoretical work underlying the code is analytical rather than numerical in nature. Highly accurate evaluation of the free vibration frequencies and mode shapes is therefore obtained. Using the latest experimentally determined parameters available, the free vibration analysis is followed by a forced vibration analysis. Tube response due to fluid turbulence and vortex shedding is determined, as well as critical fluid velocity associated with fluid-elastic instability

A preliminary stability analysis of MYRRHA Primary Heat Exchanger two-phase tube bundle

Energy Technology Data Exchange (ETDEWEB)

Castelliti, Diego [Studiecentrum voor kernenergie – Centre d’étude de l’énergie nucléaire (SCK-CEN), Boeretang 200, Mol (Belgium); GeNERG – DIME/TEC, University of Genova, Via all’Opera Pia 15/a, 16145 Genova (Italy); Lomonaco, Guglielmo, E-mail: guglielmo.lomonaco@unige.it [GeNERG – DIME/TEC, University of Genova, Via all’Opera Pia 15/a, 16145 Genova (Italy); INFN, Via Dodecaneso 33, 16146 Genova (Italy)

2016-08-15

Highlights: • MYRRHA is a pool-type LBE-cooled ADS, operable also as a critical reactor. • MYRRHA is a high priority infrastructure for nuclear research in Europe. • PHX (primary side: LBE, secondary side: two-phase water), is a MYRRHA key component. • The original MYRRHA PHX design does not offer a fully satisfying response to DWO. • The adoption of an orifice allows extending considerably the stability of MYRRHA PHX. - Abstract: The MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) project, started at SCK·CEN since 1999, aims at the construction of a pool-type sub-critical Accelerator-Driven System (ADS) which could also operate as a critical reactor. The primary system, enclosed in the primary vessel, is filled with Lead Bismuth Eutectic (LBE) which acts as primary coolant. The power is then delivered through four heat exchangers to four secondary loops. The secondary cooling fluid is two-phase water operating at relatively low pressure (16 bar). Four aero-condensers act as heat sinks, since MYRRHA design does not foresee any electricity generation. The MYRRHA Primary Heat eXchangers (PHXs) cover a role of fundamental importance in normal operation and accidental conditions, being part of the primary and secondary cooling system and of the Decay Heat Removal (DHR) system. It is thus highly relevant to understand the PHXs behavior under all the potential working conditions. In particular, the stability of the PHXs must be guaranteed under all operating conditions. System code models play an important role in understanding and predicting the behavior of the reactor in all conditions, from steady state to operational and accidental transients, and simulating all the postulated scenarios. A solid PHX design requires a complete assessment of two-phase flow instabilities in the secondary system water tube bundle and the potential implementation of a suitable stabilizing device (orifice) to reduce the impact of the perturbations along

[Comparison of dentomaxillary pantomography and periapical radiographs with horizontal tube shift in localizing the impacted teeth].

Science.gov (United States)

Wang, Sun; Fan, Lin-feng

2005-04-01

To compare the clinic value between dentomaxillary pantomography and periapical radiographs in localization of the impacted teeth. 43 impacted teeth were localized with both dentomaxillary pantomography technique and periapical radiographs with horizontal tube shift which is clinically widely used. And a comparison between the two methods was carried out using Chi square test. Both dentomaxillary pantomography and periapical radiographs with horizontal tube shift can relatively precisely demonstrate the position of the impacted teeth. The percentage of the cases which the image and the result of surgery was consistent in the two methods was 93.02% and 95.35% (P>0.05) respectively. There was no statistical difference between the two groups. Dentomaxillary pantomography can precisely localize the impacted teeth.

Gamma densitomeric measurements of gas concentrations at a heated tube bundle; Gammadensitometrische Gasgehaltsmessungen an einem beheizten Rohrbuendel

Energy Technology Data Exchange (ETDEWEB)

Franz, R.; Sprewitz, U.; Hampel, U.

2012-07-01

The contribution under consideration reports on a gamma denitometric measurement of gas concentrations in a vertical heated tube bundle which is flowed around by a fluid. Two measurement positions, two flow rates of the circulating fluid, two subcooling values and eleven heat fluxes were selected for the measurement. The authors of this contribution describe the test facility, measurement methodology, results and their interpretation. The measurement uncertainty is described in detail.

Unsteady Reynolds Averaged Navier-Stokes and Large Eddy Simulations of Flows across Staggered Tube Bundle for a VHTR Lower Plenum Design

International Nuclear Information System (INIS)

Choi, Hyeon Kyeong; Park, Jong Woon

2013-01-01

In this work, behavior of unsteady and oscillating flow through a typical tube bundle array are analyzed by unsteady computations: 2D unsteady Reynolds averaged Navier-Stokes (URANS) and 3D Large Eddy Simulation (LES) and the results are compared with existing experimental data. In order to confirm appropriateness and limitations of CFD applications in the Korean VHTR design, two types of unsteady computations are performed such as 2D unsteady Reynolds averaged Navier-Stokes (URANS) and 3D Large Eddy Simulation (LES) for the existing tube bundle array. The velocity component profiles are compared with the experimental data and it is concluded that the URANS with the standard k-ω model is reasonably appropriate for cost-effective VHTR lower plenum analysis. Nevertheless, if more accurate results are needed, the LES-Smagorinsky computation is recommended considering limitations in the time averaged RANS in capturing small eddies

Strength verification of tube bundle heat exchangers; Zum Festigkeitsnachweis von Rohrbuendelwaermeuebertragern

Energy Technology Data Exchange (ETDEWEB)

Weiss, E. [Dortmund Univ. (Germany). Chemieapparatebau; Rudolph, J. [TUeV Nord EnSys GmbH und Co. KG, Hannover (Germany)

2006-06-15

For a tube bundle heat exchanger with a solid pipe head with and without compensator, the available analytical calculation methods according to AD 2000, DIN EN 13455 part 3 Sect. 13 and Appendix J are evaluated and compared with a detailed numeric analysis. The latter is based on a combination of 3D-CAD/FEA (Autodesk {sup registered} Inventor and ANSYS {sup registered} Workbench Environment AWE) and permits an assessment of the computing time and validity in strength analyses of a complex coupled system of several components. In particular, it is investigated whether the analytical methods prescribed by the technical rules are capable of focusing on the strength problem in critical parts of the structure. It is shown that CAD-FEA coupling on a workbench basis is closer to this goal but is also more complex (modelling, variants calculation). (orig.)

Effect of spacer grids on CHF in tube bundles

International Nuclear Information System (INIS)

Jayanti, Sreenivas; Valette, Michel

2004-01-01

Spacers grids are used to support tube bundles in steam generators and in nuclear reactor fuel assemblies. These grids interface with the flow and heat transfer in a number of ways and their effect has been studied by a number of researchers. It is known that generally they have a beneficial effect on critical heat flux (CHF) in typical nuclear reactor assemblies. However, the enhancement obtained depends on the geometric characteristics of the spacer grids as well as on the parameter range in terms of pressure, local mass velocity and quality. In the present study, the problem is approached in the context of a one-dimensional three-field model. Unlike in previous approaches, no specific modeling of the constitutive laws is made to account for spacer effects and only the geometric details such as the reduction in the cross-sectional area and the hydraulic diameter are included in the calculation which is otherwise the same as that for flow through a single tube. It is shown by comparison with literature data that this approach leads to satisfactory prediction of the thermal-hydraulic effects of spacers and that the beneficial effects of spacers on dry out can be manifested only when the entrainment rate is neither too high nor too low. Their effect on reducing the post-dry out wall temperature is also limited to certain cases. The present work has been performed as part of the EDF-CEA Neptune project also supported by the Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and FRAMATOME-ANP. NEPTUNE is a new set of two phase thermalhydraulic computer codes devoted to safety analysis of nuclear power plants. (author)

Effects of roll waves on annular flow heat transfer at horizontal condenser tube

International Nuclear Information System (INIS)

Kondo, Masaya; Nakamura, Hideo; Anoda, Yoshinari; Sakashita, Akihiro

2002-01-01

Heat removal characteristic of a horizontal in-tube condensation heat exchanger is under investigation to be used for a passive containment cooling system (PCCS) of a next generation-type BWR. Flow regime observed at the inlet of the condenser tube was annular flow, and the local heat transfer rate was ∼20% larger than the prediction by the Dobson-Chato correlation. Roll waves were found to appear on the liquid film in the annular flow. The measured local condensation heat transfer rate was being closely related to the roll waves frequency. Based on these observations, a model is proposed which predicts the condensation heat transfer coefficient for annular flows around the tube inlet. The proposed model predicts well the influences of pressure, local gas-phase velocity and film thickness. (author)

The effect of diameter on vertical and horizontal flow boiling crisis in a tube cooled by Freon-12

International Nuclear Information System (INIS)

Merilo, M.; Ahmad, S.Y.

1979-03-01

The influence of test section orientation and diameter on flow boiling crisis occurring in tubes has been studied experimentally using Freon-12 as a coolant. At low mass flux the critical heat flux (CHF) was lower in horizontal flow than in vertical. As either the liquid or vapour velocity, or both, were increased the vertical and horizontal CHF results converged. Above a mass flux of 4 Mg.m -2 .s -1 the results were essentially identical. The effect of tube diameter on boiling crisis in general depends crucially on the parameters which are maintained constant when the comparison is made. (author)

Ab initio density functional theory investigation of structural and electronic properties of silicon carbide nanotube bundles

Science.gov (United States)

Moradian, Rostam; Behzad, Somayeh; Chegel, Raad

2008-10-01

By using ab initio density functional theory the structural and electronic properties of isolated and bundled (8,0) and (6,6) silicon carbide nanotubes (SiCNTs) are investigated. Our results show that for such small diameter nanotubes the inter-tube interaction causes a very small radial deformation, while band splitting and reduction of the semiconducting energy band gap are significant. We compared the equilibrium interaction energy and inter-tube separation distance of (8,0) SiCNT bundle with (10,0) carbon nanotube (CNT) bundle where they have the same radius. We found that there is a larger inter-tube separation and weaker inter-tube interaction in the (8,0) SiCNT bundle with respect to (10,0) CNT bundle, although they have the same radius.

Ab initio density functional theory investigation of structural and electronic properties of silicon carbide nanotube bundles

International Nuclear Information System (INIS)

Moradian, Rostam; Behzad, Somayeh; Chegel, Raad

2008-01-01

By using ab initio density functional theory the structural and electronic properties of isolated and bundled (8,0) and (6,6) silicon carbide nanotubes (SiCNTs) are investigated. Our results show that for such small diameter nanotubes the inter-tube interaction causes a very small radial deformation, while band splitting and reduction of the semiconducting energy band gap are significant. We compared the equilibrium interaction energy and inter-tube separation distance of (8,0) SiCNT bundle with (10,0) carbon nanotube (CNT) bundle where they have the same radius. We found that there is a larger inter-tube separation and weaker inter-tube interaction in the (8,0) SiCNT bundle with respect to (10,0) CNT bundle, although they have the same radius

Equipment for inspection and carrying out repairs, if required, for tube bundles of steam raising units

International Nuclear Information System (INIS)

Gugel, G.

1976-01-01

The equipment solves the problem of being able to inspect and possibly to repair U-tubes of a vertical steam raising unit standing on a tube floor, without draining the primary medium and bringing the test equipment and tools into the inside of the boiler first. This is achieved by leaving a considerable part of the equipment permanently in the hemispherical space under the tube floor and operating it from the outside, on the other side of the concrete shielding. An inspection tube is threaded in turn horizontally through a concrete shield, a tube duct in the heat insulation of the steam raising unit, and through a hole in the hemispherical space under the tube floor into this space. The end of an angle tube can be moved axially from outside the concrete shield and can be rotated in a semicircle above the tube axis. By interposing a, for example, 12 part distributor with 12 short, differently bent tubes 12 adjacent tubes opening into the tube floor can be controlled and tested, by axial movement of the angle tube together with the distributor, e.g. 4 x 12 other U tubes. A turbulent flow sensor, for example, can be introduced through the angle tube and distributor. In the non-operational condition the equipment is moved into a recess via a supporting angle and stopped there. (ORU) [de

Benchmark simulation of turbulent flow through a staggered tube bundle to support CFD as a reactor design tool. Part 1. SRANS CFD simulation

International Nuclear Information System (INIS)

Ridluan, Artit; Tokuhiro, Akira

2008-01-01

Time-invariant and time-variant numerical simulations of flow through a staggered tube bundle array, idealizing the lower plenum (LP) subsystem configuration of a very high temperature reactor (VHTR), were performed. In Part 1, the CFD prediction of fully periodic isothermal tube-bundle flow using steady Reynolds-averaged Navier-Stokes (SRANS) equations with common turbulence models was investigated at a Reynolds number (Re) of 1.8x10 4 , based on the tube diameter and inlet velocity. Three first-order turbulence models, standard k-ε turbulence, renormalized group (RNG) k-ε, and shear stress transport (SST) k-ω models, and a second-order turbulence model, Reynolds stress model (RSM), were considered. A comparison of CFD simulations and experiment results was made at five locations along (x,y) coordinates. The SRANS simulation showed that no universal model predicted the turbulent Reynolds stresses, and generally, the results were marginal to poor. This is because these models cannot accurately model the periodic, spatiotemporal nature of the complex wake flow structure. (author)

Electronic structure and optical properties of boron nitride nanotube bundles from first principles

Science.gov (United States)

Behzad, Somayeh

2015-06-01

The electronic and optical properties of bundled armchair and zigzag boron nitride nanotubes (BNNTs) are investigated by using density functional theory. Owing to the inter-tube coupling, the dispersions along the tube axis and in the plane perpendicular to the tube axis of BNNT bundles are significantly varied, which are characterized by the decrease of band gap, the splitting of the doubly degenerated states, the expansions of valence and conduction bands. The calculated dielectric functions of the armchair and zigzag bundles are similar to that of the isolated tubes, except for the appearance of broadened peaks, small shifts of peak positions about 0.1 eV and increasing of peak intensities.

Computerized representation of experimental data on burnout in tubes, annular channels and fuel bundles

International Nuclear Information System (INIS)

Katan, I.B.; Sal'nikova, O.V.; Vinogradov, V.N.

1983-01-01

Realization of TEFOR formate for presentation in data bases of bibliographic information obtained when studying heat exchange crisis in channels of the most widely spread types (tubes, annular channels, fuel bundles) has been described. The use of the unified formate, providing a possibility to completely describe the information from the initial source, results in standardization of data base formation in different sections of thermal physics and hydrodynamics of NPPs, permits to develop the general apparatus of bank control in the form of packet of applied programs and to use unified techniques, algorithms and programs during calculations with the use of data of the banks

Heat transfer to immersed horizontal tubes in gas fluidized bed dryers

Energy Technology Data Exchange (ETDEWEB)

Jonassen, Ola

1999-07-01

The main objective of this study was to construct heat pump fluidized bed dryers of the FHT type with improved dewatering capacity for a given size of the dryer. The use of heat exchangers immersed in the fluidized bed drying chambers is an important part of the FHT (Fluidized Bed High Temperature Heat Pump Dryer) concept. A pilot plant FHT dryer was built and successfully tested on fish meal raw material and seaweed. The plant included two fluidized bed drying chambers with immersed heat exchangers. The gain in water vapor of the drying air through the chambers was increased up to four times that of adiabatic drying. The energy saving concept was retained as a SMER ratio of 3.5 to 4.7 was measured in the same tests. Therefore optimization of the immersed heat exchangers was considered the most important single objective for this work. The optimization study of the heat exchangers was confined to the actual operating conditions for the dryers using: (1) Bubbling gas fluidized beds were used, (2) air as the only type of fluidizing gas,(3) beds at atmospheric pressure, (4) bed temperatures below 100 {sup o}C, (5) fluidized particles of Geldart classes B and D, (6) horizontal tube banks for the immersed heat exchanger and the influence of radiation heat transfer was ignored. The heat transfer study was confined to the fluidized bed side of the heat exchanger surface. It was concluded early in this work that the bubbles play a major role in generating the particle circulation inside the bed and hence also in heat transfer. Publications describing the most important bubble induced mechanisms contributing to high rates of heat transfer were found to be limited. Therefore the first part of this study was aimed at establishing a method for locating and measuring the size and rise velocity of bubbles inside the bed. The method established through this work using differential pressure measurements from two static pressure probes was used later in the study of heat transfer

Design concept and testing of an in-bundle gamma densitometer for subchannel void fraction measurements in the THTF electrically heated rod bundle

International Nuclear Information System (INIS)

Felde, D.K.

1982-04-01

A design concept is presented for an in-bundle gamma densitometer system for measurement of subchannel average fluid density and void fraction in rod or tube bundles. This report describes (1) the application of the design concept to the Thermal-Hydraulic Test Facility (THTF) electrically heated rod bundle; and (2) results from tests conducted in the THTF

Velocity distribution measurement in wire-spaced fuel pin bundle

International Nuclear Information System (INIS)

Mizuta, Hiroshi; Ohtake, Toshihide; Uruwashi, Shinichi; Takahashi, Keiichi

1974-01-01

Flow distribution measurement was made in the subchannels of a pin bundle in air flow. The present paper is interim because the target of this work is the decision of temperature of the pin surface in contact with wire spacers. The wire-spaced fuel pin bundle used for the experiment consists of 37 simulated fuel pins of stainless steel tubes, 3000 mm in length and 31.6 mm in diameter, which are wound spirally with 6 mm stainless steel wire. The bundle is wrapped with a hexagonal tube, 3500 mm in length and 293 mm in flat-to-flat distance. The bundle is fixed with knock-bar at the entrance of air flow in the hexagonal tube. The pitch of pins in the bundle is 37.6 mm (P/D=1.19) and the wrapping pitch of wire is 1100 mm (H/D=34.8). A pair of arrow-type 5-hole Pitot tubes are used to measure the flow velocity and the direction of air flow in the pin bundle. The measurement of flow distribution was made with the conditions of air flow rate of 0.33 m 3 /sec, air temperature of 45 0 C, and average Reynolds number of 15100 (average air velocity of 20.6 m/sec.). It was found that circular flow existed in the down stream of wire spacers, that axial flow velocity was slower in the subchannels, which contained wire spacers, than in those not affected by the wire, and that the flow angle to the axial velocity at the boundary of subchannels was two thirds smaller than wire wrapping angle. (Tai, I.)

Nucleate boiling at the forced flow of binary non-azeotropic mixtures in horizontal tubes

Directory of Open Access Journals (Sweden)

Mezentseva N.N.

2015-01-01

Full Text Available Analysis of experimental values of heat transfer coefficients obtained through investigation of nucleate boiling of the two-component non-azeotropic mixtures inside the horizontal smooth tubes by various authors is presented. In the zone of nucleate boiling, the experimental data are in good agreement with the calculation dependence.

Tube vibration in industrial size test heat exchanger

International Nuclear Information System (INIS)

Halle, H.; Wambsganss, M.W.

1980-03-01

Tube vibration data from tests of a specially built and instrumented, industrial-type, shell-and-tube heat exchanger are reported. The heat exchanger is nominally 0.6 m (2 ft) in dia and 3.7 m (12 ft) long. Both full tube and no-tubes-in-window bundles were tested for inlet/outlet nozzles of different sizes and with the tubes supported by seven, equally-spaced, single-segmental baffles. Prior to water flow testing, natural frequencies and damping of representative tubes were measured in air and water. Flow testing was accomplished by increasing the flow rates in stepwise fashion and also by sweeping through a selected range of flow rates. The primary variables measured and reported are tube accelerations and/or displacements and pressure drop through the bundle. Tests of the full tube bundle configuration revealed tube rattling to occur at intermediate flow rates, and fluidelastic instability, with resultant tube impacting, to occur when the flow rate exceeded a threshold level; principally, the four-span tubes were involved in the regions immediately adjacent to the baffle cut. For the range of flow rates tested, fluidelastic instability was not achieved in the no-tubes-in-window bundle; in this configuration the tubes are supported by all seven baffles and are, therefore, stiffer

Heat transfer correlations for evaporation of refrigerant mixtures flowing inside horizontal microfin tubes

Energy Technology Data Exchange (ETDEWEB)

Zhang, Xiaoyan [School of Energy Engineering, Xi' an University of Science and Technology, 58 Yanta Street, Xi' an, Shaanxi 710054 (China); School of Energy and Power Engineering, Xi' an Jiaotong University, 28 Xianning Road, Xi' an, Shaanxi 710049 (China); Yuan, Xiuling [School of Energy and Power Engineering, Xi' an Jiaotong University, 28 Xianning Road, Xi' an, Shaanxi 710049 (China)

2008-11-15

Based on the experimental results of R417A flowing inside horizontal microfin tubes, the present work deals with the development of prediction methods for evaporation heat transfer of refrigerant mixtures in microfin tube. The microfin model by Thome et al. is modified by adjusting the convective heat transfer term, and the other microfin model is developed by introducing the enhancement factor into the modified-Kattan model. The comparison of the calculations by several microfin models and the experimental results reveals that the new microfin models developed at the present study are in much better agreement with the experimental results with the reducing average deviation by 30-50% than the models by Thome et al. and Cavallini et al., and are recommended for the prediction of evaporation heat transfer coefficients for non-azeotropic refrigerant mixtures inside microfin tubes. (author)

Heat transfer correlations for evaporation of refrigerant mixtures flowing inside horizontal microfin tubes

Energy Technology Data Exchange (ETDEWEB)

Xiaoyan, Zhang [School of Energy Engineering, Xi' an University of Science and Technology, 58 Yanta Street, Xi' an, Shaanxi 710054 (China); School of Energy and Power Engineering, Xi' an Jiaotong University, 28 Xianning Road, Xi' an, Shaanxi 710049 (China)], E-mail: gqzxy@sohu.com; Xiuling, Yuan [School of Energy and Power Engineering, Xi' an Jiaotong University, 28 Xianning Road, Xi' an, Shaanxi 710049 (China)

2008-11-15

Based on the experimental results of R417A flowing inside horizontal microfin tubes, the present work deals with the development of prediction methods for evaporation heat transfer of refrigerant mixtures in microfin tube. The microfin model by Thome et al. is modified by adjusting the convective heat transfer term, and the other microfin model is developed by introducing the enhancement factor into the modified-Kattan model. The comparison of the calculations by several microfin models and the experimental results reveals that the new microfin models developed at the present study are in much better agreement with the experimental results with the reducing average deviation by 30-50% than the models by Thome et al. and Cavallini et al., and are recommended for the prediction of evaporation heat transfer coefficients for non-azeotropic refrigerant mixtures inside microfin tubes.

Heat transfer correlations for evaporation of refrigerant mixtures flowing inside horizontal microfin tubes

International Nuclear Information System (INIS)

Zhang Xiaoyan; Yuan Xiuling

2008-01-01

Based on the experimental results of R417A flowing inside horizontal microfin tubes, the present work deals with the development of prediction methods for evaporation heat transfer of refrigerant mixtures in microfin tube. The microfin model by Thome et al. is modified by adjusting the convective heat transfer term, and the other microfin model is developed by introducing the enhancement factor into the modified-Kattan model. The comparison of the calculations by several microfin models and the experimental results reveals that the new microfin models developed at the present study are in much better agreement with the experimental results with the reducing average deviation by 30-50% than the models by Thome et al. and Cavallini et al., and are recommended for the prediction of evaporation heat transfer coefficients for non-azeotropic refrigerant mixtures inside microfin tubes

Aerosol retention in the flooded steam generator bundle during SGTR

International Nuclear Information System (INIS)

Lind, Terttaliisa; Dehbi, Abdel; Guentay, Salih

2011-01-01

Research highlights: → High retention of aerosol particles in a steam generator bundle flooded with water. → Increasing particle inertia, i.e., particle size and velocity, increases retention. → Much higher retention of aerosol particles in the steam generator bundle flooded with water than in a dry bundle. → Much higher retention of aerosol particles in the steam generator bundle than in a bare pool. → Bare pool models have to be adapted to be applicable for flooded bundles. - Abstract: A steam generator tube rupture in a pressurized water reactor may cause accidental release of radioactive particles into the environment. Its specific significance is in its potential to bypass the containment thereby providing a direct pathway of the radioactivity from the primary circuit to the environment. Under certain severe accident scenarios, the steam generator bundle may be flooded with water. In addition, some severe accident management procedures are designed to minimize the release of radioactivity into the environment by flooding the defective steam generator secondary side with water when the steam generator has dried out. To extend our understanding of the particle retention phenomena in the flooded steam generator bundle, tests were conducted in the ARTIST and ARTIST II programs to determine the effect of different parameters on particle retention. The effects of particle type (spherical or agglomerate), particle size, gas mass flow rate, and the break submergence on particle retention were investigated. Results can be summarized as follows: increasing particle inertia was found to increase retention in the flooded bundle. Particle shape, i.e., agglomerate or spherical structure, did not affect retention significantly. Even with a very low submergence, 0.3 m above the tube break, significant aerosol retention took place underlining the importance of the jet-bundle interactions close to the tube break. Droplets were entrained from the water surface with

Development and Assessment of a Bundle Correction Method for CHF

International Nuclear Information System (INIS)

Hwang, Dae Hyun; Chang, Soon Heung

1993-01-01

A bundle correction method, based on the conservation laws of mass, energy, and momentum in an open subchannel, is proposed for the prediction of the critical heat flux (CHF) in rod bundles from round tube CHF correlations without detailed subchannel analysis. It takes into account the effects of the enthalpy and mass velocity distributions at subchannel level using the first dericatives of CHF with respect to the independent parameters. Three different CHF correlations for tubes (Groeneveld's CHF table, Katto correlation, and Biasi correlation) have been examined with uniformly heated bundle CHF data collected from various sources. A limited number of GHE data from a non-uniformly heated rod bundle are also evaluated with the aid of Tong's F-factor. The proposed method shows satisfactory CHF predictions for rod bundles both uniform and non-uniform power distributions. (Author)

Development of nuclear fuel. Development of CANDU advanced fuel bundle

International Nuclear Information System (INIS)

Suk, Ho Chun; Hwang, Woan; Jeong, Young Hwan; Jung, Sung Hoon

1991-07-01

In order to develop CANDU advanced fuel, the agreement of the joint research between KAERI and AECL was made on February 19, 1991. AECL conceptual design of CANFLEX bundle for Bruce reactors was analyzed and then the reference design and design drawing of the advanced fuel bundle with natural uranium fuel for CANDU-6 reactor were completed. The CANFLEX fuel cladding was preliminarily investigated. The fabricability of the advanced fuel bundle was investigated. The design and purchase of the machinery tools for the bundle fabrication for hydraulic scoping tests were performed. As a result of CANFLEX tube examination, the tubes were found to be meet the criteria proposed in the technical specification. The dummy bundles for hydraulic scoping tests have been fabricated by using the process and tools, where the process parameters and tools have been newly established. (Author)

MARS-KS Code Assessment for Condensation Heat Transfer in Horizontal Tube with the Presence of Non-Condensable Gas using Purdue Experiment

International Nuclear Information System (INIS)

Jeon, Seong Su; Lee, Byung Chul; Park, Ju Yeop; Seul, Kwang Won

2011-01-01

In South Korea, advanced power reactor plus (APR+), as a Korean specific reactor, is currently under development for the export strategy. In order to raise competitiveness of the APR+ in the world market, it is necessary to develop the original technology for the improved technology, economics, and safety features. For this purpose, a passive auxiliary feedwater system (PAFS) was adopted as an improved safety design concept of APR+: and then there have been many efforts to develop the PAFS. According to PAFS design concept, PAFS can completely replace the auxiliary feedwater system. When the design basis accident, in which feedwater is unavailable, occurs, the PAFS can remove the residual heat in the core and then prevent the core damage. In the PAFS with the horizontal type heat exchanger, two-phase natural circulation, condensation heat transfer in tube, boiling heat transfer in pool, natural convection in pool, etc. are considered as very important thermalhydraulic phenomena (see Fig. 1). Compared with the vertical heat exchanger from these phenomena, the major difference of the horizontal heat exchanger is the condensation heat transfer phenomena in the tube side. There have been many efforts to understand the condensation heat transfer with in the presence of NC gas in tube but most researches focused on the condensation heat transfer in vertical tube. Therefore the details of the condensation heat transfer in the presence of NC gas in horizontal condenser tubes are not well understood. In order to develop the safety evaluation system for APR+ PAFS, it is required to evaluate the capability and applicability of the MARS-KS code for modeling the condensation heat transfer in the horizontal tube with NC gas because many heat transfer correlations in MARS-KS are known to have much uncertainty. In particular, there is no reliable model for the condensation phenomena in horizontal tube with NC gas. In order to assess the MARS-KS code results and identify the

Heat transfer during forced convection condensation inside horizontal tube

Energy Technology Data Exchange (ETDEWEB)

Tandon, T.N. [M.M.M. Engineering College, Gorakhpur, Uttar Pradesh (India). Dept. of Mechanical Engineering; Varma, H.K.; Gupta, C.P. [Roorkee Univ., Uttar Pradesh (India). Dept. of Mechanical and Industrial Engineering

1995-03-01

This paper presents the results of an experimental investigation on heat transfer behaviour during forced convection condensation inside a horizontal tube for wavy, semi-annular and annular flows. A qualitative study was made of the effect of various parameters - refrigerant mass flux, vapour quality, condensate film temperature drop and average vapour mass velocity - on average condensing-heat transfer coefficient. Akers-Rosson correlations have been found to predict the heat transfer coefficients within {+-} 25% for the entire range of data. A closer examination of the data revealed that the nature of the relation for the heat transfer coefficient changes from annular and semi-annular flow to wavy flow. Akers-Rosson correlations with changed constant and power have been recommended for the two flow regimes. (author)

Experimental study of the condensation heat transfer characteristics of CO2 in a horizontal microfin tube with a diameter of 4.95 mm

Science.gov (United States)

Son, Chang-Hyo; Oh, Hoo-Kyu

2012-11-01

The condensation heat transfer characteristics for CO2 flowing in a horizontal microfin tube were investigated by experiment with respect to condensation temperature and mass flux. The test section consists of a 2,400 mm long horizontal copper tube of 4.6 mm inner diameter. The experiments were conducted at refrigerant mass flux of 400-800 kg/m2s, and saturation temperature of 20-30 °C. The main experimental results showed that annular flow was highly dominated the majority of condensation flow in the horizontal microfin tube. The condensation heat transfer coefficient increases with decreasing saturation temperature and increasing mass flux. The experimental data were compared against previous heat transfer correlations. Most correlations failed to predict the experimental data. However, the correlation by Cavallini et al. showed relatively good agreement with experimental data in the microfin tube. Therefore, a new condensation heat transfer correlation is proposed with mean and average deviations of 3.14 and -7.6 %, respectively.

Experiments on vibration of heat exchanger tube arrays in cross flow

International Nuclear Information System (INIS)

Blevins, R.D.; Gibert, R.J.; Villard, B.

1981-08-01

A series of tests have been made at the Commissariat a l'Energie Atomique, in cooperation with General Atomic Company, SAN DIEGO (U.S.A.) on the flow-induced vibration of heat exchanger tube bundles in cross flow. These tests were made in air on tube bundles which simulated heat exchangers in the high temperature gas cooled reactors. The tests were of two types. In the first type, an instrumented tube was inserted at various locations into a tube bundle. Measurements were made of pressure at a number of points along the tube and about the circumference of the tube. These measurements were processed to obtain the spectra of turbulent pressure fluctuations on the tube, the spanwise correlation and the lift force. The second set of tests was made on tube bundles with flexible tubes. As the flow velocity was increased, these tests clearly show an instability. Nine tube configurations were tested with both plastic and metallic tubes and the effect of tube-to-tube difference in natural frequency was investigated

Microstructure and properties of TP2 copper tube with La microalloying by horizontal continuous casting

Directory of Open Access Journals (Sweden)

Jin-hu Wu

2018-01-01

Full Text Available The TP2 copper tube was prepared with La microalloying by horizontal continuous casting (HCC. The absorptivity of La and its effects on microstructure, tensile and corrosion properties of HCC TP2 copper tube were studied by means of the inductively coupled plasma optical emission spectrometer (ICP-OES, optical microscope (OM, scanning electron microscope (SEM and potentiodynamic polarization measurements. The results show that the absorptivity of La in the HCC TP2 copper tube is about 15% under antivacuum conditions due to the good chemical activities of La. The impurity elements in copper tube such as O, S, Pb and Si can be significantly reduced, and the average columnar dendrite spacing of the copper tube can also be reduced from 2.21 mm to 0.93 mm by adding La. The ultimate tensile strength and the elongation with and without La addition are almost unchanged. However, the annual corrosion rate of the HCC TP2 copper tube is reduced from 10.18 mm•a-1 to 9.37 mm•a-1 by the purification effect of trace La.

Visualization of cross-sectional flow structure during condensation of steam in a slightly inclined horizontal tube

Energy Technology Data Exchange (ETDEWEB)

Puseya, Andree; Kim, H. [Kyung Hee University, Yongin (Korea, Republic of); Kwon, T. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

These flow characteristics called flow patterns still depend on a proper visualization technique in order to identify such local distribution. These proper distributions will have a dependence on the inclination of the tube as well, as it was demonstrated by Lips and Mayer. This work is focused on presenting an experimental investigation to visualize the cross sectional two-phase flow structure for condensation of steam in a horizontal tube and identify the liquid-gas interface using the axial-viewing technique. This innovative technique developed by Hewitt and more recently used in visualization works by Badie, permits the achievement to identify those systems in the area of interest by looking directly into the two-phase flow system during condensation of steam inside a pipe with technology such a high speed camera. An experimental work to visualize and locate the liquid-gas interface for steam condensation in horizontal tubes with slightly inclination was developed on this research The experimental results shows that the axial viewing technique works well with condensation phenomena and can be used for further developments in the field such as determination of liquid film geometry and calculation of void fraction.

New Correlation Methods of Evaporation Heat Transfer in Horizontal Microfine Tubes

Science.gov (United States)

Makishi, Osamu; Honda, Hiroshi

A stratified flow model and an annular flow model of evaporation heat transfer in horizontal microfin tubes have been proposed. In the stratified flow model, the contributions of thin film evaporation and nucleate boiling in the groove above a stratified liquid were predicted by a previously reported numerical analysis and a newly developed correlation, respectively. The contributions of nucleate boiling and forced convection in the stratified liquid region were predicted by the new correlation and the Carnavos equation, respectively. In the annular flow model, the contributions of nucleate boiling and forced convection were predicted by the new correlation and the Carnavos equation in which the equivalent Reynolds number was introduced, respectively. A flow pattern transition criterion proposed by Kattan et al. was incorporated to predict the circumferential average heat transfer coefficient in the intermediate region by use of the two models. The predictions of the heat transfer coefficient compared well with available experimental data for ten tubes and four refrigerants.

