Comparison of strongly heat-driven flow codes for unsaturated media
International Nuclear Information System (INIS)
Updegraff, C.D.
1989-08-01
Under the sponsorship of the US Nuclear Regulatory Commission, Sandia National Laboratories (SNL) is developing a performance assessment methodology for the analysis of long-term disposal of high-level radioactive waste (HLW) in unsaturated welded tuff. As part of this effort, SNL evaluated existing strongly heat-driven flow computer codes for simulating ground-water flow in unsaturated media. The three codes tested, NORIA, PETROS, and TOUGH, were compared against a suite of problems for which analytical and numerical solutions or experimental results exist. The problems were selected to test the abilities of the codes to simulate situations ranging from simple, uncoupled processes, such as two-phase flow or heat transfer, to fully coupled processes, such as vaporization caused by high temperatures. In general, all three codes were found to be difficult to use because of (1) built-in time stepping criteria, (2) the treatment of boundary conditions, and (3) handling of evaporation/condensation problems. A drawback of the study was that adequate problems related to expected repository conditions were not available in the literature. Nevertheless, the results of this study suggest the need for thorough investigations of the impact of heat on the flow field in the vicinity of an unsaturated HLW repository. Recommendations are to develop a new flow code combining the best features of these three codes and eliminating the worst ones. 19 refs., 49 figs
Method for calculating internal radiation and ventilation with the ADINAT heat-flow code
International Nuclear Information System (INIS)
Butkovich, T.R.; Montan, D.N.
1980-01-01
One objective of the spent fuel test in Climax Stock granite (SFTC) is to correctly model the thermal transport, and the changes in the stress field and accompanying displacements from the application of the thermal loads. We have chosen the ADINA and ADINAT finite element codes to do these calculations. ADINAT is a heat transfer code compatible to the ADINA displacement and stress analysis code. The heat flow problem encountered at SFTC requires a code with conduction, radiation, and ventilation capabilities, which the present version of ADINAT does not have. We have devised a method for calculating internal radiation and ventilation with the ADINAT code. This method effectively reproduces the results from the TRUMP multi-dimensional finite difference code, which correctly models radiative heat transport between drift surfaces, conductive and convective thermal transport to and through air in the drifts, and mass flow of air in the drifts. The temperature histories for each node in the finite element mesh calculated with ADINAT using this method can be used directly in the ADINA thermal-mechanical calculation
Wang, Qunzhen; Mathias, Edward C.; Heman, Joe R.; Smith, Cory W.
2000-01-01
A new, thermal-flow simulation code, called SFLOW. has been developed to model the gas dynamics, heat transfer, as well as O-ring and flow path erosion inside the space shuttle solid rocket motor joints by combining SINDA/Glo, a commercial thermal analyzer. and SHARPO, a general-purpose CFD code developed at Thiokol Propulsion. SHARP was modified so that friction, heat transfer, mass addition, as well as minor losses in one-dimensional flow can be taken into account. The pressure, temperature and velocity of the combustion gas in the leak paths are calculated in SHARP by solving the time-dependent Navier-Stokes equations while the heat conduction in the solid is modeled by SINDA/G. The two codes are coupled by the heat flux at the solid-gas interface. A few test cases are presented and the results from SFLOW agree very well with the exact solutions or experimental data. These cases include Fanno flow where friction is important, Rayleigh flow where heat transfer between gas and solid is important, flow with mass addition due to the erosion of the solid wall, a transient volume venting process, as well as some transient one-dimensional flows with analytical solutions. In addition, SFLOW is applied to model the RSRM nozzle joint 4 subscale hot-flow tests and the predicted pressures, temperatures (both gas and solid), and O-ring erosions agree well with the experimental data. It was also found that the heat transfer between gas and solid has a major effect on the pressures and temperatures of the fill bottles in the RSRM nozzle joint 4 configuration No. 8 test.
Computer code TOBUNRAD for PWR fuel bundle heat-up calculations
International Nuclear Information System (INIS)
Shimooke, Takanori; Yoshida, Kazuo
1979-05-01
The computer code TOBUNRAD developed is for analysis of ''fuel-bundle'' heat-up phenomena in a loss-of-coolant accident of PWR. The fuel bundle consists of fuel pins in square lattice; its behavior is different from that of individual pins during heat-up. The code is based on the existing TOODEE2 code which analyzes heat-up phenomena of single fuel pins, so that the basic models of heat conduction and transfer and coolant flow are the same as the TOODEE2's. In addition to the TOODEE2 features, unheated rods are modeled and radiation heat loss is considered between fuel pins, a fuel pin and other heat sinks. The TOBUNRAD code is developed by a new FORTRAN technique which makes it possible to interrupt a flow of program controls wherever desired, thereby attaching several subprograms to the main code. Users' manual for TOBUNRAD is presented: The basic program-structure by interruption method, physical and computational model in each sub-code, usage of the code and sample problems. (author)
Predicting critical heat flux in slug flow regime of uniformly heated ...
African Journals Online (AJOL)
Numerical computation code (PWR-DNBP) has been developed to predict Critical Heat Flux (CHF) of forced convective flow of water in a vertical heated channel. The code was based on the liquid sub-layer model, with the assumption that CHF occurred when the liquid film thickness between the heated surface and vapour ...
International Nuclear Information System (INIS)
Horak, W.C.; Guppy, J.G.
1984-01-01
The Super System Code (SSC) was developed at the Brookhaven National Laboratory (BNL) for the thermal hydraulic analysis of natural circulation transients, operational transients, and other system wide transients in nuclear power plants. SSC is a generic, best estimate code that models the in-vessel components, heat transport loops, plant protection systems and plant control systems. SSC also simulates the balance of plant when interfaced with the MINET code. SSC has been validated against both numerical and experimental data bases and is now used by several outside users. An important area of interest in LMFBR transient analysis is the prediction of the response of the reactor core under low flow conditions, such as experienced during a natural circulation event. Under these circumstances there are many physical phenomena which must be modeled to provide an adequate representation by a computer code simulation. The present version of SSC contains numerous models which account for most of the major phenomena. However, one area where the present model in SSC is being improved is in the representation of heat transfer and buoyancy effects under low flow operation. To properly improve the present version, the addition of models to represent certain inter-assembly effects is required
Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System
Energy Technology Data Exchange (ETDEWEB)
Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong [Kookmin Univ., Seoul (Korea, Republic of)
2007-03-15
The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow.
Development of 3-D Flow Analysis Code for Fuel Assembly using Unstructured Grid System
International Nuclear Information System (INIS)
Myong, Hyon Kook; Kim, Jong Eun; Ahn, Jong Ki; Yang, Seung Yong
2007-03-01
The flow through a nuclear rod bundle with mixing vanes are very complex and required a suitable turbulence model to be predicted accurately. Final objective of this study is to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system. In order to develop a CFD code for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system, the following researches are made: - Development of numerical algorithm for CFD code's solver - Grid and geometric connectivity data - Development of software(PowerCFD code) for fluid flow and heat transfer analysis in a nuclear fuel assembly using unstructured grid system - Modulation of software(PowerCFD code) - Development of turbulence model - Development of analysis module of RANS/LES hybrid models - Analysis of turbulent flow and heat transfer - Basic study on LES analysis - Development of main frame on pre/post processors based on GUI - Algorithm for fully-developed flow
NASA Lewis steady-state heat pipe code users manual
International Nuclear Information System (INIS)
Tower, L.K.
1992-06-01
The NASA Lewis heat pipe code has been developed to predict the performance of heat pipes in the steady state. The code can be used as a design tool on a personal computer or, with a suitable calling routine, as a subroutine for a mainframe radiator code. A variety of wick structures, including a user input option, can be used. Heat pipes with multiple evaporators, condensers, and adiabatic sections in series and with wick structures that differ among sections can be modeled. Several working fluids can be chosen, including potassium, sodium, and lithium, for which the monomer-dimer equilibrium is considered. The code incorporates a vapor flow algorithm that treats compressibility and axially varying heat input. This code facilitates the determination of heat pipe operating temperatures and heat pipe limits that may be encountered at the specified heat input and environment temperature. Data are input to the computer through a user-interactive input subroutine. Output, such as liquid and vapor pressures and temperatures, is printed at equally spaced axial positions along the pipe as determined by the user
NASA Lewis Steady-State Heat Pipe Code Architecture
Mi, Ye; Tower, Leonard K.
2013-01-01
NASA Glenn Research Center (GRC) has developed the LERCHP code. The PC-based LERCHP code can be used to predict the steady-state performance of heat pipes, including the determination of operating temperature and operating limits which might be encountered under specified conditions. The code contains a vapor flow algorithm which incorporates vapor compressibility and axially varying heat input. For the liquid flow in the wick, Darcy s formula is employed. Thermal boundary conditions and geometric structures can be defined through an interactive input interface. A variety of fluid and material options as well as user defined options can be chosen for the working fluid, wick, and pipe materials. This report documents the current effort at GRC to update the LERCHP code for operating in a Microsoft Windows (Microsoft Corporation) environment. A detailed analysis of the model is presented. The programming architecture for the numerical calculations is explained and flowcharts of the key subroutines are given
International Nuclear Information System (INIS)
Tsang, C.F.
1986-01-01
A summary is given of the authors recent studies in the verification, validation and application of a coupled heat and fluid flow code. Verification has been done against eight analytic and semi-analytic solutions. These solutions include those involving thermal buoyancy flow and fracture flow. Comprehensive field validation studies over a period of four years are discussed. The studies are divided into three stages: (1) history matching, (2) double-blind prediction and confirmation, (3) design optimization. At each stage, parameter sensitivity studies are performed. To study the applications of mathematical models, a problem proposed by the International Energy Agency (IEA) is solved using this verified and validated numerical model as well as two simpler models. One of the simpler models is a semi-analytic method assuming the uncoupling of the heat and fluid flow processes. The other is a graphical method based on a large number of approximations. Variations are added to the basic IEA problem to point out the limits of ranges of applications of each model. A number of lessons are learned from the above investigations. These are listed and discussed
CFD analysis on heat transfer in low Prandtl number fluid flows
Energy Technology Data Exchange (ETDEWEB)
Borgohain, A.; Maheshwari, N.K.; Vijayan, P.K.; Sinha, R.K., E-mail: bananta@barc.gov.in [Bhabha Atomic Research Centre, Reactor Engineering Div., Trombay, Mumbai (India)
2011-07-01
Use of Computational Fluid Dynamics (CFD) code is helpful for designing liquid metal cooled nuclear reactor systems. Before using any CFD code proper evaluation of the code is essential for simulation of heat transfer in liquid metal flow. In this paper, a review of the literature on the correlations for liquid metal heat transfer is carried out and a comparison with experimental results is performed. CFD analysis is carried out using PHOENICS-3.6 code on heat transfer in molten Lead Bismuth Eutectic (LBE) flowing through tube. Turbulent flow analyses are carried out for the evaluation of the CFD code. The CFD results are compared with the available correlations. Assessment of various turbulence models and correlations for turbulent Prandtl number in the tube geometry are carried out. From the analysis it is found that, the CFD prediction can be improved with modified turbulent Prandtl number in the turbulence models. (author)
FEHMN 1.0: Finite element heat and mass transfer code
International Nuclear Information System (INIS)
Zyvoloski, G.; Dash, Z.; Kelkar, S.
1991-04-01
A computer code is described which can simulate non-isothermal multiphase multicomponent flow in porous media. It is applicable to natural-state studies of geothermal systems and ground-water flow. The equations of heat and mass transfer for multiphase flow in porous and permeable media are solved using the finite element method. The permeability and porosity of the medium are allowed to depend on pressure and temperature. The code also has provisions for movable air and water phases and noncoupled tracers; that is, tracer solutions that do not affect the heat and mass transfer solutions. The tracers can be passive or reactive. The code can simulate two-dimensional, two-dimensional radial, or three-dimensional geometries. A summary of the equations in the model and the numerical solution procedure are provided in this report. A user's guide and sample problems are also included. The main use of FEHMN will be to assist in the understanding of flow fields in the saturated zone below the proposed Yucca Mountain Repository. 33 refs., 27 figs., 12 tabs
Flow-excursion-induced dryout at low-heat-flux
International Nuclear Information System (INIS)
Khatib-Rahbar, M.; Cazzoli, E.G.
1983-01-01
Flow-excursion-induced dryout at low-heat-flux natural-convection boiling, typical of liquid-metal fast-breeder reactors, is addressed. Steady-state calculations indicate that low-quality boiling is possible up to the point of Ledinegg instability leading to flow excursion and subsequent dryout in agreement with experimental data. A flow-regime-dependent dryout heat flux relationship based upon saturated boiling criterion is also presented. Transient analysis indicates that premature flow excursion can not be ruled out and sodium boiling is highly transient dependent. Analysis of a high-heat-flux forced convection, loss-of-flow transient shows a significantly faster flow excursion leading to dryout in excellent agreement with parallel calculations using the two-dimensional THORAX code. 17 figures
Analytical model for bottom reflooding heat transfer in light water reactors (the UCFLOOD code)
International Nuclear Information System (INIS)
Arrieta, L.; Yadigaroglu, G.
1978-08-01
The UCFLOOD code is based on mechanistic models developed to analyze bottom reflooding of a single flow channel and its associated fuel rod, or a tubular test section with internal flow. From the hydrodynamic point of view the flow channel is divided into a single-phase liquid region, a continuous-liquid two-phase region, and a dispersed-liquid region. The void fraction is obtained from drift flux models. For heat transfer calculations, the channel is divided into regions of single-phase-liquid heat transfer, nucleate boiling and forced-convection vaporization, inverted-annular film boiling, and dispersed-flow film boiling. The heat transfer coefficients are functions of the local flow conditions. Good agreement of calculated and experimental results has been obtained. A code user's manual is appended
Analysis of natural convection heat transfer and flows in internally heated stratified liquid pools
International Nuclear Information System (INIS)
Gubaidullin, A.A. Jr.; Dinh, T.N.; Sehgal, B.R.
1999-01-01
In this paper, natural convection flows and heat transfer in a liquid pool, with two superposed immiscible fluid layers, are analyzed. The objective of the study is to examine the effect of interfacial hydrodynamics and to develop a method which enables energy splitting to be evaluated in a stratified liquid pool. The thermal convection, with and without an internal heat source, in a rectangular cavity with different pairs of fluids was numerically simulated by a CFD code FLOW-3D. It was found that the code performs very well for prediction of heat transfer coefficients for different conditions. The hydrodynamic coupling between immiscible layers was found to have minor, if any, impact on the natural convection heat transfer for the conditions examined. Calculated results were used to develop, and validate, a new correlation for energy splitting and for heat transfer in stratified liquid pools
Investigation and assessment of wall heat transfer correlations in SPACE code
International Nuclear Information System (INIS)
Kim, Jung Woo; Kim, Kyung Doo; Moon, Sang Ki; Choi, Ki Yong; Park, Hyun Sik
2010-06-01
SPACE, which is a safety analysis code for nuclear power plants, has been developed to analyze the multidimensional, two-component and three-field flow. This code can be applied to safety analysis for approval which is thermal-hydraulic analysis to support the nuclear power station design, establishment of accident ease strategy, development of operating guide line, experiment plan and analysis. To do so, SPACE code has 12 wall heat transfer mode and the corresponding models and correlations to deal with the physical heat transfer phenomenon in wall surface. In this report, the physical correlation models regarding the wall heat transfer are explained and their performance is assessed against several SET
Development of an integral computer code for simulation of heat exchangers
International Nuclear Information System (INIS)
Horvat, A.; Catton, I.
2001-01-01
Heat exchangers are one of the basic installations in power and process industries. The present guidelines provide an ad-hoc solution to certain design problems. A unified approach based on simultaneous modeling of thermal-hydraulics and structural behavior does not exist. The present paper describes the development of integral numerical code for simulation of heat exchangers. The code is based on Volume Averaging Technique (VAT) for porous media flow modeling. The calculated values of the whole-section drag and heat transfer coefficients show an excellent agreement with already published values. The matching results prove the correctness of the selected approach and verify the developed numerical code used for this calculation.(author)
Numerical simulation on coolant flow and heat transfer in core
International Nuclear Information System (INIS)
Yao Zhaohui; Wang Xuefang; Shen Mengyu
1997-01-01
To simulate the coolant flow and the heat transfer characteristics of a core, a computer code, THAPMA (Thermal Hydraulic Analysis Porous Medium Analysis) has been developed. In THAPMA code, conservation equations are based on a porous-medium formulation, which uses four parameters, i.e, volume porosity, directional surface porosity, distributed resistance, and distributed heat source (sink), to model the effects of fuel rods and other internal solid structures on flow and heat transfer. Because the scheme and the solution are very important in accuracy and speed of calculation, a new difference scheme (WSUC) has been used in the energy equation, and a modified PISO solution method have been employed to simulate the steady/transient states. The code has been proved reliable and can effectively solve the transient state problem by several numerical tests. According to the design of Qinshan NPP-II, the flow and heat transfer phenomena in reactor core have been numerically simulated. The distributions of the velocity and the temperature can provide a theoretical basis for core design and safety analysis
Flowing and heat transfer characteristics of turbulent flow in typical rod bundles at rolling motion
International Nuclear Information System (INIS)
Yan Binghuo; Yu Lei; Gu Hanyang
2011-01-01
The influence mechanism of rolling motion on the flowing and heat transfer characteristics of turbulent flow in typical four rod bundles was investigated with Fluent code. The flowing and heat transfer characteristics of turbulent flow in rod bundles can be affected by rolling motion. But the flowing similarity of turbulent flow in adiabatic and non-adiabatic can not be affected. If the rolling period is small, the radial additional force can make the parameter profiles, the turbulent flowing and heat transfer change greatly. At rolling motion, as the pitch to diameter ratio decreases, especially if it is less than 1.1, the flowing and heat transfer of turbulent flow at rolling motion change significantly. The variation of pitch to diameter ratio can change the profiles of secondary flow and turbulent kinetic energy in cross-section greatly. (authors)
Gas flow environmental and heat transfer nonrotating 3D program
Geil, T.; Steinhoff, J.
1983-01-01
A complete set of benchmark quality data for the flow and heat transfer within a large rectangular turning duct is being compiled. These data will be used to evaluate and verify three dimensional internal viscous flow models and computational codes. The analytical objective is to select such a computational code and define the capabilities of this code to predict the experimental results. Details of the proper code operation will be defined and improvements to the code modeling capabilities will be formulated.
Computation of turbulent flow and heat transfer in subassemblies
International Nuclear Information System (INIS)
Slagter, W.
1979-01-01
This research is carried out in order to provide information on the thermohydraulic behaviour of fast reactor subassemblies. The research work involves the development of versatile computation methods and the evaluation of combined theoretical and experimental work on fluid flow and heat transfer in fuel rod bundles. The computation method described here rests on the application of the distributed parameter approach. The conditions considered cover steady, turbulent flow and heat transfer of incompressible fluids in bundles of bare rods. Throughout 1978 main efforts were given to the development of the VITESSE program and to the validation of the hydrodynamic part of the code. In its present version the VITESSE program is applicable to predict the fully developed turbulent flow and heat transfer in the subchannels of a bundle with bare rods. In this paper the main features of the code are described as well as the present status of development
International Nuclear Information System (INIS)
Grange, J.L.
1996-09-01
A simplified model for heat and mass transfer in the lower rainfall of a counter-flow cooling toward had to be implemented in the N3S code-cooling tower release It is built from an old code: ZOPLU. The air velocity field is calculated by N3S. The air and water temperature fields are solved by a Runge-Kutta method on a mesh in an adequate number of vertical plans. Heat exchange and drags correlations are given. And all the necessary parameters are specified. All the subroutines are described. They are taken from ZOPLU and modified in order to adapt their abilities to the N3S requirements. (author). 6 refs., 3 figs., 3 tabs., 3 appends
Conjugate heat transfer for turbulent flow in a thick walled plain pipe
Directory of Open Access Journals (Sweden)
Canli Eyub
2018-01-01
Full Text Available Laminar and turbulent flow have their own characteristics in respect of heat transfer in pipes. While conjugate heat transfer is a major concern for a thick walled pipe with laminar flow inside it, there are limited studies about a turbulent flow in a thick walled plain pipe considering the conjugate heat transfer. In order to conduct such a work by means of in-house developed code, it was desired to make a preliminary investigation with commercially available CFD codes. ANSYS CFD was selected as the tool since it has a positive reputation in the literature for reliability. Defined heat transfer problem was solved with SIMPLE and Coupled Schemes for pressure velocity coupling and results are presented accordingly.
User's Manual for the FEHM Application-A Finite-Element Heat- and Mass-Transfer Code
Energy Technology Data Exchange (ETDEWEB)
George A. Zyvoloski; Bruce A. Robinson; Zora V. Dash; Lynn L. Trease
1997-07-07
This document is a manual for the use of the FEHM application, a finite-element heat- and mass-transfer computer code that can simulate nonisothermal multiphase multicomponent flow in porous media. The use of this code is applicable to natural-state studies of geothermal systems and groundwater flow. A primary use of the FEHM application will be to assist in the understanding of flow fields and mass transport in the saturated and unsaturated zones below the proposed Yucca Mountain nuclear waste repository in Nevada. The equations of heat and mass transfer for multiphase flow in porous and permeable media are solved in the FEHM application by using the finite-element method. The permeability and porosity of the medium are allowed to depend on pressure and temperature. The code also has provisions for movable air and water phases and noncoupled tracers; that is, tracer solutions that do not affect the heat- and mass-transfer solutions. The tracers can be passive or reactive. The code can simulate two-dimensional, two-dimensional radial, or three-dimensional geometries. In fact, FEHM is capable of describing flow that is dominated in many areas by fracture and fault flow, including the inherently three-dimensional flow that results from permeation to and from faults and fractures. The code can handle coupled heat and mass-transfer effects, such as boiling, dryout, and condensation that can occur in the near-field region surrounding the potential repository and the natural convection that occurs through Yucca Mountain due to seasonal temperature changes. The code is also capable of incorporating the various adsorption mechanisms, ranging from simple linear relations to nonlinear isotherms, needed to describe the very complex transport processes at Yucca Mountain. This report outlines the uses and capabilities of the FEHM application, initialization of code variables, restart procedures, and error processing. The report describes all the data files, the input data
FEHM, Finite Element Heat and Mass Transfer Code
International Nuclear Information System (INIS)
Zyvoloski, G.A.
2002-01-01
1 - Description of program or function: FEHM is a numerical simulation code for subsurface transport processes. It models 3-D, time-dependent, multiphase, multicomponent, non-isothermal, reactive flow through porous and fractured media. It can accurately represent complex 3-D geologic media and structures and their effects on subsurface flow and transport. Its capabilities include flow of gas, water, and heat; flow of air, water, and heat; multiple chemically reactive and sorbing tracers; finite element/finite volume formulation; coupled stress module; saturated and unsaturated media; and double porosity and double porosity/double permeability capabilities. 2 - Methods: FEHM uses a preconditioned conjugate gradient solution of coupled linear equations and a fully implicit, fully coupled Newton Raphson solution of nonlinear equations. It has the capability of simulating transport using either a advection/diffusion solution or a particle tracking method. 3 - Restriction on the complexity of the problem: Disk space and machine memory are the only limitations
Wang, C. R.; Towne, C. E.; Hippensteele, S. A.; Poinsatte, P. E.