Measurements of local liquid velocity and interfacial parameters of air-water bubbly flows in a horizontal tube

International Nuclear Information System (INIS)

Yang Jian; Zhang Mingyuan; Zhang Chaojie; Su Yuliang

2002-01-01

Distribution of local kinematic parameters of air-water bubbly flows in a horizontal tube with an ID of 35 mm was investigated. The local liquid velocity was measured with a cylindrical hot film probe, and local void fraction, bubble frequency and bubble velocity were measured with a double-sensor probe. It was found that the axial liquid velocity has a same profile as that of single liquid phase flow in the lower part of the tube, and it suffers a sudden reduction in the upper part of the tube. With increasing airflow rate, the liquid velocity would increase in the lower part of the tube, and further decrease at the upper part of the tube, respectively. Most bubbles are congested at the upper part of the tube, and the void fraction and bubble frequencies have similar profile and both are asymmetrical with the tube axis with their maximum values located near the upper tube wall

Prediction of effective friction factors for single-phase flow in horizontal microfin tubes

Energy Technology Data Exchange (ETDEWEB)

Wang, H S; Rose, J W [University of London (United Kingdom). Queen Mary, Department of Engineering

2004-12-01

An experimental database, covering a wide range of tube and fin geometric dimensions, Reynolds number and including data for water, R11, and ethylene glycol has been compiled for friction factor for single-phase flow in spirally grooved, horizontal microfin tubes. The tubes (21 in all) had inside diameter at the fin root between 6.46 and 24.13 mm, fin height between 0.13 and 0.47 mm, fin pitch between 0.32 and 1.15 mm, and helix angle between 17 and 45 degrees. The Reynolds number ranged from 2.0x10{sup 3} to 1.63x10{sup 5}. Six earlier friction factor correlations, each based on restricted data sets, have been compared with the database as a whole. None was found to be in good agreement with all of the data. The Jensen and Vlakancic correlation was found to be the best and represents the database within {+-}21%. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

Numerical simulation and experimental results of horizontal tube falling film generator working in a NH3-LiNO3 absorption refrigeration system

International Nuclear Information System (INIS)

Herrera, J.V.; Garcia-Valladares, O.; Gomez, V.H.; Best, R.

2010-01-01

This paper describes the work made at the Centro de Investigacion en Energia in the development of an absorption refrigeration system for cooling and refrigeration applications with a capacity of 10 kW. The single effect unit utilizes ammonia-lithium nitrate as working pair and it is air cooled. The generator is a falling film type with horizontal tubes where the heating oil flows inside the tube bank and the ammonia-lithium nitrate solution flows as a falling film on the tube outside, where it is heated and ammonia vapor is generated. The generator consists of tree columns and four rows per column of horizontal tubes. The system was tested at controlled conditions with heating oil obtained from an electric resistance heating loop. A numerical model of the horizontal falling film generator was developed that divided the system into three different thermal elements: the flow inside the tube, the heat conduction in the tube wall and the falling film solution flow. The mathematical model was tested and validated with experimental data and a study of the influence of the heat transfer coefficient for ammonia-lithium nitrate solution in the numerical model was carried out. A comparison between experimental and numerical data for the heat flux in the system and the temperature profiles in the oil and solution flows shown a good degree of correlation.

Experimental observation of thermal-hydraulic behavior in PCCS horizontal heat exchanger

International Nuclear Information System (INIS)

Kondo, Masaya; Nakamura, Hideo; Yamamoto, Kazuhiko; Shimada, Rumi; Tokuma, Hideaki

2003-01-01

A series of thermal-hydraulic experiments have been performed using a prototypical-scale experimental facility simulating a horizontal heat exchanger of a Passive Containment Cooling System (PCCS) for next generation BWRs. The influences of multi-dimensional boiling flow in secondary water pool on primary flow in parallel tubes are investigated. The experimental results at postulated accident conditions; 0.7 MPa, steam flow rate equivalent to 1% core power with 1% non-condensable gas, show that steam condensation completes in almost the same heat transfer length in all the instrumented tubes. The secondary heat transfer coefficient is relatively small at the lower portion in the tube bundle when the flow regime is bubbly flow, and increases with elevation as the flow regime turns into churn-like flow. The primary steam flow distribution among tubes is rather insensitive to such a variation in the secondary heat transfer coefficient, since the contribution of the secondary heat transfer to the local heat resistance is 30% or less at postulated accident conditions. The influence of steam flow rate is insensitive too, while the contribution of the secondary heat transfer coefficient increases at low pressure conditions. (author)

Numerical studies on heat transfer and pressure drop characteristics of flat finned tube bundles with various fin materials

Science.gov (United States)

Peng, Y.; Zhang, S. J.; Shen, F.; Wang, X. B.; Yang, X. R.; Yang, L. J.

2017-11-01

The air-cooled heat exchanger plays an important role in the field of industry like for example in thermal power plants. On the other hand, it can be used to remove core decay heat out of containment passively in case of a severe accident circumstance. Thus, research on the performance of fins in air-cooled heat exchangers can benefit the optimal design and operation of cooling systems in nuclear power plants. In this study, a CFD (Computational Fluid Dynamic) method is implemented to investigate the effects of inlet velocity, fin spacing and tube pitch on the flow and the heat transfer characteristics of flat fins constructed of various materials (316L stainless steel, copper-nickel alloy and aluminium). A three dimensional geometric model of flat finned tube bundles with fixed longitudinal tube pitch and transverse tube pitch is established. Results for the variation of the average convective heat transfer coefficient with respect to cooling air inlet velocity, fin spacing, tube pitch and fin material are obtained, as well as for the pressure drop of the cooling air passing through finned tube. It is shown that the increase of cooling air inlet velocity results in enhanced average convective heat transfer coefficient and decreasing pressure drop. Both fin spacing and tube pitch engender positive effects on pressure drop and have negative effects on heat transfer characteristics. Concerning the fin material, the heat transfer performance of copper-nickel alloy is superior to 316L stainless steel and inferior to aluminium.

Development of S/G Lancing System for Upper Bundle Hydraulic Cleaning

International Nuclear Information System (INIS)

Jeong, Woo Tae; Kim, Suk Tae; Hong, Sung Yull

2005-01-01

Steam generators of nuclear power plants are recommended to be cleaned during plant outages. Various lancing equipments are developed for the cleaning of tube sheet area of nuclear steam generators. However, no lancing system has been developed in Korea for cleaning upper bundle area of steam generators. Therefore, we developed an upper bundle cleaning system for removing sludge deposited on the tube support plates of nuclear steam generators

The use of twisted tapes for the enhancement of heat transfer outside tube bundles

International Nuclear Information System (INIS)

Mansur, Sergio-Said

1993-01-01

A numerical and experimental investigation of the thermohydraulics of tubular heat exchangers equipped with twisted tapes outside the tubes was carried out. Experimental data for the pressure drop and flow velocity as well as flow visualization data were obtained using a simulated exchanger made of plexiglas. A porous medium type of model allowed for the numerical evaluation of the heat transfer and pressure drop in this unique geometry exchanger. The model was used on the TRIO computer code, developed by the Commissariat a l'Energie Atomique, CEA - France. The experimental data allowed for the evaluation of the flow distribution throughout the exchanger and for the determination of parameters entering the numerical model. The appropriateness of the latter for the macroscopic description of the flow was confirmed by extensive comparison with the experimental data. A comparative analysis of different types of configurations of this exchanger revealed satisfactory performance levels for the exchanger presently investigated. Finally, the flow visualization data were used to qualitatively infer the main aspects of the turbulent diffusion along the tube bundle. The twisted tapes were observed to enhance the fluid mixing process, thus providing for a more effective diffusion of momentum, mass and energy. (author) [fr

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

International Nuclear Information System (INIS)

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

2013-01-01

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

Heat transfer in tube bundles subjected to blockages. Pt. 1

International Nuclear Information System (INIS)

Khattab, M.; Mariy, A.; Habib, M.

1983-01-01

The present work is carried out on unblocked test section bundle, half blocked, single ballooning and four ballooning blockages. The hydro-thermal performance of the bundle, (4x4) stainless steel, under each of the previous cases are studied. It is found that the existance of blockages increases the eddies and swirling flow streams. Furthermore, the average heat transfer in a bundle without blockages is superior than that with blockages. The percentage decrease of the average heat transfer coefficient with blockages depends on the position and shape of the blockage. Correlations describing average heat transfer, pressure drop and friction factor are established. All experimental tests are carried out under non-boiling region. (orig.) [de

Turbulence prediction in two-dimensional bundle flows using large eddy simulation

Energy Technology Data Exchange (ETDEWEB)

Ibrahim, W.A.; Hassan, Y.A. [Texas A& M Univ., College Station, TX (United States)

1995-09-01

Turbulent flow is characterized by random fluctuations in the fluid velocity and by intense mixing of the fluid. Due to velocity fluctuations, a wide range of eddies exists in the flow field. Because these eddies carry mass, momentum, and energy, this enhanced mixing can sometimes lead to serious problems, such as tube vibrations in many engineering systems that include fluid-tube bundle combinations. Nuclear fuel bundles and PWR steam generators are existing examples in nuclear power plants. Fluid-induced vibration problems are often discovered during the operation of such systems because some of the fluid-tube interaction characteristics are not fully understood. Large Eddy Simulation, incorporated in a three dimensional computer code, became one of the promising techniques to estimate flow turbulence, predict and prevent of long-term tube fretting affecting PWR steam generators. the present turbulence investigations is a step towards more understanding of fluid-tube interaction characteristics by comparing the tube bundles with various pitch-to-diameter ratios were performed. Power spectral densities were used for comparison with experimental data. Correlations, calculations of different length scales in the flow domain and other important turbulent-related parameters were calculated. Finally, important characteristics of turbulent flow field were presented with the aid of flow visualization with tracers impeded in the flow field.

Bubble-assisted film evaporation correlation for saline water at sub-atmospheric pressures in horizontal-tube evaporator

KAUST Repository

Shahzad, Muhammad Wakil; Myat, Aung; Chun, Won Gee; Ng, Kim Choon

2013-01-01

film boiling on horizontal tubes, but working at low pressures of 0.93-3.60 kPa (corresponding solution saturation temperatures of 279-300 K) as well as seawater salinity of 15,000 to 90,000 mg/l or ppm. Owing to a dearth of literature on film

Fluid dynamic forces acting on a circular tube bundle in cross flow. Proposals of generation condition of vortex-induced vibration and correlation equation of turbulence-induced exciting force

International Nuclear Information System (INIS)

Inada, Fumio; Yoneda, Kimitoshi; Yasuo, Akira; Nishihara, Takashi

2000-01-01

In the circular tube bundle immersed in the crossflow, the exciting force induced by the turbulence and periodically discharged vortices becomes large, and it is necessary to confirm a long-term integrity to the flow induced vibration. In this report, the local fluid exciting force and the correlation length in the direction of tube axis were measured. The exciting force acting on the first row was smaller than that inside the tube bundle, and the exciting force was almost saturated at the third row. As for vortex induced vibration, there could be an influence when a dimensionless frequency was 0.4 or less. When vortex induced vibration did not affect the vibration, a correlation composed of a correlation length and power spectrum density of the local fluid exciting force were proposed, with which we could estimate the amplitude of the vibration. A computer program to estimate the vibration amplitude and maximum stress was made using the flow velocity distribution and the mode of vibration. (author)

Geometry of Quantum Principal Bundles. Pt. 1

International Nuclear Information System (INIS)

Durdevic, M.

1996-01-01

A theory of principal bundles possessing quantum structure groups and classical base manifolds is presented. Structural analysis of such quantum principal bundles is performed. A differential calculus is constructed, combining differential forms on the base manifold with an appropriate differential calculus on the structure quantum group. Relations between the calculus on the group and the calculus on the bundle are investigated. A concept of (pseudo)tensoriality is formulated. The formalism of connections is developed. In particular, operators of horizontal projection, covariant derivative and curvature are constructed and analyzed. Generalizations of the first Structure Equation and of the Bianchi identity are found. Illustrative examples are presented. (orig.)

Vertical steam generator with slab-type tube-plate with even tube bundle washing

International Nuclear Information System (INIS)

Manek, O.; Masek, V.; Motejl, V.; Quitta, R.

1980-01-01

A shielding plate supporting the tubes attached to the tube plate of a vertical steam generator is mounted above the tube plate. Tube sleeves are designed with a dimensional tolerance relative to the heat transfer tubes and the sleeve end and the tube plate end. A separate space is thus formed above the tube plate in which circulation or feed water is introduced to flow between the branch and the heat transfer tube. This provides intensive washing of heat transfer tubes at a critical point and prevents deposit formation, thus excluding heat transfer tube failures. (J.B.)

Ab initio density functional theory investigation of electronic properties of semiconducting single-walled carbon nanotube bundles

Science.gov (United States)

Moradian, Rostam; Behzad, Somayeh; Azadi, Sam

2008-09-01

By using ab initio density functional theory we investigated the structural and electronic properties of semiconducting (7, 0), (8, 0) and (10, 0) carbon nanotube bundles. The energetic and electronic evolutions of nanotubes in the bundling process are also studied. The effects of inter-tube coupling on the electronic dispersions of semiconducting carbon nanotube bundles are demonstrated. Our results show that the inter-tube coupling decreases the energy gap in semiconducting nanotubes. We found that bundles of (7, 0) and (8, 0) carbon nanotubes have metallic feature, while (10, 0) bundle is a semiconductor with an energy gap of 0.22 eV. To clarify our results the band structures of isolated and bundled nanotubes are compared.

ASSERT-PV 3.2: Advanced subchannel thermalhydraulics code for CANDU fuel bundles

International Nuclear Information System (INIS)

Rao, Y.F.; Cheng, Z.; Waddington, G.M.; Nava-Dominguez, A.

2014-01-01

Highlights: • Introduction to a new version of the Canadian subchannel code, ASSERT-PV 3.2. • Enhanced models for flow-distribution, CHF and post-dryout heat transfer prediction. • Model changes focused on unique features of horizontal CANDU bundles. • Detailed description of model changes for all major thermalhydraulics models. • Discussion on rationale and limitation of the model changes. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The most recent release version, ASSERT-PV 3.2 has enhanced phenomenon models for improved predictions of flow distribution, dryout power and CHF location, and post-dryout (PDO) sheath temperature in horizontal CANDU fuel bundles. The focus of the improvements is mainly on modeling considerations for the unique features of CANDU bundles such as horizontal flows, small pitch to diameter ratios, high mass fluxes, and mixed and irregular subchannel geometries, compared to PWR/BWR fuel assemblies. This paper provides a general introduction to ASSERT-PV 3.2, and describes the model changes or additions in the new version to improve predictions of flow distribution, dryout power and CHF location, and PDO sheath temperatures in CANDU fuel bundles

ASSERT-PV 3.2: Advanced subchannel thermalhydraulics code for CANDU fuel bundles

Energy Technology Data Exchange (ETDEWEB)

Rao, Y.F., E-mail: raoy@aecl.ca; Cheng, Z., E-mail: chengz@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca; Nava-Dominguez, A., E-mail: navadoma@aecl.ca

2014-08-15

Highlights: • Introduction to a new version of the Canadian subchannel code, ASSERT-PV 3.2. • Enhanced models for flow-distribution, CHF and post-dryout heat transfer prediction. • Model changes focused on unique features of horizontal CANDU bundles. • Detailed description of model changes for all major thermalhydraulics models. • Discussion on rationale and limitation of the model changes. - Abstract: Atomic Energy of Canada Limited (AECL) has developed the subchannel thermalhydraulics code ASSERT-PV for the Canadian nuclear industry. The most recent release version, ASSERT-PV 3.2 has enhanced phenomenon models for improved predictions of flow distribution, dryout power and CHF location, and post-dryout (PDO) sheath temperature in horizontal CANDU fuel bundles. The focus of the improvements is mainly on modeling considerations for the unique features of CANDU bundles such as horizontal flows, small pitch to diameter ratios, high mass fluxes, and mixed and irregular subchannel geometries, compared to PWR/BWR fuel assemblies. This paper provides a general introduction to ASSERT-PV 3.2, and describes the model changes or additions in the new version to improve predictions of flow distribution, dryout power and CHF location, and PDO sheath temperatures in CANDU fuel bundles.

Condensation of refrigerants in horizontal microfin tubes: comparison of prediction methods for heat transfer

Energy Technology Data Exchange (ETDEWEB)

Wang, H S; Honda, H [Kyushu University, Fukuoka (Japan). Institute of Advanced Material Study

2003-06-01

A comparison was made between the predictions of previously proposed empirical correlations and theoretical model and available experimental data for the heat transfer coefficient during condensation of refrigerants in horizontal microfin tubes. The refrigerants tested were R11, R123, R134a, R22 and R410A. Experimental data for six tubes with the tube inside diameter at fin root of 6.49-8.8 8 mm, the fin height of 0.16-0.24 mm, fin pitch of 0.34-0.53 mm and helix angle of groove of 12-20{sup o} were adopted. The r.m.s. error of the predictions for all tubes and all refrigerants decreased in the order of the correlations proposed by Luu and Bergies [ASHRAE Trans. 86 (1980) 293], Cavallini et al. Condensation of new refrigerants inside smooth and enhanced tubes. In: Proc. 19th Int. Cong. Refrigeration, vol. IV, Hague, The Netherlands, 1995. p. 105-114, Shikazono et al. [Trans. Jap. Sco. Mech. Engrs. 64 (1995) 196], Kedzierski and Goncalves [J. Enhanced Heat Transfer 6 (1999) 16], Yu and Koyama [Yu J, Koyama S. Condensation heat transfer of pure refrigerants in microfin tubes. In: Proc. Int. Refrigeration Conference at Purdue Univ., West Lafayette, USA, 1998. p. 325-330], and the theoretical model proposed by Wang et al. [Int. J. Heat Mass Transfer 45 (2002) 1513]. (author)

Modelling disassembled fuel bundles using CATHENA MOD-3.5a under LOCA/LOECC conditions

Energy Technology Data Exchange (ETDEWEB)

Lei, Q M; Sanderson, D B; Dutton, R [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Labs.

1996-12-31

CATHENA MOD-3.5a is a multipurpose thermalhydraulic computer code developed primarily to analyse postulated loss-of-coolant scenarios for CANDU nuclear reactors. The code contains a generalized heat transfer package that enables it to model the behaviour of a fuel channel in great detail. Throughout the development of the CATHENA code, considerable effort has been devoted to evaluating, validating and documenting its overall capability as a design and safety assessment tool. Specific attention has focused on its ability to predict fuel channel behaviour under postulated accident conditions. This paper describes an investigation of CATHENA`s ability to predict the thermal-chemical responses of a fuel channel in which the 37-element bundles were assumed to disassemble and rearrange into a closed-packed stack of elements at the bottom of the pressure tube. A representative disassembled bundle geometry was modelled during a simulated loss-of-coolant accident scenario using CATHENA MOD-3.5a/Rev 0, with superheated steam being the only coolant available. Thermal conduction in the radial and circumferential directions was calculated for individual fuel elements, the pressure tube, and the calandria tube. Radiation view factors for the intact and disassembled bundle geometries were calculated using a CATHENA utility program. Inter-element metal-to-metal contact was accounted for using the CATHENA solid-solid contact model. An offset pressure-tube configuration, representing a partially sagged pressure tube, and the effect of steam starvation on the exothermic zirconium-steam reaction, were included in the CATHENA model. The CATHENA-predicted results show a dramatic suppression of heat generation from the zirconium-steam reaction when bundle disassembly is initiated. The predicted results show a smaller temperature increase in the fuel sheaths and the pressure tube for the disassembled bundle geometry, compared to the temperature excursion for the intact bundle. (author

Prediction of evaporation heat transfer coefficient based on gas-liquid two-phase annular flow regime in horizontal microfin tubes

International Nuclear Information System (INIS)

Wang Yueshe; Wang Yanling; Wang, G.-X.; Honda, Hiroshi

2009-01-01

A physical model of gas-liquid two-phase annular flow regime is presented for predicting the enhanced evaporation heat transfer characteristics in horizontal microfin tubes. The model is based on the equivalence of a periodical distortion of the disturbance wave in the substrate layer. Corresponding to the stratified flow model proposed previously by authors, the dimensionless quantity Fr 0 = G/[gd e ρ v (ρ l - ρ v )] 0.5 may be used as a measure for determining the applicability of the present theoretical model, which was used to restrict the transition boundary between the stratified-wavy flow and the annular/intermittent flows. Comparison of the prediction of the circumferential average heat transfer coefficient with available experimental data for four tubes and three refrigerants reveals that a good agreement is obtained or the trend is better than that of the previously developed stratified flow model for Fr 0 > 4.0 as long as the partial dry out of tube does not occur. Obviously, the developed annular model is applicable and reliable for evaporation in horizontal microfin tubes under the case of high heat flux and high mass flux.

Prediction of evaporation heat transfer coefficient based on gas-liquid two-phase annular flow regime in horizontal microfin tubes

Energy Technology Data Exchange (ETDEWEB)

Wang Yueshe, E-mail: wangys@mail.xjtu.edu.cn [State Key Laboratory of Multiphase Flow in Power Engineering, Xi' an Jiaotong University, Xi' an 710049 (China); Yanling, Wang [State Key Laboratory of Multiphase Flow in Power Engineering, Xi' an Jiaotong University, Xi' an 710049 (China); Wang, G -X [Mechanical Engineering Department, The University of Akron, Akron, OH 44325-3903 (United States); Honda, Hiroshi [Kyushu University, 337 Kasuya-machi, Kasuya-gun, Kukuoka 811-2307 (Japan)

2009-10-15

A physical model of gas-liquid two-phase annular flow regime is presented for predicting the enhanced evaporation heat transfer characteristics in horizontal microfin tubes. The model is based on the equivalence of a periodical distortion of the disturbance wave in the substrate layer. Corresponding to the stratified flow model proposed previously by authors, the dimensionless quantity Fr{sub 0} = G/[gd{sub e}{rho}{sub v}({rho}{sub l} - {rho}{sub v})]{sup 0.5} may be used as a measure for determining the applicability of the present theoretical model, which was used to restrict the transition boundary between the stratified-wavy flow and the annular/intermittent flows. Comparison of the prediction of the circumferential average heat transfer coefficient with available experimental data for four tubes and three refrigerants reveals that a good agreement is obtained or the trend is better than that of the previously developed stratified flow model for Fr{sub 0} > 4.0 as long as the partial dry out of tube does not occur. Obviously, the developed annular model is applicable and reliable for evaporation in horizontal microfin tubes under the case of high heat flux and high mass flux.

Study on velocity field in a wire wrapped fuel pin bundle of sodium cooled reactor. Detailed velocity distribution in a subchannel

International Nuclear Information System (INIS)

Sato, Hiroyuki; Kobayashi, Jun; Miyakoshi, Hiroyuki; Kamide, Hideki

2009-01-01

A sodium cooled fast reactor is designed to attain a high burn-up core in a feasibility study on commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, deformation of fuel pin due to the swelling and thermal bowing may decrease local flow velocity via change of flow area in the subassembly and influence the heat removal capability. Therefore, it is of importance to obtain the flow velocity distribution in a wire wrapped pin bundle. A 2.5 times enlarged 7-pin bundle water model was applied to investigate the detailed velocity distribution in an inner subchannel surrounded by 3 pins with wrapping wire. The test section consisted of a hexagonal acrylic duct tube and fluorinated resin pins which had nearly the same refractive index with that of water and a high light transmission rate. The velocity distribution in an inner subchannel with the wrapping wire was measured by PIV (Particle Image Velocimetry) through the front and lateral sides of the duct tube. In the vertical velocity distribution in a narrow space between the pins, the wrapping wire decreased the velocity downstream of the wire and asymmetric flow distribution was formed between the pin and wire. In the horizontal velocity distribution, swirl flow around the wrapping wire was obviously observed. The measured velocity data are useful for code validation of pin bundle thermalhydraulics. (author)

Cecil gives in-bundle access for inspection and lancing [steam generators

International Nuclear Information System (INIS)

Trovato, S.A.; Ruggieri, S.K.

1989-01-01

Cecil (Consolidated Edison Combined Inspection and Lancing System) is a robotic device which makes it possible to take inspection and sludge lancing equipment deep inside steam generator tube bundles. Cecil is teleoperated to perform tube bundle inspections, sludge sampling and sludge lancing. The first field test of Cecil at Indian Point 2 reactor, successfully demonstrated its capability for high quality inspection, and its potential for improved sludge removal, both with reduced personnel radiation exposure. (U.K.)

Device for the selective positioning of a component on a tube plate

International Nuclear Information System (INIS)

1974-01-01

The invention relates to a device for the selective positioning of a component on a tube plate. It particularly applies to the positioning of a guide tube head successively opposite all the tubes of the tube bundle of a nuclear reactor steam generator. The large number of tubes in the tube bundle of the steam generator in a pressure water nuclear power station must be checked periodically for any likely corrosion. This check is effected with a Foucault current probe which is inserted in each tube in turn and is connected to a probe signal processing unit. The probe is placed in a flexible guide tube brought in turn in front of each tube of the bundle to be checked. The invention concerns a device to move the opening of a tube guide for a Foucault current detector over the entire surface of the tube plate, thereby providing access to all the tubes whilst limiting the interventions to a single positioning and a single withdrawal of the apparatus for testing all the bundle. Between the two interventions at the beginning and end of the operation, all displacements are remote controlled from outside the dangerous radioacive area [fr

Characteristics of two-phase flow pattern transitions and pressure drop of five refrigerants in horizontal circular small tubes

Energy Technology Data Exchange (ETDEWEB)

Pamitran, A.S. [Department of Mechanical Engineering, University of Indonesia, Kampus Baru UI, Depok 16424 (Indonesia); Choi, Kwang-Il [Graduate School, Chonnam National University, San 96-1, Dunduk-Dong, Yeosu, Chonnam 550-749 (Korea); Oh, Jong-Taek [Department of Refrigeration and Air Conditioning Engineering, Chonnam National University, San 96-1, Dunduk-Dong, Yeosu, Chonnam 550-749 (Korea); Hrnjak, Pega [Department of Mechanical Science and Engineering, ACRC, University of Illinois at Urbana-Champaign, 1206 West Green Street, Urbana, IL 61801 (United States)

2010-05-15

An experimental investigation on the characteristics of two-phase flow pattern transitions and pressure drop of R-22, R-134a, R-410A, R-290 and R-744 in horizontal small stainless steel tubes of 0.5, 1.5 and 3.0 mm inner diameters is presented. Experimental data were obtained over a heat flux range of 5-40 kW/m{sup 2}, mass flux range of 50-600 kg/(m{sup 2} s), saturation temperature range of 0-15 C, and quality up to 1.0. Experimental data were evaluated with Wang et al. and Wojtan et al. [Wang, C.C., Chiang, C.S., Lu, D.C., 1997. Visual observation of two-phase flow pattern of R-22, R-134a, and R-407C in a 6.5-mm smooth tube. Exp. Therm. Fluid Sci. 15, 395-405; Wojtan, L., Ursenbacher, T., Thome, J.R., 2005. Investigation of flow boiling in horizontal tubes: part I - a new diabatic two-phase flow pattern map. Int. J. Heat Mass Transfer 48, 2955-2969.] flow pattern maps. The effects of mass flux, heat flux, saturation temperature and inner tube diameter on the pressure drop of the working refrigerants are reported. The experimental pressure drop was compared with the predictions from some existing correlations. A new two-phase pressure drop model that is based on a superposition model for two-phase flow boiling of refrigerants in small tubes is presented. (author)

Large eddy simulation of bundle turbulent flows

International Nuclear Information System (INIS)

Hassan, Y.A.; Barsamian, H.R.

1995-01-01

Large eddy simulation may be defined as simulation of a turbulent flow in which the large scale motions are explicitly resolved while the small scale motions are modeled. This results into a system of equations that require closure models. The closure models relate the effects of the small scale motions onto the large scale motions. There have been several models developed, the most popular is the Smagorinsky eddy viscosity model. A new model has recently been introduced by Lee that modified the Smagorinsky model. Using both of the above mentioned closure models, two different geometric arrangements were used in the simulation of turbulent cross flow within rigid tube bundles. An inlined array simulations was performed for a deep bundle (10,816 nodes) as well as an inlet/outlet simulation (57,600 nodes). Comparisons were made to available experimental data. Flow visualization enabled the distinction of different characteristics within the flow such as jet switching effects in the wake of the bundle flow for the inlet/outlet simulation case, as well as within tube bundles. The results indicate that the large eddy simulation technique is capable of turbulence prediction and may be used as a viable engineering tool with the careful consideration of the subgrid scale model. (author)

Heat transfer in the post-dryout region of vertical and horizontal tubes uniformly heated

International Nuclear Information System (INIS)

Kastner, W.; Koehler, W.; Kraetzer, W.

1983-11-01

Increased knowledge of the heat transfer in the post-dryout region is required for novel design of environmentally acceptable power plant technologies (e.g. fluidized bed combustion) and further development of proved steam generators. In particular, the influence of tube orientation and diameter are of consequence. Relating to the onset of critical boiling conditions and the heat transfer in the post-dryout region these aspects were investigated performing 357 tests which cover the operating conditions of fossil fired steam generators. In certain regions of parameters significant differences of the heat transfer behaviour of horizontal and vertical steam generator tubes occured. The experimental results were analysed and compared with theoretical models which were taken from the literature or developed within the frame of this project. (orig.) [de

SCADOP: Phenomenological modeling of dryout in nuclear fuel rod bundles

Energy Technology Data Exchange (ETDEWEB)

Dasgupta, Arnab, E-mail: arnie@barc.gov.in; Chandraker, D.K., E-mail: dineshkc@barc.gov.in; Vijayan, P.K., E-mail: vijayanp@barc.gov.in

2015-11-15

Highlights: • Phenomenological model for annular flow dryout is presented. • The model evaluates initial entrained fraction using a new methodology. • The history effect in annular flow is predicted and validated. • Rod bundle dryout is predicted using subchannel methodology. • Model is validated against experimental dryout data in tubes and rod bundles. - Abstract: Analysis and prediction of dryout is of important consequence to safety of nuclear fuel clusters of boiling water type of reactors. Traditionally, experimental correlations are used for dryout predictions. Since these correlations are based on operating parameters and do not aim to model the underlying phenomena, there has been a proliferation of the correlations, each catering to some specific bundle geometry under a specific set of operating conditions. Moreover, such experiments are extremely costly. In general, changes in tested bundle geometry for improvement in thermal-hydraulic performance would require re-experimentation. Understanding and modeling the basic processes leading to dryout in flow boiling thus has great incentive. Such a model has the ability to predict dryout in any rod bundle geometry, unlike the operating parameter based correlation approach. Thus more informed experiments can be carried out. A good model can, reduce the number of experiments required during the iterations in bundle design. In this paper, a phenomenological model as indicated above is presented. The model incorporates a new methodology to estimate the Initial Entrained Fraction (IEF), i.e., entrained fraction at the onset of annular flow. The incorporation of this new methodology is important since IEF is often assumed ad-hoc and sometimes also used as a parameter to tune the model predictions to experimental data. It is highlighted that IEF may be low under certain conditions against the general perception of a high IEF due to influence of churn flow. It is shown that the same phenomenological model is

The examination of the ruptured Zircaloy-2 pressure tube from Pickering NGS Unit 2

International Nuclear Information System (INIS)

Cheadle, B.A.; Smith, A.D.; Baskin, C.C.

1985-07-01

On 1983 August 01 a Zircaloy-2 pressure tube in Pickering NGS Unit 2 ruptured. All the fuel channel components, the fuel bundles, pressure tube, end fittings, garter springs and calandria tubes were shipped to Chalk River Nuclear Laboratories for examination to determine the cause of the rupture. The examination showed that the rupture initiated at a series of hydride blisters on the outside surface of the pressure tube. The blisters formed because of the garter spring spacers between the pressure tube and calandria tube was about one metre out of position. This allowed the horizontal pressure tube to sag by creep and touch the cool calandria tube. The resulting thermal gradients in the pressure tube concentrated the hydrogen and deuterium at the cool zones and blisters of solid hydride formed. Cracks initiated at several of the blisters and linked together to form a partial through wall critical crack which initiated the final rupture. The video presentation shows how the examination of the fuel channel components was conducted in underwater bays and shielded cells and explains the sequence of events that caused the rupture

Steam generator for a pressurized-water coolant nuclear reactor

International Nuclear Information System (INIS)

Schroeder, H.J.; Berger, W.

1975-01-01

A description is given of a steam generator which has a vertical cylindrical housing having a steam output outlet, a horizontal tube sheet closing the lower end of this housing, and an inverted U-shaped tube bundle inside of the housing and having vertical inlet and outlet legs with their ends mounted in the tube sheet. Beneath the tube sheet there are inlet and outlet manifolds for the respective ends of the tube bundle so that pressurized-water coolant from a pressurized-water coolant nuclear reactor can be circulated through the tube bundle

Effect of liquid nitrogen flow rate on solidification of stagnant water in a horizontal tube

International Nuclear Information System (INIS)

Ibrahim, S.M.

1995-01-01

Five experiments are conducted to study the effect of liquid nitrogen flow rate on the solidification of stagnant water inside a horizontal stainless steel tube of inner diameter 19.6 cm and 12 mm thick. This tube simulates the down-comer of the nuclear reactor ET-R R-1. The apparatus design is mentioned more detail description. The results show that for the first experiment where the liquid nitrogen flow rate is 30 1/hr, the progress of solidification of water has stopped at a diameter of 12 cm. By increasing the flow rate from 30 1/hr to 40,50 and 60 1/hr, the time of freezing the water inside the tube is decreased from 86 to 67 and 60 minutes respectively. By increasing the liquid nitrogen flow rate to 70 1/hr, there is no much effect on the time of frozen. In all experiments, where the solidification is happened, the ice block formed inside the tube is subjected to a pressure of 3 at mg least, and is succeed to withstand this pressure without any leak. 7 figs

Modelling nuclear fuel vibrations in horizontal CANDU reactors

International Nuclear Information System (INIS)

Jagannath, D.V.; Oldaker, I.E.