1997-01-01
This study investigated the Navier-Stokes computations of the surface heat transfer coefficients of a transition duct flow. A transition duct from an axisymmetric cross section to a non-axisymmetric cross section, is usually used to connect the turbine exit to the nozzle. As the gas turbine inlet temperature increases, the transition duct is subjected to the high temperature at the gas turbine exit. The transition duct flow has combined development of hydraulic and thermal entry length. The design of the transition duct required accurate surface heat transfer coefficients. The Navier-Stokes computational method could be used to predict the surface heat transfer coefficients of a transition duct flow. The Proteus three-dimensional Navier-Stokes numerical computational code was used in this study. The code was first studied for the computations of the turbulent developing flow properties within a circular duct and a square duct. The code was then used to compute the turbulent flow properties of a transition duct flow. The computational results of the surface pressure, the skin friction factor, and the surface heat transfer coefficient were described and compared with their values obtained from theoretical analyses or experiments. The comparison showed that the Navier-Stokes computation could predict approximately the surface heat transfer coefficients of a transition duct flow.
Fluid dynamics and heat transfer methods for the TRAC code
International Nuclear Information System (INIS)
Reed, W.H.; Kirchner, W.L.
1977-01-01
A computer code called TRAC is being developed for analysis of loss-of-coolant accidents and other transients in light water reactors. This code involves a detailed, multidimensional description of two-phase flow coupled implicitly through appropriate heat transfer coefficients with a simulation of the temperature field in fuel and structural material. Because TRAC utilizes about 1000 fluid mesh cells to describe an LWR system, whereas existing lumped parameter codes typically involve fewer than 100 fluid cells, new highly implicit difference techniques are developed that yield acceptable computing times on modern computers. Several test problems for which experimental data are available, including blowdown of single pipe and loop configurations with and without heated walls, have been computed with TRAC. Excellent agreement with experimental results has been obtained
User's manual for the FEHM application - A finite-element heat- and mass-transfer code
International Nuclear Information System (INIS)
Zyvoloski, G.A.; Robinson, B.A.; Dash, Z.V.; Trease, L.L.
1997-07-01
The use of this code is applicable to natural-state studies of geothermal systems and groundwater flow. A primary use of the FEHM application will be to assist in the understanding of flow fields and mass transport in the saturated and unsaturated zones below the proposed Yucca Mountain nuclear waste repository in Nevada. The equations of heat and mass transfer for multiphase flow in porous and permeable media are solved in the FEHM application by using the finite-element method. The permeability and porosity of the medium are allowed to depend on pressure and temperature. The code also has provisions for movable air and water phases and noncoupled tracers; that is, tracer solutions that do not affect the heat- and mass-transfer solutions. The tracers can be passive or reactive. The code can simulate two-dimensional, two-dimensional radial, or three-dimensional geometries. In fact, FEHM is capable of describing flow that is dominated in many areas by fracture and fault flow, including the inherently three-dimensional flow that results from permeation to and from faults and fractures. The code can handle coupled heat and mass-transfer effects, such as boiling, dryout, and condensation that can occur in the near-field region surrounding the potential repository and the natural convection that occurs through Yucca Mountain due to seasonal temperature changes. This report outlines the uses and capabilities of the FEHM application, initialization of code variables, restart procedures, and error processing. The report describes all the data files, the input data, including individual input records or parameters, and the various output files. The system interface is described, including the software environment and installation instructions
Study of flow control by localized volume heating in hypersonic boundary layers
Keller, M. A.; Kloker, M. J.; Kirilovskiy, S. V.; Polivanov, P. A.; Sidorenko, A. A.; Maslov, A. A.
2014-12-01
Boundary-layer flow control is a prerequisite for a safe and efficient operation of future hypersonic transport systems. Here, the influence of an electric discharge—modeled by a heat-source term in the energy equation—on laminar boundary-layer flows over a flat plate with zero pressure gradient at Mach 3, 5, and 7 is investigated numerically. The aim was to appraise the potential of electro-gasdynamic devices for an application as turbulence generators in the super- and hypersonic flow regime. The results with localized heat-source elements in boundary layers are compared to cases with roughness elements serving as classical passive trips. The numerical simulations are performed using the commercial code ANSYS FLUENT (by ITAM) and the high-order finite-difference DNS code NS3D (by IAG), the latter allowing for the detailed analysis of laminar flow instability. For the investigated setups with steady heating, transition to turbulence is not observed, due to the Reynolds-number lowering effect of heating.
FLASH: A finite element computer code for variably saturated flow
International Nuclear Information System (INIS)
Baca, R.G.; Magnuson, S.O.
1992-05-01
A numerical model was developed for use in performance assessment studies at the INEL. The numerical model, referred to as the FLASH computer code, is designed to simulate two-dimensional fluid flow in fractured-porous media. The code is specifically designed to model variably saturated flow in an arid site vadose zone and saturated flow in an unconfined aquifer. In addition, the code also has the capability to simulate heat conduction in the vadose zone. This report presents the following: description of the conceptual frame-work and mathematical theory; derivations of the finite element techniques and algorithms; computational examples that illustrate the capability of the code; and input instructions for the general use of the code. The FLASH computer code is aimed at providing environmental scientists at the INEL with a predictive tool for the subsurface water pathway. This numerical model is expected to be widely used in performance assessments for: (1) the Remedial Investigation/Feasibility Study process and (2) compliance studies required by the US Department of Energy Order 5820.2A
Simulation of heat and mass transfer in boiling water with the Melodif code
International Nuclear Information System (INIS)
Freydier, P.; Chen, O.; Olive, J.; Simonin, O.
1991-04-01
The Melodif code is developed at Electricite de France, Research and Development Division. It is an eulerian two dimensional code for the simulation of turbulent two phase flows (a three dimensional code derived from Melodif, ASTRID, is currently being prepared). Melodif is based on the two fluid model, solving the equations of conservation for mass, momentum and energy, for both phases. In such a two fluid model, the description of interfacial transfers between phases is a crucial issue. The model used applies to a dominant continuous phase, and a dispersed phase. A good description of interfacial momentum transfer exists in the standard MELODIF code: the drag force, the apparent mass force... are taken into account. An important factor for interfacial transfers is the interfacial area per volume unit. With the assumption of spherical gas bubbles, an equation has been written for this variable. In the present wok, a model has been tested for interfacial heat and mass transfer in the case of boiling water: it is assumed that mass transfer is controlled by heat transfer through the latent massic energy taken in the phase that vaporizes (or condenses). This heat and mass transfer model has been tested in various configurations: - a cylinder with water flowing inside, is being heated. Boiling takes place near the wall, while bubbles migrating to the core of the flow recondense. This roughly simulates a sub-cooled boiling phenomenon. - a box containing liquid water is depressurized. Boiling takes place in the whole volume of the fluid. The Melodif code can simulate this configuration due to the implicitation of the relation between interphase mass transfer and the pressure variable
Evaluation of Advanced Models for PAFS Condensation Heat Transfer in SPACE Code
Energy Technology Data Exchange (ETDEWEB)
Bae, Byoung-Uhn; Kim, Seok; Park, Yu-Sun; Kang, Kyung Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahn, Tae-Hwan; Yun, Byong-Jo [Pusan National University, Busan (Korea, Republic of)
2015-10-15
The PAFS (Passive Auxiliary Feedwater System) is operated by the natural circulation to remove the core decay heat through the PCHX (Passive Condensation Heat Exchanger) which is composed of the nearly horizontal tubes. For validation of the cooling and operational performance of the PAFS, PASCAL (PAFS Condensing Heat Removal Assessment Loop) facility was constructed and the condensation heat transfer and natural convection phenomena in the PAFS was experimentally investigated at KAERI (Korea Atomic Energy Research Institute). From the PASCAL experimental result, it was found that conventional system analysis code underestimated the condensation heat transfer. In this study, advanced condensation heat transfer models which can treat the heat transfer mechanisms with the different flow regimes in the nearly horizontal heat exchanger tube were analyzed. The models were implemented in a thermal hydraulic safety analysis code, SPACE (Safety and Performance Analysis Code for Nuclear Power Plant), and it was evaluated with the PASCAL experimental data. With an aim of enhancing the prediction capability for the condensation phenomenon inside the PCHX tube of the PAFS, advanced models for the condensation heat transfer were implemented into the wall condensation model of the SPACE code, so that the PASCAL experimental result was utilized to validate the condensation models. Calculation results showed that the improved model for the condensation heat transfer coefficient enhanced the prediction capability of the SPACE code. This result confirms that the mechanistic modeling for the film condensation in the steam phase and the convection in the condensate liquid contributed to enhance the prediction capability of the wall condensation model of the SPACE code and reduce conservatism in prediction of condensation heat transfer.
Fluid dynamics and heat transfer methods for the TRAC code
International Nuclear Information System (INIS)
Reed, W.H.; Kirchner, W.L.
1977-01-01
A computer code called TRAC is being developed for analysis of loss-of-coolant accidents and other transients in light water reactors. This code involves a detailed, multidimensional description of two-phase flow coupled implicitly through appropriate heat transfer coefficients with a simulation of the temperature field in fuel and structural material. Because TRAC utilizes about 1000 fluid mesh cells to describe an LWR system, whereas existing lumped parameter codes typically involve fewer than 100 fluid cells, we have developed new highly implicit difference techniques that yield acceptable computing times on modern computers. Several test problems for which experimental data are available, including blowdown of single pipe and loop configurations with and without heated walls, have been computed with TRAC. Excellent agreement with experimental results has been obtained. (author)
Preliminary Numerical Analysis of Convective Heat Transfer Loop Using MARS Code
Energy Technology Data Exchange (ETDEWEB)
Lee, Yongjae; Seo, Gwang Hyeok; Jeun, Gyoodong; Kim, Sung Joong [Hanyang Univ., Seoul (Korea, Republic of)
2014-05-15
The MARS has been developed adopting two major modules: RELAP5/MOD3 (USA) for one-dimensional (1D) two-fluid model for two-phase flows and COBRA-TF code for a three-dimensional (3D), two-fluid, and three-field model. In addition to the MARS code, TRACE (USA) is a modernized thermal-hydraulics code designed to consolidate and extend the capabilities of NRC's 3 legacy safety code: TRAC-P, TRAC-B and RELAP. CATHARE (French) is also thermal-hydraulic system analysis code for Pressurized Water Reactor (PWR) safety. There are several researches on comparing experimental data with simulation results by the MARS code. Kang et al. conducted natural convection heat transfer experiments of liquid gallium loop, and the experimental data were compared to MARS simulations. Bang et al. examined the capability of the MARS code to predict condensation heat transfer experiments with a vertical tube containing a non-condensable gas. Moreover, Lee et al. adopted MELCOR, which is one of the severe accident analysis codes, to evaluate several strategies for the severe accident mitigation. The objective of this study is to conduct the preliminary numerical analysis for the experimental loop at HYU using the MARS code, especially in order to provide relevant information on upcoming experiments for the undergraduate students. In this study, the preliminary numerical analysis for the convective heat transfer loop was carried out using the MARS Code. The major findings from the numerical simulations can be summarized as follows. In the calculations of the outlet and surface temperatures, the several limitations were suggested for the upcoming single-phase flow experiments. The comparison work for the HTCs shows validity for the prepared input model. This input could give useful information on the experiments. Furthermore, the undergraduate students in department of nuclear engineering, who are going to be taken part in the experiments, could prepare the program with the input, and will
Preliminary Numerical Analysis of Convective Heat Transfer Loop Using MARS Code
International Nuclear Information System (INIS)
Lee, Yongjae; Seo, Gwang Hyeok; Jeun, Gyoodong; Kim, Sung Joong
2014-01-01
The MARS has been developed adopting two major modules: RELAP5/MOD3 (USA) for one-dimensional (1D) two-fluid model for two-phase flows and COBRA-TF code for a three-dimensional (3D), two-fluid, and three-field model. In addition to the MARS code, TRACE (USA) is a modernized thermal-hydraulics code designed to consolidate and extend the capabilities of NRC's 3 legacy safety code: TRAC-P, TRAC-B and RELAP. CATHARE (French) is also thermal-hydraulic system analysis code for Pressurized Water Reactor (PWR) safety. There are several researches on comparing experimental data with simulation results by the MARS code. Kang et al. conducted natural convection heat transfer experiments of liquid gallium loop, and the experimental data were compared to MARS simulations. Bang et al. examined the capability of the MARS code to predict condensation heat transfer experiments with a vertical tube containing a non-condensable gas. Moreover, Lee et al. adopted MELCOR, which is one of the severe accident analysis codes, to evaluate several strategies for the severe accident mitigation. The objective of this study is to conduct the preliminary numerical analysis for the experimental loop at HYU using the MARS code, especially in order to provide relevant information on upcoming experiments for the undergraduate students. In this study, the preliminary numerical analysis for the convective heat transfer loop was carried out using the MARS Code. The major findings from the numerical simulations can be summarized as follows. In the calculations of the outlet and surface temperatures, the several limitations were suggested for the upcoming single-phase flow experiments. The comparison work for the HTCs shows validity for the prepared input model. This input could give useful information on the experiments. Furthermore, the undergraduate students in department of nuclear engineering, who are going to be taken part in the experiments, could prepare the program with the input, and will
User`s manual for the FEHM application -- A finite-element heat- and mass-transfer code
Energy Technology Data Exchange (ETDEWEB)
Zyvoloski, G.A.; Robinson, B.A.; Dash, Z.V.; Trease, L.L.
1997-07-01
The use of this code is applicable to natural-state studies of geothermal systems and groundwater flow. A primary use of the FEHM application will be to assist in the understanding of flow fields and mass transport in the saturated and unsaturated zones below the proposed Yucca Mountain nuclear waste repository in Nevada. The equations of heat and mass transfer for multiphase flow in porous and permeable media are solved in the FEHM application by using the finite-element method. The permeability and porosity of the medium are allowed to depend on pressure and temperature. The code also has provisions for movable air and water phases and noncoupled tracers; that is, tracer solutions that do not affect the heat- and mass-transfer solutions. The tracers can be passive or reactive. The code can simulate two-dimensional, two-dimensional radial, or three-dimensional geometries. In fact, FEHM is capable of describing flow that is dominated in many areas by fracture and fault flow, including the inherently three-dimensional flow that results from permeation to and from faults and fractures. The code can handle coupled heat and mass-transfer effects, such as boiling, dryout, and condensation that can occur in the near-field region surrounding the potential repository and the natural convection that occurs through Yucca Mountain due to seasonal temperature changes. This report outlines the uses and capabilities of the FEHM application, initialization of code variables, restart procedures, and error processing. The report describes all the data files, the input data, including individual input records or parameters, and the various output files. The system interface is described, including the software environment and installation instructions.
International Nuclear Information System (INIS)
Dershowitz, W; Herbert, A.; Long, J.
1989-03-01
The hydrology of the SCV site will be modelled utilizing discrete fracture flow models. These models are complex, and can not be fully cerified by comparison to analytical solutions. The best approach for verification of these codes is therefore cross-verification between different codes. This is complicated by the variation in assumptions and solution techniques utilized in different codes. Cross-verification procedures are defined which allow comparison of the codes developed by Harwell Laboratory, Lawrence Berkeley Laboratory, and Golder Associates Inc. Six cross-verification datasets are defined for deterministic and stochastic verification of geometric and flow features of the codes. Additional datasets for verification of transport features will be documented in a future report. (13 figs., 7 tabs., 10 refs.) (authors)
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
Simplified numerical model for predicting onset of flow instability in parallel heated channels
International Nuclear Information System (INIS)
Noura Rassoul; El-Khider Si-Ahmed; Tewfik Hamidouche; Anis Bousbia-Salah
2005-01-01
Full text of publication follows: Flow instabilities are undesirable phenomena in heated channels since change in flow rate affects the local heat transfer characteristics and may results in premature burnout. For instance, two-phase flow excursion (Ledinegg) instability in boiling channels is of great concern in the design and operation of numerous practical systems especially the MTR fuel type Research Reactors. For heated parallel channels, the negative-sloped segment of the pressure drop-flow rate characteristics (demand curve) of a boiling channel becomes negative. Such instability can lead to significant reduction in channel flow, thereby causing premature burnout of the heated channel before the CHF point. Furthermore, as a consequence of this flow decrease, different types of flow instabilities that may appear can also induce (density wave) flow oscillations of constant amplitude or diverging amplitude. The present work focuses on a numerical simulation of pressure drop in forced convection boiling in vertical narrow and parallel uniformly heated channels. The objective is to determine the point of Onset of flow instability by varying input flow rate without any consideration to density wave oscillations. By the way, the axial void distribution is provided. The numerical model is based on the finite difference method which transform the partial differential conservation equations of Mass, Momentum and Energy, in algebraic equations. Closure relationships as the drift flux model and other constitutive equations are considered to determine the channel pressure drop under steady state boiling conditions. The model validation is performed by confronting the calculations with the Oak Ridge National Laboratory Thermal Hydraulic Test Loop (THTL) experimental data set. Further verification of this model is performed by code-to code verification using the results of RELAP5/Mod 3.2 code. (authors)
TOUGH Simulations of the Updegraff's Set of Fluid and Heat Flow Problems
Energy Technology Data Exchange (ETDEWEB)
Moridis, G.J.; Pruess (editor), K.
1992-11-01
The TOUGH code [Pruess, 1987] for two-phase flow of water, air, and heat in penneable media has been exercised on a suite of test problems originally selected and simulated by C. D. Updegraff [1989]. These include five 'verification' problems for which analytical or numerical solutions are available, and three 'validation' problems that model laboratory fluid and heat flow experiments. All problems could be run without any code modifications (*). Good and efficient numerical performance, as well as accurate results were obtained throughout. Additional code verification and validation problems from the literature are briefly summarized, and suggestions are given for proper applications of TOUGH and related codes.
Solving implicit multi-mesh flow and conjugate heat transfer problems with RELAP-7
International Nuclear Information System (INIS)
Zou, L.; Peterson, J.; Zhao, H.; Zhang, H.; Andrs, D.; Martineau, R.
2013-01-01
The fully implicit simulation capability of RELAP-7 to solve multi-mesh flow and conjugate heat transfer problems for reactor system safety analysis is presented. Compared to general single-mesh simulations, the reactor system safety analysis-type of code has unique challenges due to its highly simplified, interconnected, one-dimensional, and zero-dimensional flow network describing multiple physics with significantly different time and length scales. To use the Jacobian-free Newton Krylov-type of solver, preconditioning is generally required for the Krylov method. The uniqueness of the reactor safety analysis-type of code in treating the interconnected flow network and conjugate heat transfer also introduces challenges in providing preconditioning matrix. Typical flow and conjugate heat transfer problems involved in reactor safety analysis using RELAP-7, as well as the special treatment on the preconditioning matrix are presented in detail. (authors)
Implementation and testing of the CFDS-FLOW3D code
International Nuclear Information System (INIS)
Smith, B.L.
1994-03-01
FLOW3D is a multi-purpose, transient fluid dynamics and heat transfer code developed by Computational Fluid Dynamics Services (CFDS), a branch of AEA Technology, based at Harwell. The code is supplied with a SUN-based operating environment consisting of an interactive grid generator SOPHIA and a post-processor JASPER for graphical display of results. Both SOPHIA and JASPER are extensions of the support software originally written for the ASTEC code, also promoted by CFDS. The latest release of FLOW3D contains well-tested turbulence and combustion models and, in a less-developed form, a multi-phase modelling potential. This document describes briefly the modelling capabilities of FLOW3D (Release 3.2) and outlines implementation procedures for the VAX, CRAY and CONVEX computer systems. Additional remarks are made concerning the in-house support programs which have been specially written in order to adapt existing ASTEC input data for use with FLOW3D; these programs operate within a VAX-VMS environment. Three sample calculations have been performed and results compared with those obtained previously using the ASTEC code, and checked against other available data, where appropriate. (author) 35 figs., 3 tabs., 42 refs
Directory of Open Access Journals (Sweden)
Taymaz Imdat
2015-01-01
Full Text Available The Lattice Boltzmann Method is applied to computationally investigate the laminar flow and heat transfer of an incompressible fluid with constant material properties in a two-dimensional channel with a built-in bluff body. In this study, a triangular prism is taken as the bluff body. Not only the momentum transport, but also the energy transport is modeled by the Lattice Boltzmann Method. A uniform lattice structure with a single time relaxation rule is used. For obtaining a higher flexibility on the computational grid, interpolation methods are applied, where the information is transferred from the lattice structure to the computational grid by Lagrange interpolation. The flow is investigated for different Reynolds numbers, while keeping the Prandtl number at the constant value of 0.7. The results show how the presence of a triangular prism effects the flow and heat transfer patterns for the steady-state and unsteady-periodic flow regimes. As an assessment of the accuracy of the developed Lattice Boltzmann code, the results are compared with those obtained by a commercial Computational Fluid Dynamics code. It is observed that the present Lattice Boltzmann code delivers results that are of similar accuracy to the well-established Computational Fluid Dynamics code, with much smaller computational time for the prediction of the unsteady phenomena.
Development of platform to compare different wall heat transfer packages for system analysis codes
International Nuclear Information System (INIS)
Kim, Min-Gil; Lee, Won Woong; Lee, Jeong Ik; Shin, Sung Gil
2016-01-01
System thermal hydraulic (STH) analysis code is used for analyzing and evaluating the safety of a designed nuclear system. The system thermal hydraulic analysis code typically solves mass, momentum and energy conservation equations for multiple phases with sets of selected empirical constitutive equations to close the problem. Several STH codes are utilized in academia, industry and regulators, such as MARS-KS, SPACE, RELAP5, COBRA-TF, TRACE, and so on. Each system thermal hydraulic code consists of different sets of governing equations and correlations. However, the packages and sets of correlations of each code are not compared quantitatively yet. Wall heat transfer mode transition maps of SPACE and MARS-KS have a little difference for the transition from wall nucleate heat transfer mode to wall film heat transfer mode. Both codes have the same heat transfer packages and correlations in most region except for wall film heat transfer mode. Most of heat transfer coefficients calculated for the range of selected variables of SPACE are the same with those of MARS-KS. For the intervals between 500K and 540K of wall temperature, MARS-KS selects the wall film heat transfer mode and Bromley correlation but SPACE select the wall nucleate heat transfer mode and Chen correlation. This is because the transition from nucleate boiling to film boiling of MARS-KS is earlier than SPACE. More detailed analysis of the heat transfer package and flow regime package will be followed in the near future
Conjugate Compressible Fluid Flow and Heat Transfer in Ducts
Cross, M. F.
2011-01-01
A computational approach to modeling transient, compressible fluid flow with heat transfer in long, narrow ducts is presented. The primary application of the model is for analyzing fluid flow and heat transfer in solid propellant rocket motor nozzle joints during motor start-up, but the approach is relevant to a wide range of analyses involving rapid pressurization and filling of ducts. Fluid flow is modeled through solution of the spatially one-dimensional, transient Euler equations. Source terms are included in the governing equations to account for the effects of wall friction and heat transfer. The equation solver is fully-implicit, thus providing greater flexibility than an explicit solver. This approach allows for resolution of pressure wave effects on the flow as well as for fast calculation of the steady-state solution when a quasi-steady approach is sufficient. Solution of the one-dimensional Euler equations with source terms significantly reduces computational run times compared to general purpose computational fluid dynamics packages solving the Navier-Stokes equations with resolved boundary layers. In addition, conjugate heat transfer is more readily implemented using the approach described in this paper than with most general purpose computational fluid dynamics packages. The compressible flow code has been integrated with a transient heat transfer solver to analyze heat transfer between the fluid and surrounding structure. Conjugate fluid flow and heat transfer solutions are presented. The author is unaware of any previous work available in the open literature which uses the same approach described in this paper.
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.