1976-01-01

Flow-induced fuel vibrations in the pressure tubes of CANDU reactors are of vital interest to designers because fretting damage may result. Computer simulation is being used to study how bundles vibrate and to identify bundle design features which will reduce vibration and hence fretting. (author)

Numerical investigation of heat transfer in upward flows of supercritical water in circular tubes and tight fuel rod bundles

International Nuclear Information System (INIS)

Yang Jue; Oka, Yoshiaki; Ishiwatari, Yuki; Liu Jie; Yoo, Jaewoon

2007-01-01

Heat transfer in upward flows of supercritical water in circular tubes and in tight fuel rod bundles is numerically investigated by using the commercial CFD code STAR-CD 3.24. The objective is to have more understandings about the phenomena happening in supercritical water and for designs of supercritical water cooled reactors. Some turbulence models are selected to carry out numerical simulations and the results are compared with experimental data and other correlations to find suitable models to predict heat transfer in supercritical water. The comparisons are not only in the low bulk temperature region, but also in the high bulk temperature region. The two-layer model (Hassid and Poreh) gives a better prediction to the heat transfer than other models, and the standard k-ε high Re model with the standard wall function also shows an acceptable predicting capability. Three-dimensional simulations are carried out in sub-channels of tight square lattice and triangular lattice fuel rod bundles at supercritical pressure. Results show that there is a strong non-uniformity of the circumferential distribution of the cladding surface temperature, in the square lattice bundle with a small pitch-to-diameter ratio (P/D). However, it does not occur in the triangular lattice bundle with a small P/D. It is found that this phenomenon is caused by the large non-uniformity of the flow area in the cross-section of sub-channels. Some improved designs are numerically studied and proved to be effective to avoid the large circumferential temperature gradient at the cladding surface

Effect of surface roughness on heat transfer from horizontal immersed tubes in a fluidized bed

International Nuclear Information System (INIS)

Grewal, N.S.; Saxena, S.C.

1979-01-01

Experimental results of the total heat transfer coefficient between 12.7 mm dia copper tubes with four different rough surfaces and glass beads of three different sizes as taken in a 0.305 m x 0.305 m square fluidized bed as a function of fluidizing velocity are reported. The comparison of results for the rough and technically smooth tubes suggests that the heat transfer coefficient strongly depends on the ratio of pitch (P/sub f/) to the average particle diameter (d/sub p/), where P/sub f/ is the distance between the two corresponding points on consecutive threads or knurls. By the proper choice of (P/sub f//d/sub p/) ratio, the maximum total heat transfer coefficient for V-thread tubes (h/sub w/fb) can be increased by as much as 40 percent over the value for a smooth tube with the same outside diameter. However, for values of (P/sub f//d/sub p/) less than 0.95, the maximum heat transfer coefficient for the V-thread rough tubes is smaller than the smooth tube having the same outside diameter. The qualitative variation of the heat transfer coefficient for rough tubes with (P/sub f//d) is explained on the basis of the combined effect of contact geometry between the solid particles and the heat transfer surface, and the solids renewal rate at the surface. The present findings are critically compared with somewhat similar investigations from the literature on the heat transfer from horizontal or vertical rough tubes and tubes with small fins

Modeling of fuel bundle vibration and the associated fretting wear in a CANDU fuel channel

International Nuclear Information System (INIS)

Mohany, A.; Hassan, M.

2011-01-01

In this paper a numerical model is developed to predict the vibration response of a CANDU® fuel bundle and the associated fretting wear in the surrounding pressure tube. One excitation mechanism is considered in this model; turbulence-induced excitation caused by coolant flow inside the fuel channel. The numerical model can be easily adapted to include the effects of seismic events, fuel bundle impact during refuelling and start-up of the reactor, and the acoustic pressure pulsations caused by the primary heat transport (PHT) pumps. The simulation is performed for a typical CANDU fuel bundle with 37 fuel elements. The clearances between the buttons of the inner fuel elements, and between the bearing pads of the outer fuel elements and the pressure tube were measured from an actual fuel bundle. Some variability among the measured clearance values was observed. Therefore, probability density functions of the measured clearance values were established and the simulation was performed for the probabilistic distribution of the clearance values. The contact between the fuel bundle and the pressure tube is modeled using pseudo-force contact method. The proposed modelling technique can be used in future CANDU reactors to avoid fuel and pressure tube fretting damage due to the aforementioned excitation mechanisms. (author)

Expansion lyre-shaped tube

International Nuclear Information System (INIS)

Andro, Jean.

1973-01-01

The invention relates the expansion lyre-shaped tube portions formed in dudgeoned tubular bundles between two bottom plates. An expansion lyre comprises at least two sets of tubes of unequal lengths coplanar and symmetrical with respect to the main tube axis, with connecting portions between the tubes forming said sets. The invention applies to apparatus such as heat exchangers, heaters, superheaters or breeders [fr

Steady state heat transfer of helium cooled cable bundles

International Nuclear Information System (INIS)

Khalil, A.

1982-01-01

In the present study nucleate and film boiling heat transfer characteristics of horizontal conductor bundles are investigated at steady state conditions. The effect of gaps between wires, number of wires, wire position, wire size and bundle orientation on the departure from nucleate boiling and transition to film boiling is studied. For gaps close to the bubble departure diameter, the critical heat flux can approach up to 90% of the single wire value. Consequently, the maximum stable current for a given bundle can be significantly increased above the single conductor value for the same cross-sectional area. (author)

Intermediate heat exchanger tube vibration induced by cross and parallel mixed flow

International Nuclear Information System (INIS)

Kawamura, Koji

1986-01-01

The characteristics of pool type LMFBR intermediate heat exchanger (IHX) tube vibrations induced by cross and parallel mixed flow were basically investigated. Secondary coolant in IHX tube bundle is mixed flow of parallel jit flow along the tube axis through flow holes in baffle plates and cross flow. By changing these two flow rate, flow distributions vary in the tube bundle. Mixed flow also induces vibrations which cause fretting wear and fatigue of tube. It is therefore very important to evaluate the tube vibration characteristics for estimating the tube integrity. The results show that the relationships between tube vibrations and flow distributions in the tube bundle were cleared, and mixed flow induced tube vibration could be evaluated on the base of the characteristics of both parallel and cross flow induced vibration. From these investigations it could be concluded that the characteristics of tube vibration for various flow distributions can be systematically evaluated. (author)

Experimental Investigation of Natural Convection into a Horizontal Annular Tube with Porous Medium Effects

Directory of Open Access Journals (Sweden)

Saad Najeeb Shehab

2018-12-01

Full Text Available In this work, an experimental investigation has been done for heat transfer by natural-convection through a horizontal concentric annulus with porous media effects. The porous structure in gap spacing consists of a glass balls and replaced by plastic (PVC balls with different sizes. The outer surface of outer tube is isothermally cooled while the outer surface of inner tube is heated with constant heat flux condition. The inner tube is heated with different supplied electrical power levels. Four different radius ratios of annulus are used. The effects of porous media material, particles size and annulus radius ratio on heat dissipation in terms of average Nusselt number have been analyzed. The experimental results show that the average Nusselt number increases with increasing annulus radius ratio and particle diameter for same porous media material. Furthermore, two empirical correlations of average Nusselt number with average Rayleigh number for glass and PVC particles are developed. The present experimental results are compared with previously works and good correspondence is showed.

Fabrication of novel SnO2 nanofibers bundle and their optical properties

International Nuclear Information System (INIS)

Butt, Faheem K.; Cao, Chuanbao; Khan, Waheed S.; Ali, Zulfiqar; Mahmood, Tariq; Ahmed, R.; Hussain, Sajad; Nabi, Ghulam

2012-01-01

Here we report on the synthesis of novel SnO 2 nanofibers bundle (NFB) by using ball milled Fe powders via chemical vapor deposition (CVD). The reaction was carried out in a horizontal tube furnace (HTF) at 1100 °C under Ar flow. The as prepared product was characterized by X-ray diffraction (XRD), scanning electron microscopy, energy dispersive X-ray spectroscopy, transmission electron microscopy, high resolution transmission electron microscopy and selected area electron diffraction (SAED). The microscopy analysis reveals the existence of tubular structure that might be formed by the accumulation of nanofibers. The Raman spectrum reveals that the product is rutile SnO 2 with additional peaks ascribed to defects or oxygen vacancies. Room temperature Photoluminescence (PL) spectrum exhibits three emission bands at 369, 450 and 466.6 nm. Using optical absorbance data, a direct optical bandgap of 3.68 eV was calculated. -- Graphical abstract: Novel SnO 2 nanofibers bundle (NFB) fabricated via CVD method. Field emission scanning electron microscopy image of novel SnO 2 NFB and their room temperature PL emission. Highlights: ► Synthesis of novel SnO 2 nanofibers bundle at 1100 °C under partial flow of Ar gas. ► A VLS mechanism is proposed for the formation of SnO 2 nanofibers. ► The PL spectrum exhibits three emission bands at 369, 450 and 466.6 nm. ► A direct optical bandgap of 3.68 eV was calculated.

CANDU fuel bundle deformation modelling with COMSOL multiphysics

International Nuclear Information System (INIS)

Bell, J.S.; Lewis, B.J.

2012-01-01

Highlights: ► The deformation behaviour of a CANDU fuel bundle was modelled. ► The model has been developed on a commercial finite-element platform. ► Pellet/sheath interaction and end-plate restraint effects were considered. ► The model was benchmarked against the BOW code and a variable-load experiment. - Abstract: A model to describe deformation behaviour of a CANDU 37-element bundle has been developed under the COMSOL Multiphysics finite-element platform. Beam elements were applied to the fuel elements (composed of fuel sheaths and pellets) and endplates in order to calculate the bowing behaviour of the fuel elements. This model is important to help assess bundle-deformation phenomena, which may lead to more restrictive coolant flow through the sub-channels of the horizontally oriented bundle. The bundle model was compared to the BOW code for the occurrence of a dry-out patch, and benchmarked against an out-reactor experiment with a variable load on an outer fuel element.

Field synergy characteristics in condensation heat transfer with non-condensable gas over a horizontal tube

Directory of Open Access Journals (Sweden)

Junxia Zhang

2017-05-01

Full Text Available Field synergy characteristics in condensation heat transfer with non-condensable gas (NCG over a horizontal tube were numerically simulated. Consequently, synergy angles between velocity and pressure or temperature gradient fields, gas film layer thickness, and induced velocity and shear stress on gas–liquid interface were obtained. Results show that synergy angles between velocity and temperature gradient fields are within 73.2°–88.7° and ascend slightly with the increment in mainstream velocity and that the synergy is poor. However, the synergy angle between velocity and pressure gradient fields decreases intensively with the increase in mainstream velocity at θ ≤ 30°, thereby improving the pressure loss. As NCG mass fraction increases, the gas film layer thickness enlarges and the induced velocity and shear stress on gas–liquid interface decreases. The synergy angles between velocity and temperature gradient fields increase, and the synergy angles between velocity and pressure gradient fields change at θ = 70°, decrease at θ 70°. When the horizontal tube circumference angle increases, the synergy angles between velocity and temperature or pressure gradient fields decrease, the synergy between velocity and pressure fields enhances, and the synergy between velocity and temperature fields degrades.

Tube to limit a bundle of penetrating rays

International Nuclear Information System (INIS)

Arauner, A.

1980-01-01

The tube of the X-ray therapy equipment, which is adjustable in width, guarantees the same dose rate in the edge zone for all tube widths which can be set. The tube for this purpose consists of wall segments, which can be individually moved via shafts at right angles to their own axis and the axis of symmetry of the tube. The inclination of the wall elements to the axis of symmetry is adjustable by levers running in a curve. (RW) [de

Investigation of velocity distribution in an inner subchannel of wire wrapped fuel pin bundle of sodium-cooled fast reactor

International Nuclear Information System (INIS)

Nishimura, Masahiro; Kamide, Hideki; Ohshima, Hiroyuki; Kobayashi, Jun; Sato, Hiroyuki

2011-01-01

A sodium cooled fast reactor is designed to attain a high burn-up of core fuel in commercialized fast reactor cycle systems. In high burn-up fuel subassemblies, deformation of fuel pin due to the swelling and thermal bowing may decrease local flow velocity via change of flow area in the subassembly and influence the heat removal capability. Therefore, it is important to obtain the detail of flow velocity distribution in a wire wrapped pin bundle. In this study, water experiments were carried out to investigate the detailed velocity distribution in a subchannel of nominal pin geometry as the first step. These basic data are not only useful for understanding of pin bundle thermal hydraulics but also a code validation. A wire-wrapped 3-pin bundle water model was applied to investigate the detailed velocity distribution in the subchannel which is surrounded by 3 pins with wrapping wire. The test section consists of an irregular hexagonal acrylic duct tube and three pins made of fluorinated resin pins which has nearly the same refractive index with that of water and a high light transmission rate. This enables to visualize the central subchannel through the pins. The velocity distribution in the central subchannel with the wrapping wire was measured by PIV (Particle Image Velocimetry) through a side wall of the duct tube. Typical flow velocity conditions in the pin bundle were 0.36m/s (Re=2,700) and 1.6m/s (Re=13,500). Influence of the wrapping wire on the velocity distributions in vertical and horizontal directions was confirmed. A clockwise swirl flow around the wire was found in subchannel. Significant differences were not recognized between the two cases of Re=2,700 and 13,500 concerning flow patterns. (author)

CANFLEX fuel bundle cross-flow endurance test (test report)

International Nuclear Information System (INIS)

Hong, Sung Deok; Chung, C. H.; Chang, S. K.; Kim, B. D.

1997-04-01

As part of the normal refuelling sequence of CANDU nuclear reactor, both new and irradiated bundles can be parked in the cross-flow region of the liner tubes. This situation occurs normally for a few minutes. The fuel bundle which is subjected to the cross-flow should be capable of withstanding the consequences of cross flow for normal periods, and maintain its mechanical integrity. The cross-flow endurance test was conducted for CANFLEX bundle, latest developed nuclear fuel, at CANDU-Hot Test Loop. The test was carried out during 4 hours at the inlet cross-flow region. After the test, the bundle successfully met all acceptance criteria after the 4 hours cross-flow test. (author). 2 refs., 3 tabs

CANFLEX fuel bundle cross-flow endurance test (test report)

Energy Technology Data Exchange (ETDEWEB)

Hong, Sung Deok; Chung, C. H.; Chang, S. K.; Kim, B. D.

1997-04-01

As part of the normal refuelling sequence of CANDU nuclear reactor, both new and irradiated bundles can be parked in the cross-flow region of the liner tubes. This situation occurs normally for a few minutes. The fuel bundle which is subjected to the cross-flow should be capable of withstanding the consequences of cross flow for normal periods, and maintain its mechanical integrity. The cross-flow endurance test was conducted for CANFLEX bundle, latest developed nuclear fuel, at CANDU-Hot Test Loop. The test was carried out during 4 hours at the inlet cross-flow region. After the test, the bundle successfully met all acceptance criteria after the 4 hours cross-flow test. (author). 2 refs., 3 tabs.

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

International Nuclear Information System (INIS)

Kang, Myeong Gie

2012-01-01

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

Numerical investigation of supercritical LNG convective heat transfer in a horizontal serpentine tube

Science.gov (United States)

Han, Chang-Liang; Ren, Jing-Jie; Dong, Wen-Ping; Bi, Ming-Shu

2016-09-01

The submerged combustion vaporizer (SCV) is indispensable general equipment for liquefied natural gas (LNG) receiving terminals. In this paper, numerical simulation was conducted to get insight into the flow and heat transfer characteristics of supercritical LNG on the tube-side of SCV. The SST model with enhanced wall treatment method was utilized to handle the coupled wall-to-LNG heat transfer. The thermal-physical properties of LNG under supercritical pressure were used for this study. After the validation of model and method, the effects of mass flux, outer wall temperature and inlet pressure on the heat transfer behaviors were discussed in detail. Then the non-uniformity heat transfer mechanism of supercritical LNG and effect of natural convection due to buoyancy change in the tube was discussed based on the numerical results. Moreover, different flow and heat transfer characteristics inside the bend tube sections were also analyzed. The obtained numerical results showed that the local surface heat transfer coefficient attained its peak value when the bulk LNG temperature approached the so-called pseudo-critical temperature. Higher mass flux could eliminate the heat transfer deteriorations due to the increase of turbulent diffusion. An increase of outer wall temperature had a significant influence on diminishing heat transfer ability of LNG. The maximum surface heat transfer coefficient strongly depended on inlet pressure. Bend tube sections could enhance the heat transfer due to secondary flow phenomenon. Furthermore, based on the current simulation results, a new dimensionless, semi-theoretical empirical correlation was developed for supercritical LNG convective heat transfer in a horizontal serpentine tube. The paper provided the mechanism of heat transfer for the design of high-efficiency SCV.

Effect of crevice environment PH on corrosion damage of horizontal steam generator tubes

International Nuclear Information System (INIS)

Brozova, A.; Burda, J.; Splichal, K.

2002-01-01

In support of a project on lifetime calculation experiments were carried out to evaluate the resistance to environmentally assisted cracking (EAC) of steam generator tubes during operation. Estimations of the incubation period for crack initiation and the threshold K value, K Iscc , and the crack growth rate were made to predict evolution of damage in tube walls. The paper summarizes results of experiments of C ring specimen for the initiation testing and results of SENT (single edge notch tensile) specimen for the crack growth rate (CGR) testing. The specimens were exposed to concentrated environments at elevated temperatures simulating crevice environments in secondary side crevices in horizontal steam generators. The results show that the material of SG tubes is sensitive to transgranular environmentally assisted cracking in the three basic concentrated environments used, alkaline, neutral and acid. The most corrosive medium was the acid environment. The crack initiated practically immediately after acid environment exposure. The initiation process takes a long time in neutral and alkaline environments. The K Iscc values for environmentally assisted crack growth rate in alkaline and neutral concentrated environment were essentially the same. The crack growth rate was slightly higher for the neutral environment than for the alkaline one. Fracture patterns for the both environments were similar. (author)

CFD simulation of flow and heat transfer in Canadian SCWR bundles

International Nuclear Information System (INIS)

Podila, K.; Rao, Y.F.

2014-01-01

Within the Generation-IV (Gen-IV) International Forum, Atomic Energy of Canada Limited (AECL) is leading the effort in developing a conceptual design for the Canadian supercritical water-cooled reactor (SCWR). AECL proposed a new fuel bundle design with two rings of fuel elements placed between central flow tube and the pressure tube. In line with the scope of the conceptual design, the objective of the present CFD work is to aid in developing a bundle heat transfer correlation for the Canadian SCWR fuel bundle design. This paper presents results from an ongoing effort in determining the conditions favorable for possible occurrence of heat transfer deterioration (HTD) in the supercritical bundle flows. In the current investigation, a bare-rod bundle geometry was tested for the proposed fuel bundle design at 23.5, 25 and 28 MPa using STAR-CCM+ CFD code. Taking advantage of the design symmetry of the fuel bundle, only 1/32 of the computational domain was simulated. The SST k-ω turbulence model along with y + <1 was used in the simulations. For lower mass flow simulations, the increase of inlet temperature and operational pressure was found effective in reducing the occurrence of HTD. For higher mass flow simulations, normal heat transfer behaviour was observed except for the lower pressure range (23.5MPa). Ultimately, the goal of this study is to aid the development of a criterion for the onset of HTD in the proposed SCWR bundles, which is planned in the next phase of the project. (author)

Critical heat flux in tubes and tight hexagonal rod lattices

International Nuclear Information System (INIS)

Erbacher, F.J.; Cheng Xu; Zeggel, W.

1994-01-01

The critical heat flux (CHF) in small-diameter tubes and in tight hexagonal 7-rod and 37-rod bundles was investigated in the KRISTA test facility, using Freon 12 as the working fluid. The measurements in tubes showed that the influence of the tube diameter on CHF cannot be described as suggested by earlier publications with sufficient accuracy. CHF in bundles is lower than in tubes under comparable conditions. The influence of spacers (grid spacers, wire wraps) on CHF was found to be governed by local steam qualities. A comparison of the test results with some CHF prediction methods showed that the look-up table method reproduces the test results in circular tubes most accurately. Combined with CHF look-up tables, subchannel analysis and Ahmad's fluid-to-fluid scaling law, Freon experiments have proven to be a suitable tool for CHF prediction in water-cooled rod bundles. (orig.) [de

The use of titanium for condenser tube bundles

International Nuclear Information System (INIS)

Dobrovitch, N.

2002-01-01

In a power plant, the condenser is a strategic heat exchanger with regards to the efficiency of the steam turbine and its reliability guarantees the performance and continuous operation of the plant. Until the early 1980's, copper alloys were routinely used in condenser tubes, thanks to their high heat transfer rates. Yet numerous problems arose from the use of this material, such as stress cracking corrosion, ammoniacal corrosion, fouling, erosion, dezincification, abrasion, erosion-corrosion,... and lately the problem of inadequateness of copper with nuclear steam generators (in nuclear power plant the abrasion problem of the copper alloy tubes created a deposit problem in the steam generator conducting to the replacement of all the condensers). The trend was then to consider new tube materials, such stainless steel and titanium, firstly for particular operating conditions and now for most of the projects, with several objectives, such as: 1) improve the reliability (titanium in particular can bring major improvements such as higher water velocities promoting better heat coefficients, excellent resistance to abrasion, erosion and corrosion thereby improving resistance to fouling; 2) find more cost-effective solutions. The first investment is higher but money is saved on maintenance costs and on time reliability of the material. Titanium tube manufacturing has greatly evolved for the last 20 years. Tubes are mostly welded tubes from ASTM SB 338 grade 1 made on a continuous manufacturing line. All manufacturing operations (welding, annealing, non-destructive testing) are fully automated to produce high quality tubes in large quantities. The most common way to attach tubes to a tubesheet is to roller expand them. (A.C.)

Some applications on tangent bundle with Kaluza-Klein metric

Directory of Open Access Journals (Sweden)

Murat Altunbaş

2017-01-01

Full Text Available In this paper, differential equations of geodesics; parallelism, incompressibility and closeness conditions of the horizontal and complete lift of the vector fields are investigated with respect to Kaluza-Klein metric on tangent bundle.

Establishment and assessment of CHF data base for square-lattice rod bundles

International Nuclear Information System (INIS)

Hwang, Dae Hyun; Seo, K. W.; Kim, K. K.; Zee, S. Q.

2002-02-01

A CHF data base is constructed for square-lattice rod bundles, and assessed with various existing CHF prediction models. The CHF data base consists of 10725 data points obtained from 147 test bundles with uniform axial power distributions and 29 test bundles with non-uniform axial power distributions. The local thermal-hydraulic conditions in the subchannels are calculated by employing a subchannel analysis code MATRA. The influence of turbulent mixing parameter on CHF is evaluated quantitatively for selected test bundles with representative cross sectional configurations. The performance of various CHF prediction models including empirical correlations for round tubes or rod bundles, theoretical DNB models such as sublayer dryout model and bubble crowding model, and CHF lookup table for round tubes, are assessed for the localized rod bundle CHF data base. In view of the analysis result, it reveals that the 1995 AECL-IPPE CHF lookup table method is one of promising models in the aspect of the prediction accuracy and the applicable range. As the result of analysis employing the CHF lookup table for 9113 data points with uniform axial heat profile, the mean and the standard deviation of P/M are calculated as 1.003 and 0.115 by HBM, 1.022 and 0.319 by DSM respectively

Cooperative microexcitations in 2+1D chain-bundle dusty plasma liquids

International Nuclear Information System (INIS)

Io, C.-W.; Chan, C.-L.; Lin I

2010-01-01

Through direct visualization at the discrete level, the microexcitations in cold 2+1D dusty plasma liquids formed by negatively charged dusts suspended in low pressure gaseous discharges were experimentally investigated, in which the downward ion flow wake field induces strong vertical coupling and chain bundle structure. It is found that the horizontal structure and motion are similar to those of the two-dimensional liquid. Different types of basic cooperative chain excitations: straight vertical chains with small amplitude jittering, chain tilting-restraightening, bundle twisting-restraightening, and chain breaking-reconnection, are observed. The region with good (poor) horizontal structural order prefers the straight (tilted or broken) chains with little (large) titling and tilting rate.

Theoretical analysis of film condensation in horizontal microfin tubes; Microfin tsuki suihei kannai gyoshuku no riron kaiseki

Energy Technology Data Exchange (ETDEWEB)

Honda, H; Wang, H [Kyushu University, Fukuoka (Japan). Institute of Advanced Material Study; Nozu, S [Okamaya Prefectural University, Okayama (Japan). Faculty of Computer Science and System Engineering

2000-10-25

A theoretical study has been made of film condensation in helically-grooved, horizontal microfin tubes. The annular flow regime and the stratified flow regime were considered. For the annular flow regime, a previously developed theoretical model was applied. For the stratified flow regime, the height of stratified condensate was estimated by a modified Taitel and Dukler model. For the upper part of the tube exposed to the vapor flow, numerical calculation of Laminar film condensation considering the combined effects of gravity and surface tension forces was conducted. The heat transfer coefficient at the lower part of the tube was estimated by an empirical equation for the internally finned tubes developed by Carnavos. The theoretical predictions of the circumferential average heat transfer coefficient by the two theoretical models were compared with available experimental data for four refrigerants and four tubes. Generally, the annular flow model gave a higher heat transfer coefficient than the stratified flow model in the high quality region, whereas the stratified flow model gave a higher heat transfer coefficient in the low quality region. For tubes with fin heights of 0.16 {approx} 0.24 mm, most of the experimental data agreed within {+-} 20% with the higher of the two theoretical predictions. (author)

Quantitative study of bundle size effect on thermal conductivity of single-walled carbon nanotubes

Science.gov (United States)

Feng, Ya; Inoue, Taiki; An, Hua; Xiang, Rong; Chiashi, Shohei; Maruyama, Shigeo

2018-05-01

Compared with isolated single-walled carbon nanotubes (SWNTs), thermal conductivity is greatly impeded in SWNT bundles; however, the measurement of the bundle size effect is difficult. In this study, the number of SWNTs in a bundle was determined based on the transferred horizontally aligned SWNTs on a suspended micro-thermometer to quantitatively study the effect of the bundle size on thermal conductivity. Increasing the bundle size significantly degraded the thermal conductivity. For isolated SWNTs, thermal conductivity was approximately 5000 ± 1000 W m-1 K-1 at room temperature, three times larger than that of the four-SWNT bundle. The logarithmical deterioration of thermal conductivity resulting from the increased bundle size can be attributed to the increased scattering rate with neighboring SWNTs based on the kinetic theory.

Technology of double casing tubes & a binary cycle system for hole cleaning for CBM multi-branch horizontal wells

Directory of Open Access Journals (Sweden)

Yong Yang

2017-03-01

Full Text Available At present, the aeration-assisted cutting-carrying technology is faced with complexities in the drilling of CBM multi-branch horizontal wells. For example, the aerating pressure is hardly maintained, and the borehole instability may happen. In view of these prominent problems, the technology of double casing tubes & a binary cycle system suitable for CBM multi-branch horizontal wells was developed according to the Venturi principle by means of parasitic tube insufflation which is used for well control simulation system. Then, a multiphase flow finite element model was established for the fluid-cutting particle system in this drilling condition. This technology was tested in field. Double-casing tubes cementing is adopted in this technology and a jet generator is installed at the bottom of the inner casing. In the process of drilling, the drilling fluid injected through double intermediate casing annulus is converted by the jet generator into a high-efficiency steering water jet, which, together with the water jet generated by the bit nozzle, increases the fluid returning rate in the inner annulus space. It is indicated from simulation results that the cutting-carrying effect is the best when the included angle between the nozzle of the jet generator and the vertical direction is 30°. Besides, the influential laws of cutting size, primary cycle volume, accessory cycle volume and drilling velocity on hole cleaning are figured out. It is concluded that this technology increases the flow rate of drilling fluid in annulus space, the returning rate of drilling fluid significantly and the cutting-carrying capacity. It is currently one of the effective hole cleaning technologies for CBM multi-branch horizontal wells where fresh water is taken as the drilling fluid.

Tube bundle system studies at Signal Peak Energy Bull Mountains #1 Mine.

Science.gov (United States)

Zipf, R K; Ochsner, R; Krog, R; Marchewka, W; Valente, M; Jensen, R

2014-03-01

A tube bundle system (TBS) is a mechanical system for continuously drawing gas samples through tubes from multiple monitoring points located in an underground coal mine for analysis and display on the surface. The U.S. National Institute for Occupational Safety and Health (NIOSH), in collaboration with Signal Peak Energy (SPE), LLC, Bull Mountains No. 1 Mine, operated a TBS during mining of two bleederless, longwall panels. This paper describes the gas analysis data and its interpretation. As verified by the TBS, coal at the SPE mine tends to oxidize slowly. It was known that a reservoir of low-oxygen concentration atmosphere developed about 610 m (2,000 ft) behind the longwall face. A bleederless ventilation system facilitates formation of an inert atmosphere in this longwall gob and decreases the likelihood of spontaneous combustion. Connections of the mine atmosphere to the surface through subsidence cracks could allow airflow into the longwall gob, revive coal oxidation and increase spontaneous combustion risk. The atmospheric composition of the sealed areas was homogeneous, except in the immediate vicinity of suspected ingassing points. The TBS verified that gases within the partially sealed, bleederless longwall gob expanded into the longwall tailgate area when barometric pressure decreased. The concentration of carbon dioxide in the back return airflow at the longwall tailgate was observed to increase by a factor of three and possibly up to 10 times the typical background concentration of 0.5 to 1.0%, depending on the size of the longwall gob and the magnitude of barometric pressure decrease. TBS have the inherent disadvantage of slow response time due to travel time of the gas samples and sequential gas analyses. A TBS or similar continuous monitoring system could be beneficial in detecting and providing warning of potentially hazardous gas concentrations, if the slow response time of the system is always understood.

Ab initio density functional theory investigation of crystalline bundles of polygonized single-walled silicon carbide nanotubes

Energy Technology Data Exchange (ETDEWEB)

Moradian, Rostam; Behzad, Somayeh; Chegel, Raad [Physics Department, Faculty of Science, Razi University, Kermanshah (Iran, Islamic Republic of)], E-mail: moradian.rostam@gmail.com

2008-11-19

By using ab initio density functional theory, the structural characterizations and electronic properties of two large-diameter (13, 13) and (14, 14) armchair silicon carbide nanotube (SiCNT) bundles are investigated. Full structural optimizations show that the cross sections of these large-diameter SiCNTs in the bundles have a nearly hexagonal shape. The effects of inter-tube coupling on the electronic dispersions of large-diameter SiCNT bundles are demonstrated. By comparing the band structures of the triangular lattices of (14, 14) SiCNTs with nearly hexagonal and circular cross sections we found that the polygonization of the tubes in the bundle leads to a further dispersion of the occupied bands and an increase in the bandgap by 0.18 eV.

Ab initio density functional theory investigation of crystalline bundles of polygonized single-walled silicon carbide nanotubes

International Nuclear Information System (INIS)

Moradian, Rostam; Behzad, Somayeh; Chegel, Raad

2008-01-01

By using ab initio density functional theory, the structural characterizations and electronic properties of two large-diameter (13, 13) and (14, 14) armchair silicon carbide nanotube (SiCNT) bundles are investigated. Full structural optimizations show that the cross sections of these large-diameter SiCNTs in the bundles have a nearly hexagonal shape. The effects of inter-tube coupling on the electronic dispersions of large-diameter SiCNT bundles are demonstrated. By comparing the band structures of the triangular lattices of (14, 14) SiCNTs with nearly hexagonal and circular cross sections we found that the polygonization of the tubes in the bundle leads to a further dispersion of the occupied bands and an increase in the bandgap by 0.18 eV.

Ab initio density functional theory investigation of crystalline bundles of polygonized single-walled silicon carbide nanotubes

Science.gov (United States)

Moradian, Rostam; Behzad, Somayeh; Chegel, Raad

2008-11-01

By using ab initio density functional theory, the structural characterizations and electronic properties of two large-diameter (13, 13) and (14, 14) armchair silicon carbide nanotube (SiCNT) bundles are investigated. Full structural optimizations show that the cross sections of these large-diameter SiCNTs in the bundles have a nearly hexagonal shape. The effects of inter-tube coupling on the electronic dispersions of large-diameter SiCNT bundles are demonstrated. By comparing the band structures of the triangular lattices of (14, 14) SiCNTs with nearly hexagonal and circular cross sections we found that the polygonization of the tubes in the bundle leads to a further dispersion of the occupied bands and an increase in the bandgap by 0.18 eV.

An in vitro biomechanical comparison of anterior cruciate ligament reconstruction: single bundle versus anatomical double bundle techniques

Directory of Open Access Journals (Sweden)

Sandra Umeda Sasaki

2008-01-01

Full Text Available INTRODUCTION: Anterior cruciate ligament ruptures are frequent, especially in sports. Surgical reconstruction with autologous grafts is widely employed in the international literature. Controversies remain with respect to technique variations as continuous research for improvement takes place. One of these variations is the anatomical double bundle technique, which is performed instead of the conventional single bundle technique. More recently, there has been a tendency towards positioning the two bundles through double bone tunnels in the femur and tibia (anatomical reconstruction. OBJECTIVES: To compare, through biomechanical tests, the practice of anatomical double bundle anterior cruciate ligament reconstruction with a patellar graft to conventional single bundle reconstruction with the same amount of patellar graft in a paired experimental cadaver study. METHODS: Nine pairs of male cadaver knees ranging in age from 44 to 63 years were randomized into two groups: group A (single bundle and group B (anatomical reconstruction. Each knee was biomechanically tested under three conditions: intact anterior cruciate ligament, reconstructed anterior cruciate ligament, and injured anterior cruciate ligament. Maximum anterior dislocation, rigidity, and passive internal tibia rotation were recorded with knees submitted to a 100 N horizontal anterior dislocation force applied to the tibia with the knees at 30, 60 and 90 degrees of flexion. RESULTS: There were no differences between the two techniques for any of the measurements by ANOVA tests. CONCLUSION: The technique of anatomical double bundle reconstruction of the anterior cruciate ligament with bone-patellar tendon-bone graft has a similar biomechanical behavior with regard to anterior tibial dislocation, rigidity, and passive internal tibial rotation.

Reliability of horizontal and vertical tube shift techniques in the localisation of supernumerary teeth.

Science.gov (United States)

Mallineni, S K; Anthonappa, R P; King, N M

2016-12-01

To assess the reliability of the vertical tube shift technique (VTST) and horizontal tube shift technique (HTST) for the localisation of unerupted supernumerary teeth (ST) in the anterior region of the maxilla. A convenience sample of 83 patients who attended a major teaching hospital because of unerupted ST was selected. Only non-syndromic patients with ST and who had complete clinical and radiographic and surgical records were included in the study. Ten examiners independently rated the paired set of radiographs for each technique. Chi-square test, paired t test and kappa statistics were employed to assess the intra- and inter-examiner reliability. Paired sets of 1660 radiographs (830 pairs for each technique) were available for the analysis. The overall sensitivity for VTST and HTST was 80.6 and 72.1% respectively, with slight inter-examiner and good intra-examiner reliability. Statistically significant differences were evident between the two localisation techniques (p HTST in the anterior region of the maxilla.

CFD analysis of flow and heat transfer in Canadian supercritical water reactor bundle

International Nuclear Information System (INIS)

Podila, K.; Rao, Y.F.

2015-01-01

Highlights: • Flow and heat transfer in SCWR fuel bundle design by AECL is studied using CFD. • Bare-rod bundle geometry is tested at 23.5, 25 and 28 MPa using STAR-CCM+ code. • SST k–ω low-Re model was used to study occurrence of heat transfer deterioration. - Abstract: Within the Gen-IV International Forum, AECL is leading the effort in developing a conceptual design for the Canadian SCWR. AECL proposed a new fuel bundle design with two rings of fuel elements placed between central flow tube and the pressure tube. In line with the scope of the conceptual design, the objective of the present CFD work is to aid in developing a bundle heat transfer correlation for the Canadian SCWR fuel bundle design. This paper presents results from an ongoing effort in determining the conditions favorable for occurrence of HTD in the supercritical bundle flows. In the current investigation, bare-rod bundle geometry was tested for the proposed fuel bundle design at 23.5, 25 and 28 MPa using STAR-CCM+ CFD code. Taking advantage of the design symmetry of the fuel bundle, only 1/32 of the computational domain was simulated. The low-Reynolds number modification of SST k–ω turbulence model along with y + < 1 was used in the simulations. For lower mass flow simulations, the increase of inlet temperature and operational pressure was found effective in reducing the occurrence of HTD. For higher mass flow simulations, normal heat transfer behaviour was observed except for the lower pressure range (23.5 MPa)

New models of droplet deposition and entrainment for prediction of CHF in cylindrical rod bundles

Energy Technology Data Exchange (ETDEWEB)

Zhang, Haibin, E-mail: hb-zhang@xjtu.edu.cn [School of Chemical Engineering and Technology, Xi’an Jiaotong University, Xi’an 710049 (China); Department of Chemical Engineering, Imperial College, London SW7 2BY (United Kingdom); Hewitt, G.F. [Department of Chemical Engineering, Imperial College, London SW7 2BY (United Kingdom)

2016-08-15

Highlights: • New models of droplet deposition and entrainment in rod bundles is developed. • A new phenomenological model to predict the CHF in rod bundles is described. • The present model is well able to predict CHF in rod bundles. - Abstract: In this paper, we present a new set of model of droplet deposition and entrainment in cylindrical rod bundles based on the previously proposed model for annuli (effectively a “one-rod” bundle) (2016a). These models make it possible to evaluate the differences of the rates of droplet deposition and entrainment for the respective rods and for the outer tube by taking into account the geometrical characteristics of the rod bundles. Using these models, a phenomenological model to predict the CHF (critical heat flux) for upward annular flow in vertical rod bundles is described. The performance of the model is tested against the experimental data of Becker et al. (1964) for CHF in 3-rod and 7-rod bundles. These data include tests in which only the rods were heated and data for simultaneous uniform and non-uniform heating of the rods and the outer tube. It was shown that the predicted CHFs by the present model agree well with the experimental data and with the experimental observation that dryout occurred first on the outer rods in 7-rod bundles. It is expected that the methodology used will be generally applicable in the prediction of CHF in rod bundles.

Hydraulic design considerations for a multi-tube sodium economizer

International Nuclear Information System (INIS)

Hassberger, J.A.; McConnell, P.M.; Olson, W.H.

1975-01-01

Operating experience gained from tests shows that flow distribution effects can severely affect the thermal performance of high effectiveness, low pressure drop sodium heat exchangers. It has been shown that design efforts for such devices must include proper consideration of potential causes of flow maldistribution within the tube bundle. Furthermore, it has been demonstrated that fairly simple design features can be capable of eliminating detrimental flow fields in the tube bundle

Development of the tube bundle structure for fluid-structure interaction analysis model

International Nuclear Information System (INIS)

Yoon, Kyung Ho; Kim, Jae Yong

2010-02-01

Tube bundle structures within a Boiler or heat exchanger are laid the fluid-structure, thermal-structure and fluid-thermal-structure coupled boundary condition. In these complicated boundary conditions, Fluid-structure interaction (FSI) occurs when fluid flow causes deformation of the structure. This deformation, in turn, changes the boundary conditions for the fluid flow. The structural analysis discipline, and then independently analyzed each other. However, the fluid dynamic force effect the behavior of the structure, and the vibration amplitude of the structure to fluid. FSI analysis model was separately created fluid and structure model, and then defined the fsi boundary condition, and simultaneously analyzed in one domain. The analysis results were compared with those of the experimental method for validating the analysis model. Flow-induced vibration test was executed with single rod configuration. The vibration amplitudes of a fuel rod were measured by the laser vibro-meter system in x and y-direction. The analyses results were not closely with the test data, but the trend was very similar with the test result. In fsi coupled analysis case, the turbulent model was very important with the reliability of the accuracy of the analysis model. Therefore, the analysis model will be needed to further study

Thermophysical properties of multi-wall carbon nanotube bundles at elevated temperatures up to 830 K

International Nuclear Information System (INIS)

Wang, Xinwei; Wang, Jianmei; Huang, Xiaopeng; Eres, Gyula

2011-01-01

In this paper we discuss the results of thermal transport measurements in multi-wall carbon nanotube (MWCNT) bundles at elevated temperatures. A novel generalized electrothermal technique (GET) was developed for measuring the thermal diffusivity ( ) and conductivity (k) of MWCNT bundles. The results show that the feeding current has a negligible effect on the thermal properties. The measured k is larger than the reported values for unaligned bundles, and is comparable to that of typical aligned arrays. Compared with experimental and theoretical data for individual CNTs, k of the MWCNT bundles is two to three orders of magnitude lower, suggesting that the thermal transport in CNT bundles is dominated by the thermal contact resistance of tube-to-tube junctions. The effective density for the two MWCNT bundles, which is difficult to measure using other techniques, was determined to be 116 kg/m3 and 234 kg/m3, respectively. The temperature dependences of and k at temperatures up to 830 K was obtained. slightly decreases with temperature while k exhibits a small increase with temperature up to 500 K and then decreases. For the first time, the behavior of specific heat cp(T) for CNTs above room temperature was determined. The specific heat is close to graphite at 300-400 K but is lower than that for graphite above 400 K, indicating that the behavior of phonons in MWCNT bundles is dominated by boundary scattering rather than by the three-phonon Umklapp process. The length of the mean curvature between two adjacent tube contact points in these bundles is estimated to be on the order of micrometer to millimeter. The analysis of the radiation heat loss suggests that it needs to be considered when measuring the thermophysical properties of micro/nano wires of high aspect ratios at elevated temperatures, especially for individual CNTs due to their extremely small diameter.

Characterization of two-phase flow regimes in horizontal tubes using 81mKr tracer experiments.

Science.gov (United States)

Oriol, Jean; Leclerc, Jean Pierre; Berne, Philippe; Gousseau, Georges; Jallut, Christian; Tochon, Patrice; Clement, Patrice

2008-10-01

The diagnosis of heat exchangers on duty with respect to flow mal-distributions needs the development of non-intrusive inlet-outlet experimental techniques in order to perform an online fault diagnosis. Tracer experiments are an example of such techniques. They can be applied to mono-phase heat exchangers but also to multi-phase ones. In this case, the tracer experiments are more difficult to perform. In order to check for the capabilities of tracer experiments to be used for the flow mal-distribution diagnosis in the case of multi-phase heat exchangers, we present here a preliminary study on the simplest possible system: two-phase flows in a horizontal tube. (81m)Kr is used as gas tracer and properly collimated NaI (TI) crystal scintillators as detectors. The specific shape of the tracer response allows two-phase flow regimes to be characterized. Signal analysis allows the estimation of the gas phase real average velocity and consequently of the liquid phase real average velocity as well as of the volumetric void fraction. These results are compared successfully to those obtained with liquid phase tracer experiments previously presented by Oriol et al. 2007. Characterization of the two-phase flow regimes and liquid dispersion in horizontal and vertical tubes using coloured tracer and no intrusive optical detector. Chem. Eng. Sci. 63(1), 24-34, as well as to those given by correlations from literature.

Characterization of two-phase flow regimes in horizontal tubes using 81mKr tracer experiments

International Nuclear Information System (INIS)

Oriol, Jean; Leclerc, Jean Pierre; Berne, Philippe; Gousseau, Georges; Jallut, Christian; Tochon, Patrice; Clement, Patrice

2008-01-01

The diagnosis of heat exchangers on duty with respect to flow mal-distributions needs the development of non-intrusive inlet-outlet experimental techniques in order to perform an online fault diagnosis. Tracer experiments are an example of such techniques. They can be applied to mono-phase heat exchangers but also to multi-phase ones. In this case, the tracer experiments are more difficult to perform. In order to check for the capabilities of tracer experiments to be used for the flow mal-distribution diagnosis in the case of multi-phase heat exchangers, we present here a preliminary study on the simplest possible system: two-phase flows in a horizontal tube. 81m Kr is used as gas tracer and properly collimated NaI (TI) crystal scintillators as detectors. The specific shape of the tracer response allows two-phase flow regimes to be characterized. Signal analysis allows the estimation of the gas phase real average velocity and consequently of the liquid phase real average velocity as well as of the volumetric void fraction. These results are compared successfully to those obtained with liquid phase tracer experiments previously presented by Oriol et al. 2007. Characterization of the two-phase flow regimes and liquid dispersion in horizontal and vertical tubes using coloured tracer and no intrusive optical detector. Chem. Eng. Sci. 63(1), 24-34, as well as to those given by correlations from literature

An experimental study on two-phase pressure drop in small diameter horizontal, downward inclined and vertical tubes

Directory of Open Access Journals (Sweden)

Autee Arun

2015-01-01

Full Text Available An experimental study of two-phase pressure drop in small diameter tubes orientated horizontally, vertically and at two other downward inclinations of θ= 300 and θ = 600 is described in this paper. Acrylic transparent tubes of internal diameters 4.0, 6.0, and 8.0 mm with lengths of 400 mm were used as the test section. Air-water mixture was used as the working fluid. Two-phase pressure drop was measured and compared with the existing correlations. These correlations are commonly used for calculation of pressure drop in macro and mini-microchannels. It is observed that the existing correlations are inadequate in predicting the two-phase pressure drop in small diameter tubes. Based on the experimental data, a new correlation has been proposed for predicting the two-phase pressure drop. This correlation is developed by modification of Chisholm parameter C by incorporating different parameters. It was found that the proposed correlation predicted two-phase pressure drop at satisfactory level.

Method for confirming flow pattern of gas-water flow in horizontal tubes under rolling state

International Nuclear Information System (INIS)

Luan Feng; Yan Changqi

2008-01-01

An experimental study on the flow patterns of gas-water flow was carried out in horizontal tubes under rolling state. It was found that the pressure drop of two phase flow was with an obvious periodical characteristic. The flow pattern of the gas-water flow was distinguished according to the characteristics of the pressure drop in this paper. It was proved that the characteristics of the pressure drop can distinguish the flow pattern of gas-water flow correctly through comparing with the result of careful observation and high speed digital camera. (authors)

Heat transfer characteristics for evaporation of R417A flowing inside horizontal smooth and internally grooved tubes

Energy Technology Data Exchange (ETDEWEB)

Xiaoyan, Zhang [School of Energy and Power Engineering, Xi' an Jiaotong University, 28 Xianning Road, Xi' an, Shaanxi 710049 (China); School of Energy Engineering, Xi' an University of Science and Technology, 58 Yanta Street, Xi' an, Shaanxi 710054 (China)], E-mail: gqzxy@sohu.com; Xingqun, Zhang; Yunguang, Chen; Xiuling, Yuan [School of Energy and Power Engineering, Xi' an Jiaotong University, 28 Xianning Road, Xi' an, Shaanxi 710049 (China)

2008-06-15

The experimental study on evaporation heat transfer of R417A (R125/R134a/R600) flowing inside horizontal smooth and two internally grooved tubes with different geometrical parameters was conducted with the mass flow rate range from 176 to 344 kg m{sup -2} s{sup -1}, heat flux from 11 to 32 kW m{sup -2}, evaporation temperature from 0 to 5.5 deg. C and vapor quality from 0.2 to 1. Based on the experimental results, the mechanism and role of the mass flow rate, heat flux, vapor quality and enhanced surface influencing the evaporation heat transfer coefficients were analyzed and discussed. In comparison to R22, the evaporation heat transfer coefficients for R417A were lower and much lower in the internally grooved tubes than in the smooth tube. The present experimental results are also compared with the existing correlations, and the modified Kattan model is found to be in much better agreement with the experimental results than the Kattan model. The Koyama and Wellsandt microfin models all tend to over predict the evaporation heat transfer coefficients rather strongly for R417A inside internally grooved tubes.

Heat transfer characteristics for evaporation of R417A flowing inside horizontal smooth and internally grooved tubes

Energy Technology Data Exchange (ETDEWEB)

Zhang, Xiaoyan [School of Energy and Power Engineering, Xi' an Jiaotong University, 28 Xianning Road, Xi' an, Shaanxi 710049 (China); School of Energy Engineering, Xi' an University of Science and Technology, 58 Yanta Street, Xi' an, Shaanxi 710054 (China); Zhang, Xingqun; Chen, Yunguang; Yuan, Xiuling [School of Energy and Power Engineering, Xi' an Jiaotong University, 28 Xianning Road, Xi' an, Shaanxi 710049 (China)

2008-06-15

The experimental study on evaporation heat transfer of R417A (R125/R134a/R600) flowing inside horizontal smooth and two internally grooved tubes with different geometrical parameters was conducted with the mass flow rate range from 176 to 344 kg m{sup -2} s{sup -1}, heat flux from 11 to 32 kW m{sup -2}, evaporation temperature from 0 to 5.5{sup o}C and vapor quality from 0.2 to 1. Based on the experimental results, the mechanism and role of the mass flow rate, heat flux, vapor quality and enhanced surface influencing the evaporation heat transfer coefficients were analyzed and discussed. In comparison to R22, the evaporation heat transfer coefficients for R417A were lower and much lower in the internally grooved tubes than in the smooth tube. The present experimental results are also compared with the existing correlations, and the modified Kattan model is found to be in much better agreement with the experimental results than the Kattan model. The Koyama and Wellsandt microfin models all tend to over predict the evaporation heat transfer coefficients rather strongly for R417A inside internally grooved tubes. (author)

Heat transfer characteristics for evaporation of R417A flowing inside horizontal smooth and internally grooved tubes

International Nuclear Information System (INIS)

Zhang Xiaoyan; Zhang Xingqun; Chen Yunguang; Yuan Xiuling

2008-01-01

The experimental study on evaporation heat transfer of R417A (R125/R134a/R600) flowing inside horizontal smooth and two internally grooved tubes with different geometrical parameters was conducted with the mass flow rate range from 176 to 344 kg m -2 s -1 , heat flux from 11 to 32 kW m -2 , evaporation temperature from 0 to 5.5 deg. C and vapor quality from 0.2 to 1. Based on the experimental results, the mechanism and role of the mass flow rate, heat flux, vapor quality and enhanced surface influencing the evaporation heat transfer coefficients were analyzed and discussed. In comparison to R22, the evaporation heat transfer coefficients for R417A were lower and much lower in the internally grooved tubes than in the smooth tube. The present experimental results are also compared with the existing correlations, and the modified Kattan model is found to be in much better agreement with the experimental results than the Kattan model. The Koyama and Wellsandt microfin models all tend to over predict the evaporation heat transfer coefficients rather strongly for R417A inside internally grooved tubes

Natural convection heat transfer within horizontal spent nuclear fuel assemblies

International Nuclear Information System (INIS)

Canaan, R.E.

1995-12-01

Natural convection heat transfer is experimentally investigated in an enclosed horizontal rod bundle, which characterizes a spent nuclear fuel assembly during dry storage and/or transport conditions. The basic test section consists of a square array of sixty-four stainless steel tubular heaters enclosed within a water-cooled rectangular copper heat exchanger. The heaters are supplied with a uniform power generation per unit length while the surrounding enclosure is maintained at a uniform temperature. The test section resides within a vacuum/pressure chamber in order to subject the assembly to a range of pressure statepoints and various backfill gases. The objective of this experimental study is to obtain convection correlations which can be used in order to easily incorporate convective effects into analytical models of horizontal spent fuel systems, and also to investigate the physical nature of natural convection in enclosed horizontal rod bundles in general. The resulting data consist of: (1) measured temperatures within the assembly as a function of power, pressure, and backfill gas; (2) the relative radiative contribution for the range of observed temperatures; (3) correlations of convective Nusselt number and Rayleigh number for the rod bundle as a whole; and (4) correlations of convective Nusselt number as a function of Rayleigh number for individual rods within the array

Analytical solution of velocity for ammonia-water horizontal falling-film flow

International Nuclear Information System (INIS)

Zhang, Qiang; Gao, Yide

2016-01-01

Highlights: • We built a new falling-film flow model that analyzed the film flow characteristics. • We have obtained a new formula of film thickness over the horizontal tube. • We derived analysis solution to analyze the effect of inertial force to velocity in the entrance region of liquid film. • It described the characters of the ammonia-waterfalling-film film over the horizontal tube. • It is good for falling-film absorption, generation and evaporation to optimizing the design parameters and further improving the capabilities. - Abstract: A new horizontal tube falling film velocity model was built and calculated to analyze the problem of film flow conditions. This model also analyzed the film thickness distribution in horizontal tube falling film flow and considered the effect of the inertial force on velocity. The film thickness and velocity profile can be obtained based on the principle of linear superposition, a method of separation of variables that introduces the effect of variable inertial force on the velocity profile in the process of falling-film absorption. The film flow condition and the film thickness distribution at different fluid Reynolds numbers (Re) and tube diameters were calculated and compared with the results of the Crank–Nicolson numerical solution under the same conditions. The results show that the film flow condition out of a horizontal tube and that the film thickness increases with the fluid Re. At a specific Re and suitable tube diameter, the horizontal tube reaches a more uniform film. Finally, the analysis results have similar trend with the experimental and numerical predicted data in literature.

Development of inspection equipment for fuel bundles of CANDU-PHWR using R981 underwater radiation tolerant camera

Energy Technology Data Exchange (ETDEWEB)

Koo, Dae-Seo; Cho, Moon-Sung; Jo, Chang-Keun; Jun, Ji-Su; Jung, Jong Yeob; Park, Kwang-June; Suk, Ho-Chun

2005-03-15

The inspection equipment of fuel bundles was developed, which could perform visual inspection and dimensional measurement on fuel bundles of CANDU-PHWR, to evaluate, analyze the defective behavior of fuel bundles and inner surface of pressure tubes of inherent two-phase flow over 24kg/s in CANDU-6. The R981 radiation tolerant camera system with pan and tilt function was ordered and manufactured, which was waterproof, shielding radiation in underwater 10m in depth. The performance test, of the system ,due to camera-object distance was carried out in air/underwater atmosphere. The results of performance test of R981 radiation tolerant camera system are good. The inspection equipment of fuel bundles using R981 radiation tolerant camera system and underwater-radiation tolerant LVDT sensor(D5/200AW) was fabricated, which could perform visual inspection and dimensional measurement on fuel bundles of CANDU-PHWR with measurement accuracy 10{mu}m. This equipment will be utilizable integrity evaluation of fuel bundles which are irradiated in pressure tube of CANDU-PHWR.

A new Theory for frequencies computation of overhead lines with bundle conductors.

OpenAIRE

dubois, Hervé; Dal Maso, Filipo; Lilien, Jean-Louis

1991-01-01

Vertical, horizontal and torsional mechanical frequencies are studies for both single and bundle conductor lines. Models and tests are presented. These data are of particular impact on galloping phenomenon. Peer reviewed

R404A condensing under forced flow conditions inside smooth, microfin and cross-hatched horizontal tubes

Energy Technology Data Exchange (ETDEWEB)

Infante Ferreira, C A; Nan, X [Delft University of Technology (Netherlands). Laboratory for Refrigeration and Indoor Climate Control; Newell, T A; Chato, J C [University of Illinois, Urbana, IL (United States). Dept. of Mechanical and Industrial Engineering

2003-06-01

Two-phase heat transfer coefficient characteristics of R404A condensing under forced flow conditions inside smooth, microfin and cross-hatched horizontal tubes are experimentally investigated. Experimental parameters include a lubricating polyol ester oil concentration varied from 0 to 4%. The test runs were done at average inlet saturated condensing temperatures of 40{sup o}C. The inlet vapor was kept at saturation (quality = 1.0). The mass fluxes were between 200 and 600 kg/m{sup 2}s, and the heat fluxes were selected to obtain a quality of 0.0 at the outlet of the test section, varying from 5 to 45 kW/m{sup 2}. The heat transfer enhancement factor varied between 1.8 and 2.4 for both microfin and cross-hatched tubes. The larger values applied for larger mass fluxes for the cross-hatched tube and smaller mass fluxes for the microfin tube. Enhancement factors increased as oil concentration increased up to oil concentrations of 2%. For higher oil concentrations the enhancement decreased especially at high mass fluxes, the cross-hatched tube being less sensitive to oil contamination. Pressure drop in the test section increased by approximately 25% as the oil concentration increased from 0 to 4%. The results from the experiments are compared with those calculated from correlations reported in the literature. Moreover, modified correlations for the condensation heat transfer coefficient are proposed for practical applications. (author)

Heat transfer characteristics of horizontal steam generators under natural circulation conditions

International Nuclear Information System (INIS)

Hyvaerinen, J.

1996-01-01

This paper deals with the heat transfer characteristics of horizontal steam generators, particularly under natural circulation (decay heat removal) conditions on the primary side. Special emphasis is on the inherent features of horizontal steam generator behaviour. A mathematical model of the horizontal steam generator primary side is developed and qualitative results are obtained analytically. A computer code, called HSG, is developed to solve the model numerically, and its predictions are compared with experimental data. The code is employed to obtain for VVER 440 steam generators quantitative results concerning the dependence of primary-to-secondary heat transfer efficiency on the primary side flow rate, temperature and secondary level. It turns out that the depletion of the secondary inventory leads to an inherent limitation of the decay energy removal in VVER steam generators. The limitation arises as a consequence of the steam generator tube bundle geometry. As an example, it is shown that the grace period associated with pressurizer safety valve opening during a station black-out is 2 1/2-3 hours instead of the 5-6 hours reported in several earlier studies. (However, the change in core heat-up timing is much less-about 1 h at most.) The heat transfer limitation explains the fact that, in the Greifswald VVER 440 station black-out accident in 1975, the steam generators never boiled dry. In addition, the stability of single-phase natural circulation is discussed and insights on the modelling of horizontal steam generators with general-purpose thermal-hydraulic system codes are also presented. (orig.)

Prediction of Heat Removal Capacity of Horizontal Condensation Heat Exchanger submerged in Pool

Energy Technology Data Exchange (ETDEWEB)

Jeon, Seong-Su; Hong, Soon-Joon [FNC Tech., Yongin (Korea, Republic of); Cho, Hyoung-Kyu [Seoul National University, Seoul (Korea, Republic of); Park, Goon-Cherl [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2014-10-15

As representative passive safety systems, there are the passive containment cooling system (PCCS) of ESBWR, the emergency condenser system (ECS) of the SWR-1000, the passive auxiliary feed-water system (PAFS) of the APR+ and etc. During the nuclear power plant accidents, these passive safety systems can cool the nuclear system effectively via the heat transfer through the steam condensation, and then mitigate the accidents. For the optimum design and the safety analysis of the passive safety system, it is essential to predict the heat removal capacity of the heat exchanger well. The heat removal capacity of the horizontal condensation heat exchanger submerged in a pool is determined by a combination of a horizontal in-tube condensation heat transfer and a boiling heat transfer on the horizontal tube. Since most correlations proposed in the previous nuclear engineering field were developed for the vertical tube, there is a certain limit to apply these correlations to the horizontal tube. Therefore, this study developed the heat transfer model for the horizontal Ushaped condensation heat exchanger submerged in a pool to predict well the horizontal in-tube condensation heat transfer, the boiling heat transfer on the horizontal tube and the overall heat removal capacity of the heat exchanger using the best-estimate system analysis code, MARS.

Analytical study of condensation heat transfer on titanium tube with super-hydrophobic surface

Energy Technology Data Exchange (ETDEWEB)

Ji, Dae Yun; Park, Hyun Gyu; Lee, Kwon Yeong [Handong Global University, Pohang (Korea, Republic of)

2016-05-15

There are many nuclear or fossil power plants which occupy more than 85% among entire power plants in the world. These plants release heat through condenser into nature. The condenser is an important component for cooling the working fluid after the turbine. Its performance is related with material and size of its tubes. To have good performance or to reduce condenser size, it is important to increase condensation heat transfer coefficient on condenser tubes. Ma et al. executed heat transfer experiment in dropwise condensation with non-condensable gas, and studied how the amount of air and pressure difference affect condensation heat transfer coefficient. The more non-condensable gas existed, the condensation heat transfer coefficient was decreased. Shen et al. studied condensation heat transfer at horizontal bundle tubes. Several variables such as coolant velocity, saturated pressure, and surface conditions were studied. As a result, surface modified brass tube and stainless tube showed higher condensation heat transfer coefficient as much as 1.3 and 1.4 times comparing with their bare tubes, in 70 kPa vacuum condition respectively. Rausch et al. studied dropwise condensation on ion-implanted titanium surface. Experimental study is performed to evaluate the performance of surface modified titanium tube in vacuum state. SAM coating is used to make super-hydrophobic surface of titanium tube. Preliminary analysis were performed considering filmwise and dropwise condensations, respectively. Experiment facility is almost prepared and the test result will be shown soon.

Development of Bundle Position-Wise Linear Model for Predicting the Pressure Tube Diametral Creep in CANDU Reactors

International Nuclear Information System (INIS)

Lee, Jae Yong; Na, Man Gyun

2011-01-01

Diametral creep of the pressure tube (PT) is one of the principal aging mechanisms governing the heat transfer and hydraulic degradation of a heat transport system. PT diametral creep leads to diametral expansion that affects the thermal hydraulic characteristics of the coolant channels and the critical heat flux. Therefore, it is essential to predict the PT diametral creep in CANDU reactors, which is caused mainly by fast neutron irradiation, reactor coolant temperature and so forth. The currently used PT diametral creep prediction model considers the complex interactions between the effects of temperature and fast neutron flux on the deformation of PT zirconium alloys. The model assumes that long-term steady-state deformation consists of separable, additive components from thermal creep, irradiation creep and irradiation growth. This is a mechanistic model based on measured data. However, this model has high prediction uncertainty. Recently, a statistical error modeling method was developed using plant inspection data from the Bruce B CANDU reactor. The aim of this study was to develop a bundle position-wise linear model (BPLM) to predict PT diametral creep employing previously measured PT diameters and HTS operating conditions. There are twelve bundles in a fuel channel and for each bundle, a linear model was developed by using the dependent variables, such as the fast neutron fluxes and the bundle temperatures. The training data set was selected using the subtractive clustering method. The data of 39 channels that consist of 80 percent of a total of 49 measured channels from Units 2, 3 and 4 were used to develop the BPLM models. The remaining 10 channels' data were used to test the developed BPLM models. The BPLM was optimized by the maximum likelihood estimation method. The developed BPLM to predict PT diametral creep was verified using the operating data gathered from the Units 2,3 and 4 in Korea. Two error components for the BPLM, which are the epistemic

Conceptual optimization using genetic algorithms for tube in tube structures

International Nuclear Information System (INIS)

Pârv, Bianca Roxana; Hulea, Radu; Mojolic, Cristian

2015-01-01

The purpose of this article is to optimize the tube in tube structural systems for tall buildings under the horizontal wind loads. It is well-known that the horizontal wind loads is the main criteria when choosing the structural system, the types and the dimensions of structural elements in the majority of tall buildings. Thus, the structural response of tall buildings under the horizontal wind loads will be analyzed for 40 story buildings and a total height of 120 meters; the horizontal dimensions will be 30m × 30m for the first two optimization problems and 15m × 15m for the third. The optimization problems will have the following as objective function the cross section area, as restrictions the displacement of the building< the admissible displacement (H/500), and as variables the cross section dimensions of the structural elements

Characterization of two-phase flow regimes in horizontal tubes using {sup 81m}Kr tracer experiments

Energy Technology Data Exchange (ETDEWEB)

Oriol, Jean [LPAC, CEA Grenoble, 17, rue des Martyrs, 38054 Grenoble Cedex 9 (France); Leclerc, Jean Pierre [Laboratoire des Sciences du Genie Chimique (LSGC), Nancy-Universite, CNRS, BP 20451, F-54001 Nancy (France)], E-mail: leclerc@ensic.inpl-nancy.fr; Berne, Philippe; Gousseau, Georges [L2T, CEA Grenoble, 17, rue des Martyrs, 38054 Grenoble Cedex 9 (France); Jallut, Christian [Universite de Lyon, Universite Lyon 1, LAGEP, UMR CNRS 5007, ESCPE, 43 Bd du 11 novembre 1918, 69622 Villeurbanne Cedex (France); Tochon, Patrice; Clement, Patrice [GRETh, CEA Grenoble, 17, rue des Martyrs, 38054 Grenoble Cedex 9 (France)

2008-10-15

The diagnosis of heat exchangers on duty with respect to flow mal-distributions needs the development of non-intrusive inlet-outlet experimental techniques in order to perform an online fault diagnosis. Tracer experiments are an example of such techniques. They can be applied to mono-phase heat exchangers but also to multi-phase ones. In this case, the tracer experiments are more difficult to perform. In order to check for the capabilities of tracer experiments to be used for the flow mal-distribution diagnosis in the case of multi-phase heat exchangers, we present here a preliminary study on the simplest possible system: two-phase flows in a horizontal tube. {sup 81m}Kr is used as gas tracer and properly collimated NaI (TI) crystal scintillators as detectors. The specific shape of the tracer response allows two-phase flow regimes to be characterized. Signal analysis allows the estimation of the gas phase real average velocity and consequently of the liquid phase real average velocity as well as of the volumetric void fraction. These results are compared successfully to those obtained with liquid phase tracer experiments previously presented by Oriol et al. 2007. Characterization of the two-phase flow regimes and liquid dispersion in horizontal and vertical tubes using coloured tracer and no intrusive optical detector. Chem. Eng. Sci. 63(1), 24-34, as well as to those given by correlations from literature.

Eddy current testing device for metallic tubes at least locally curved

International Nuclear Information System (INIS)

Pigeon, Marcel; Vienot, Claude.

1975-01-01

Steam generators, condensers and heat exchangers generally consist of metallic tube bundles, the tubes having a complex geometry. The invention concerns an Eddy current testing device for metallic tubes at least locally curved, operating by translation of a probe inside the tubes [fr

Safety analysis report of the irradiation test of Type-B bundle

Energy Technology Data Exchange (ETDEWEB)

Lee, Choong Sung; Lim, I. C.; Lee, B. C.; Ryu, J. S.; Kim, H. R

2000-06-01

The HANARO fuel, U{sub 3}Si-A1, has been developed by AECL and tested in NRU reactor. In the course of the fuel qualification tests, only one case was performed under the higher power condition than maximum linear power which was expected in the design stage. The Korea regulatory body, KINS imposed that HANARO shall be operated at the power level less than 24MW which is 80% of the design full power until HANARO shows the repetitive performance of the fuel at the power condition abov e 112.8KW/m. To resolve this imposition, KAERI designed two types of special test bundles: two non-instrumented(Type-A) and one instrumented(Type-B) test bundles. Two Type-A bundles were irradiated in HANARO: one of them has finished PIE and the other is under PIE. Type-B bundle was loaded in the core during 1.32 day at 1996, but outstanding FIV(flow induced vibration) was observed at the pool top because of long guide tube attached to the top of the bundle. The successful installation of the chimney fastener to fix the guide tube resulted in conducting the irradiation test of Type-B bundle again. The test will start at mid- July, 2000. In order to safely do the Type-B irradiation test, the safety analysis for the nuclear, mechanical and thermal-hydraulic aspects was performed. The reactivity worth and the maximum 1 near power predicted by VENTURE are 6.3mk/k and 121.6kW/m, respectively. Thermal margins for normal and transient conditions using MATRA-h, are assessed to satisfy the safety criteria.

Fixing device for a tube bundle especially for steam generator

International Nuclear Information System (INIS)

Fournier, Y.

1983-01-01

The helical tubes in concentric layers are maintained by a device comprising longitudinal rods with concave cylindrical slots to hold the tubes of the same cylindrical layer. The rods are radially disposed for every cylindrical layers. The tubes are fixed on the rods by fixation elements between two successive rods on the same radius. The tube is maintained in the slot isostatically by three points [fr

Possible first occurrence of external corrosion on alloy 600TT tubes in France

International Nuclear Information System (INIS)

Boccanfuso, M.; Thebault, Y.; Massini, B.; Bigne, L.

2015-01-01

During the last decade, in different countries, several occurrences of external corrosion have been identified on steam generator (SG) tube bundles equipped with thermally treated 600 alloy. In France, this feedback leads EDF to enhance the SG inspection program. Nevertheless, until now, no damage of this type was reported. Recently, during in-service inspection at the Cattenom plant on a SG equipped with alloy 600TT tubes, Eddy current tests have highlighted a signal that could be related to external corrosion. The tube was removed and sent to the EDF hot laboratory for destructive examinations. Various exams were performed at different scales to characterize the causes of this NDT signal, the material properties and the residual stresses. The assessments carried out on the tube conclude that the source of the damage is external intergranular stress corrosion cracking, also called ODSCC (Outside Diameter Stress Corrosion Cracking) making it the first occurrence on the tube bundles made of alloy 600TT in the French fleet. This first case of 600 TT ODSCC in France is an unexpected and particular one, because of its altitude in the full mechanical rolling area. This is reinforced by the low number of occurrences noted to date (only one after nearly 30 years of operation of alloy 600TT tube bundles). International (Biblis) OPEX had identified recent IGSCC with cracks initiated and propagated in the tubesheet. For this case, the scenario considered requires highly restrictive conditions (tube in the sludge zone and on the periphery of the tube bundle, including the tube lane) and may explain the singular nature of the Cattenom tube

Gentilly 2 steam generators Spring 2000 outage: tubesheet waterlance cleaning and inspection; upper bundle inspection

International Nuclear Information System (INIS)

Akeroyd, J.K.; Plante, S.