EFLOD code for reflood heat transfer
International Nuclear Information System (INIS)
Gay, R.R.
1979-01-01
A computer code called EFLOD has been developed for simulation of the heat transfer and hydrodynamics of a nuclear power reactor during the reflood phase of a loss-of-coolant accident. EFLOD models the downcomer, lower plenum, core, and upper plenum of a nuclear reactor vessel using seven control volumes assuming either homogeneous or unequal-velocity, unequal-temperature (UVUT) models of two-phase flow, depending on location within the vessel. The moving control volume concept in which a single control volume models the quench region in the core and moves with the core liquid level was developed and implemented in EFLOD so that three control volumes suffice to model the core region. A simplified UVUT model that assumes saturated liquid above the quench front was developed to handle the nonhomogeneous flow situation above the quench region. An explicit finite difference routine is used to model conduction heat transfer in the fuel, gap, and cladding regions of the fuel rod. In simulation of a selected FLECHT-SET experimental run, EFLOD successfully predicted the midplane maximum temperature and turnaround time as well as the time-dependent advance of the core liquid level. However, the rate of advancement of the quench level and the ensuing liquid entrainment were overpredicted during the early part of the transient
International Nuclear Information System (INIS)
Zyvoloski, G.A.; Robinson, B.A.; Dash, Z.V.; Trease, L.L.
1997-07-01
The mathematical models and numerical methods employed by the FEHM application, a finite-element heat- and mass-transfer computer code that can simulate nonisothermal multiphase multi-component flow in porous media, are described. The use of this code is applicable to natural-state studies of geothermal systems and groundwater flow. A primary use of the FEHM application will be to assist in the understanding of flow fields and mass transport in the saturated and unsaturated zones below the proposed Yucca Mountain nuclear waste repository in Nevada. The component models of FEHM are discussed. The first major component, Flow- and Energy-Transport Equations, deals with heat conduction; heat and mass transfer with pressure- and temperature-dependent properties, relative permeabilities and capillary pressures; isothermal air-water transport; and heat and mass transfer with noncondensible gas. The second component, Dual-Porosity and Double-Porosity/Double-Permeability Formulation, is designed for problems dominated by fracture flow. Another component, The Solute-Transport Models, includes both a reactive-transport model that simulates transport of multiple solutes with chemical reaction and a particle-tracking model. Finally, the component, Constitutive Relationships, deals with pressure- and temperature-dependent fluid/air/gas properties, relative permeabilities and capillary pressures, stress dependencies, and reactive and sorbing solutes. Each of these components is discussed in detail, including purpose, assumptions and limitations, derivation, applications, numerical method type, derivation of numerical model, location in the FEHM code flow, numerical stability and accuracy, and alternative approaches to modeling the component
GAM-HEAT -- a computer code to compute heat transfer in complex enclosures
International Nuclear Information System (INIS)
Cooper, R.E.; Taylor, J.R.; Kielpinski, A.L.; Steimke, J.L.
1991-02-01
The GAM-HEAT code was developed for heat transfer analyses associated with postulated Double Ended Guillotine Break Loss Of Coolant Accidents (DEGB LOCA) resulting in a drained reactor vessel. In these analyses the gamma radiation resulting from fission product decay constitutes the primary source of energy as a function of time. This energy is deposited into the various reactor components and is re- radiated as thermal energy. The code accounts for all radiant heat exchanges within and leaving the reactor enclosure. The SRS reactors constitute complex radiant exchange enclosures since there are many assemblies of various types within the primary enclosure and most of the assemblies themselves constitute enclosures. GAM-HEAT accounts for this complexity by processing externally generated view factors and connectivity matrices, and also accounts for convective, conductive, and advective heat exchanges. The code is applicable for many situations involving heat exchange between surfaces within a radiatively passive medium. The GAM-HEAT code has been exercised extensively for computing transient temperatures in SRS reactors with specific charges and control components. Results from these computations have been used to establish the need for and to evaluate hardware modifications designed to mitigate results of postulated accident scenarios, and to assist in the specification of safe reactor operating power limits. The code utilizes temperature dependence on material properties. The efficiency of the code has been enhanced by the use of an iterative equation solver. Verification of the code to date consists of comparisons with parallel efforts at Los Alamos National Laboratory and with similar efforts at Westinghouse Science and Technology Center in Pittsburgh, PA, and benchmarked using problems with known analytical or iterated solutions. All comparisons and tests yield results that indicate the GAM-HEAT code performs as intended
International Nuclear Information System (INIS)
Chen, K.F.; Olson, C.A.
1983-01-01
One reliable method that can be used to verify the solution scheme of a computer code is to compare the code prediction to a simplified problem for which an analytic solution can be derived. An analytic solution for the axial pressure drop as a function of the flow was obtained for the simplified problem of homogeneous equilibrium two-phase flow in a vertical, heated channel with a cosine axial heat flux shape. This analytic solution was then used to verify the predictions of the CONDOR computer code, which is used to evaluate the thermal-hydraulic performance of boiling water reactors. The results show excellent agreement between the analytic solution and CONDOR prediction
International Nuclear Information System (INIS)
Borges, Eduardo M.; Sabundjian, Gaiane
2015-01-01
Identifying the flow regimes and the heat transfer modes is important for the analysis of accidents such as the Loss-of-Coolant Accident (LOCA). The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used in the RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 100cm 2 -rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR - A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. (author)
International Nuclear Information System (INIS)
Bang, K. H.; Lee, J. Y.; Yoo, S. O.; Kim, M. W.; Kim, H. J.
2002-01-01
Three-dimensional analyses of fluid flow and heat transfer has been performed in this study. The simulation of SPEL experimental work and comparison with experimental data has been carried out to verify the analyses models. Moreover, to verify the CANDU-6 reactor type, analyses of fluid flow and heat transfer in the calandria under the condition of steady state has been performed using FLUENT code, which is the conventional code for a three-dimensional analyses of fluid flow and heat transfer for moderator integrity assessment in PHWR thermal-hydraulics. It is found that the maximum temperature in the moderator is 347K (74 ), so that the moderator has the enough subcoolability to ensure the integrity of pressure tube during LOCA conditions
Two-phase flow regimes and mechanisms of critical heat flux under subcooled flow boiling conditions
International Nuclear Information System (INIS)
Le Corre, Jean-Marie; Yao, Shi-Chune; Amon, Cristina H.
2010-01-01
A literature review of critical heat flux (CHF) experimental visualizations under subcooled flow boiling conditions was performed and systematically analyzed. Three major types of CHF flow regimes were identified (bubbly, vapor clot and slug flow regime) and a CHF flow regime map was developed, based on a dimensional analysis of the phenomena and available experimental information. It was found that for similar geometric characteristics and pressure, a Weber number (We)/thermodynamic quality (x) map can be used to predict the CHF flow regime. Based on the experimental observations and the review of the available CHF mechanistic models under subcooled flow boiling conditions, hypothetical CHF mechanisms were selected for each CHF flow regime, all based on a concept of wall dry spot overheating, rewetting prevention and subsequent dry spot spreading. Even though the selected concept has not received much attention (in term or theoretical developments and applications) as compared to other more popular DNB models, its basis have often been cited by experimental investigators and is considered by the authors as the 'most-likely' mechanism based on the literature review and analysis performed in this work. The selected modeling concept has the potential to span the CHF conditions from highly subcooled bubbly flow to early stage of annular flow and has been numerically implemented and validated in bubbly flow and coupled with one- and three-dimensional (CFD) two-phase flow codes, in a companion paper. [Le Corre, J.M., Yao, S.C., Amon, C.H., in this issue. A mechanistic model of critical heat flux under subcooled flow boiling conditions for application to one and three-dimensional computer codes. Nucl. Eng. Des.].
Effect of non-equilibrium flow chemistry and surface catalysis on surface heating to AFE
Stewart, David A.; Henline, William D.; Chen, Yih-Kanq
1991-01-01
The effect of nonequilibrium flow chemistry on the surface temperature distribution over the forebody heat shield on the Aeroassisted Flight Experiment (AFE) vehicle was investigated using a reacting boundary-layer code. Computations were performed by using boundary-layer-edge properties determined from global iterations between the boundary-layer code and flow field solutions from a viscous shock layer (VSL) and a full Navier-Stokes solution. Surface temperature distribution over the AFE heat shield was calculated for two flight conditions during a nominal AFE trajectory. This study indicates that the surface temperature distribution is sensitive to the nonequilibrium chemistry in the shock layer. Heating distributions over the AFE forebody calculated using nonequilibrium edge properties were similar to values calculated using the VSL program.
Plate heat exchanger - inertia flywheel performance in loss of flow transient
International Nuclear Information System (INIS)
Abou-El-Maaty, Talal; Abd-El-Hady, Amr
2009-01-01
One of the most versatile types of heat exchangers used is the plate heat exchanger. It has principal advantages over other heat exchangers in that plates can be added and/or removed easily in order to change the area available for heat transfer and therefore its overall performance. The cooling systems of Egypt's second research reactor (ETRR 2) use this type of heat exchanger for cooling purposes in its primary core cooling and pool cooling systems. In addition to the change in the number of heat exchanger cooling channels, the effect of changing the amount of mass flow rate on the heat exchanger performance is an important issues in this study. The inertia flywheel mounted on the primary core cooling system pump with the plate heat exchanger plays an important role in the case of loss of flow transients. The PARET code is used to simulate the effect of loss of flow transients on the reactor core. Hence, the core outlet temperature with the pump-flywheel flow coast down is fed into the plate heat exchanger model developed to estimate the total energy transferred to the cooling tower, the primary side heat exchanger temperature variation, the transmitted heat exchanger power, and the heat exchanger effectiveness. In addition, the pressure drop in both, the primary side and secondary side of the plate heat exchanger is calculated in all simulated transients because their values have limits beyond which the heat exchanger is useless. (orig.)
Magnus: A New Resistive MHD Code with Heat Flow Terms
Navarro, Anamaría; Lora-Clavijo, F. D.; González, Guillermo A.
2017-07-01
We present a new magnetohydrodynamic (MHD) code for the simulation of wave propagation in the solar atmosphere, under the effects of electrical resistivity—but not dominant—and heat transference in a uniform 3D grid. The code is based on the finite-volume method combined with the HLLE and HLLC approximate Riemann solvers, which use different slope limiters like MINMOD, MC, and WENO5. In order to control the growth of the divergence of the magnetic field, due to numerical errors, we apply the Flux Constrained Transport method, which is described in detail to understand how the resistive terms are included in the algorithm. In our results, it is verified that this method preserves the divergence of the magnetic fields within the machine round-off error (˜ 1× {10}-12). For the validation of the accuracy and efficiency of the schemes implemented in the code, we present some numerical tests in 1D and 2D for the ideal MHD. Later, we show one test for the resistivity in a magnetic reconnection process and one for the thermal conduction, where the temperature is advected by the magnetic field lines. Moreover, we display two numerical problems associated with the MHD wave propagation. The first one corresponds to a 3D evolution of a vertical velocity pulse at the photosphere-transition-corona region, while the second one consists of a 2D simulation of a transverse velocity pulse in a coronal loop.
International Nuclear Information System (INIS)
Yun, B. J.; Song, C. H.; Splawski, A.; Lo, S.