2000-01-01

A review of the secondary side maintenance activities recently completed during the Gentilly 2 Annual Spring 2000 Maintenance Outage. Activities included: 1) Tubesheet intertube waterlance cleaning and visual inspection, 2) First tube support plate, in-bundle visual inspection of the hot leg, and 3) Upper bundle tube support plate visual inspection. A description of the waterlancing and inspection equipment and setup in the RB at Gentilly 2 is provided. Several innovative techniques were successfully employed and yielded savings in critical path duration, labour and personnel radiation dose. These included accessing the SG tubesheet region through one handhole only and sludge removal utilizing the SG blowdown system. Plant personnel judged tubesheet sludge removal successful. Before and after results of the cleaning process along with samples of the visual inspection results are provided. Inspection of the first support plate, which was a repeat of an inspection done in 1997, was conducted along with an in-bundle inspection of the upper tube supports. Results are presented along with a discussion of the implications for future steam generator maintenance. (author)

Flow condensation pressure drop characteristics of R410A-oil mixture inside small diameter horizontal microfin tubes

Energy Technology Data Exchange (ETDEWEB)

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

2010-11-15

Flow condensation pressure drop characteristics of R410A-oil mixture inside small diameter (5.0 mm and 4.0 mm O.D.) horizontal microfin tubes were investigated experimentally covering nominal oil concentrations from 0% to 5%. The research results indicate that, comparing with the frictional pressure drop of pure R410A, the frictional pressure drop of R410A-oil mixture may decrease by maximum of 18% when the vapor quality is lower than 0.6, and increase by maximum of 13% when the vapor quality is higher than 0.6. A new frictional pressure drop correlation for R410A-oil mixture flow condensation inside microfin tubes is developed based on the refrigerant-oil mixture properties, and can agree with 94% of the experimental data within a deviation of -30% to +30%. (author)

Contribution to the heat transfer analysis of substitute refrigerants in evaporator tubes with smooth or enhanced tube surfaces

Energy Technology Data Exchange (ETDEWEB)

Kattan, N

1997-12-31

The substitution of CFC refrigerants in refrigeration systems, heat pumps and organic Rankine cycles for heat recovery, requests a good knowledge of heat transfer properties of substitute fluids. A new test facility has been built at the Laboratory for Industrial Energy Systems (LENI) to contribute to this international effort. It consists of two sets of concentric tubes allowing either annular or inside tube convective boiling with a counter current water flow heating to be studied. A new data base including heat transfer coefficients and pressure drop measurements for four new refrigerants (R123, R134A, R402A and R404A) and three older refrigerants (R11, R12 and R502) has been collected. Flow boiling measurements covered a broad range of mass velocities, vapor qualities and heat fluxes. Some of the tests included plain tubes and others enhanced surface tubes (microfilms from Wieland) in horizontal and vertical orientations. An improved Wilson plot technique, that covers both the transition and turbulent flow regimes of the water flowing in the annular channel for the inside tube boiling tests, is proposed to overcome the severe limitations of conventional Wilson plots, to improve accuracy and to facilitate data processing. Mean flow boiling heat transfer coefficients were measured for R12 and R134A evaporating inside a horizontal plain tube and for R11 and R123 evaporating inside a horizontal plain tube. Local flow boiling heat transfer coefficients were measured for : R134A, R123, R404A and R502 evaporating inside a horizontal plain tube, for R134A and R123 evaporating inside a horizontal microfin tube and for R134 evaporating inside a vertical microfin tube. In addition microfin heat transfer augmentation relative to plain tube test data was investigated. The measured heat transfer coefficients were compared to different existing inside tube flow boiling correlations. (author) figs., tabs., refs.

Evaporation heat transfer of carbon dioxide at low temperature inside a horizontal smooth tube

Science.gov (United States)

Yoon, Jung-In; Son, Chang-Hyo; Jung, Suk-Ho; Jeon, Min-Ju; Yang, Dong-Il

2017-05-01

In this paper, the evaporation heat transfer coefficient of carbon dioxide at low temperature of -30 to -20 °C in a horizontal smooth tube was investigated experimentally. The test devices consist of mass flowmeter, pre-heater, magnetic gear pump, test section (evaporator), condenser and liquid receiver. Test section is made of cooper tube. Inner and outer diameter of the test section is 8 and 9.52 mm, respectively. The experiment is conducted at mass fluxes from 100 to 300 kg/m2 s, saturation temperature from -30 to -20 °C. The main results are summarized as follows: In case that the mass flux of carbon dioxide is 100 kg/m2 s, the evaporation heat transfer coefficient is almost constant regardless of vapor quality. In case of 200 and 300 kg/m2 s, the evaporation heat transfer coefficient increases steadily with increasing vapor quality. For the same mass flux, the evaporation heat transfer coefficient increases as the evaporation temperature of the refrigerant decreases. In comparison of heat transfer correlations with the experimental result, the evaporation heat transfer correlations do not predict them exactly. Therefore, more accurate heat transfer correlation than the previous one is required.

An advanced tube wear and fatigue workstation to predict flow induced vibrations of steam generator tubes

International Nuclear Information System (INIS)

Gay, N.; Baratte, C.; Flesch, B.

1997-01-01

Flow induced tube vibration damage is a major concern for designers and operators of nuclear power plant steam generators (SG). The operating flow-induced vibrational behaviour has to be estimated accurately to allow a precise evaluation of the new safety margins in order to optimize the maintenance policy. For this purpose, an industrial 'Tube Wear and Fatigue Workstation', called 'GEVIBUS Workstation' and based on an advanced methodology for predictive analysis of flow-induced vibration of tube bundles subject to cross-flow has been developed at Electricite de France. The GEVIBUS Workstation is an interactive processor linking modules as: thermalhydraulic computation, parametric finite element builder, interface between finite element model, thermalhydraulic code and vibratory response computations, refining modelling of fluid-elastic and random forces, linear and non-linear dynamic response and the coupled fluid-structure system, evaluation of tube damage due to fatigue and wear, graphical outputs. Two practical applications are also presented in the paper; the first simulation refers to an experimental set-up consisting of a straight tube bundle subject to water cross-flow, while the second one deals with an industrial configuration which has been observed in some operating steam generators i.e., top tube support plate degradation. In the first case the GEVIBUS predictions in terms of tube displacement time histories and phase planes have been found in very good agreement with experiment. In the second application the GEVIBUS computation showed that a tube with localized degradation is much more stable than a tube located in an extended degradation zone. Important conclusions are also drawn concerning maintenance. (author)

Reactor physics assessment of modified 37-element CANDU fuel bundles

International Nuclear Information System (INIS)

Pristavu, R.; Rizoiu, A.

2016-01-01

Reducing the central element diameter in order to improve the total flow area of CANDU fuel bundle and redistribute the power density of all remaining elements was studied in Canada and Korea when considering the effect of aging pressure tube diametral creep. The aim of this paper is to study the modified bundle behavior using the transport codes WIMS and DRAGON. In calculations, a WIMS nuclear data library on 172 energy groups was used. 2-D transport calculations were performed with WIMS and DRAGON, leading to similar results in estimated cell parameters. Additionally, 3-D DRAGON calculations were carried on in order to evaluate the local flux distribution shift, as well as the incremental cross sections for supercells containing modified CANDU bundles and reactivity devices. The overall effect of using modified fuel bundles was meaningless for both cell and supercell parameters, thus ensuring this possibility of fuel improvement for thermal-hydraulic purposes only. (authors)

Enhanced Evaporation and Condensation in Tubes

Science.gov (United States)

Honda, Hiroshi

A state-of-the-art review of enhanced evaporation and condensation in horizontal microfin tubes and micro-channels that are used for air-conditioning and refrigeration applications is presented. The review covers the effects of flow pattern and geometrical parameters of the tubes on the heat transfer performance. Attention is paid to the effect of surface tension which leads to enhanced evaporation and condensation in the microfin tubes and micro-channels. A review of prior efforts to develop empirical correlations of the heat transfer coefficient and theoretical models for evaporation and condensation in the horizontal microfin tubes and micro-channels is also presented.

A steam superheater exchanger provided with two coaxial casings and an horizontal axis

International Nuclear Information System (INIS)

Marjollet, Jacques; Palacio, Gerard; Tondeur, Gerard.

1976-01-01

This invention concerns the general lay-out of an horizontal axis separator-superheater for supplying steam to a high power turbine, particularly for a nuclear power station. The invention significantly reduces the length of the pipework connecting the superheated steam outlet and its inlet to the turbine. For this, the outer casing is provided with a coaxial internal annular sleeve in which are housed, one above the other, the separator and the bundle of superheater tubes through which circulates the water emulsion to be separated and steam to be superheated. At the end of its treatment, the superheated steam spreads out in the space between the sleeve and the outer casing from whence it can be drawn off at any point of its periphery, thus making it possible to choose an extraction point as near as possible to the inlet of the turbine to be fed [fr

Evaluation of a new laser-resistant fabric and copper foil-wrapped endotracheal tube.

Science.gov (United States)

Sosis, M B; Braverman, B; Caldarelli, D D

1996-07-01

The risk of an endotracheal tube's combustion during laser airway surgery necessitates the use of special anesthetic techniques and equipment to prevent this complication. This study was designed to evaluate the Laser-Trach(TM), a new laser-resistant rubber endotracheal tube for use during laser airway surgery. The Laser-Trach endotracheal tubes that were evaluated were size 6.0 mm internal diameter (ID) red rubber endotracheal tubes which had been commercially wrapped by Kendall-Sheridan (Mansfield, Mass.) with copper foil tape and overwrapped with fabric. The fabric layer was saturated with water prior to our tests, as recommended by the manufacturer. The Laser-Trach endotracheal tubes were compared with plain (bare) size 6.0 mm ID Rusch red rubber endotracheal tubes. The tubes under study were positioned horizontally on wet towels in air and had 5 L x min(-1) of oxygen flowing through them. They were subjected to continuous laser radiation at 40 W from either a CO2 or an Nd-YAG laser. The Nd-YAG laser was propagated via a 600-micron fiber bundle. Each laser was directed perpendicularly at the shaft of the endotracheal tube being studied, and its output was continued until a blowtorch fire occurred or 60 seconds had elapsed. Sixty seconds of CO2 laser fire did not ignite any of the eight Laser-Trach endotracheal tubes tested. However, blowtorch ignition of all eight bare rubber tubes tested occurred after 0.87 +/- 0.21 (mean +/- SD) seconds of CO2 laser fire. Nd-YAG laser contact with the Laser-Trach endotracheal tubes caused the perforation and blowtorch ignition of all eight tubes tested after 18.79 +/- 7.83 seconds. This was a significantly (Presistant to the C02 laser. However, this endotracheal tube is not recommended for use with the Nd-YAG laser.

Manufacturing of 37-element fuel bundles for PHWR 540 - new approach

Energy Technology Data Exchange (ETDEWEB)

Arora, U.K.; Sastry, V.S.; Banerjee, P.K.; Rao, G.V.S.H.; Jayaraj, R.N. [Nuclear Fuel Complex, Dept. Atomic Energy, Government of India, Hyderabad (India)

2003-07-01

Nuclear Fuel Complex (NFC), established in early seventies, is a major industrial unit of Department of Atomic Energy. NFC is responsible for the supply of fuel bundles to all the 220 MWe PHWRs presently in operation. For supplying fuel bundles for the forthcoming 540 MWe PHWRs, NEC is dovetailing 37-element fuel bundle manufacturing facilities in the existing plants. In tune with the philosophy of self-reliance, emphasis is given to technology upgradation, higher customer satisfaction and application of modern quality control techniques. With the experience gained over the years in manufacturing 19-element fuel bundles, NEC has introduced resistance welding of appendages on fuel tubes prior to loading of UO{sub 2} pellets, use of bio-degradable cleaning agents, simple diagnostic tools for checking the equipment condition, on line monitoring of variables, built-in process control methods and total productive maintenance concepts in the new manufacturing facility. Simple material handling systems have been contemplated for handling of the fuel bundles. This paper highlights the flow-sheet adopted for the process, design features of critical equipment and the methodology for fabricating the 37-element fuel bundles, 'RIGHT FIRST TIME'. (author)

Manufacturing of 37-element fuel bundles for PHWR 540 - new approach

International Nuclear Information System (INIS)

Arora, U.K.; Sastry, V.S.; Banerjee, P.K.; Rao, G.V.S.H.; Jayaraj, R.N.

2003-01-01

Nuclear Fuel Complex (NFC), established in early seventies, is a major industrial unit of Department of Atomic Energy. NFC is responsible for the supply of fuel bundles to all the 220 MWe PHWRs presently in operation. For supplying fuel bundles for the forthcoming 540 MWe PHWRs, NEC is dovetailing 37-element fuel bundle manufacturing facilities in the existing plants. In tune with the philosophy of self-reliance, emphasis is given to technology upgradation, higher customer satisfaction and application of modern quality control techniques. With the experience gained over the years in manufacturing 19-element fuel bundles, NEC has introduced resistance welding of appendages on fuel tubes prior to loading of UO 2 pellets, use of bio-degradable cleaning agents, simple diagnostic tools for checking the equipment condition, on line monitoring of variables, built-in process control methods and total productive maintenance concepts in the new manufacturing facility. Simple material handling systems have been contemplated for handling of the fuel bundles. This paper highlights the flow-sheet adopted for the process, design features of critical equipment and the methodology for fabricating the 37-element fuel bundles, 'RIGHT FIRST TIME'. (author)

Experimental and numerical study on thermal-hydraulic performance of wing-shaped-tubes-bundle equipped with winglet vortex generators

Science.gov (United States)

Abdelatief, Mohamed A.; Sayed Ahmed, Sayed Ahmed E.; Mesalhy, Osama M.

2018-03-01

The present work evaluates, experimentally and numerically, by the aid of commercial code FLUENT 6.3.26, the effects of relative locations (ΔX or ΔY), heights (hw), and span-angle (θ) of winglet-vortex-generators (WVGs) on thermal-hydraulic performance enhancement for down-stream and/or up-stream wing-shaped-tubes bundle heat exchangers for air Re ranging from 1.85 × 103 to 9.7 × 103 while water Re = 5 × 102. hw is set as 5 mm, 7.5 mm and 10 mm. For tube down-stream, θ is set as 0° (Base-line-case) and from 5° to 45° clockwise common-flow up (CFUp) and counterclockwise common-flow down (CFDn) while for tube up-stream it is set as -5°, -10° and -15° CFUp. Results show that the increase of θ counterclockwise-(CFDn) or clockwise-(CFUp) leads to increase the values of Nu number. Using WVGs with (+5 ° ≤ θ ≤ +45°) results in increasing Nu number by about from 34 to 48% comparing with that of base-line-case. The lowest values of drag coefficient ( f) for tube down-stream are obtained at +5° CFDn and -15° CFUp with respect to the base-line case. For tube up-stream, Nu number increases by increasing θ from 0° to -5° and the values of Nu number for θ varying from -5° to -15° have no significant changes. ( f) increases with hw and has negligible effect on ha. Furthermore, optimization analyses of θ and longitudinal fin (LF) are utilized, in four cases, for finding the optimum combination and maximum efficiency. The highest values of heat transfer parameters such as effectiveness (ɛ), area goodness factor (G) and efficiency index (η) and the lowest values of fluid-flow parameters like ( f) and hence the best efficiency, are achieved for -15° CFUp down-stream, ("case 3" of -15° CFUp down-stream and 6 mm LF height) and +5° CFDn down-stream. Correlations of Nu number, ( f) and (ɛ) as a function of θ and Re for the studied cases are performed.

Influence of thermal buoyancy on vertical tube bundle thermal density head predictions under transient conditions

International Nuclear Information System (INIS)

Lin, H.C.; Kasza, K.E.

1984-01-01

The thermal-hydraulic behavior of an LMFBR system under various types of plant transients is usually studied using one-dimensional (1-D) flow and energy transport models of the system components. Many of the transient events involve the change from a high to a low flow with an accompanying change in temperature of the fluid passing through the components which can be conductive to significant thermal bouyancy forces. Thermal bouyancy can exert its influence on system dynamic energy transport predictions through alterations of flow and thermal distributions which in turn can influence decay heat removal, system-response time constants, heat transport between primary and secondary systems, and thermal energy rejection at the reactor heat sink, i.e., the steam generator. In this paper the results from a comparison of a 1-D model prediction and experimental data for vertical tube bundle overall thermal density head and outlet temperature under transient conditions causing varying degrees of thermal bouyancy are presented. These comparisons are being used to generate insight into how, when, and to what degree thermal buoyancy can cause departures from 1-D model predictions

Flux and power distributions in BWR multi-bundle fuel arrays

International Nuclear Information System (INIS)

Cheng, H.S.

1976-02-01

Multi-bundle calculations have been performed in order to shed some light on an abnormal TIP trace recently discovered in a BWR/3. Transport theory was employed to perform the calculations with ENDF/B-IV data. The results indicate that a strong variation of the TIP reading does exist along the narrow water gap of a BWR due to the steep gradient of the thermal neutron flux; the maxima occurring at the intersections of the water gaps and the minima in between. Using this characteristic behavior of the TIP reading, together with the observed normal TIP trace, the abnormal behavior of the affected TIP trace exhibiting three peaks along the channel was roughly simulated. The calculations confirmed that the observed TIP trace anomaly was caused by the severe bending of the affected instrument tube as was actually discovered. The effect of hot water intrusion into the TIP guide tube, as well as that of loading the new 8 x 8 reload bundles, was also evaluated

Vibrations of tube arrays in transversal flow

International Nuclear Information System (INIS)

Gibert, R.J.; Doyen, R.

1981-08-01

In this study the local forces per unit length acting in a tube in a single row and in bundle have been measured. Their modification by a given harmonic motion of the tube itself or of an adjacent tube has been particularly studied. Some complementary experiments have been performed to extend the whirling coefficient tabulation and also to precise the effect of the upstream velocity profile on the whirling critical velocities [fr

Dryout power of a CANFLEX bundle string with raised bearing pads

International Nuclear Information System (INIS)

Leung, L.K.H.; Dimmick, G.R.; Bullock, D.E.; Inch, W.W.R.; Jun, J.S.; Suk, H.C.

2001-01-01

Dryout power data have been obtained with CANFLEX bundle strings equipped with raised bearing pads (1.7 mm and 1.8 mm height as compared to 1.4 mm in the current Mk-IV design) at Stern Laboratories. The experiment covered a wide range of steam-water flow conditions in three flow tubes simulating uncrept, and 3.3% and 5.1% crept profiles. The dryout power follows consistent parametric trends: it increases with increasing mass-flow rate, and decreases with increasing pressure, inlet-fluid temperature and channel creep. Local and boiling-length-average (BLA) critical-heat-flux (CHF) values were evaluated from the dryout-power measurements. The dryout power and BLA CHF values of the high bearing-pad bundles are higher than those of the low bearing-pad bundles at the same channel inlet flow conditions. On average, the dryout powers for bundles with 1.7 mm and 1.8 mm bearing pads are about 8% and 10%, respectively, higher than those for the bundle with 1.4 mm bearing pads. Compared to the 37-element bundle, an enhancement in dryout power is shown with CANFLEX bundles for all bearing-pad heights, at flow conditions of interest for reactor licensing. The average dryout power enhancement varies from 4% for the CANFLEX bundle with 1.4 mm bearing pads in the uncrept channel to 27% for the CANFLEX bundle with 1.8 mm bearing pads in the 5.1% crept channel. (author)

Structural integrity evaluations of CANDU pressure tubes

International Nuclear Information System (INIS)

Radu, Vasile

2003-01-01

The core of a CANDU-6 pressurized heavy water reactor consists of some hundred horizontal pressure tubes that are manufactured from a Zr-2.5%Nb alloy and which contain the fuel bundles. These tubes are susceptible to a damaging phenomenon known as Delayed Hydride Cracking (DHC). The Zr-2.5%Nb alloy is susceptible to DHC phenomenon when there is diffusion of hydrogen atoms to a service-induced flaws, followed by the hydride platelets formation on the certain crystallographic planes in the matrix material. Finally, the development of hydride regions at the flaw-tip will happened. These hydride regions are able to fracture under stress-temperature conditions (DHC initiation) and the cracks can extend and grow by DHC mechanism. Some studies have been focused on the potential to initiate DHC at the blunt flaws in a CANDU reactor pressure tube and a methodology for structural integrity evaluation was developed. The methodology based on the Failure Assessment Diagrams (FAD's) consists in an integrated graphical plot, where the fracture failure and plastic collapse are simultaneously evaluated by means of two non-dimensional variables (K r and L r ). These two variables represent the ratio of the applied value of either stress or stress intensity factor and the resistance parameter of corresponding magnitude (yield stress or fracture toughness, respectively). Once the plotting plane is determined by the variables K r and L r , the procedure defines a critical failure line that establishes the safe area. The paper will demonstrate the possibility to perform structural integrity evaluations by means of Failure Assessment Diagrams for flaws occurring in CANDU pressure tubes. (author)

The Comparison Analysis of Thermalhydraulic Behavior Between A Reference 37-element Bundle and A Modified 37-element Bundle

International Nuclear Information System (INIS)

Ryu, Eui-Seung; You, Sung-Chang

2014-01-01

As pressure tube diameter creep increase, the coolant flows through some of the interior subchannels of the fuel bundle are reduced and consequently reduces the Critical Heat Flux (CHF). For this reason, Canadian Utilities have performed the project that developing the new fuel design (modified 37-element bundle) to increase critical heat flux. The modified 37-element (37M) bundle has the same overall geometry as the reference 37-element (37R) bundle that is using in the Wolsong units now but the center element diameter has been reduced from 13.06mm to 11.5mm. The reduction in center element diameter of the 37M bundle design increase the flow of center areas to improve the cooling and thus to enhance CHF. The CHF experiments with 37M bundle string simulator in un-crept and crept (3.3%, 5.1% peak creep) flow channels were completed at Stern Laboratories in 2008. A substantially large increase in dryout-power was observed for the 37M bundle compared to the 37R bundle, particularly in the 5.1% crept channel. As a result of the experiments, Ontario Power Generation (OPG) and Bruce Power (BP) have increased the operational margin with this CHF correlation and has fully refueled the 37M fuel on some units or almost done on the other units. KHNP also has performed the project to refuel the 37M bundle which is the same design with OPG and BP recently. This paper summarizes the comparison assessment of Thermalhydraulic (T/H) behavior for 37M bundle and 37R bundle with their own correlations and geometry parameters. This analysis performed with the thermal hydraulic code (NUCIRC) and the site measured data at the Wolsong Unit2. Tests to evaluate the CHF performance with the 37M fuel bundle have been conducted in 2008 using the un-crept, 3.3% crept and 5.1% crept flow channels in the CHF Test facility at Stern Laboratories. In addition pressure drop tests have been performed at the same time. The changes of geometry from 37R bundle to 37M bundle reduced the center element

Pressure drop characteristics in tight-lattice bundles for reduced-moderation water reactors

International Nuclear Information System (INIS)

Tamai, Hidesada; Kureta, Masatoshi; Yoshida, Hiroyuki; Akimoto, Hajime

2004-01-01

The reduced-moderation water reactor (RMWR) consists of several distinctive structures; a triangular tight-lattice configuration and a double-flat core. In order to design the RMWR core from the point of view of thermal-hydraulics, an evaluation method on pressure drop characteristics in the rod bundles at the tight-lattice configuration is required. In this study, calculated results by the Martinelli-Nelson's and Hancox's correlations were compared with experimental results in 4 x 5 rod bundles and seven-rod bundles. Consequently, the friction loss in two-phase flows becomes smaller at the tight-lattice configuration with the hydraulic diameter less than about 3 mm. This reason is due to the difference of the configuration between the multi-rod bundle and the circular tube and due to the effect of the small hydraulic diameter on the two-phase multiplier. (author)

Steady-state heat transfer in an inverted U-tube steam generator

International Nuclear Information System (INIS)

Boucher, T.J.

1986-01-01

Experimental results are presented involving U-tube steam generator tube bundle local heat transfer and fluid conditions during steady-state, full-power operations performed at high temperatures and pressures with conditions typical of a pressurized water reactor (15.0 MPa primary pressure, 600 K hot-leg fluid temperatures, 6.2 MPa secondary pressure). The MOD-2C facility represents the state-of-the-art in measurement of tube local heat transfer data and average tube bundle secondary fluid density at several elevations, which allows an estimate of the axial heat transfer and void distributions during steady-state and transient operations. The method of heat transfer data reduction is presented and the heat flux, secondary convective heat transfer coefficient, and void fraction distributions are quantified for steady-state, full-power operations

Steady-state heat transfer in an inverted U-tube steam generator

International Nuclear Information System (INIS)

Boucher, T.J.

1987-01-01

Experimental results are presented involving U-tube steam generator tube bundle local heat transfer and fluid conditions during stead-state, full-power operations performed at high temperatures and pressures with conditions typical of a pressurized water reactor (15.0 MPa primary pressure, 600 K steam generator inlet plenum fluid temperatures, 6.2 MPa secondary pressure). The Semiscale (MOD-2C facility represents the state-of-the-art in measurement of tube local heat transfer data and average tube bundle secondary fluid density at several elevations, which allows an estimate of the axial heat transfer and void distributions during steady-state and transient operations. The method of heat transfer data reduction is presented and the heat flux, secondary convective heat transfer coefficient, and void fraction distributions are quantified for steady-state, full-power operations

New insights into controlling tube-bundle fouling using alternative amines

Energy Technology Data Exchange (ETDEWEB)

Turner, C.W.; Klimas, S.J.; Guzonas, D.A.; Fruzzetti, K. [Atomic Energy of Canada Ltd. (Canada); Frattini, P.L. [Electric Power Research Inst. (United States)

2002-07-01

A volatile amine is added to the secondary heat-transport system of a nuclear power plant to reduce the rate of corrosion and corrosion product transport in the feedwater and to protect steam generator (SG) crevices and materials exposed to steam condensate. Volatility and base strength of the amine at the SG operating temperature are two important considerations when choosing the optimum amine (or mixture of amines) for corrosion control in the steam cycle. The investigation has found that the rate of tube-bundle fouling is strongly dependent upon the surface chemistry of the corrosion products. For example, the fouling rates of fully oxidized iron oxides, such as hematite and lepidocrocite, are at least an order of magnitude greater than the fouling rate of magnetite under identical operating conditions. The difference is related to the sign of the surface charge on the corrosion products at temperature. The choice of amine for pH-control also influences the fouling rate. This was originally thought to be a surface-charge effect as well, but recent tests have suggested that it is related to the role that the amine plays in governing the rate of deposit consolidation on the heat-transfer surface. Amines that promote a high rate of deposit consolidation result in a low rate of deposit removal and a high fouling rate. Conversely, amines that tend to inhibit deposit consolidation produce a higher rate of deposit removal and a lower fouling rate. Dimethyl-amine and dodecyl-amine have been identified as two amines that inhibit the rate of deposit consolidation and, consequently, result in fouling rates that are up to 5 times lower than rates measured for amines that promote consolidation. A significant difference between morpholine (high fouling rate) and dimethyl-amine (low fouling rate) is that the latter desorbs more slowly from the surface of magnetite. How to account for a correlation between slow desorption kinetics and lower rate constants for deposition and

New insights into controlling tube-bundle fouling using alternative amines

International Nuclear Information System (INIS)

Turner, C.W.; Klimas, S.J.; Guzonas, D.A.; Fruzzetti, K.; Frattini, P.L.

2002-01-01

A volatile amine is added to the secondary heat-transport system of a nuclear power plant to reduce the rate of corrosion and corrosion product transport in the feedwater and to protect steam generator (SG) crevices and materials exposed to steam condensate. Volatility and base strength of the amine at the SG operating temperature are two important considerations when choosing the optimum amine (or mixture of amines) for corrosion control in the steam cycle. The investigation has found that the rate of tube-bundle fouling is strongly dependent upon the surface chemistry of the corrosion products. For example, the fouling rates of fully oxidized iron oxides, such as hematite and lepidocrocite, are at least an order of magnitude greater than the fouling rate of magnetite under identical operating conditions. The difference is related to the sign of the surface charge on the corrosion products at temperature. The choice of amine for pH-control also influences the fouling rate. This was originally thought to be a surface-charge effect as well, but recent tests have suggested that it is related to the role that the amine plays in governing the rate of deposit consolidation on the heat-transfer surface. Amines that promote a high rate of deposit consolidation result in a low rate of deposit removal and a high fouling rate. Conversely, amines that tend to inhibit deposit consolidation produce a higher rate of deposit removal and a lower fouling rate. Dimethyl-amine and dodecyl-amine have been identified as two amines that inhibit the rate of deposit consolidation and, consequently, result in fouling rates that are up to 5 times lower than rates measured for amines that promote consolidation. A significant difference between morpholine (high fouling rate) and dimethyl-amine (low fouling rate) is that the latter desorbs more slowly from the surface of magnetite. How to account for a correlation between slow desorption kinetics and lower rate constants for deposition and

An apparatus with a horizontal capillary tube intended for measurement of the surface tension of supercooled liquids

Science.gov (United States)

Vinš, Václav; Hošek, Jan; Hykl, Jiří; Hrubý, Jan

2015-05-01

New experimental apparatus for measurement of the surface tension of liquids under the metastable supercooled state has been designed and assembled in the study. The measuring technique is similar to the method employed by P.T. Hacker [NACA TN 2510] in 1951. A short liquid thread of the liquid sample was sucked inside a horizontal capillary tube partly placed in a temperature-controlled glass chamber. One end of the capillary tube was connected to a setup with inert gas which allowed for precise tuning of the gas overpressure in order of hundreds of Pa. The open end of the capillary tube was precisely grinded and polished before the measurement in order to assure planarity and perpendicularity of the outer surface. The liquid meniscus at the open end was illuminated by a laser beam and observed by a digital camera. Application of an increasing overpressure of the inert gas at the inner meniscus of the liquid thread caused variation of the outer meniscus such that it gradually changed from concave to flat and subsequently convex shape. The surface tension at the temperature of the inner meniscus could be evaluated from the overpressure corresponding to exactly planar outer meniscus. Detailed description of the new setup together with results of the preliminary tests is provided in the study.

The experiments of resonance tube method and resonator method ...

Indian Academy of Sciences (India)

Take a 50 ml measuring cylinder or a big test tube and hold it vertical using a stand. Add some water in the tube so that the horizontal surface of water is above the rounded bottom. Stick a strip of graph paper or ruler on the tube and mark a horizontal line corresponding to the water level. Keep the speaker pointing towards.

Natural convection heat transfer for a staggered array of heated, horizontal cylinders within a rectangular enclosure

Energy Technology Data Exchange (ETDEWEB)

Triplett, C.E.

1996-12-01

This thesis presents the results of an experimental investigation of natural convection heat transfer in a staggered array of heated cylinders, oriented horizontally within a rectangular enclosure. The main purpose of this research was to extend the knowledge of heat transfer within enclosed bundles of spent nuclear fuel rods sealed within a shipping or storage container. This research extends Canaan`s investigation of an aligned array of heated cylinders that thermally simulated a boiling water reactor (BWR) spent fuel assembly sealed within a shipping or storage cask. The results are presented in terms of piecewise Nusselt-Rayleigh number correlations of the form Nu = C(Ra){sup n}, where C and n are constants. Correlations are presented both for individual rods within the array and for the array as a whole. The correlations are based only on the convective component of the heat transfer. The radiative component was calculated with a finite-element code that used measured surface temperatures, rod array geometry, and measured surface emissivities as inputs. The correlation results are compared to Canaan`s aligned array results and to other studies of natural convection in horizontal tube arrays.

Natural convection heat transfer for a staggered array of heated, horizontal cylinders within a rectangular enclosure

International Nuclear Information System (INIS)

Triplett, C.E.

1996-12-01

This thesis presents the results of an experimental investigation of natural convection heat transfer in a staggered array of heated cylinders, oriented horizontally within a rectangular enclosure. The main purpose of this research was to extend the knowledge of heat transfer within enclosed bundles of spent nuclear fuel rods sealed within a shipping or storage container. This research extends Canaan's investigation of an aligned array of heated cylinders that thermally simulated a boiling water reactor (BWR) spent fuel assembly sealed within a shipping or storage cask. The results are presented in terms of piecewise Nusselt-Rayleigh number correlations of the form Nu = C(Ra) n , where C and n are constants. Correlations are presented both for individual rods within the array and for the array as a whole. The correlations are based only on the convective component of the heat transfer. The radiative component was calculated with a finite-element code that used measured surface temperatures, rod array geometry, and measured surface emissivities as inputs. The correlation results are compared to Canaan's aligned array results and to other studies of natural convection in horizontal tube arrays

Critical power experiment with a tight-lattice 37-rod bundle

International Nuclear Information System (INIS)

Kureta, Masatoshi; Tamai, Hidesada; Ohnuki, Akira; Sato, Takashi; Liu, Wei; Akimoto, Hajime

2006-01-01

Since most of critical power or CHF data have been collected in tube, annulus, or BWR geometries under BWR flow conditions, critical power data for highly tight and triangular lattice bundles under low mass velocity are indispensable for thermal-hydraulic design of Reduced-Moderation Water Reactor. Large-scale thermal-hydraulic experiments which use a basic 37-rod bundle test section (rod diameter: 13.0 mm, gap width between rods: 1.3 mm) were therefore carried out in this study within range of 2-9 MPa in pressure and 150-1,000 kg/(m 2 ·s) in mass velocity. Fundamental characteristics of boiling transition were investigated through effects of flow parameter on critical power and those of rod number. It was confirmed that the fundamental characteristics in 37-rod bundle are similar to those in 7-rod bundle and in case of the BWR geometry. The results of the transverse non-uniform power distribution test and subchannel analysis suggest that the critical power becomes higher when the transverse local quality distribution closes to uniform. (author)

Gasification in a revolving tube

International Nuclear Information System (INIS)

Speicher, R.F.

1981-01-01

The concept of a method for allothermal coal gasification is to refine raw lignite from the Rhine area to high-quality synthesis gas or reduction gas without extracting the water utilizing nuclear process heat in a heated revolving bundle of tubes. Computational models are described for the macroscopic course of events in parallel flow gasification. In the design of the test plant, the principle of drag-in and transport of the tube drier was applied. (DG) [de

Unusual behavior of growing pollen tubes in the ovary of plum culture (Prunus domestica L.

Directory of Open Access Journals (Sweden)

Đorđević Milena

2010-01-01

Full Text Available Unusual behavior of growing pollen tubes in different combinations of pollination was observed in the ovary of the plum (Prunus domestica L. cv 'Čačanska Lepotica'. It primarily refers to several issues, i.e. the curling up of pollen tubes within the micropyle, the growth of two pollen tubes into the nucellus of an ovule, the occurrence of a bundle above the nucellar cap and fluorescence of the part of the embryo sac containing the egg apparatus. Upon the growth of pollen tubes into the nucellus of the ovule, subsequently penetrating pollen tubes form a bundle either above the micropyle entrance or above the nucellus. Branching and bending of pollen tubes by 180o upon their growth into the micropyle was also observed.

Development of the tube bundle structure for fluid-structure interaction analysis model - Intermediate Report -

International Nuclear Information System (INIS)

Yoon, Kyung Ho; Kim, Jae Yong; Lee, Kang Hee; Lee, Young Ho; Kim, Hyung Kyu

2009-07-01

Tube bundle structures within a Boiler or heat exchanger are laid the fluid-structure, thermal-structure and fluid-thermal-structure coupled boundary condition. In these complicated boundary conditions, Fluid-structure interaction (FSI) occurs when fluid flow causes deformation of the structure. This deformation, in turn, changes the boundary conditions for the fluid flow. The structural analysis have been executed as follows. First of all, divide the fluid and structural analysis discipline, and then independently analyzed each other. However, the fluid dynamic force effect the behavior of the structure, and the vibration amplitude of the structure to fluid. FSI analysis model was separately created fluid and structure model, and then defined the fsi boundary condition, and simultaneously analyzed in one domain. The analysis results were compared with those of the experimental method for validating the analysis model. Flow-induced vibration test was executed with single rod configuration. The vibration amplitudes of a fuel rod were measured by the laser vibro-meter system in x and y-direction. The analyses results were not closely with the test data, but the trend was very similar with the test result. In fsi coupled analysis case, the turbulent model was very important with the reliability of the accuracy of the analysis model. Therefore, the analysis model will be needed to further study

Determination of a cross-sectional void fraction in a tube bundle using a single beam gamma densitometer

International Nuclear Information System (INIS)

Guichard, J.; Mezoul, B.; Peturaud, P.; Thomas, B.

1991-06-01

In order to qualify 3-dimensional two-phase flow computer codes modelling average flows in tube bundles, cross-section average void fractions must be measured over sub-channels. On the VATICAN mockup, such void fractions(integrated on the mockup thickness) are determined using a single (narrow) beam gamma densitometer. But to avoid a refined exploration of each measurement mesh, for each test, empirical calibration curves have been developed in a regular mesh of the mockup, in axial flow conditions. These calibration curves, which evaluate the sought cross-sectional value as a function of a chordal void fraction (right in the inter-rod gap) depend only on heat flux density and pressure. The data are consistent with the ARMAND-MASSENA and LELLOUCHE-ZOLOTAR slip correlations, and they are fitted by 3rd degree polynomials, for each heat flux density investigated, with a good accuracy. Unfortunately, preliminary testing and analysis indicate that the use of these calibration curves in subcooled boiling and transverse mixing zones might result in significant uncertainties and errors

Reinforcement of single-walled carbon nanotube bundles by intertube bridging

Science.gov (United States)

Kis, A.; Csányi, G.; Salvetat, J.-P.; Lee, Thien-Nga; Couteau, E.; Kulik, A. J.; Benoit, W.; Brugger, J.; Forró, L.

2004-03-01

During their production, single-walled carbon nanotubes form bundles. Owing to the weak van der Waals interaction that holds them together in the bundle, the tubes can easily slide on each other, resulting in a shear modulus comparable to that of graphite. This low shear modulus is also a major obstacle in the fabrication of macroscopic fibres composed of carbon nanotubes. Here, we have introduced stable links between neighbouring carbon nanotubes within bundles, using moderate electron-beam irradiation inside a transmission electron microscope. Concurrent measurements of the mechanical properties using an atomic force microscope show a 30-fold increase of the bending modulus, due to the formation of stable crosslinks that effectively eliminate sliding between the nanotubes. Crosslinks were modelled using first-principles calculations, showing that interstitial carbon atoms formed during irradiation in addition to carboxyl groups, can independently lead to bridge formation between neighbouring nanotubes.

Particle deposition modeling in the secondary side of a steam generator bundle model

Energy Technology Data Exchange (ETDEWEB)

Mukin, Roman, E-mail: roman.mukin@psi.ch; Dehbi, Abdel, E-mail: abdel.dehbi@psi.ch

2016-04-01

A steam generator (SG) tube rupture (SGTR) model is studied in this paper. This model based on a experimental facility called Aerosol Trapping In a Steam Generator (ARTIST), which is a model of a scaled steam generator tube bundle consisting of 270 tubes and a guillotine tube to address aerosol deposition phenomena on two different scales: near the tube break, where the gas velocities and turbulence are very intensive, and far away from the break, where the flow velocities are three orders of magnitude lower. Owing to complexity of the flow, 3D simulations with highly resolved computational mesh near the break were done. First, the flow inside an isolated tube with a guillotine tube break has been studied in the framework of Reynolds Averaged Navier Stokes (RANS) approach. The next part is devoted to the simulation of an inclined gas jet entering the SG tube bundle via the guillotine tube breach with more advanced CFD tools. In particular, Detached Eddy Simulation (DES) and RANS are applied to tackle the wide range of flow scales. Flow field velocity comparison showed that DES results are reproducing wavy structure of the flow field in far field from the break observed in experiment. Particle transport and deposition is modelled by Lagrangian continuous random walk (CRW) model, which has been developed and validated previously. It is found that the DES combined with the CRW to supply fluctuating velocity components predicts deposition rates that are generally within the scatter of the measured data. Monodisperse, spherical SiO{sub 2} particles with AMMD = 1.4 μm were used as aerosol particles in simulations. To be economically feasible, the computations were made with the open source CFD code OpenFOAM. Comparison of the calculated flow with the experimental axial velocity distribution data at different vertical levels has been performed.

Analysis of coolant flow in central tube of WWER-440 fuel assemblies

International Nuclear Information System (INIS)

Zsiros, G.; Toth, S.; Attila Aszodi, A.

2011-01-01

Three dimensional computational fluid dynamics model has been built to investigate the coolant flow in the central tube of the WWER-440 fuel assemblies. The model was verified based on measured data of the Kurchatov Institute. With the model calculations were performed for two fuel assemblies used in PAKS NPP. One of them has symmetrical and another has inclined pin power profile. Ratios of the outlet mass fluxes of the central tube to the inlet mass fluxes of the rod bundle were determined. Heat up ratios of the tube and rod bundle flows were calculated too. Sensitivity of the results on the assembly power distribution, inlet temperature and mass flow rate was investigated. The results of these simulations can be used as boundary conditions of central tube in studies of coolant mixing in fuel assembly heads. (Authors)

Staking solutions to tube vibration problems (developed by Technos et Compagnie - FRANCE)

International Nuclear Information System (INIS)

Hewitt, E.W.; Bizard, A.; Horn, M.J.

1989-01-01

Electric generating plant steam surface condensers have been prone to vibration induced tube failures. One common and effective method for stopping this vibration has been to insert stakes into the bundle to provide additional support. Stakes have been fabricated of a variety of rigid and semi-rigid materials of fixed dimensions. Installation difficulties and problems of incomplete tube support have been associated with this approach. New developments in the application of plastic technology has offered another approach. Stakes made of plastic tubes which are flattened, by evacuation, at the time of manufacture may now be easily inserted into the tube bundle. After insertion, the vacuum is released and the memory of the plastic causes the stakes to expand and assume their original form. The spring force of the plastic cradles the adjacent condenser tubes and stops the vibration. Developed for Electricite de France (EDF), the stakes are currently installed in 19 units of the French utility system, and two units in the United States

Effect of superficial velocity on vaporization pressure drop with propane in horizontal circular tube

Science.gov (United States)

Novianto, S.; Pamitran, A. S.; Nasruddin, Alhamid, M. I.

2016-06-01

Due to its friendly effect on the environment, natural refrigerants could be the best alternative refrigerant to replace conventional refrigerants. The present study was devoted to the effect of superficial velocity on vaporization pressure drop with propane in a horizontal circular tube with an inner diameter of 7.6 mm. The experiments were conditioned with 4 to 10 °C for saturation temperature, 9 to 20 kW/m2 for heat flux, and 250 to 380 kg/m2s for mass flux. It is shown here that increased heat flux may result in increasing vapor superficial velocity, and then increasing pressure drop. The present experimental results were evaluated with some existing correlations of pressure drop. The best prediction was evaluated by Lockhart-Martinelli (1949) with MARD 25.7%. In order to observe the experimental flow pattern, the present results were also mapped on the Wang flow pattern map.

Bundles to prevent ventilator-associated pneumonia: how valuable are they?

Science.gov (United States)

Wip, Charity; Napolitano, Lena

2009-04-01

To review the value of care bundles to prevent ventilator-associated pneumonia (VAP). The Ventilator Bundle contains four components, elevation of the head of the bed to 30-45 degrees, daily 'sedation vacation' and daily assessment of readiness to extubate, peptic ulcer disease prophylaxis, and deep venous thrombosis prophylaxis, aimed to improve outcome in mechanically ventilated patients, but not all are associated with VAP prevention. Daily spontaneous awakening and breathing trials are associated with early liberation from mechanical ventilation and VAP reduction. Although a small prospective, randomized clinical study documented that the semirecumbent position was associated with a significant reduction in VAP, more recent studies have documented that the semirecumbent position is difficult to maintain in mechanically ventilated patients and may not impact VAP reduction. Prophylaxis for peptic ulcer disease and deep venous thrombosis do not directly impact VAP reduction. Other methods to reduce VAP, such as oral care and hygiene, chlorhexidine in the posterior pharynx, and specialized endotracheal tubes (continuous aspiration of subglottic secretions, silver-coated), should be considered for inclusion in a revised Ventilator Bundle more specifically aimed at VAP prevention. The Ventilator Bundle is an effective method to reduce VAP rates in ICUs. The ventilator bundle should be modified and expanded to include specific processes of care that have been definitively demonstrated to be effective in VAP reduction or a specific VAP bundle created to focus on VAP prevention.

Application of the Critical Heat Flux Look-Up Table to Large Diameter Tubes

Directory of Open Access Journals (Sweden)

M. El Nakla

2013-01-01

Full Text Available The critical heat flux look-up table was applied to a large diameter tube, namely 67 mm inside diameter tube, to predict the occurrence of the phenomenon for both vertical and horizontal uniformly heated tubes. Water was considered as coolant. For the vertical tube, a diameter correction factor was directly applied to the 1995 critical heat flux look-up table. To predict the occurrence of critical heat flux in horizontal tube, an extra correction factor to account for flow stratification was applied. Both derived tables were used to predict the effect of high heat flux and tube blockage on critical heat flux occurrence in boiler tubes. Moreover, the horizontal tube look-up table was used to predict the safety limits of the operation of boiler for 50% allowable heat flux.

Wastage of Steam Generator Tubes by Sodium-Water Reaction

International Nuclear Information System (INIS)

Jeong, Ji Young; Kim, Jong Man; Kim, Tae Joon; Choi, Jong Hyeun; Kim, Byung Ho; Lee, Yong Bum; Park, Nam Cook

2010-01-01

The Korea Advanced LIquid MEtal Reactor (KALIMER) steam generator is a helical coil, vertically oriented, shell-and-tube type heat exchanger with fixed tube-sheet. The conceptual design and outline drawing of the steam generator are shown in Figure 1. Flow is counter-current, with sodium on the shell side and water/steam on the tube side. Sodium flow enters the steam generator through the upper inlet nozzles and then flows down through the tube bundle. Feedwater enters the steam generator through the feedwater nozzles at the bottom of steam generator. Therefore, if there is a hole or a crack in a heat transfer tube, a leakage of water/steam into the sodium may occur, resulting in a sodium-water reaction. When such a leak occurs, so-called 'wastage' is the result which may cause damage to or a failure of the adjacent tubes. If a steam generator is operated for some time in this condition, it is possible that it might create an intermediate leak state which would then give rise to the problems of a multi-target wastage in a very short time. Therefore, it is very important to predict these phenomena quantitatively from the view of designing a steam generator and its leak detection systems. For this, multi-target wastage tests for modified 9Cr-1Mo steel tube bundle by intermediate leaks are being prepared

Experimental investigation of flooding in air-water counter-current flow with a vertical adiabatic multi-rod bundle

International Nuclear Information System (INIS)

Chung, Bub Dong; Kim, Hho Jung; Cha, Jong Hee; Cho, Sung Jae; Chun, Moon Hyun

1991-01-01

The process of flooding phenomenon in a vertical adiabatic 3 x 3 tube bundle flow channel has been studied experimentally. A series of tests was performed, using three types of tube bundle differing only in the number of spacer grids attached, to investigate the effects of spacer grids and multi-flow channel interactions on the air-water counter-current flow limitations. Experimentally determined flooding points at various water film Reynolds numbers for three different test sections are presented in graphical form and compared with entrainment criterion for co-current flow and instability criteria. In addition, empirical flooding correlations of the Kutateladze type are obtained for each type of test section using liquid penetration data

Detection of muon bundles at large zenith angles

International Nuclear Information System (INIS)

Aynutdinov, V.M.; Castellina, A.; Chernov, D.V.; Ezubchenko, A.A.; Fulgione, W.; Kindin, V.V.; Kokoulin, R.P.; Kompaniets, K.G.; Konovalov, A.A.; Mannocchi, G.; Petrukhin, A.A.; Rodin, Yu. N.; Saavedra, O.; Shutenko, V.V.; Trinchero, G.; Vernetto, S.; Vonsovsky, N.N.; Yanson, E.E.; Yashin, I.I.

1999-01-01

The large area coordinate detector (DECOR) represents a multilayer system of plastic streamer tube chamber modules surrounding the Cherenkov water calorimeter NEVOD. Experimental data collected during the test expositions of first DECOR supermodules (8 vertical planes with 8.4 m 2 working area) have been analysed, angular and spatial reconstruction accuracies have been estimated. The procedure of the selection of events corresponding to detection of parallel tracks (muon bundles originated in the atmosphere) is discussed

Tube Length and Water Flow

Directory of Open Access Journals (Sweden)

Ben Ruktantichoke

2011-06-01

Full Text Available In this study water flowed through a straight horizontal plastic tube placed at the bottom of a large tank of water. The effect of changing the length of tubing on the velocity of flow was investigated. It was found that the Hagen-Poiseuille Equation is valid when the effect of water entering the tube is accounted for.

Fluid-elastic force measurements acting on a tube bundle in two-phase cross flow

International Nuclear Information System (INIS)

Inada, Fumio; Kawamura, Koji; Yasuo, Akira

1996-01-01

Fluid-elastic force acting on a square tube bundle of P/D = 1.47 in air-water two-phase cross flow was measured to investigate the characteristics and to clarify whether the fluid elastic vibration characteristics could be expressed using two-phase mixture characteristics. Measured fluid elastic forces were separated into fluid-elastic force coefficients such as added mass, added stiffness, and added damping coefficient. The added damping coefficient was separated into a two-phase damping and a flow-dependent component as in previous research (Carlucci, 1981 and 1983; Pettigrew, 1994). These coefficients were nondimensionalized with two-phase mixture characteristics such as void fraction, mixture density and mixture velocity, which were obtained using the drift-flux model with consideration given to the model. The result was compared with the result obtained with the homogeneous model. It was found that fluid-elastic force coefficients could be expressed with two-phase flow mixture characteristics very well in the experimental result, and that better result can be derived using the slip model as compared to the homogeneous model. Added two-phase flow, which could be expressed as a function of void fraction, where two-phase damping was nondimensionalized with the relative velocity between the gas and liquid phases used as a reference velocity. Using these, the added stiffness coefficient and flow-dependent component of damping could be expressed very well as a function of nondimensional mixture velocity

Pressure loss in two-phase flow through a microchannel rod bundle

International Nuclear Information System (INIS)

Smith, A.C.; Hamm, L.L.; Qureshi, Z.; Steeper, T.J.

1998-01-01

The purpose of the microchannel rod bundle two-phase flow test described here was to provide data for benchmarking safety analyses for the accelerator production of tritium (APT). The objective was to obtain pressure loss data for a typical accelerator target rod bundle over a wide range of two-phase flow conditions. The test rod bundle assembly was fabricated for single-phase pressure drop tests conducted at Los Alamos National Laboratory (LANL) and subsequently used for the two-phase flow testing described here. The results for a typical case are given. These results fall generally in the slug flow regime for the horizontal flow results of Fukano and Kariyasaki for a 1.0-mm circular channel. Fukano and Kariyasaki found that surface tension effects were dominant in the 1-mm channel and report no churn regime. The results were also compared with the flow regime maps given by Triplett et al. for flow in discrete microchannels. Triplett employed both circular and trapezoidal channels, the latter to approximate the rod bundle interstitial flow channel shape. It was found that the rod bundle flow fell across the slug-to-churn flow regime transition reported by Triplett. This is consistent with the expectation that cross flow among channels would result in turbulent mixing and would suppress the formation of large discrete bubbles

Study on drop pressure and flow distribution of double-tube heat exchanger

International Nuclear Information System (INIS)

Liu Junqiang; Chen Minghui; Hu Yumin; Li Rizhu; Kong Dechun; Zhang Weijie

2007-01-01

The parallel connection channel pressure drop characters of the double-tube bundle heat exchange were experimentally investigated in this paper in order to find out how the flow of the heat exchanger is distributed and then to optimize the structure of heat exchanger according to the flow distribution. A double-tube bundle heat exchanger was built according to the similarity criteria. The experiment system was also built to test the optimization of the heat exchanger. The experiment results reveal that the calculating model is reliable and decreasing pipe space to optimize the heat exchanger is reasonable. (authors)

Condensation of refrigerants in horizontal, spirally grooved microfin tubes: Numerical analysis of heat transfer in the annular flow regime

Energy Technology Data Exchange (ETDEWEB)

Nozu, S; Honda, H

2000-02-01

A method is presented for estimating the condensation heat transfer coefficient in a horizontal, spirally grooved microfin tube. Based on the flow observation study performed by the authors, a laminar film condensation model in the annular flow regime is proposed. The model assumes that all the condensate flow occurs through the grooves. The condensate film is segmented into thin and thick film regions. In the thin film region formed on the fin surface, the condensate is assumed to be drained by the combined surface tension and vapor shear forces. In the thick film region formed in the groove, on the other hand, the condensate is assumed to be driven by the vapor shear force. The present and previous local heat transfer data including four fluids (CFC11, HCFC22, HCFC123, and HFCl34a) and three microfin tubes are found to agree with the present predictions to a mean absolute deviation of 15.1%.

Enhanced field emission properties of carbon nanotube bundles confined in SiO2 pits

Science.gov (United States)

Lim, Yu Dian; Grapov, Dmitry; Hu, Liangxing; Kong, Qinyu; Tay, Beng Kang; Labunov, Vladimir; Miao, Jianmin; Coquet, Philippe; Aditya, Sheel

2018-02-01

It has been widely reported that carbon nanotubes (CNTs) exhibit superior field emission (FE) properties due to their high aspect ratios and unique structural properties. Among the various types of CNTs, random growth CNTs exhibit promising FE properties due to their reduced inter-tube screening effect. However, growing random growth CNTs on individual catalyst islands often results in spread out CNT bundles, which reduces overall field enhancement. In this study, significant improvement in FE properties in CNT bundles is demonstrated by confining them in microfabricated SiO2 pits. Growing CNT bundles in narrow (0.5 μm diameter and 2 μm height) SiO2 pits achieves FE current density of 1-1.4 A cm-2, which is much higher than for freestanding CNT bundles (76.9 mA cm-2). From the Fowler Nordheim plots, confined CNT bundles show a higher field enhancement factor. This improvement can be attributed to the reduced bundle diameter by SiO2 pit confinement, which yields bundles with higher aspect ratios. Combining the obtained outcomes, it can be conclusively summarized that confining CNTs in SiO2 pits yields higher FE current density due to the higher field enhancement of confined CNTs.

A critical heat flux approach for square rod bundles using the 1995 Groeneveld CHF table and bundle data of heat transfer research facility

International Nuclear Information System (INIS)

Lee, M.

2000-01-01

The critical heat flux (CHF) approach using CHF look-up tables has become a widely accepted CHF prediction technique. In these approaches, the CHF tables are developed based mostly on the data bank for flow in circular tubes. A set of correction factors was proposed by Groeneveld et al. [Groeneveld, D.C., Cheng, S.C., Doan, T. (1986)] to extend the application of the CHF table to other flow situations including flow in rod bundles. The proposed correction factors are based on a limited amount of data not specified in the original paper. The CHF approach of Groeneveld and co-workers is extensively used in the thermal hydraulic analysis of nuclear reactors. In 1996, Groeneveld et al. proposed a new CHF table to predict CHF in circular tubes [Groeneveld, D.C., et al., 1996. The 1995 look-up table for Critical Heat Flux. Nucl. Eng. Des. 163(1), 23]. In the present study, a set of correction factors is developed to extend the applicability of the new CHF table to flow in rod bundles of square array. The correction factors are developed by minimizing the statistical parameters of the ratio of the measured and predicted bundle CHF data from the Heat Transfer Research Facility. The proposed correction factors include: the hydraulic diameter factor (K hy ), the bundle factor (K bf ), the heated length factor (K hl ), the grid spacer factor (K sp ), the axial flux distribution factors (K nu ), the cold wall factor (K cw ) and the radial power distribution factor (K rp ). The value of constants in these correction factors is different when the heat balance method (HBM) and direct substitution method (DSM) are adopted to predict the experimental results of HTRF. With the 1995 Groeneveld CHF Table and the proposed correction factors, the average relative error is 0.1 and 0.0% for HBM and DSM, respectively, and the root mean square (RMS) error is 31.7% in DSM and 17.7% in HBM for 9852 square array data points of HTRF. (orig.)

Measurement and analysis of the re-wetting front velocity during quench cooling of hot horizontal tubes

Energy Technology Data Exchange (ETDEWEB)

Takrouri, Kifah, E-mail: takroukj@mcmaster.ca [Department of Engineering Physics, McMaster University, 1280 Main Street West, Hamilton, Ontario L8S 4L7 (Canada); Luxat, John, E-mail: luxatj@mcmaster.ca [Department of Engineering Physics, McMaster University, 1280 Main Street West, Hamilton, Ontario L8S 4L7 (Canada); Hamed, Mohamed [Thermal Processing Laboratory (TPL), Department of Mechanical Engineering, McMaster University, 1280 Main Street West, Hamilton, Ontario L8S 4L7 (Canada)

2017-01-15

Highlights: • Two phase flow & re-wetting front velocity were studied for quench of hot tubes. • The velocity decreased as temperature difference between tube and coolant decreased. • Increasing surface curvature was found to decrease the re-wetting front velocity. • Increasing tube thermal conductivity decreased the velocity. • Correlations were developed to predict the front velocity. - Abstract: When a liquid is put into contact with a hot dry surface, there exists a maximum temperature called the re-wetting temperature below which the liquid is in actual contact with the surface. Re-wetting occurs after destabilization of a vapor film that exists between the hot surface and the liquid. If re-wetting is established at a location on the hot surface, a wet patch appears at that location and starts to spread to cover and cool the entire surface. The outer edge of the wet patch is called the re-wetting front and can proceed only if the surface ahead of it cools down to the re-wetting temperature. Study of re-wetting heat transfer is very important in nuclear reactor safety for limiting the extent of core damage during the early stages of severe accidents after loss of coolant accidents LOCA and is essential for predicting the rate at which the coolant cools an overheated core. One of the important parameters in re-wetting cooling is the velocity at which the re-wetting front moves on the surface. In this study, experimental tests were carried out to investigate the re-wetting front velocity on hot horizontal cylindrical tubes being cooled by a vertical rectangular water multi-jet system. Effects of initial surface temperature in the range 400–740 °C, water subcooling in the range 15–80 °C and jet velocity in the range 0.17–1.43 m/s on the re-wetting front velocity were investigated. The two-phase flow behavior was observed by using a high-speed camera. The re-wetting front velocity was found to increase by increasing water subcooling, decreasing

The use of titanium for condenser tube bundles

International Nuclear Information System (INIS)

Dobrowitch, Nicolas

2003-01-01

In a power plant, the condenser is a strategic heat exchanger with regards to the efficiency of the steam turbine and its reliability guarantees the performance and continuous operation of the plant. Until the early 1980s, copper alloys were routinely used in condenser tubes, thanks to their high heat transfer rates. Yet, numerous problems arose from the use of this material, such as stress corrosion cracking, ammoniacal corrosion, fouling, erosion, dezincification, abrasion, erosion-corrosion, etc. and lately the problem of the inadequacy of copper with nuclear steam generators. The trend was then to consider new tube materials, such as stainless steel and titanium, at first for particular operating conditions and now for most of the projects, with several objectives, such as: improving reliability (titanium in particular can bring major improvements including higher water velocities promoting better heat transfer coefficients, and excellent resistance to abrasion, erosion and corrosion thereby improving resistance to fouling); finding more cost-effective solutions. The first investment is higher but money is saved on maintenance costs and on time reliability of the material. (orig.)

Detection of muon bundles at large zenith angles

Energy Technology Data Exchange (ETDEWEB)

Aynutdinov, V.M.; Castellina, A.; Chernov, D.V.; Ezubchenko, A.A.; Fulgione, W.; Kindin, V.V.; Kokoulin, R.P.; Kompaniets, K.G.; Konovalov, A.A.; Mannocchi, G.; Petrukhin, A.A.; Rodin, Yu. N.; Saavedra, O.; Shutenko, V.V.; Trinchero, G.; Vernetto, S.; Vonsovsky, N.N.; Yanson, E.E.; Yashin, I.I

1999-03-01

The large area coordinate detector (DECOR) represents a multilayer system of plastic streamer tube chamber modules surrounding the Cherenkov water calorimeter NEVOD. Experimental data collected during the test expositions of first DECOR supermodules (8 vertical planes with 8.4 m{sup 2} working area) have been analysed, angular and spatial reconstruction accuracies have been estimated. The procedure of the selection of events corresponding to detection of parallel tracks (muon bundles originated in the atmosphere) is discussed.

The Preliminary Study for Numerical Computation of 37 Rod Bundle in CANDU Reactor

International Nuclear Information System (INIS)

Jeon, Yu Mi; Bae, Jun Ho; Park, Joo Hwan

2010-01-01

A typical CANDU 6 fuel bundle consists of 37 fuel rods supported by two endplates and separated by spacer pads at various locations. In addition, the bearing pads are brazed to each outer fuel rod with the aim of reducing the contact area between the fuel bundle and the pressure tube. Although the recent progress of CFD methods has provided opportunities for computing the thermal-hydraulic phenomena inside of a fuel channel, it is yet impossible to reflect the detailed shape of rod bundle on the numerical computation due to a lot of computing mesh and memory capacity. Hence, the previous studies conducted a numerical computation for smooth channels without considering spacers, bearing pads. But, it is well known that these components are an important factor to predict the pressure drop and heat transfer rate in a channel. In this study, the new computational method is proposed to solve the complex geometry such as a fuel rod bundle. In front of applying the method to the problem of 37 rod bundle, the validity and the accuracy of the method are tested by applying the method to the simple geometry. Based on the present result, the calculation for the fully shaped 37-rod bundle is scheduled for the future works

Spatial Variation of Hydrodynamic Mass Coefficients for Tube Bundle in a Cylindrical Shell

Energy Technology Data Exchange (ETDEWEB)

Yang, Keum Hee; Ryu, Ki Wahn [Chonbuk National University, Jeonju (Korea, Republic of); Park, Chi Yong [KEPCO Research Institute, Daejeon (Korea, Republic of)

2011-05-15

Wear of the steam generator (SG) tubes affects the performance of nuclear power plants. Generally, the problem is caused by excessive flow-induced vibration (FIV). In analyzing the FIV, many researchers have used a uniform added mass coefficient for all of the SG tubes. However, the outermost SG tubes have more structural problems than inside tubes. The purpose of this study is to find out the added mass coefficients of each tube in a cylindrical shell

Influence of oil on flow condensation heat transfer of R410A inside 4.18 mm and 1.6 mm inner diameter horizontal smooth tubes

Energy Technology Data Exchange (ETDEWEB)

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

2010-01-15

The influence of oil on condensation heat transfer of R410A inside 4.18 mm and 1.6 mm inner diameter horizontal smooth tubes is investigated experimentally. The experimental condensing temperature is 40 C, and nominal oil concentration range is from 0% to 5%. The test results indicate that the presence of oil deteriorates the heat transfer, and the deterioration effect becomes obvious with the increase of oil concentration. At oil concentration of 5%, the heat transfer coefficient decreases by maximum 24.9% and 28.5% for 4.18 mm and 1.6 mm tubes, respectively. A new correlation for heat transfer coefficients of R410A-oil mixture flow condensation inside smooth tubes is proposed, which agrees with all the experimental data within a deviation of -30% {proportional_to} +20%. (author)

Numerical Study on the Heat Transfer of Carbon Dioxide in Horizontal Straight Tubes under Supercritical Pressure

Science.gov (United States)

Yang, Mei

2016-01-01

Cooling heat transfer of supercritical CO2 in horizontal straight tubes with wall is numerically investigated by using FLUENT. The results show that almost all models are able to present the trend of heat transfer qualitatively, and the stand k−ε with enhanced wall treatment model shows the best agreement with the experimental data, followed by LB low Re turbulence model. Then further studies are discussed on velocity, temperature and turbulence distributions. The parameters which are defined as the criterion of buoyancy effect on convection heat transfer are introduced to judge the condition of the fluid. The relationships among the inlet temperature, outlet temperature, the mass flow rate, the heat flux and the diameter are discussed and the difference between the cooling and heating of CO2 are compared. PMID:27458729

Numerical Study on the Heat Transfer of Carbon Dioxide in Horizontal Straight Tubes under Supercritical Pressure.

Directory of Open Access Journals (Sweden)

Mei Yang

Full Text Available Cooling heat transfer of supercritical CO2 in horizontal straight tubes with wall is numerically investigated by using FLUENT. The results show that almost all models are able to present the trend of heat transfer qualitatively, and the stand k-ε with enhanced wall treatment model shows the best agreement with the experimental data, followed by LB low Re turbulence model. Then further studies are discussed on velocity, temperature and turbulence distributions. The parameters which are defined as the criterion of buoyancy effect on convection heat transfer are introduced to judge the condition of the fluid. The relationships among the inlet temperature, outlet temperature, the mass flow rate, the heat flux and the diameter are discussed and the difference between the cooling and heating of CO2 are compared.

FIMBRIN1 is involved in lily pollen tube growth by stabilizing the actin fringe.

Science.gov (United States)

Su, Hui; Zhu, Jinsheng; Cai, Chao; Pei, Weike; Wang, Jiaojiao; Dong, Huaijian; Ren, Haiyun

2012-11-01

An actin fringe structure in the subapex plays an important role in pollen tube tip growth. However, the precise mechanism by which the actin fringe is generated and maintained remains largely unknown. Here, we cloned a 2606-bp full-length cDNA encoding a deduced 77-kD fimbrin-like protein from lily (Lilium longiflorum), named FIMBRIN1 (FIM1). Ll-FIM1 was preferentially expressed in pollen and concentrated at actin fringe in the subapical region, as well as in longitudinal actin-filament bundles in the shank of pollen tubes. Microinjection of Ll-FIM1 antibody into lily pollen tubes inhibited tip growth and disrupted the actin fringe. Furthermore, we verified the function of Ll-FIM1 in the fim5 mutant of its closest relative, Arabidopsis thaliana. Pollen tubes of fim5 mutants grew with a larger diameter in early stages but could recover into normal forms in later stages, despite significantly slower growth rates. The actin fringe of the fim5 mutants, however, was impaired during both early and late stages. Impressively, stable expression of fim5pro:GFP:Ll-FIM1 rescued the actin fringe and the growth rate of Arabidopsis fim5 pollen tubes. In vitro biochemical analysis showed that Ll-FIM1 could bundle actin filaments. Thus, our study has identified a fimbrin that may stabilize the actin fringe by cross-linking actin filaments into bundles, which is important for proper tip growth of lily pollen tubes.

Higher order jet prolongations type gauge natural bundles over vector bundles

Directory of Open Access Journals (Sweden)

Jan Kurek

2004-05-01

Full Text Available Let $rgeq 3$ and $mgeq 2$ be natural numbers and $E$ be a vector bundle with $m$-dimensional basis. We find all gauge natural bundles ``similar" to the $r$-jet prolongation bundle $J^rE$ of $E$. We also find all gauge natural bundles ``similar" to the vector $r$-tangent bundle $(J^r_{fl}(E,R_0^*$ of $E$.

Characterization of flaws in a tube bundle mock-up for reliability studies

International Nuclear Information System (INIS)

Kupperman, D.S.; Bakhtiari, S.

1997-01-01

As part of an assessment of in-service inspection of steam generator tubes, the authors will assemble a steam generator mock-up for round robin studies and use as a test bed in evaluating emerging technologies. Progress is reported on the characterization of flaws that will be part of the mock-up. Eddy current and ultrasonic techniques are being evaluated as a means to characterize the flaws in the mock-up tubes before final assembly. Twenty Inconel 600 tubes with laboratory-grown cracks, typical of those to be used in the mock-up, were provided by Pacific Northwest National Laboratory for laboratory testing. After the tubes were inspected with eddy current and ultrasonic techniques, they were destructively analyzed to establish the actual depths, lengths, and profiles of the cracks. The analysis of the results will allow the best techniques to be used for characterizing the flaws in the mock-up tubes

Characterization of flaws in a tube bundle mock-up for reliability studies

International Nuclear Information System (INIS)

Kupperman, D.S.; Bakhtiari, S.

1996-10-01

As part of an assessment of in-service inspection of steam generator tubes, the authors will assemble a steam generator mock-up for round robin studies and use as a test bed in evaluating emerging technologies. Progress is reported on the characterization of flaws that will be part of the mock-up. Eddy current and ultrasonic techniques are being evaluated as a means to characterize the flaws in the mock-up tubes before final assembly. Twenty Inconel 600 tubes with laboratory-grown cracks, typical of those to be used in the mock-up, were provided by Pacific Northwest National Laboratory for laboratory testing. After the tubes were inspected with eddy current and ultrasonic techniques, they were destructively analyzed to establish the actual depths, lengths, and profiles of the cracks. The analysis of the results will allow the best techniques to be used for characterizing the flaws in the mock-up tubes

Effect of tube-support interaction on the dynamic responses of heat exchanger tubes

International Nuclear Information System (INIS)

Shin, Y.S.; Jendrzejczyk, J.A.; Wambsganss, M.W.