2010-01-01
Subcooled boiling is one of the crucial phenomena for the design, operation and safety analysis of a nuclear power plant. It occurs due to the thermally nonequilibrium state in the two-phase heat transfer system. Many complicated phenomena such as a bubble generation, a bubble departure, a bubble growth, and a bubble condensation are created by this thermally nonequilibrium condition in the subcooled boiling flow. However, it has been revealed that most of the existing best estimate safety analysis codes have a weakness in the prediction of the subcooled boiling phenomena in which multi-dimensional flow behavior is dominant. In recent years, many investigators are trying to apply CFD (Computational Fluid Dynamics) codes for an accurate prediction of the subcooled boiling flow. In the CFD codes, evaporation heat flux from heated wall is one of the key parameters to be modeled for an accurate prediction of the subcooled boiling flow. The evaporate heat flux for the CFD codes is expressed typically as follows, q' e = πD 3 d /6 ρ g h fg fN' where, D d , f ,N' are bubble departure size, bubble departure frequency and active nucleation site density, respectively. In the most of the commercial CFD codes, Tolubinsky bubble departure size model, Kurul and Podowski active nucleation site density model and Ceumem-Lindenstjerna bubble departure frequency model are adopted as a basic wall boiling model. However, these models do not consider their dependency on the flow, pressure and fluid type. In this paper, an advanced wall boiling model was proposed in order to improve subcooled boiling model for the CFD codes
Magnetic heat pump flow director
Howard, Frank S. (Inventor)
1995-01-01
A fluid flow director is disclosed. The director comprises a handle body and combed-teeth extending from one side of the body. The body can be formed of a clear plastic such as acrylic. The director can be used with heat exchangers such as a magnetic heat pump and can minimize the undesired mixing of fluid flows. The types of heat exchangers can encompass both heat pumps and refrigerators. The director can adjust the fluid flow of liquid or gas along desired flow directions. A method of applying the flow director within a magnetic heat pump application is also disclosed where the comb-teeth portions of the director are inserted into the fluid flow paths of the heat pump.
Thermal-hydraulic analysis of PWR core including intermediate flow mixers with the THYC code
International Nuclear Information System (INIS)
Mur, J.; Meignin, J.C.
1997-07-01
Departure from nucleate boiling (DNB) is one of the major limiting factors of pressurized water reactors (PWRs). Safety requires that occurrence of DNB should be precluded under normal or incidental operating conditions. The thermal-hydraulic THYC code developed by EDF is described. The code is devoted to heat and mass transfer in nuclear components. Critical Heat Flux (CHF) is predicted from local thermal-hydraulic parameters such as pressure, mass flow rate, and quality. A three stage methodology to evaluate thermal margins in order to perform standard core design is described. (K.A.)
Thermal-hydraulic analysis of PWR core including intermediate flow mixers with the THYC code
Energy Technology Data Exchange (ETDEWEB)
Mur, J. [Electricite de France (EDF), 78 - Chatou (France); Meignin, J.C. [Electricite de France (EDF), 69 - Villeurbanne (France)
1997-07-01
Departure from nucleate boiling (DNB) is one of the major limiting factors of pressurized water reactors (PWRs). Safety requires that occurrence of DNB should be precluded under normal or incidental operating conditions. The thermal-hydraulic THYC code developed by EDF is described. The code is devoted to heat and mass transfer in nuclear components. Critical Heat Flux (CHF) is predicted from local thermal-hydraulic parameters such as pressure, mass flow rate, and quality. A three stage methodology to evaluate thermal margins in order to perform standard core design is described. (K.A.) 8 refs.
Spike Code Flow in Cultured Neuronal Networks.
Tamura, Shinichi; Nishitani, Yoshi; Hosokawa, Chie; Miyoshi, Tomomitsu; Sawai, Hajime; Kamimura, Takuya; Yagi, Yasushi; Mizuno-Matsumoto, Yuko; Chen, Yen-Wei
2016-01-01
We observed spike trains produced by one-shot electrical stimulation with 8 × 8 multielectrodes in cultured neuronal networks. Each electrode accepted spikes from several neurons. We extracted the short codes from spike trains and obtained a code spectrum with a nominal time accuracy of 1%. We then constructed code flow maps as movies of the electrode array to observe the code flow of "1101" and "1011," which are typical pseudorandom sequence such as that we often encountered in a literature and our experiments. They seemed to flow from one electrode to the neighboring one and maintained their shape to some extent. To quantify the flow, we calculated the "maximum cross-correlations" among neighboring electrodes, to find the direction of maximum flow of the codes with lengths less than 8. Normalized maximum cross-correlations were almost constant irrespective of code. Furthermore, if the spike trains were shuffled in interval orders or in electrodes, they became significantly small. Thus, the analysis suggested that local codes of approximately constant shape propagated and conveyed information across the network. Hence, the codes can serve as visible and trackable marks of propagating spike waves as well as evaluating information flow in the neuronal network.
Axial flow heat exchanger devices and methods for heat transfer using axial flow devices
Koplow, Jeffrey P.
2016-02-16
Systems and methods described herein are directed to rotary heat exchangers configured to transfer heat to a heat transfer medium flowing in substantially axial direction within the heat exchangers. Exemplary heat exchangers include a heat conducting structure which is configured to be in thermal contact with a thermal load or a thermal sink, and a heat transfer structure rotatably coupled to the heat conducting structure to form a gap region between the heat conducting structure and the heat transfer structure, the heat transfer structure being configured to rotate during operation of the device. In example devices heat may be transferred across the gap region from a heated axial flow of the heat transfer medium to a cool stationary heat conducting structure, or from a heated stationary conducting structure to a cool axial flow of the heat transfer medium.
Energy Technology Data Exchange (ETDEWEB)
Borges, Eduardo M.; Sabundjian, Gaiane, E-mail: gdgian@ipen.br, E-mail: borges.em@hotmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)
2015-07-01
Identifying the flow regimes and the heat transfer modes is important for the analysis of accidents such as the Loss-of-Coolant Accident (LOCA). The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used in the RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 100cm{sup 2}-rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR - A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. (author)
A code to study the water flow in a thermal test loop
International Nuclear Information System (INIS)
Saunier, Jean-Pierre; Duffourt, Nicole; Lago, Bernard
1965-01-01
A first part reports the theoretical and analytical formulation of a flow within a specific circuit used in a thermal test installation. Equations in the different parts of the circuit are developed, and their resolution for integration into a computation code is described, including boundary conditions, constants and input functions (cell characteristics, fluid characteristics, heat transfer, friction, time slicing). The second part reports an extension of this theoretical and analytical development and code development to a two-branch circuit
Modelling of fluid flow and heat transfer in a reciprocating compressor
Tuhovcak, J.; Hejcik, J.; Jicha, M.
2015-08-01
Efficiency of reciprocating compressor is strongly dependent on several parameters. The most important are valve behaviour and heat transfer. Valves affect the flow through the suction and discharge line. Heat flow from the walls to working fluid increases the work of the cycle. Understanding of these phenomena inside the compressor is a necessary step in the development process. Commercial CFD tools offer wide range of opportunities how to simulate the flow inside the reciprocating compressor nowadays, however they are too demanding in terms of computational time and mesh creation. Several approaches using various correlation equation exist to describe the heat transfer inside the cylinder, however none of them was validated by measurements due to the complicated settings. The goal of this paper is to show a comparison between these correlations using in-house code based on energy balance through the cycle.
OPR1000 RCP Flow Coastdown Analysis using SPACE Code
Energy Technology Data Exchange (ETDEWEB)
Lee, Dong-Hyuk; Kim, Seyun [KHNP CRI, Daejeon (Korea, Republic of)
2016-10-15
The Korean nuclear industry developed a thermal-hydraulic analysis code for the safety analysis of PWRs, named SPACE(Safety and Performance Analysis Code for Nuclear Power Plant). Current loss of flow transient analysis of OPR1000 uses COAST code to calculate transient RCS(Reactor Coolant System) flow. The COAST code calculates RCS loop flow using pump performance curves and RCP(Reactor Coolant Pump) inertia. In this paper, SPACE code is used to reproduce RCS flowrates calculated by COAST code. The loss of flow transient is transient initiated by reduction of forced reactor coolant circulation. Typical loss of flow transients are complete loss of flow(CLOF) and locked rotor(LR). OPR1000 RCP flow coastdown analysis was performed using SPACE using simplified nodalization. Complete loss of flow(4 RCP trip) was analyzed. The results show good agreement with those from COAST code, which is CE code for calculating RCS flow during loss of flow transients. Through this study, we confirmed that SPACE code can be used instead of COAST code for RCP flow coastdown analysis.
Spike Code Flow in Cultured Neuronal Networks
Directory of Open Access Journals (Sweden)
Shinichi Tamura
2016-01-01
Full Text Available We observed spike trains produced by one-shot electrical stimulation with 8 × 8 multielectrodes in cultured neuronal networks. Each electrode accepted spikes from several neurons. We extracted the short codes from spike trains and obtained a code spectrum with a nominal time accuracy of 1%. We then constructed code flow maps as movies of the electrode array to observe the code flow of “1101” and “1011,” which are typical pseudorandom sequence such as that we often encountered in a literature and our experiments. They seemed to flow from one electrode to the neighboring one and maintained their shape to some extent. To quantify the flow, we calculated the “maximum cross-correlations” among neighboring electrodes, to find the direction of maximum flow of the codes with lengths less than 8. Normalized maximum cross-correlations were almost constant irrespective of code. Furthermore, if the spike trains were shuffled in interval orders or in electrodes, they became significantly small. Thus, the analysis suggested that local codes of approximately constant shape propagated and conveyed information across the network. Hence, the codes can serve as visible and trackable marks of propagating spike waves as well as evaluating information flow in the neuronal network.
International Nuclear Information System (INIS)
Dash, Z.V.; Robinson, B.A.; Zyvoloski, G.A.
1997-07-01
The requirements, design, and verification and validation of the software used in the FEHM application, a finite-element heat- and mass-transfer computer code that can simulate nonisothermal multiphase multicomponent flow in porous media, are described. The test of the DOE Code Comparison Project, Problem Five, Case A, which verifies that FEHM has correctly implemented heat and mass transfer and phase partitioning, is also covered
Magnetic Heat Pump Containing Flow Diverters
Howard, Frank S.
1995-01-01
Proposed magnetic heat pump contains flow diverters for suppression of undesired flows. If left unchecked, undesired flows mix substantial amounts of partially heated and partially cooled portions of working fluid, effectively causing leakage of heat from heated side to cooled side. By reducing leakage of heat, flow diverters increase energy efficiency of magnetic heat pump, potentially offering efficiency greater than compressor-driven refrigerator.
Documentation of the heat conduction code TRANCO
International Nuclear Information System (INIS)
Callahan, G.D.
1975-01-01
A transient heat conduction code used for thermal, thermoelastic, thermoelastic/plastic, and thermo/viscoelastic analyses is presented. The code can solve two-dimensional X-Y and axially symmetric R-theta-z thermal problems with the following conditions: constant temperature, constant flux, convective, or adiabatic boundary conditions; time-dependent or constant internal heat generation; and anisotropic thermal conductivities
International Nuclear Information System (INIS)
Milora, S.L.; Combs, S.K.; Foster, C.A.