1977-01-01

Operating heat exchangers have experienced tube damages due to excessive flow-induced vibration. The relatively small inherent tube-to-baffle hole clearances associated with manufacturing tolerances in heat exchangers affect the tube vibrational characteristics. In attempting a theoretical analysis, questions arise as to the effects of tube-baffle impacting on dynamic responses. Experiments were performed to determine the effects of tube-baffle impacting in vertical/horizontal tube orientation, and in air/water medium on the vibrational characteristics (resonant frequencies, mode shapes, and damping) and displacement response amplitudes of a seven-span tube model. The tube and support conditions were prototypic, and overall length approximately one-third that of a straight tube segment of the steam generator designed for the CRBR. The test results were compared with the analytical results based on the multispan beam with ''knife-edge'' supports

Pressure drop-flow rate curves for single-phase steam in Combustion Engineering type steam generator U-tubes during severe accidents

Energy Technology Data Exchange (ETDEWEB)

Fynan, Douglas A.; Ahn, Kwang-Il, E-mail: kiahn@kaeri.re.kr

2016-12-15

Highlights: • Pressure drop-flow rate curves for superheated steam in U-tubes were generated. • Forward flow of hot steam is favored in the longer and taller U-tubes. • Reverse flow of cold steam is favored in short U-tubes. • Steam generator U-tube bundle geometry and tube diameter are important. • Need for correlation development for natural convention heat transfer coefficient. - Abstract: Characteristic pressure drop-flow rate curves are generated for all row numbers of the OPR1000 steam generators (SGs), representative of Combustion Engineering (CE) type SGs featuring square bend U-tubes. The pressure drop-flow rate curves are applicable to severe accident natural circulations of single-phase superheated steam during high pressure station blackout sequences with failed auxiliary feedwater and dry secondary side which are closely related to the thermally induced steam generator tube rupture event. The pressure drop-flow rate curves which determine the recirculation rate through the SG tubes are dependent on the tube bundle geometry and hydraulic diameter of the tubes. The larger CE type SGs have greater variation of tube length and height as a function of row number with forward flow of steam favored in the longer and taller high row number tubes and reverse flow favored in the short low row number tubes. Friction loss, natural convection heat transfer coefficients, and temperature differentials from the primary to secondary side are dominant parameters affecting the recirculation rate. The need for correlation development for natural convection heat transfer coefficients for external flow over tube bundles currently not modeled in system codes is discussed.

Fabrication of CANFLEX bundle kit for irradiation test in NRU

International Nuclear Information System (INIS)

Cho, Moon Sung; Kwon, Hyuk Il; Ji, Chul Goo; Chang, Ho Il; Sim, Ki Seob; Suk, Ho Chun.

1997-10-01

CANFLEX bundle kit was prepared at KAERI for the fabrication of complete bundle at AECL. Completed bundle will be used for irradiation test in NRU. Provisions in the 'Quality Assurance Manual for HWR Fuel Projects,' 'Manufacturing Plan' and 'Quality Verification, Inspection and Test Plan' were implemented as appropriately for the preparation of CANFLEX kit. A set of CANFLEX kit consist of 43 fuel sheath of two different sizes with spacers, bearing pads and buttons attached, 2 pieces of end plates and 86 pieces of end caps with two different sizes. All the documents utilized as references for the fabrication such as drawings, specifications, operating instructions, QC instructions and supplier's certificates are specified in this report. Especially, suppliers' certificates and inspection reports for the purchased material as well as KAERI's inspection report are integrated as attachments to this report. Attached to this report are supplier's certificates and KAERI inspection reports for the procured materials and KAERI QC inspection reports for tubes, pads, spacers, buttons, end caps, end plates and fuel sheath. (author). 37 refs

Heat transfer, pressure drop and flow patterns during flow boiling of R407C in a horizontal microfin tube

Science.gov (United States)

Rollmann, P.; Spindler, K.; Müller-Steinhagen, H.

2011-08-01

The heat transfer, pressure drop and flow patterns during flow boiling of R407C in a horizontal microfin tube have been investigated. The microfin tube is made of copper with a total fin number of 55 and a helix angle of 15°. The fin height is 0.24 mm and the inner tube diameter at fin root is 8.95 mm. The test tube is 1 m long. It is heated electrically. The experiments have been performed at saturation temperatures between -30°C and +10°C. The mass flux was varied between 25 and 300 kg/m2/s, the heat flux from 20,000 W/m2 down to 1,000 W/m2. The vapour quality was kept constant at 0.1, 0.3, 0.5, 0.7 at the inlet and 0.8, 1.0 at the outlet, respectively. The measured heat transfer coefficient is compared with the correlations of Cavallini et al., Shah as well as Zhang et al. Cavallini's correlation contains seven experimental constants. After fitting these constants to our measured values, the correlation achieves good agreement. The measured pressure drop is compared to the correlations of Pierre, Kuo and Wang as well as Müller-Steinhagen and Heck. The best agreement is achieved with the correlation of Kuo and Wang. Almost all values are calculated within an accuracy of ±30%. The flow regimes were observed. It is shown, that changes in the flow regime affect the heat transfer coefficient significantly.

Experimental study of heat transfer in a transverse flow around the heat exchanger tubes bank by lead

International Nuclear Information System (INIS)

Berezin, A.N.; Grabezhnaya, V.A.; Mikheev, A.S.; Parfenov, A.S.

2014-01-01

The results of the work to determine the heat transfer coefficient in crossflow by lead of pipes are presented. The study was conducted at supercritical pressure in the water circuit. There was a significant inequality in the distribution of the heat flow in different rows of the bundle of heat exchange tubes of corridor location at crossflow their lead. The experimentally determined heat transfer coefficients from the lead differ substantially from those generally accepted recommendations for the calculation of heat transfer at cross flow of rod bundle by liquid metal. The experimental results are close to those obtained earlier on the model with cross flow of heat exchanger tubes bundle by lead alloy with bismuth [ru

Source Terms for HFIR Beam Tube Shielding Analyses, and a Complete Shielding Analysis of the HB-3 Tube

International Nuclear Information System (INIS)

Bucholz, J.A.

2000-01-01

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory is in the midst of a massive upgrade program to enhance experimental facilities. The reactor presently has four horizontal experimental beam tubes, all of which will be replaced or redesigned. The HB-2 beam tube will be enlarged to support more guide tubes, while the HB-4 beam tube will soon include a cold neutron source

Source Terms for HFIR Beam Tube Shielding Analyses, and a Complete Shielding Analysis of the HB-3 Tube

Energy Technology Data Exchange (ETDEWEB)

Bucholz, J.A.

2000-07-01

The High Flux Isotope Reactor (HFIR) at the Oak Ridge National Laboratory is in the midst of a massive upgrade program to enhance experimental facilities. The reactor presently has four horizontal experimental beam tubes, all of which will be replaced or redesigned. The HB-2 beam tube will be enlarged to support more guide tubes, while the HB-4 beam tube will soon include a cold neutron source.

Simulating the fluid-structure interaction of a flexible tube in an array of rigid tubes

International Nuclear Information System (INIS)

Warnica, D.; Maleki, M.; Hariri, A.; Feldman, H.

2011-01-01

Two important single-phase mechanisms for flow-induced vibration of heat-exchanger tube bundles were used to demonstrate the capabilities of commercial software to simulate unsteady fluid-structure interactions (FSI). Reasonable agreement was obtained between the FSI simulations and experimental data for the onset of fluid elastic instability. There was also reasonable agreement between the FSI simulations and empirical correlations for the dynamic tube response to random turbulence excitation. Additional benefits of performing FSI simulations were the ability to characterize important features of the unsteady flow fields and hydrodynamic parameters such as viscous damping coefficients, which would otherwise require elaborate experimental measurements. (author)

CFD thermal-hydraulic analysis of a CANDU fuel channel with SEU43 type fuel bundle

International Nuclear Information System (INIS)

Catana, A.; Prisecaru, Ilie; Dupleac, D.; Danila, Nicolae

2009-01-01

This paper presents the numerical investigation of a CANDU fuel channel using CFD (Computational Fluid Dynamics) methodology approach, when SEU43 fuel bundles are used. Comparisons with STD37 fuel bundles are done in order to evaluate the influence of geometrical differences of the fuel bundle types on fluid flow properties. We adopted a strategy to analyze only the significant segments of fuel channel, namely : - the fuel bundle junctions with adjacent segments; - the fuel bundle spacer planes with adjacent segments; - the fuel bundle segments with turbulence enhancement buttons; - and the regular segments of fuel bundles. The computer code used is an academic version of FLUENT code, available from UPB. The complex flow domain of fuel bundles contained in pressure tube and operating conditions determine a high turbulence flow and in some parts of fuel channel also a multi-phase flow. Numerical simulation of the flow in the fuel channel has been achieved by solving the equations for conservation of mass, momentum and energy. For turbulence model the standard k-model is employed although other turbulence models can be used. In this paper we do not consider heat generation and heat transfer capabilities of CFD methods. Boundary conditions for CFD analysis are provided by system and sub-channel analysis. In this paper the discussion is focused on some flow parameters behaviour at the bundle junction, spacer's plane configuration, etc. of a SEU43 fuel bundle in conditions of a typical CANDU 6 fuel channel starting from some experience gained in a previous work. (authors)

Experimental results of the consequences of sodium water reactions at the bottom tube plate region of straight tube steam generators

International Nuclear Information System (INIS)

Ruloff, G.

1990-01-01

Experience with sodium water reactions has shown, that the course of such a steam generator accident depends strongly on its place in the steam generator. For the EFR steam generators we have to differentiate between: weld region at the upper tube plate (gas space); bundle region; weld region at the bottom tube plate. This paper describes results of a running tests program simulating the bottom tube plate area. One main part of these tests is the investigation of the influence of wastage protection shrouds between the tubes in the weld region to avoid a fast leak propagation and to give time for leak detection and mastering of the accidents. (author). 10 figs, 2 tabs

Validation of the assert subchannel code: Prediction of CHF in standard and non-standard Candu bundle geometries

International Nuclear Information System (INIS)

Carver, M.B.; Kiteley, J.C.; Zhou, R.Q.N.; Junop, S.V.; Rowe, D.S.

1993-01-01

The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of CANDU PHWR fuel channels, and to provide a detailed prediction of critical heat flux (CHF) distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of prediting CHF at these local conditions, makes it a unique tool for predicting CHF for situations outside the existing experimental data base. In particular, ASSERT is an appropriate tool to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. This paper discusses the numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology. The evolutionary validation plan is discussed, and early validation exercises are summarized. The paper concentrates, however, on more recent validation exercises in standard and non-standard geometries

The investigation of added masses and damping factors for vibrations of tube and tube bundles in fluid

International Nuclear Information System (INIS)

Sinyavskii, V.F.; Fedotovskii, V.S.; Kukhtin, A.B.

1977-01-01

The vibrations of single cylinders in fluid being surrounded by the solid walls of different form as well as the bundles of cylindric rods have been considered in this report. A model is proposed for hydrodynamic damping of vibrations and the analytic solution of a problem concerning damping of cylinder vibrations in fluid surrounded by a concentric shell. It has been shown that the fluid viscosity and vibration frequency influence the value of the fluid added mass and the damping factor of vibrations

Experimental study of nonequilibrium post-chf heat transfer in rod bundles

International Nuclear Information System (INIS)

Unal, C.; Tuzla, K.; Badr, O.; Neti, S.; Chen, J.

1986-01-01

Verifications and improvements of nonequilibrium heat transfer models, for post-critical-heat-flux convective boiling, has been greatly affected by the lack of experimental data regarding the degree of thermodynamic nonequilibrium. Recent studies had been successful in measuring vapor superheats in a vertical single tube. This paper extends the nonequilibrium convective boiling data to a rod bundle geometry. Vapor superheat measurements were obtained in a rod bundle with nine heated rods and a heated shroud. Tests were carried out with water at low mass fluxes with a wide range of dryout conditions. Significant nonequilibrium was observed, with vapor superheats of up to 600 0 C. Parametric effects of mass flux, heat flux and inlet conditions on vapor superheat are presented

Tube-AVB gap measurements using an eddy current rotating probe

International Nuclear Information System (INIS)

Badson, F.; Chiron, D.; Trumpff, B.

1988-01-01

The wears of tubes due to flow induced vibrations have been observed after a few years of operating PWR steam generators (SG). The vibration and wear are intimately related to the gap between tubes and anti-vibration bars (AVB's) located in the bundle. The authors report the development of an eddy current (EC) method for the measurement of this gap. The method is based on using an EC probe rotating in the tube. Since for each measurement zone the tube is interacting with two AVB's the use of a rotating EC probe is necessary to perform separate and accurate measurements of each tube-AVB gap

Horizontal steam generator thermal-hydraulics

Energy Technology Data Exchange (ETDEWEB)

Ubra, O. [SKODA Praha Company, Prague (Czechoslovakia); Doubek, M. [Czech Technical Univ., Prague (Czechoslovakia)

1995-09-01

Horizontal steam generators are typical components of nuclear power plants with pressure water reactor type VVER. Thermal-hydraulic behavior of horizontal steam generators is very different from the vertical U-tube steam generator, which has been extensively studied for several years. To contribute to the understanding of the horizontal steam generator thermal-hydraulics a computer program for 3-D steady state analysis of the PGV-1000 steam generator has been developed. By means of this computer program, a detailed thermal-hydraulic and thermodynamic study of the horizontal steam generator PGV-1000 has been carried out and a set of important steam generator characteristics has been obtained. The 3-D distribution of the void fraction and 3-D level profile as functions of load and secondary side pressure have been investigated and secondary side volumes and masses as functions of load and pressure have been evaluated. Some of the interesting results of calculations are presented in the paper.

Two-phase heat transfer and pressure drop of LNG during saturated flow boiling in a horizontal tube

Science.gov (United States)

Chen, Dongsheng; Shi, Yumei

2013-12-01

Two-phase heat transfer and pressure drop of LNG (liquefied natural gas) have been measured in a horizontal smooth tube with an inner diameter of 8 mm. The experiments were conducted at inlet pressures from 0.3 to 0.7 MPa with a heat flux of 8-36 kW m-2, and mass flux of 49.2-201.8 kg m-2 s-1. The effect of vapor quality, inlet pressure, heat flux and mass flux on the heat transfer characteristic are discussed. The comparisons of the experimental data with the predicted value by existing correlations are analyzed. Zou et al. (2010) correlation shows the best accuracy with 24.1% RMS deviation among them. Moreover four frictional pressure drop methods are also chosen to compare with the experimental database.

Steam generator with U-tube bank arranged within an oblong pressure vessel

International Nuclear Information System (INIS)

Beckmann, G.; Fritz, K.

1976-01-01

This steam generator equipped with a U-tube bundle differs substantially from standard types because of its operational condition. The boiler is at a tilt of 45 0 , the piping base, the inlet, and the outlet for the primary medium are arranged at the top. This improves the heat flow of the secondary medium within the boiler. The steam room placed near the piping base is enlarged on the hot side of the U-tube bundle due to the tilt of the water level, allowing drying and overheating of the steam without additional mounting of water separators and special overheaters. The additional space obtained by this construction is estimated at 6%. (FW) [de

Numerical simulation of cross-flow-induced fluidelastic vibration of tube arrays and comparison with experimental results

International Nuclear Information System (INIS)

Eisinger, F.L.; Rao, M.S.M.; Steininger, D.A.; Haslinger, K.H.

1995-01-01

Tube arrays exposed to air, gas or liquid cross-flow can vibrate due to vortex-shedding, turbulence, or fluidelastic instability. The major emphasis of this paper is on the phenomenon of fluidelastic instability (or fluidelastic vibration). A numerical model is applied to the simulation of fluidelastic vibration of representative tubes in a tube bundle, based on S. S. Chen's unsteady flow theory. The results are validated against published data based on linear cases. The model is then applied to a nonlinear structure of a U-bend tube bundle with clearances at supports, and the computed results compared to those obtained by experimental testing. The numerical studies were performed using the ABAQUS-EPGEN finite element code using a special subroutine incorporating fluidelastic forces. It is shown that the results of both the linear and nonlinear modeling are in good agreement with experimental data

The Atiyah bundle and connections on a principal bundle

Indian Academy of Sciences (India)

be the fiber bundle constructed as in (1.1) for the universal principal G-bundle. In a work in progress, we hope to show that the universal G-connection can be realized as a fiber bundle over C(EG). Turning this ... a G-invariant vector field on EG|U . In other words, we get a bijective linear map between. A(EG)(U) (the space of ...

Behavior of a bundle of fast fuel pins under irradiation

International Nuclear Information System (INIS)

Marbach, G.; Millet, P.; Robert, J.; Languille, A.

1979-01-01

In the French design of fuel elements for fast reactors, great deformation of pins can bring about interaction with the hexagonal tube through the spacer wires. The change in such bundles is described here when the diameter of the cladding increases and the outcome of this reaction (bending and ovalization of pins) is calculated with a simplified model. It is shown that the results achieved agree well with the experimental observations [fr

Relation between medium fluid temperature and centroid subchannel temperatures of a nuclear fuel bundle mock-up

International Nuclear Information System (INIS)

Carvalho Tofani, P. de.

1986-01-01

The subchannel method used in nuclear fuel bundle thermal-hydraulic analysis lies in the statement that subchannel fluid temperatures are taken at mixed mean values. However, the development of mixing correlations and code assessment procedures are, sometimes in the literature, based upon the assumption of identity between lumped and local (subchannel centroid) temperature values. The present paper is concerned with the presentation of an approach for correlating lumped to centroid subchannel temperatures, based upon previously formulated models by the author, applied, applied to a nine heated tube bundle experimental data set. (Author) [pt

Relation between medium fluid temperature and centroid subchannel temperatures of a nuclear fuel bundle mock-up

International Nuclear Information System (INIS)

Carvalho Tofani, P. de.

1986-01-01

The subchannel method used in nuclear fuel bundle thermal-hydraulic analysis lies in the statement that subchannel fluid temperatures are taken at mixed mean values. However, the development of mixing correlations and code assessment procedures are, sometimes in the literature, based upon the assumption of identity between lumped and local (subchannel centroid) temperature values. The present paper is concerned with the presentation of an approach for correlating lumped to centroid subchannel temperatures, based upon previously formulated models by the author, applied to a nine heated tube bundle experimental data set. (Author) [pt

Modelling of a cross flow evaporator for CSP application

DEFF Research Database (Denmark)

Sørensen, Kim; Franco, Alessandro; Pelagotti, Leonardo

2016-01-01

) applications. Heat transfer and pressure drop prediction methods are an important tool for design and modelling of diabatic, two-phase, shell-side flow over a horizontal plain tubes bundle for a vertical up-flow evaporator. With the objective of developing a model for a specific type of cross flow evaporator...... the available correlations for the definition of two-phase flow heat transfer, void fraction and pressure drop in connection with the operation of steam generators, focuses attention on a comparison of the results obtained using several different models resulting by different combination of correlations......Heat exchangers consisting of bundles of horizontal plain tubes with boiling on the shell side are widely used in industrial and energy systems applications. A recent particular specific interest for the use of this special heat exchanger is in connection with Concentrated Solar Power (CSP...

Validation of the ASSERT subchannel code for prediction of CHF in standard and non-standard CANDU bundle geometries

International Nuclear Information System (INIS)

Kiteley, J.C.; Carver, M.B.; Zhou, Q.N.

1993-01-01

The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of CANDU PHWR fuel channels, and to provide a detailed prediction of critical heat flux distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of predicting critical heat flux (CHF) at these local conditions makes it a unique tool for predicting CHF for situations outside the existing experimental data base. In particular, ASSERT is the only tool available to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. This paper discusses the numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology. The evolutionary validation plan is discussed, and early validation exercises are summarized. The paper concentrates, however, on more recent validation exercises in standard and non-standard geometries. 28 refs., 12 figs

Validation of the ASSERT subchannel code: Prediction of critical heat flux in standard and nonstandard CANDU bundle geometries

International Nuclear Information System (INIS)

Carver, M.B.; Kiteley, J.C.; Zhou, R.Q.N.; Junop, S.V.; Rowe, D.S.

1995-01-01

The ASSERT code has been developed to address the three-dimensional computation of flow and phase distribution and fuel element surface temperatures within the horizontal subchannels of Canada uranium deuterium (CANDU) pressurized heavy water reactor fuel channels and to provide a detailed prediction of critical heat flux (CHF) distribution throughout the bundle. The ASSERT subchannel code has been validated extensively against a wide repertoire of experiments; its combination of three-dimensional prediction of local flow conditions with a comprehensive method of predicting CHF at these local conditions makes it a unique tool for predicting CHF for situations outside the existing experimental database. In particular, ASSERT is an appropriate tool to systematically investigate CHF under conditions of local geometric variations, such as pressure tube creep and fuel element strain. The numerical methodology used in ASSERT, the constitutive relationships incorporated, and the CHF assessment methodology are discussed. The evolutionary validation plan is also discussed and early validation exercises are summarized. More recent validation exercises in standard and nonstandard geometries are emphasized

Experimental modeling of flow-induced vibration of multi-span U-tubes in a CANDU steam generator

International Nuclear Information System (INIS)

Mohany, A.; Feenstra, P.; Janzen, V.P.; Richard, R.

2009-01-01

Flow-induced vibration of the tubes in a nuclear steam generator is a concern for designers who are trying to increase the life span of these units. The dominant excitation mechanisms are fluidelastic instability and random turbulence excitation. The outermost U-bend region of the tubes is of greatest concern because the flow is almost perpendicular to the tube axis and the unsupported span is relatively long. The support system in this region must be well designed in order to minimize fretting wear of the tubes at the support locations. Much of the previous testing was conducted on straight single-span or cantilevered tubes in cross-flow. However, the dynamic response of steam generator multi-span U-tubes with clearance supports is expected to be different. Accurate modeling of the tube dynamics is important to properly simulate the dynamic interaction of the tube and supports. This paper describes a test program that was developed to measure the dynamic response of a bundle of steam generator U-tubes with Anti-Vibration Bar (AVB) supports, subjected to Freon two-phase cross-flow. The tube bundle has similar geometrical conditions to those expected for future CANDU steam generators. Future steam generators will be larger than previous CANDU steam generators, nearly twice the heat transfer area, with significant changes in process conditions in the U-bend region, such as increased steam quality and a broader range of flow velocities. This test program was initiated at AECL to demonstrate that the tube support design for future CANDU steam generators will meet the stringent requirements associated with a 60 year design life. The main objective of the tests is to address the issue of in-plane and out-of-plane fluidelastic instability and random turbulent excitation of a U-tube bundle with Anti-Vibration Bar (AVB) supports. Details of the test rig, measurement techniques and preliminary instrumentation results are described in the paper. (author)

Polyelectrolyte bundles

Energy Technology Data Exchange (ETDEWEB)

Limbach, H J; Sayar, M; Holm, C [Max-Planck-Institut fuer Polymerforschung, Ackermannweg 10, 55128 Mainz (Germany)

2004-06-09

Using extensive molecular dynamics simulations we study the behaviour of polyelectrolytes with hydrophobic side chains, which are known to form cylindrical micelles in aqueous solution. We investigate the stability of such bundles with respect to hydrophobicity, the strength of the electrostatic interaction and the bundle size. We show that for the parameter range relevant for sulfonated poly(para-phenylenes) (PPP) one finds a stable finite bundle size. In a more generic model we also show the influence of the length of the precursor oligomer on the stability of the bundles. We also point out that our model has close similarities to DNA solutions with added condensing agents, hinting at the possibility that the size of DNA aggregates is, under certain circumstances, thermodynamically limited.

Polyelectrolyte bundles

International Nuclear Information System (INIS)

Limbach, H J; Sayar, M; Holm, C

2004-01-01

Using extensive molecular dynamics simulations we study the behaviour of polyelectrolytes with hydrophobic side chains, which are known to form cylindrical micelles in aqueous solution. We investigate the stability of such bundles with respect to hydrophobicity, the strength of the electrostatic interaction and the bundle size. We show that for the parameter range relevant for sulfonated poly(para-phenylenes) (PPP) one finds a stable finite bundle size. In a more generic model we also show the influence of the length of the precursor oligomer on the stability of the bundles. We also point out that our model has close similarities to DNA solutions with added condensing agents, hinting at the possibility that the size of DNA aggregates is, under certain circumstances, thermodynamically limited

Polyelectrolyte bundles

Science.gov (United States)

Limbach, H. J.; Sayar, M.; Holm, C.

2004-06-01

Using extensive Molecular Dynamics simulations we study the behavior of polyelectrolytes with hydrophobic side chains, which are known to form cylindrical micelles in aqueous solution. We investigate the stability of such bundles with respect to hydrophobicity, the strength of the electrostatic interaction, and the bundle size. We show that for the parameter range relevant for sulfonated poly-para-phenylenes (PPP) one finds a stable finite bundle size. In a more generic model we also show the influence of the length of the precursor oligomer on the stability of the bundles. We also point out that our model has close similarities to DNA solutions with added condensing agents, hinting to the possibility that the size of DNA aggregates is under certain circumstances thermodynamically limited.

Characterization of bundled and individual triple-walled carbon nanotubes by resonant Raman spectroscopy.

Science.gov (United States)

Hirschmann, Thomas Ch; Araujo, Paulo T; Muramatsu, Hiroyuki; Zhang, Xu; Nielsch, Kornelius; Kim, Yoong Ahm; Dresselhaus, Mildred S

2013-03-26

The optical characterization of bundled and individual triple-walled carbon nanotubes was studied for the first time in detail by using resonant Raman spectroscopy. In our approach, the outer tube of a triple-walled carbon nanotube system protects the two inner tubes (or equivalently the inner double-walled carbon nanotube) from external environment interactions making them a partially isolated system. Following the spectral changes and line-widths of the radial breathing modes and G-band by performing laser energy dependent Raman spectroscopy, it is possible to extract important information as regards to the electronic and vibrational properties, tube diameters, wall-to-wall distances, radial breathing mode, and G-band resonance evolutions as well as high-curvature intertube interactions in isolated double- and triple-walled carbon nanotube systems.

FIMBRIN1 Is Involved in Lily Pollen Tube Growth by Stabilizing the Actin Fringe[C][W][OA

Science.gov (United States)

Su, Hui; Zhu, Jinsheng; Cai, Chao; Pei, Weike; Wang, Jiaojiao; Dong, Huaijian; Ren, Haiyun

2012-01-01

An actin fringe structure in the subapex plays an important role in pollen tube tip growth. However, the precise mechanism by which the actin fringe is generated and maintained remains largely unknown. Here, we cloned a 2606-bp full-length cDNA encoding a deduced 77-kD fimbrin-like protein from lily (Lilium longiflorum), named FIMBRIN1 (FIM1). Ll-FIM1 was preferentially expressed in pollen and concentrated at actin fringe in the subapical region, as well as in longitudinal actin-filament bundles in the shank of pollen tubes. Microinjection of Ll-FIM1 antibody into lily pollen tubes inhibited tip growth and disrupted the actin fringe. Furthermore, we verified the function of Ll-FIM1 in the fim5 mutant of its closest relative, Arabidopsis thaliana. Pollen tubes of fim5 mutants grew with a larger diameter in early stages but could recover into normal forms in later stages, despite significantly slower growth rates. The actin fringe of the fim5 mutants, however, was impaired during both early and late stages. Impressively, stable expression of fim5pro:GFP:Ll-FIM1 rescued the actin fringe and the growth rate of Arabidopsis fim5 pollen tubes. In vitro biochemical analysis showed that Ll-FIM1 could bundle actin filaments. Thus, our study has identified a fimbrin that may stabilize the actin fringe by cross-linking actin filaments into bundles, which is important for proper tip growth of lily pollen tubes. PMID:23150633

Strategic Aspects of Bundling

International Nuclear Information System (INIS)

Podesta, Marion

2008-01-01

The increase of bundle supply has become widespread in several sectors (for instance in telecommunications and energy fields). This paper review relates strategic aspects of bundling. The main purpose of this paper is to analyze profitability of bundling strategies according to the degree of competition and the characteristics of goods. Moreover, bundling can be used as price discrimination tool, screening device or entry barriers. In monopoly case bundling strategy is efficient to sort consumers in different categories in order to capture a maximum of surplus. However, when competition increases, the profitability on bundling strategies depends on correlation of consumers' reservations values. (author)

Scalable synthesis of aligned carbon nanotubes bundles using green natural precursor: neem oil

Directory of Open Access Journals (Sweden)

Kumar Rajesh

2011-01-01

Full Text Available Abstract Practical application of aligned carbon nanotubes (ACNTs would have to be determined by a matter of its economical and large-scale preparation. In this study, neem oil (also named Margoaa oil, extracted from the seeds of the neem--Azadirachta indica was used as carbon source to fabricate the bundles of ACNTs. ACNTs have been synthesized by spray pyrolysis of neem oil and ferrocene mixture at 825°C. The major components of neem oil are hydrocarbon with less amount of oxygen, which provided the precursor species in spray pyrolysis growth of CNTs. The bundles of ACNTs have been grown directly inside the quartz tube. The as-grown ACNTs have been characterized through Raman spectroscopy, scanning and transmission electron microscopic (SEM/TEM techniques. SEM images reveal that the bundles of ACNTs are densely packed and are of several microns in length. High-resolution TEM analysis reveals these nanotubes to be multi-walled CNTs. These multi-walled CNTs were found to have inner diameter between 15 and 30 nm. It was found that present technique gives high yield with high density of bundles of ACNTs.

Counter-current flow in a vertical to horizontal tube with obstructions

Energy Technology Data Exchange (ETDEWEB)

Tye, P.; Matuszkiewicz, A.; Teyssedou, A. [Institut de Genie Nucleaire, Quebec (Canada)] [and others

1995-09-01

This paper presents experimental results on counter-current flow and flooding in an elbow between a vertical and a horizontal run. The experimental technique used allowed not only the flooding limit to be determined, but also the entire partial delivery region to be studied as well. The influence that various size orifices placed in the horizontal run have on both the delivered liquid flow rates and on the flooding limits is also examined. It is observed that both the flooding limits and the delivered liquid flow rates decrease with decreasing orifice size. Further, it is also observed that the mechanisms that govern the partial delivery of the liquid are significantly different when an orifice is present in the horizontal leg as compared to the case when no orifice is present.

The modelling of condensation in horizontal tubes and the comparison with experimental data

Directory of Open Access Journals (Sweden)

Bryk Rafał

2017-01-01

Full Text Available The condensation in horizontal tubes plays an important role in determining the operation mode of passive safety systems of modern nuclear power plants. In this paper, two different approaches for modelling of this phenomenon are compared and verified against experimental data. The first approach is based on the flow regime map developed by Tandon. Depending on the regime, the heat transfer coefficient is calculated according to corresponding semi-empirical correlation. The second approach uses a general, fully empirical correlation proposed by Shah. Both models are developed with utilization of the object-oriented, equation-based Modelica language and the open-source Open-Modelica environment. The results are compared with data obtained during a large scale integral test, simulating a Loss of Coolant Accident scenario performed at the dedicated Integral Test Facility Karlstein (INKA which was built at the Components Testing Department of AREVA in Karlstein, Germany. The INKA facility was designed to test the performance of the passive safety systems of KERENA, the new AREVA boiling water reactor design. INKA represents the KERENA containment with a volume scaling of 1:24. Components heights and levels over the ground are in the full scale. The comparison of simulations results shows a good agreement.

Life management of Zr 2.5% Nb pressure tube through estimation of fracture properties by cyclic ball indentation technique

International Nuclear Information System (INIS)

Chatterjee, S.; Madhusoodanan, K.; Rama Rao, A.

2015-01-01

In Pressurised Heavy Water Reactors (PHWRs) fuel bundles are located inside horizontal pressure tubes. Pressure tubes made up of Zr 2.5 wt% Nb undergo degradation during in-service environmental conditions. Measurement of mechanical properties of degraded pressure tubes is important for assessing its fitness for further service in the reactor. The only way to accomplish this important objective is to develop a system based on insitu measurement technique. Considering the importance of such measurement, an In-situ Property Measurement System (IProMS) based on cyclic ball indentation technique has been designed and developed indigenously. The remotely operable system is capable of carrying out indentation trial on the inside surface of the pressure tube and to estimate important mechanical properties like yield strength, ultimate tensile strength, hardness etc. It is known that fracture toughness is one of the important life limiting parameters of the pressure tube. Hence, five spool pieces of Zr 2.5 wt% Nb pressure tube of different mechanical properties have been used for estimation of fracture toughness by ball indentation method. Curved Compact Tension (CCT) specimens were also prepared from the five spool pieces for measurement of fracture toughness from conventional tests. The conventional fracture toughness values were used as reference data. A methodology has been developed to estimate the fracture properties of Zr 2.5 wt% Nb pressure tube material from the analysis of the ball indentation test data. This paper highlights the comparison between tensile properties measured from conventional tests and IProMS trials and relates the fracture toughness parameters measured from conventional tests with the IProMS estimated fracture properties like Indentation Energy to Fracture. (author)

Nonabelian bundle 2-gerbes

OpenAIRE

Jurco, Branislav

2009-01-01

We define 2-crossed module bundle 2-gerbes related to general Lie 2-crossed modules and discuss their properties. A 2-crossed module bundle 2-gerbe over a manifold is defined in terms of a so called 2-crossed module bundle gerbe, which is a crossed module bundle gerbe equipped with an extra sructure. It is shown that string structures can be described and classified using 2-crossed module bundle 2-gerbes.

Recommended method to prevent leakage of titanium tube in condenser

International Nuclear Information System (INIS)

Wang Jun

2010-01-01

Qinshan Phase III is located at the estuary area of Qiantang River, where contains much slit and sand in the seawater. Since the units were put into operation, tube bundles in the condenser have been scratched, damaged or blocked by hard foreign materials, and outside wall thickness reduced and broken due to various reasons. Many tube bundles are discarded. In order to effectively prevent the re-occurrence of such problem and eliminate the existing defects, equipment management personnels of Qinshan Phase III work together with experts both from home and abroad, and perfom root-analysis for various cause of defects. After the problem root is identified, a serious of specific and effective measures are taken to prevent and eliminate the problem and reached a good effect. This paper herein is written for comments and reference. (authors)

Indoor solar thermal energy saving time with phase change material in a horizontal shell and finned-tube heat exchanger.

Science.gov (United States)

Paria, S; Sarhan, A A D; Goodarzi, M S; Baradaran, S; Rahmanian, B; Yarmand, H; Alavi, M A; Kazi, S N; Metselaar, H S C

2015-01-01

An experimental as well as numerical investigation was conducted on the melting/solidification processes of a stationary phase change material (PCM) in a shell around a finned-tube heat exchanger system. The PCM was stored in the horizontal annular space between a shell and finned-tube where distilled water was employed as the heat transfer fluid (HTF). The focus of this study was on the behavior of PCM for storage (charging or melting) and removal (discharging or solidification), as well as the effect of flow rate on the charged and discharged solar thermal energy. The impact of the Reynolds number was determined and the results were compared with each other to reveal the changes in amount of stored thermal energy with the variation of heat transfer fluid flow rates. The results showed that, by increasing the Reynolds number from 1000 to 2000, the total melting time decreases by 58%. The process of solidification also will speed up with increasing Reynolds number in the discharging process. The results also indicated that the fluctuation of gradient temperature decreased and became smooth with increasing Reynolds number. As a result, by increasing the Reynolds number in the charging process, the theoretical efficiency rises.

Polycation induced actin bundles.

Science.gov (United States)

Muhlrad, Andras; Grintsevich, Elena E; Reisler, Emil

2011-04-01

Three polycations, polylysine, the polyamine spermine and the polycationic protein lysozyme were used to study the formation, structure, ionic strength sensitivity and dissociation of polycation-induced actin bundles. Bundles form fast, simultaneously with the polymerization of MgATP-G-actins, upon the addition of polycations to solutions of actins at low ionic strength conditions. This indicates that nuclei and/or nascent filaments bundle due to attractive, electrostatic effect of polycations and the neutralization of repulsive interactions of negative charges on actin. The attractive forces between the filaments are strong, as shown by the low (in nanomolar range) critical concentration of their bundling at low ionic strength. These bundles are sensitive to ionic strength and disassemble partially in 100 mM NaCl, but both the dissociation and ionic strength sensitivity can be countered by higher polycation concentrations. Cys374 residues of actin monomers residing on neighboring filaments in the bundles can be cross-linked by the short span (5.4Å) MTS-1 (1,1-methanedyl bismethanethiosulfonate) cross-linker, which indicates a tight packing of filaments in the bundles. The interfilament cross-links, which connect monomers located on oppositely oriented filaments, prevent disassembly of bundles at high ionic strength. Cofilin and the polysaccharide polyanion heparin disassemble lysozyme induced actin bundles more effectively than the polylysine-induced bundles. The actin-lysozyme bundles are pathologically significant as both proteins are found in the pulmonary airways of cystic fibrosis patients. Their bundles contribute to the formation of viscous mucus, which is the main cause of breathing difficulties and eventual death in this disorder. Copyright © 2011 Elsevier B.V. All rights reserved.

CFD investigation of vertical rod bundles of supercritical water-cooled nuclear reactor

International Nuclear Information System (INIS)

Shang Zhi

2009-01-01

The commercial CFD code STAR-CD v4.02 is used as the numerical simulation tool for the supercritical water-cooled nuclear reactor (SCWR). The numerical simulation is based on the real full 3D rod bundles' geometry of the nuclear reactors. For satisfying the near-wall resolution of y + ≤ 1, the structure mesh with the stretched fine mesh near wall is employed. The validation of the numerical simulation for mesh generation strategy and the turbulence model for the heat transfer of supercritical water is carried out to compare with 3D tube experiments. After the validation, the same mesh generation strategy and the turbulence model are employed to study three types of the geometry frame of the real rod bundles. Through the numerical investigations, it is found that the different arrangement of the rod bundles will induce the different temperature distribution at the rods' walls. The wall temperature distributions are non-uniform along the wall and the values depend on the geometry frame. At the same flow conditions, downward flow gets higher wall temperature than upward flow. The hexagon geometry frame has the smallest wall temperature difference comparing with the others. The heat transfer is controlled by P/D ratio of the bundles.

Tube-in-shell heat exchangers

International Nuclear Information System (INIS)

Richardson, J.

1976-01-01

Tube-in-shell heat exchangers normally comprise a bundle of parallel tubes within a shell container, with a fluid arranged to flow through the tubes in heat exchange with a second fluid flowing through the shell. The tubes are usually end supported by the tube plates that separate the two fluids, and in use the tube attachments to the tube plates and the tube plates can be subject to severe stress by thermal shock and frequent inspection and servicing are required. Where the heat exchangers are immersed in a coolant such as liquid Na such inspection is difficult. In the arrangement described a longitudinally extending central tube is provided incorporating axially spaced cylindrical tube plates to which the opposite ends of the tubes are attached. Within this tube there is a tubular baffle that slidably seals against the wall of the tube between the cylindrical tube plates to define two co-axial flow ducts. These ducts are interconnected at the closed end of the tube by the heat exchange tubes and the baffle comprises inner and outer spaced walls with the interspace containing Ar. The baffle is easily removable and can be withdrawn to enable insertion of equipment for inspecting the wall of the tube and tube attachments and to facilitate plugging of defective tubes. Cylindrical tube plates are believed to be superior for carrying pressure loads and resisting the effects of thermal shock. Some protection against thermal shock can be effected by arranging that the secondary heat exchange fluid is on the tube side, and by providing a thermal baffle to prevent direct impingement of hot primary fluid on to the cylindrical tube plates. The inner wall of the tubular baffle may have flexible expansible region. Some nuclear reactor constructions incorporating such an arrangement are described, including liquid metal reactors. (U.K.)

Coupling of standard condensing nuclear power stations to horizontal aluminium tubes multieffect distillation plants

International Nuclear Information System (INIS)

Adar, J.

1977-01-01

No large nuclear back-pressure turbines are available to-day. Standard condensing nuclear turbines could operate continuously with a back-pressure of up to 7'' Hg, exhausting huge amounts of steam at 56degC-64degC with a loss of electricity production of only 6%-10%. The horizontal aluminium tube multieffect distillation process developed by 'Israel Desalination Engineering Ltd' is very suitable for the use of such low-grade heat. A special flash-chamber loop constitutes a positive barrier against any possible contamination being carried over by the steam exhausted from the turbine to the desalination plant. The operation is designed to be flexible so that the power plant can be operated either in conjunction with the desalination plant or as a single purpose plant. Flow sheets, heat and mass balances have been prepared for eight different combinations of plants. Only standard equipment is being used in the power plant. The desalination plant consists of 6 to 12 parallel double lines, each of them similar to a large prototype now being designed. Water production varies between 50 and 123 MGD and water cost between 90 and 137 c/1000 gallons. Costs are based on actual bids

DEVELOPMENT OF COILED TUBING STRESS ANALYSIS

Directory of Open Access Journals (Sweden)

Davorin Matanović

1998-12-01

Full Text Available The use of coiled tubing is increasing rapidly with drilling of horizontal wells. To satisfy all requirements (larger mechanical stresses, larger fluid capacities the production of larger sizes and better material qualities was developed. Stresses due to axial forces and pressures that coiled tubing is subjected are close to its performance limits. So it is really important to know and understand the behaviour of coiled tubing to avoid its break, burst or collapse in the well.

Thyc, a 3D thermal-hydraulic code for rod bundles. Recent developments and validation tests

International Nuclear Information System (INIS)

Caremoli, C.; Rascle, P.; Aubry, S.; Olive, J.

1993-09-01

PWR or LMFBR cores or fuel assemblies, PWR steam generators, condensers, tubular heat exchangers, are basic components of a nuclear power plant involving two-phase flows in tube or rod bundles. A deep knowledge of the detailed flow patterns on the shell side is necessary to evaluate DNB margins in reactor cores, singularity effects (grids, wire spacers, support plates, baffles), corrosion on steam generator tube sheet, bypass effects and vibration risks. For that purpose, Electricite de France has developed, since 1986, a general purpose code named THYC (Thermal HYdraulic Code) designed to study three-dimensional single and two phase flows in rod or tube bundles (pressurized water reactor cores, steam generators, condensers, heat exchangers). It considers the three-dimensional domain to contain two kinds of components: fluid and solids. The THYC model is obtained by space-time averaging of the instantaneous equations (mass, momentum and energy) of each phase over control volumes including fluid and solids. This paper briefly presents the physical model and the numerical method used in THYC. Then, validation tests (comparison with experiments) and applications (coupling with three-dimensional neutronics code and DNB predictions) are presented. They emphasize the last developments and new capabilities of the code. (authors). 10 figs., 3 tabs., 21 refs

Generator module architecture for a large solid oxide fuel cell power plant

Science.gov (United States)

Gillett, James E.; Zafred, Paolo R.; Riggle, Matthew W.; Litzinger, Kevin P.

2013-06-11

A solid oxide fuel cell module contains a plurality of integral bundle assemblies, the module containing a top portion with an inlet fuel plenum and a bottom portion receiving air inlet feed and containing a base support, the base supports dense, ceramic exhaust manifolds which are below and connect to air feed tubes located in a recuperator zone, the air feed tubes passing into the center of inverted, tubular, elongated, hollow electrically connected solid oxide fuel cells having an open end above a combustion zone into which the air feed tubes pass and a closed end near the inlet fuel plenum, where the fuel cells comprise a fuel cell stack bundle all surrounded within an outer module enclosure having top power leads to provide electrical output from the stack bundle, where the fuel cells operate in the fuel cell mode and where the base support and bottom ceramic air exhaust manifolds carry from 85% to all 100% of the weight of the stack, and each bundle assembly has its own control for vertical and horizontal thermal expansion control.

Bruce and Darlington power pulse and pressure tube integrity programs -status 1995

Energy Technology Data Exchange (ETDEWEB)

Field, G J [Atomic Energy of Canada Ltd., Mississauga, ON (Canada); Wylie, J [Ontario Hydro, Tiverton, ON (Canada). Bruce Nuclear Generating Station-A

1996-12-31

The optimum solution to pressure tube fretting at the inlet of the Bruce and Darlington channels, a concern which became very serious following inspections in early 1992, is to remove the inlet bundle and operate with a 12 fuel bundle channel. During analysis of this operating mode a `power pulse` was identified which could occur during an inlet header break where all the fuel in the channel moved rapidly to the inlet of the channel. The pulse was unacceptable and the units were derated until solutions could be implemented. A number of solutions were identified and each station has begun implementation of their specific solution. Implementation has not been without problems and this paper provides a status report on the progress to date of the long bundle implementation solution for Bruce B and Darlington and the fuelling with the flow solution being implemented at Bruce A. Both types of solution have a significant impact on the original concern, fretting of the pressure tube. (author). 1 ref., 6 figs.

FLECHT-SEASET 21-rod bundle flow blockage heat transfer during reflood

International Nuclear Information System (INIS)

Loftus, M.; Hochreiter, L.; Lee, N.

1983-01-01

The effect of various flow blockage shapes and distributions during a PWR reflood was investigated using six 21-rod bundles with full length, internally heated, cosine power-shaped electrical rods. The flow blockage shapes, simulating the fuel rod clad ballooning, were made of thin-wall stainless steel tubes hydroformed into a short, concentric shape and along, nonconcentric shape. The blockage sleeves were distributed both coplanar, with all sleeves located at the same elevation, and non-coplanar. The initial and boundary conditions were varied to include parametric effects of pressure, inlet water temperature, and primarily, flooding rate. The initial mid-plane rod temperature was 871 0 C (1600 0 F) in all tests. Rod and vapor temperature measurements were made throughout the rod bundle with emphasis on the blockage region. The rod heat transfer downstream of the blockage was found to be greater for rods in a blocked bundle than for similar rods in an unblocked bundle. The heat transfer improvement decreases both with time after flood initiation and as the distance increased downstream of the blockage. The improvement in the heat transfer is attributed primarily to the breakup of the water droplets entrained in the steam flow. The smaller droplets subsequently evaporate and desuperheat the steam, which then improves the heat transfer between the rods and the steam in and downstream of the blockage zone

Benchmark simulation of turbulent flow through a staggered tube bundle to support CFD as a reactor design tool. Part 2. URANS CFD simulation

International Nuclear Information System (INIS)

Ridluan, Artit; Tokuhiro, Akira

2008-01-01

In Part II, we described the unsteady flow simulation and proposed a modification of a traditional turbulence flow model. Computational fluid dynamics (CFD) simulations of an isothermal, fully periodic flow across a tube bundle using unsteady Reynolds averaged Navier-Stokes (URANS) equations, with turbulence models such as the Reynolds stress model (RSM) were investigated at a Reynolds number of 1.8x10 4 , based on the tube diameter and inlet velocity. As noted in Part I, CFD simulation and experimental results were compared at five positions along (x,y) coordinates. The steady RANS simulation showed that four diverse turbulence models were efficient for predicting the Reynolds stresses, and generally, SRANS results were marginal to poor, using a consistent evaluation terminology. In the URANS simulation, we modeled the turbulent flow field in a manner similar to the approach used for large eddy simulation (LES). The time-dependent URANS results showed that the simulation reproduces the dynamic stability as characterized by transverse oscillatory flow structures in the near-wake region. In particular, the inclusion of terms accounting for the time scales associated with the production range and dissipation rate of turbulence generates unsteady statistics of the mean and fluctuation flow. In spite of this, the model implemented produces better agreement with a benchmark data set and is thus recommended. (author)

Pressure vessels and methods of sealing leaky tubes disposed in pressure vessels

International Nuclear Information System (INIS)

Larson, G.C.

1980-01-01

This invention relates to pressure vessels and to methods of sealing leaky tubes in them and is especially applicable to pressure vessels in the form of sheet-and-tube type heat exchangers constructed with a large number of relatively small diameter tubes grouped in a bundle. To seal off a leaky tube in such a heat exchanger an explosive activated plug in the form of a hollow metal body is used, inserted at each end of the tube to be sealed. Using the arrangement of pressure vessel and associated tube sheets and the explosive activated plug method of sealing a leaky tube as described in this invention it is claimed that distortion of the adjacent tubes and the tube sheets is reduced when the explosive activated plugs are detonated. (U.K.)

Ion-irradiation-induced defects in bundles of carbon nanotubes

International Nuclear Information System (INIS)

Salonen, E.; Krasheninnikov, A.V.; Nordlund, K.

2002-01-01

We study the structure and formation yields of atomic-scale defects produced by low-dose Ar ion irradiation in bundles of single-wall carbon nanotubes. For this, we employ empirical potential molecular dynamics and simulate ion impact events over an energy range of 100-1000 eV. We show that the most common defects produced at all energies are vacancies on nanotube walls, which at low temperatures are metastable but long-lived defects. We further calculate the spatial distribution of the defects, which proved to be highly non-uniform. We also show that ion irradiation gives rise to the formations of inter-tube covalent bonds mediated by carbon recoils and nanotube lattice distortions due to dangling bond saturation. The number of inter-tube links, as well as the overall damage, linearly grows with the energy of incident ions

Nefness of adjoint bundles for ample vector bundles

Directory of Open Access Journals (Sweden)

Hidetoshi Maeda

1995-11-01

Full Text Available Let E be an ample vector bundle of rank >1 on a smooth complex projective variety X of dimension n. This paper gives a classification of pairs (X,E whose adjoint bundles K_X+det E are not nef in the case when  r=n-2.

Simulation of steam condensation in the presence of noncondensable gases in horizontal condenser tubes using RELAP5 for advanced nuclear reactors

International Nuclear Information System (INIS)

Macedo, Luiz Alberto; Torres, Walmir Maximo

2009-01-01

Horizontal heat exchangers are used in advanced light water nuclear reactors in their passive cooling systems, such as residual heat removal (RHRS) and passive containment cooling system (PCCS). Condensation studies of steam and noncondensable gases mixtures in these heat exchangers are very important due to the phenomena multidimensional nature and the condensate stratification effects. This work presents a comparison between simulation results and experimental data in steady state conditions for some inlet pressure, steam and noncondensable gases (air) inlet mass fractions. The test section is three meters long and consists of two concentric tubes containing pressure, temperature and flow rate sensors. The internal tube, called condenser, contains steam-air mixture flow and external tube is a counter current cooler with water flow rate at low temperature. This test section was modeled and simulations were performed with RELAP5 code. Experimental tests were carried out for 200 to 400 kPa inlet pressure and 5, 10, 15 and 20% of inlet air mass fractions. Comparisons between experimental data and simulation results are presented for 200 and 400 kPa pressure conditions and showed good agreement. However, for 400 kPa inlet steam pressure and inlet air mass fractions above 5%, the simulated temperatures are lower than the experimental data at the final third from the inlet condenser tube, indicating a code overestimation of heat transfer coefficient. New correlations for heat transfer coefficient in these steam-air conditions must be theoretical and experimentally studied and implemented in RELAP5 code for better representing the condensation phenomena. (author)

Assessment of CHF characteristics at subcooled conditions for the CANFLEX bundle

International Nuclear Information System (INIS)

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

2013-01-01

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

Inner tubes cutting method by electrical arc saw

International Nuclear Information System (INIS)

Thome, P.

1990-01-01

The research program deals on the definition of tools used for dismantling steam generator tubes bundle of PWR and on tool used for cutting pipes of great diameter by using the process of cutting by electrical arc saw. The remote tools are used for cutting by the interior pipes of contamined circuits [fr

Condensation heat transfer characteristics of R410A-oil mixture in 5 mm and 4 mm outside diameter horizontal microfin tubes

Energy Technology Data Exchange (ETDEWEB)

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

2010-10-15

Condensation heat transfer characteristics of R410A-oil mixture in 5 mm and 4 mm outside diameter horizontal microfin tubes were investigated experimentally. The experimental condensing temperature is 40 C, and nominal oil concentration range is from 0% to 5%. The test results indicate that the presence of oil deteriorates the heat transfer. The deterioration effect is negligible at nominal oil concentration of 1%, and becomes obvious with the increase of nominal oil concentration. At 5% nominal oil concentration, the heat transfer coefficient of R410A-oil mixture is found to have a maximum reduction of 25.1% and 23.8% for 5 mm and 4 mm tubes, respectively. The predictabilities of the existing condensation heat transfer correlations were verified with the experimental data, and Yu and Koyama correlation shows the best predictability. By replacing the pure refrigerant properties with the mixture's properties, Yu and Koyama correlation has a deviation of -15% to + 20% in predicting the local condensation heat transfer coefficient of R410A-oil mixture. (author)

Bubble-assisted film evaporation correlation for saline water at sub-atmospheric pressures in horizontal-tube evaporator

KAUST Repository

Shahzad, Muhammad Wakil

2013-01-01

In falling film evaporators, the overall heat transfer coefficient is controlled by film thickness, velocity, liquid properties and the temperature differential across the film layer. This article presents the heat transfer behavior for evaporative film boiling on horizontal tubes, but working at low pressures of 0.93-3.60 kPa (corresponding solution saturation temperatures of 279-300 K) as well as seawater salinity of 15,000 to 90,000 mg/l or ppm. Owing to a dearth of literature on film-boiling at these conditions, the article is motivated by the importance of evaporative film boiling in the desalination processes such as the multi-effect distillation (MED) or multi-stage flashing (MSF): It is observed that in addition to the above-mentioned parameters, evaporative heat transfer of seawater is affected by the emergence of micro-bubbles within the thin film layer, particularly when the liquid saturation temperatures drop below 298 K (3.1 kPa). Such micro bubbles are generated near to the tube wall surfaces and they enhanced the heat transfer by two or more folds when compared with the predictions of conventional evaporative film boiling. The appearance of micro-bubbles is attributed to the rapid increase in the specific volume of vapor, i.e., dv/dT, at low saturation temperature conditions. A new correlation is thus proposed in this article and it shows good agreement to the measured data with an experimental uncertainty of 8% and regression RMSE of 3.5%. © 2012 Elsevier Ltd. All rights reserved.

Evaluation of bundle duct interaction by out of pile compressive test of FBR bundles. FFTF type bundle

Energy Technology Data Exchange (ETDEWEB)

Tanaka, Kosuke; Yamamoto, Yuji; Nagamine, Tsuyoshi; Maeda, Koji [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center

2000-10-01

Bundle duct interaction (BDI) caused by expansion of fuel pin bundle becomes one of the main limiting factors for fuel life times. Then, it is important for the design of fast reactor fuel assembly to understand the BDI behavior in detail. In order to understand the BDI behavior, out of pile compressive tests were conducted for FFTF type bundle by use of X-ray CT equipment. In these compressive tests, two type bundles with different accuracy of initial wire position were conducted. The objective of this test is to evaluate the influence of the initial error from standard position of wire at the same axial position. The locations of the pins and the duct flats are analyzed from CT image data. Quantitative evaluation was performed at the CT image data and discussed the bundle deformation status under BDI condition. Following results are obtained. 1) The accuracy of initial wire position is strongly depends on the pin-to-duct contact behavior. In the case of bundle with large error from standard position, pin-to-duct contact is delayed. 2) The BDI mitigation of the bundle with small error from standard wire position is following: The elastic ovality is the dominant deformation in mild BDI condition, then the wire dispersion and pin dispersion are occurred in severe BDI condition. 3) The BDI mitigation of the bundle with large error from standard wire position is following: The elastic ovality and local bowing of pins with large error from standard wire position are occurred in mild BDI condition, then pin dispersion is occurred around pins with large error from standard wire position, finally wire dispersion is occurred in severe BDI condition. 4) The existence of pins with large error from standard wire position is effective to delay the pin-to-duct contact, but the existence of these pins is possible to contact of pin- to- pin. (author)

Nuclear power plant and apparatus for superheating steam

International Nuclear Information System (INIS)

Schluderberg, D.C.

1983-01-01

The invention consists of an apparatus for superheating steam, the apparatus comprising a horizontally disposed generally cylindrical elongate shell, inlet means in the shell for receiving steam, outlet means in the shell for discharching the steam, and a bundle of inclined tubes positioned in the flow path of the steam, each of the tubes having a length which is less than the diameter of the shell and opening into and extending in an upward direction from an outlet header to an inlet header, the inlet header beeing connected to a source of vapor, and the outlet header beeing connected to a condensate drain, characterised in that the test bundle comprises two banks of the tubes, the angle at which each of the tubes of one of the banks extends relative to a vertical longitudinal centerplane, the tubes of one of the banks terminate at and open into the inlet header, and the tubes of the other banks terminate at an open into another inlet header

Characterization of Dosimetry of the BMRR Horizontal Thimble Tubes and Broad Beam Facility.

Energy Technology Data Exchange (ETDEWEB)

Hu,J.P.; Reciniello, R.N.; Holden, N.E.

2008-05-05

The Brookhaven Medical Research Reactor was a 5 mega-watt, light-water cooled and heavy-graphite moderated research facility. It has two shutter-equipped treatment rooms, three horizontally extended thimble tubes, and an ex-core broad beam facility. The three experimental thimbles, or activation ports, external to the reactor tank were designed for several uses, including the investigations on diagnostic and therapeutic methods using radioactive isotopes of very short half-life, the analysis of radiation exposure on tissue-equivalent materials using a collimated neutron beam, and the evaluation of dose effects on biological cells to improve medical treatment. At the broad beam facility where the distribution of thermal neutrons was essential uniform, a wide variety of mammalian whole-body exposures were studied using animals such as burros or mice. Also studied at the broad beam were whole-body phantom experiments, involving the use of a neutron or photon beam streaming through a screen to obtain the flux spectrum suitable for dose analysis on the sugar-urea-water mixture, a tissue-equivalent material. Calculations of the flux and the dose at beam ports based on Monte Carlo particle-transport code were performed, and measurements conducted at the same tally locations were made using bare or cadmium-covered gold foils. Analytical results, which show good agreement with measurement data, are presented in the paper.

Analysis of fuel handling system for fuel bundle safety during station blackout in 500 MWe PHWR unit of India

International Nuclear Information System (INIS)

Madhuresh, R.; Nagarajan, R.; Jit, I.; Sanatkumar, A.

1996-01-01

Situations of Station Blackout (SBO) i.e. postulated concurrent unavailability of Class Ill and Class IV power, could arise for a long period, while on-power refuelling or other fuel handling operations are in progress with the hot irradiated fuel bundles being anywhere in the system from the Reactor Building to the Spent Fuel Storage Bay. The cooling provisions for these fuel bundles are diverse and specific to the various stages of fuel handling operations and are either on Class Ill or on Class II power with particular requirements of instrument air. Therefore, during SBO, due to the limited availability of Class II power and instrument air, it becomes difficult to maintain cooling to these fuel bundles. However, some minimal cooling is essential, to ensure the safety of the bundles. As discussed in the paper, safety of these fuel bundles in the system and/or for those lying in the liner tube region of the reactor end fitting is ensured, during SBO, by resorting to passive means like 'stay-put', 'gravity- fill', 'D 2 0- steaming' etc. for cooling the bundles. The paper also describes various consequences emanating from these cooling schemes. (author). 6 refs., 2 tabs., 8 figs

Indoor Solar Thermal Energy Saving Time with Phase Change Material in a Horizontal Shell and Finned-Tube Heat Exchanger

Directory of Open Access Journals (Sweden)

S. Paria

2015-01-01

Full Text Available An experimental as well as numerical investigation was conducted on the melting/solidification processes of a stationary phase change material (PCM in a shell around a finned-tube heat exchanger system. The PCM was stored in the horizontal annular space between a shell and finned-tube where distilled water was employed as the heat transfer fluid (HTF. The focus of this study was on the behavior of PCM for storage (charging or melting and removal (discharging or solidification, as well as the effect of flow rate on the charged and discharged solar thermal energy. The impact of the Reynolds number was determined and the results were compared with each other to reveal the changes in amount of stored thermal energy with the variation of heat transfer fluid flow rates. The results showed that, by increasing the Reynolds number from 1000 to 2000, the total melting time decreases by 58%. The process of solidification also will speed up with increasing Reynolds number in the discharging process. The results also indicated that the fluctuation of gradient temperature decreased and became smooth with increasing Reynolds number. As a result, by increasing the Reynolds number in the charging process, the theoretical efficiency rises.

Safety significance of steam generator tube degradation mechanisms

Energy Technology Data Exchange (ETDEWEB)

Roussel, G; Mignot, P [AIB-Vincotte Nuclear - AVN, Brussels (Belgium)

1991-07-01

Steam generator (SG) tube bundle is a part of the Reactor Coolant Pressure Boundary (RCPB): this means that its integrity must be maintained. However, operating experience shows various types of tube degradation to occur in the SG tubing, which may lead to SG tube leaks or SG tube ruptures and create a loss of primary system coolant through the SG, therefore providing a direct path to the environment outside the primary containment structure. In this paper, the major types of known SG tube degradations are described and analyzed in order to assess their safety significance with regard to SG tube integrity. In conclusion: The operational reliability and the safety of the PWR steam generator s requires a sufficient knowledge of the degradation mechanisms to determine the amount of degradation that a tube can withstand and the time that it may remain in operation. They also require the availability of inspection techniques to accurately detect and characterize the various degradations. The status of understanding of the major types of degradation summarized in this paper shows and justifies why efforts are being performed to improve the management of the steam generator tube defects.

The Preliminary Study for Numerical Computation of 37 Rod Bundle in CANDU Reactor

International Nuclear Information System (INIS)

Jeon, Yu Mi; Park, Joo Hwan

2010-09-01

A typical CANDU 6 fuel bundle consists of 37 fuel rods supported by two endplates and separated by spacer pads at various locations. In addition, the bearing pads are brazed to each outer fuel rod with the aim of reducing the contact area between the fuel bundle and the pressure tube. Although the recent progress of CFD methods has provided opportunities for computing the thermal-hydraulic phenomena inside of a fuel channel, it is yet impossible to reflect numerical computations on the detailed shape of rod bundle due to challenges with computing mesh and memory capacity. Hence, the previous studies conducted a numerical computation for smooth channels without considering spacers and bearing pads. But, it is well known that these components are an important factor to predict the pressure drop and heat transfer rate in a channel. In this study, the new computational method is proposed to solve complex geometry such as a fuel rod bundle. Before applying a solution to the problem of the 37 rod bundle, the validity and the accuracy of the method are tested by applying the method to simple geometry. The split channel method has been proposed with the aim of computing the fully shaped CANDU fuel channel with detailed components. The validity was tested by applying the method to the single channel problem. The average temperature have similar values for the considered two methods, while the local temperature shows a slight difference by the effect of conduction heat transfer in the solid region of a rod. Based on the present result, the calculation for the fully shaped 37-rod bundle is scheduled for future work

Experimental investigation on TBAB clathrate hydrate slurry flows in a horizontal tube: Forced convective heat transfer behaviors

Energy Technology Data Exchange (ETDEWEB)

Wenji, Song [Guangzhou Institute of Energy Conversion, CAS, No. 2 Nengyuan Road, Tianhe District, Guangzhou 510640 (China); Key Laboratory of Renewable Energy and Gas Hydrate, CAS, No. 2 Nengyuan Road, Tianhe District, Guangzhou 510640 (China); Graduate School of Chinese Academy of Sciences, Beijing 100039 (China); Rui, Xiao; Chong, Huang; Shihui, He; Kaijun, Dong; Ziping, Feng [Guangzhou Institute of Energy Conversion, CAS, No. 2 Nengyuan Road, Tianhe District, Guangzhou 510640 (China); Key Laboratory of Renewable Energy and Gas Hydrate, CAS, No. 2 Nengyuan Road, Tianhe District, Guangzhou 510640 (China)

2009-11-15

Tetra-n-butyl-ammonium bromide (TBAB) clathrate hydrate slurry (CHS) is one kind of secondary refrigerants, which is promising to be applied into air-conditioning or latent-heat transportation systems as a thermal storage or cold carrying medium for energy saving. It is a solid-liquid two phase mixture which is easy to produce and has high latent heat and good fluidity. In this paper, the heat transfer characteristics of TBAB slurry were investigated in a horizontal stainless steel tube under different solid mass fractions and flow velocities with constant heat flux. One velocity region of weakened heat transfer was found. Moreover, TBAB CHS was treated as a kind of Bingham fluids, and the influences of the solid particles, flow velocity and types of flow on the forced convective heat transfer coefficients of TBAB CHS were investigated. At last, criterial correlations of Nusselt number for laminar and turbulent flows in the form of power function were summarized, and the error with experimental results was within {+-}20%. (author)

Neutron radiography of irradiated zircaloy coupons of pressure tubes from PHWR`s

Energy Technology Data Exchange (ETDEWEB)

Gangotra, S; Ouseph, P M; Tamhane, A B; Singh, H N; Sahoo, K C [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.

1994-12-31

The Indian Pressurised Heavy Water Reactors (PHWR`s) are of CANDU type, consisting of 304 zircaloy-2 pressure tubes. These pressure tubes contain the fuel bundles, where the heat is generated and is removed by the heavy water flowing through these pressure tubes at high temperature and pressure. These pressure tubes are surrounded by the calandria tubes, and are separated from them by a pair of garter springs. Over a period of time, as a result of the irradiation creep and assisted by the displacement of the garter springs, the hot pressure tube may come in contact with the cold calandria tube. This would result in the hydrogen migrating to the cold contact location and formation of hydride blisters. These blisters could eventually rupture the pressure tube by the DHC (delayed hydrogen cracking) mechanism. 2 refs., 2 figs.

Numerical research of the swirling supersonic gas flows in the self-vacuuming vortex tube

Science.gov (United States)

Volov, V. T.; Lyaskin, A. S.

2018-03-01

This article presents the results of simulation for a special type of vortex tubes – self-vacuuming vortex tube (SVVT), for which extreme values of temperature separation and vacuum are realized. The main results of this study are the flow structure in the SVVT and energy loss estimations on oblique shock waves, gas friction, instant expansion and organization of vortex bundles in SVVT.

The WUW ML bundle detector A flow through detector for alpha-emitters

CERN Document Server

Wenzel, U; Lochny, M

1999-01-01

Using conventional laboratory ware, we designed and manufactured a flow through cell for monitoring alpha-bearing solutions. The cell consists of a bundle of thermoplastic, transparent tubes coated with a thin layer of the meltable scintillator MELTILEX sup T sup M at the inner surface. With appropriate energy windows set, the detector can suppress beta-particles to a great extent due to its geometrical dimensions. For pure alpha-solutions, the detection limits are 5 Bq/ml, for composite nuclide mixtures, the detector is capable to monitor the decontamination of medium active waste (<=10 sup 7 Bq/ml) down to 100 Bq alpha/g solution. At a throughput of 1 ml/s, the pressure build-up amounts to approx 2 bar. We have developed a quality control program to ensure the regularity of the individual bundle loops.

Dynamic subgrid scale model of large eddy simulation of cross bundle flows

International Nuclear Information System (INIS)

Hassan, Y.A.; Barsamian, H.R.

1996-01-01

The dynamic subgrid scale closure model of Germano et. al (1991) is used in the large eddy simulation code GUST for incompressible isothermal flows. Tube bundle geometries of staggered and non-staggered arrays are considered in deep bundle simulations. The advantage of the dynamic subgrid scale model is the exclusion of an input model coefficient. The model coefficient is evaluated dynamically for each nodal location in the flow domain. Dynamic subgrid scale results are obtained in the form of power spectral densities and flow visualization of turbulent characteristics. Comparisons are performed among the dynamic subgrid scale model, the Smagorinsky eddy viscosity model (that is used as the base model for the dynamic subgrid scale model) and available experimental data. Spectral results of the dynamic subgrid scale model correlate better with experimental data. Satisfactory turbulence characteristics are observed through flow visualization

Relative desorption of boiling crisis in rod bundles

International Nuclear Information System (INIS)

Bobkov, V.P.

1997-01-01

Results of describing critical heat fluxes rod bundles are presented on base of applying a generalization of the available massive of data on CHF in spherical tubes, performed on the base of a new model, developed by the physics and Power Institute specialists, as well as on the base of results of analysing comprehensive experimental material accumulated in the data bank of the Thermophysical Data Center of the PPI Ratios, allowing one to predict the values of the critical heat flux in a wide range of mode and geometry parameters under energy release with cross section variations and cross section geometry distortion are presented

Confocal microlaparoscope for imaging the fallopian tube

Science.gov (United States)

Wu, Tzu-Yu; Rouse, Andrew R.; Chambers, Setsuko K.; Hatch, Kenneth D.; Gmitro, Arthur F.

2014-11-01

Recent evidence suggests that ovarian cancer can originate in the fallopian tube. Unlike many other cancers, poor access to the ovary and fallopian tubes has limited the ability to study the progression of this deadly disease and to diagnosis it during the early stage when it is most amenable to therapy. A rigid confocal microlaparoscope system designed to image the epithelial surface of the ovary in vivo was previously reported. A new confocal microlaparoscope with an articulating distal tip has been developed to enable in vivo access to human fallopian tubes. The new microlaparoscope is compatible with 5-mm trocars and includes a 2.2-mm-diameter articulating distal tip consisting of a bare fiber bundle and an automated dye delivery system for fluorescence confocal imaging. This small articulating device should enable the confocal microlaparoscope to image early stage ovarian cancer arising inside the fallopian tube. Ex vivo images of animal tissue and human fallopian tube using the new articulating device are presented along with in vivo imaging results using the rigid confocal microlaparoscope system.

Bundle Branch Block

Science.gov (United States)

... known cause. Causes can include: Left bundle branch block Heart attacks (myocardial infarction) Thickened, stiffened or weakened ... myocarditis) High blood pressure (hypertension) Right bundle branch block A heart abnormality that's present at birth (congenital) — ...

A vapor generator equipped with an advanced drain device for the secondary side of the tubes plate

International Nuclear Information System (INIS)

Valadon, C.

1995-01-01

A draining design is proposed for the tube plate secondary side in a PWR type reactor, that does not interfere with the water flush 'street' thus allowing for an easy inspection and maintenance in the lower part of the tube bundle. The draining system is composed of a main groove on the upper side of the tube plate, which is connected to draining means situated outside the vapor generator. 6 fig