1984-11-01
An unsteady, two-dimensional heat conduction code has been used to study the performance of swirl-flow-based neutral particle beam targets. The model includes the effects of two-phase heat transfer and asymmetric heating of tubular elements. The calorimeter installed in the Medium Energy Test Facility, which has been subjected to 30-s neutral beam pulses with incident heat flux intensities of greater than or equal to 5 kW/cm 2 , has been modeled. The numerical results indicate that local heat fluxes in excess of 7 kW/cm 2 occur at the water-cooled surface on the side exposed to the beam. This exceeds critical heat flux limits for uniformly heated tubes wih straight flow by approximately a factor of 5. The design of a plasma limiter based on swirl flow heat transfer is presented
Evaluation method for two-phase flow and heat transfer in a feed-water heater
International Nuclear Information System (INIS)
Takamori, Kazuhide; Minato, Akihiko
1993-01-01
A multidimensional analysis code for two-phase flow using a two-fluid model was improved by taking into consideration the condensation heat transfer, film thickness, and film velocity, in order to develop an evaluation method for two-phase flow and heat transfer in a feed-water heater. The following results were obtained by a two-dimensional analysis of a feed-water heater for a power plant. (1) In the model, the film flowed downward in laminar flow due to gravity, with droplet entrainment and deposition. For evaluation of the film thickness, Fujii's equation was used in order to account for forced convection of steam flow. (2) Based on the former experimental data, the droplet deposition coefficient and droplet entrainment rate of liquid film were determined. When the ratio at which the liquid film directly flowed from an upper heat transfer tube to a lower heat transfer tube was 0.7, the calculated total heat transfer rate agreed with the measured value of 130 MW. (3) At the upper region of a heat transfer tube bundle where film thickness was thin, and at the outer region of a heat transfer tube bundle where steam velocity was high, the heat transfer rate was large. (author)
An analytical solution to the heat transfer problem in thick-walled hunt flow
International Nuclear Information System (INIS)
Bluck, Michael J; Wolfendale, Michael J
2017-01-01
Highlights: • Convective heat transfer in Hunt type flow of a liquid metal in a rectangular duct. • Analytical solution to the H1 constant peripheral temperature in a rectangular duct. • New H1 result demonstrating the enhancement of heat transfer due to flow distortion by the applied magnetic field. • Analytical solution to the H2 constant peripheral heat flux in a rectangular duct. • New H2 result demonstrating the reduction of heat transfer due to flow distortion by the applied magnetic field. • Results are important for validation of CFD in magnetohydrodynamics and for implementation of systems code approaches. - Abstract: The flow of a liquid metal in a rectangular duct, subject to a strong transverse magnetic field is of interest in a number of applications. An important application of such flows is in the context of coolants in fusion reactors, where heat is transferred to a lead-lithium eutectic. It is vital, therefore, that the heat transfer mechanisms are understood. Forced convection heat transfer is strongly dependent on the flow profile. In the hydrodynamic case, Nusselt numbers and the like, have long been well characterised in duct geometries. In the case of liquid metals in strong magnetic fields (magnetohydrodynamics), the flow profiles are very different and one can expect a concomitant effect on convective heat transfer. For fully developed laminar flows, the magnetohydrodynamic problem can be characterised in terms of two coupled partial differential equations. The problem of heat transfer for perfectly electrically insulating boundaries (Shercliff case) has been studied previously (Bluck et al., 2015). In this paper, we demonstrate corresponding analytical solutions for the case of conducting hartmann walls of arbitrary thickness. The flow is very different from the Shercliff case, exhibiting jets near the side walls and core flow suppression which have profound effects on heat transfer.
Development of a computer code for thermohydraulic analysis of a heated channel in transients
International Nuclear Information System (INIS)
Jafari, J.; Kazeminejad, H.; Davilu, H.
2004-01-01
This paper discusses the thermohydraulic analysis of a heated channel of a nuclear reactor in transients by a computer code that has been developed by the writer. The considered geometry is a channel of a nuclear reactor with cylindrical or planar fuel rods. The coolant is water and flows from the outer surface of the fuel rod. To model the heat transfer in the fuel rod, two dimensional time dependent conduction equations has been solved by combination of numerical methods, O rthogonal Collocation Method in radial direction and finite difference method in axial direction . For coolant modelling the single phase time dependent energy equation has been used and solved by finite difference method . The combination of the first module that solves the conduction in the fuel rod and a second one that solved the energy balance in the coolant region constitute the computer code (Thyc-1) to analysis thermohydraulic of a heated channel in transients. The Orthogonal collocation method maintains the accuracy and computing time of conventional finite difference methods, while the computer storage is reduced by a factor of two. The same problem has been modelled by RELAP5/M3 system code to asses the validity of the Thyc-1 code. The good agreement of the results qualifies the developed code
Validation of Heat Transfer and Film Cooling Capabilities of the 3-D RANS Code TURBO
Shyam, Vikram; Ameri, Ali; Chen, Jen-Ping
2010-01-01
The capabilities of the 3-D unsteady RANS code TURBO have been extended to include heat transfer and film cooling applications. The results of simulations performed with the modified code are compared to experiment and to theory, where applicable. Wilcox s k-turbulence model has been implemented to close the RANS equations. Two simulations are conducted: (1) flow over a flat plate and (2) flow over an adiabatic flat plate cooled by one hole inclined at 35 to the free stream. For (1) agreement with theory is found to be excellent for heat transfer, represented by local Nusselt number, and quite good for momentum, as represented by the local skin friction coefficient. This report compares the local skin friction coefficients and Nusselt numbers on a flat plate obtained using Wilcox's k-model with the theory of Blasius. The study looks at laminar and turbulent flows over an adiabatic flat plate and over an isothermal flat plate for two different wall temperatures. It is shown that TURBO is able to accurately predict heat transfer on a flat plate. For (2) TURBO shows good qualitative agreement with film cooling experiments performed on a flat plate with one cooling hole. Quantitatively, film effectiveness is under predicted downstream of the hole.
SCDAP/RELAP5 Modeling of Heat Transfer and Flow Losses in Lower Head Porous Debris
International Nuclear Information System (INIS)
Siefken, Larry James; Coryell, Eric Wesley; Paik, Seungho; Kuo, Han Hsiung
1999-01-01
Designs are described for implementing models for calculating the heat transfer and flow losses in porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonporous debris that results from core material slumping into the lower head. Currently, the COUPLE model has the capability to model convective and radiative heat transfer from the surfaces of nonporous debris in a detailed manner and to model only in a simplistic manner the heat transfer from porous debris. In order to advance beyond the simplistic modeling for porous debris, designs are developed for detailed calculations of heat transfer and flow losses in porous debris. Correlations are identified for convective heat transfer in porous debris for the following modes of heat transfer; (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, and (5) film boiling. Interphase heat transfer is modeled in an approximate manner. Designs are described for models to calculate the flow losses and interphase drag of fluid flowing through the interstices of the porous debris, and to apply these variables in the momentum equations in the RELAP5 part of the code. Since the models for heat transfer and flow losses in porous debris in the lower head are designed for general application, a design is also described for implementation of these models to the analysis of porous debris in the core region. A test matrix is proposed for assessing the capability of the implemented models to calculate the heat transfer and flow losses in porous debris. The implementation of the models described in this report is expected to improve the COUPLE code calculation of the temperature distribution in porous debris and in the lower head that supports the debris. The implementation of these models is also expected to improve the calculation of the temperature and flow distribution in porous debris in the core region
International Nuclear Information System (INIS)
Son, Hyung M.; Suh, Kune Y.
2012-01-01
Highlights: ► Performed experiment for the upward SCO 2 flow surrounded by highly conducting metal. ► Selected dimensionless groups representing the property variations and buoyancy. ► Developed the heat transfer correlation for the mixed thermal boundary condition. ► Wrote a finite element heat transfer code to find the appropriate correlation. ► Coupled the 1D convection and 2D heat conduction via heat transfer coefficient. - Abstract: This paper presents heat transfer characteristics of supercritical carbon dioxide flow inside vertical circular pipe surrounded by highly conducting material, and develops an adequate tool to test the performance of available heat transfer correlations with. The possible situations are illustrated for the nuclear power plant to which the above-mentioned geometric configuration might be applicable. An experimental loop with vertical circular geometry is designed and constructed to test the upward flow in supercritical state when the axial heat transfer is enhanced by the surrounding metals, resulting in a wall boundary condition between the constant heat flux and temperature. The set of correlations and important findings are critically reviewed from extensive literature survey. Incorporating nondimensional groups resorting to past insights from the available literature, a convective heat transfer correlation is proposed. The optimization procedure is described which utilizes a random walk method along with the in-house finite element heat transfer code to determine the coefficients of the proposed heat transfer correlation. The proposed methodology can be applied to evaluation of heat transfer when the heat transfer coefficient data cannot directly be determined from the experiment.
International Nuclear Information System (INIS)
2014-08-01
The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in
Energy Technology Data Exchange (ETDEWEB)
NONE
2014-08-15
The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in
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)
Core 2D. A code for non-isothermal water flow and reactive solute transport. Users manual version 2
Energy Technology Data Exchange (ETDEWEB)
Samper, J.; Juncosa, R.; Delgado, J.; Montenegro, L. [Universidad de A Coruna (Spain)
2000-07-01
Understanding natural groundwater quality patterns, quantifying groundwater pollution and assessing the effects of waste disposal, require modeling tools accounting for water flow, and transport of heat and dissolved species as well as their complex interactions with solid and gases phases. This report contains the users manual of CORE ''2D Version V.2.0, a COde for modeling water flow (saturated and unsaturated), heat transport and multicomponent Reactive solute transport under both local chemical equilibrium and kinetic conditions. it is an updated and improved version of CORE-LE-2D V0 (Samper et al., 1988) which in turns is an extended version of TRANQUI, a previous reactive transport code (ENRESA, 1995). All these codes were developed within the context of Research Projects funded by ENRESA and the European Commission. (Author)
Core2D. A code for non-isothermal water flow and reactive solute transport. Users manual version 2
International Nuclear Information System (INIS)
Samper, J.; Juncosa, R.; Delgado, J.; Montenegro, L.
2000-01-01
Understanding natural groundwater quality patterns, quantifying groundwater pollution and assessing the effects of waste disposal, require modeling tools accounting for water flow, and transport of heat and dissolved species as well as their complex interactions with solid and gases phases. This report contains the users manual of CORE ''2D Version V.2.0, a COde for modeling water flow (saturated and unsaturated), heat transport and multicomponent Reactive solute transport under both local chemical equilibrium and kinetic conditions. it is an updated and improved version of CORE-LE-2D V0 (Samper et al., 1988) which in turns is an extended version of TRANQUI, a previous reactive transport code (ENRESA, 1995). All these codes were developed within the context of Research Projects funded by ENRESA and the European Commission. (Author)
Core 2D. A code for non-isothermal water flow and reactive solute transport. Users manual version 2
Energy Technology Data Exchange (ETDEWEB)
Samper, J; Juncosa, R; Delgado, J; Montenegro, L [Universidad de A Coruna (Spain)
2000-07-01
Understanding natural groundwater quality patterns, quantifying groundwater pollution and assessing the effects of waste disposal, require modeling tools accounting for water flow, and transport of heat and dissolved species as well as their complex interactions with solid and gases phases. This report contains the users manual of CORE ''2D Version V.2.0, a COde for modeling water flow (saturated and unsaturated), heat transport and multicomponent Reactive solute transport under both local chemical equilibrium and kinetic conditions. it is an updated and improved version of CORE-LE-2D V0 (Samper et al., 1988) which in turns is an extended version of TRANQUI, a previous reactive transport code (ENRESA, 1995). All these codes were developed within the context of Research Projects funded by ENRESA and the European Commission. (Author)
Introduction of thermal-hydraulic analysis code and system analysis code for HTGR
International Nuclear Information System (INIS)
Tanaka, Mitsuhiro; Izaki, Makoto; Koike, Hiroyuki; Tokumitsu, Masashi
1984-01-01
Kawasaki Heavy Industries Ltd. has advanced the development and systematization of analysis codes, aiming at lining up the analysis codes for heat transferring flow and control characteristics, taking up HTGR plants as the main object. In order to make the model of flow when shock waves propagate to heating tubes, SALE-3D which can analyze a complex system was developed, therefore, it is reported in this paper. Concerning the analysis code for control characteristics, the method of sensitivity analysis in a topological space including an example of application is reported. The flow analysis code SALE-3D is that for analyzing the flow of compressible viscous fluid in a three-dimensional system over the velocity range from incompressibility limit to supersonic velocity. The fundamental equations and fundamental algorithm of the SALE-3D, the calculation of cell volume, the plotting of perspective drawings and the analysis of the three-dimensional behavior of shock waves propagating in heating tubes after their rupture accident are described. The method of sensitivity analysis was added to the analysis code for control characteristics in a topological space, and blow-down phenomena was analyzed by its application. (Kako, I.)
Energy Technology Data Exchange (ETDEWEB)
Lis, J [Central Electricity Research Laboratories, Leatherhead, Surrey (United Kingdom)
1984-07-01
This paper describes the experimental and analytical investigations of the gas-side heat transfer and flow characteristics of steam generators in the AGR stations carried out by CERL. The majority of the experimental work on heat transfer and flow characteristics of close-packed tube arrangements in cross-flow of gases is carried out in a pressurised heat exchanger rig. The rig is operated on-line by a dedicated PDP 11/40 computer over the range of Reynolds number 10{sup 4} to 3x10{sup 5}. Atmospheric wind tunnels employing either small or large scale models of the specific sections of steam generators are used for a variety of supplementary and development studies. Various measurements techniques and, in particular, LDA and hot wire anemometry employed in these studies are described. The more important aspects of various investigations are illustrated by typical results. In order to ensure the efficient operation and integrity of steam generators under asymmetric boundary conditions a MIX suite of 2-dimensional codes has been developed. The codes calculate the gas and water/steam flow and temperature distributions in each channel of the steam generator taking into account thermal mixing in the gas as it passes through the generator. Application of the MIX codes to the solution of various operational problems is illustrated by typical examples and the continuing exercise of validating the codes against plant operational data is discussed. (author)
International Nuclear Information System (INIS)
Lis, J.
1984-01-01
This paper describes the experimental and analytical investigations of the gas-side heat transfer and flow characteristics of steam generators in the AGR stations carried out by CERL. The majority of the experimental work on heat transfer and flow characteristics of close-packed tube arrangements in cross-flow of gases is carried out in a pressurised heat exchanger rig. The rig is operated on-line by a dedicated PDP 11/40 computer over the range of Reynolds number 10 4 to 3x10 5 . Atmospheric wind tunnels employing either small or large scale models of the specific sections of steam generators are used for a variety of supplementary and development studies. Various measurements techniques and, in particular, LDA and hot wire anemometry employed in these studies are described. The more important aspects of various investigations are illustrated by typical results. In order to ensure the efficient operation and integrity of steam generators under asymmetric boundary conditions a MIX suite of 2-dimensional codes has been developed. The codes calculate the gas and water/steam flow and temperature distributions in each channel of the steam generator taking into account thermal mixing in the gas as it passes through the generator. Application of the MIX codes to the solution of various operational problems is illustrated by typical examples and the continuing exercise of validating the codes against plant operational data is discussed. (author)
MINET: transient analysis of fluid-flow and heat-transfer networks
International Nuclear Information System (INIS)
Van Tuyle, G.J.; Guppy, J.G.; Nepsee, T.C.
1983-01-01
MINET, a computer code developed for the steady-state and transient analysis of fluid-flow and heat-transfer networks, is described. The code is based on a momentum integral network method, which offers significant computational advantages in the analysis of large systems, such as the balance of plant in a power-generating facility. An application is discussed in which MINET is coupled to the Super System Code (SSC), an advanced generic code for the transient analysis of loop- or pool-type LMFBR systems. In this application, the ability of the Clinch River Breeder Reactor Plant to operate in a natural circulation mode following an assumed loss of all electric power, was assessed. Results from the MINET portion of the calculations are compared against those generated independently by the Clinch River Project, using the DEMO code
Heat exchanger with oscillating flow
Scotti, Stephen J. (Inventor); Blosser, Max L. (Inventor); Camarda, Charles J. (Inventor)
1993-01-01
Various heat exchange apparatuses are described in which an oscillating flow of primary coolant is used to dissipate an incident heat flux. The oscillating flow may be imparted by a reciprocating piston, a double action twin reciprocating piston, fluidic oscillators or electromagnetic pumps. The oscillating fluid flows through at least one conduit in either an open loop or a closed loop. A secondary flow of coolant may be used to flow over the outer walls of at least one conduit to remove heat transferred from the primary coolant to the walls of the conduit.
Two-phase flow characteristics analysis code: MINCS
International Nuclear Information System (INIS)
Watanabe, Tadashi; Hirano, Masashi; Akimoto, Masayuki; Tanabe, Fumiya; Kohsaka, Atsuo.
1992-03-01
Two-phase flow characteristics analysis code: MINCS (Modularized and INtegrated Code System) has been developed to provide a computational tool for analyzing two-phase flow phenomena in one-dimensional ducts. In MINCS, nine types of two-phase flow models-from a basic two-fluid nonequilibrium (2V2T) model to a simple homogeneous equilibrium (1V1T) model-can be used under the same numerical solution method. The numerical technique is based on the implicit finite difference method to enhance the numerical stability. The code structure is highly modularized, so that new constitutive relations and correlations can be easily implemented into the code and hence evaluated. A flow pattern can be fixed regardless of flow conditions, and state equations or steam tables can be selected. It is, therefore, easy to calculate physical or numerical benchmark problems. (author)
RELAP5/MOD2 benchmarking study: Critical heat flux under low-flow conditions
International Nuclear Information System (INIS)
Ruggles, E.; Williams, P.T.
1990-01-01
Experimental studies by Mishima and Ishii performed at Argonne National Laboratory and subsequent experimental studies performed by Mishima and Nishihara have investigated the critical heat flux (CHF) for low-pressure low-mass flux situations where low-quality burnout may occur. These flow situations are relevant to long-term decay heat removal after a loss of forced flow. The transition from burnout at high quality to burnout at low quality causes very low burnout heat flux values. Mishima and Ishii postulated a model for the low-quality burnout based on flow regime transition from churn turbulent to annular flow. This model was validated by both flow visualization and burnout measurements. Griffith et al. also studied CHF in low mass flux, low-pressure situations and correlated data for upflows, counter-current flows, and downflows with the local fluid conditions. A RELAP5/MOD2 CHF benchmarking study was carried out investigating the performance of the code for low-flow conditions. Data from the experimental study by Mishima and Ishii were the basis for the benchmark comparisons
International Nuclear Information System (INIS)
Harvego, E. A.; Siefken, L. J.
2000-01-01
The SCDAP/RELAP5 code is being developed at the Idaho National Engineering and Environmental Laboratory under the primary sponsorship of the U.S. Nuclear Regulatory Commission (NRC) to provide best-estimate transient simulations of light water reactor coolant systems during severe accidents. This paper describes the modeling approach used in the SCDAP/RELAP5 code to calculate fluid heat transfer and flow losses through porous debris that has accumulated in the vessel lower head and core regions during the latter stages of a severe accident. The implementation of heat transfer and flow loss correlations into the code is discussed, and calculations performed to assess the validity of the modeling approach are described. The different modes of heat transfer in porous debris include: (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, (5) film boiling, and (6) transition from film boiling to convection to vapor. The correlations for flow losses in porous debris include frictional and form losses. The correlations for flow losses were integrated into the momentum equations in the RELAP5 part of the code. Since RELAP5 is a very general non-homogeneous non-equilibrium thermal-hydraulics code, the resulting modeling methodology is applicable to a wide range of debris thermal-hydraulic conditions. Assessment of the SCDAP/RELAP5 debris bed thermal-hydraulic models included comparisons with experimental measurements and other models available in the open literature. The assessment calculations, described in the paper, showed that SCDAP/RELAP5 is capable of calculating the heat transfer and flow losses occurring in porous debris regions that may develop in a light water reactor during a severe accident
Stokes flow heat transfer in an annular, rotating heat exchanger
International Nuclear Information System (INIS)
Saatdjian, E.; Rodrigo, A.J.S.; Mota, J.P.B.
2011-01-01
The heat transfer rate into highly viscous, low thermal-conductivity fluids can be enhanced significantly by chaotic advection in three-dimensional flows dominated by viscous forces. The physical effect of chaotic advection is to render the cross-sectional temperature field uniform, thus increasing both the wall temperature gradient and the heat flux into the fluid. A method of analysis for one such flow-the flow in the eccentric, annular, rotating heat exchanger-and a procedure to determine the best heat transfer conditions, namely the optimal values of the eccentricity ratio and time-periodic rotating protocol, are discussed. It is shown that in continuous flows, such as the one under consideration, there exists an optimum frequency of the rotation protocol for which the heat transfer rate is a maximum. - Highlights: → The eccentric, annular, rotating heat exchanger is studied for periodic Stokes flow. → Counter-rotating the inner tube with a periodic velocity enhances the heat transfer. → The heat-transfer enhancement under such conditions is due to chaotic advection. → For a given axial flow rate there is a frequency that maximizes the heat transfer. → There is also an optimum value of the eccentricity ratio.
Analysis of the Onset of Flow Instability in rectangular heated channel using drift flux model
International Nuclear Information System (INIS)
El-Hadjen, H.; Balistrou, M.; Hamidouche, T.; Bousbia-Salah, A.
2005-01-01
Two-phase flow excursion (Ledinegg) instability in boiling channels is of great concern in the design and operation of numerous practical systems especially in Research Reactors. Such instability can lead to significant reduction in channel flow, thereby causing premature burnout of the heated channel before the CHF point. The present work focuses on a simulation of pressure drop in forced convection boiling in vertical narrow and parallel uniformly heated channels. The objective is to determine the point of Onset of Flow Instability (OFI) by varying input flow rate. The axial void distribution is also provided. The numerical model is based on the finite difference method which transforms the partial differential conservation equation of mass, momentum and energy, in algebraic equations. Closure relationships based upon the drift flux model and other constitutive equations are considered to determine the channel pressure drop under steady state boiling conditions. The model validation is performed by confronting the calculations with the Oak Ridge National Laboratory thermal Hydraulic Test Loop (THTL) experimental data set. Further verification of this model is performed by code- to code verification using the results of RELAP5/Mod 3.2 code. (author)
Simplified computational simulation of liquid metal behaviour in turbulent flow with heat transfer
International Nuclear Information System (INIS)
Costa, E.B. da.
1992-09-01
The present work selected the available bibliography equations and empirical relationships to the development of a computer code to obtain the turbulent velocity and temperature profiles in liquid metal tube flow with heat generation. The computer code is applied to a standard problem and the results are considered satisfactory, at least from the viewpoint of qualitative behaviour. (author). 50 refs, 21 figs, 3 tabs
Influences of buoyancy and thermal boundary conditions on heat transfer with naturally-induced flow
International Nuclear Information System (INIS)
Jackson, J.D.; Li, J.
2002-01-01
A fundamental study is reported of heat transfer from a vertical heated tube to air which is induced naturally upwards through it by the action of buoyancy. Measurements of local heat transfer coefficient were made using a specially designed computer-controlled power supply and measurement system for conditions of uniform wall temperature and uniform wall heat flux. The effectiveness of heat transfer proved to be much lower than for conditions of forced convection. It was found that the results could be correlated satisfactorily when presented in terms of dimensionless parameters similar to those used for free convection heat transfer from vertical surfaces provided that the heat transfer coefficients were evaluated using local fluid bulk temperature calculated utilising the measured values of flow rate induced through the system. Additional experiments were performed' with pumped flow. These covered the entire mixed convection region. It was found that the data for naturally-induced flow mapped onto the pumped flow data when presented in terms of Nusselt number ratio (mixed to forced) and buoyancy parameter. Computational simulations of the experiments were performed using an advanced computer code which incorporated a buoyancy-influenced, variable property, developing wall shear flow formulation and a low Reynolds number k-ε turbulence model. These reproduced observed behaviour quite well. (author)
Interpretation of lunar heat flow data
International Nuclear Information System (INIS)
Conel, J.E.; Morton, J.B.
1975-01-01
Lunar heat flow observations at the Apollo 15 and 17 sites can be interpreted to imply bulk U concentrations for the Moon of 5 to 8 times those of normal chondrites and 2 to 4 times terrestrial values inferred from the Earth's heat flow and the assumption of thermal steady state between surface heat flow and heat production. A simple model of nearsurface structure that takes into account the large difference in (highly insulating) regolith thickness between mare and highland provinces is considered. This model predicts atypically high local values of heat flow near the margins of mare regions--possibly a factor of 10 or so higher than the global average. A test of the proposed model using multifrequency microwave techniques appears possible wherein heat flow traverse measurements are made across mare-highland contacts. The theoretical considerations discussed here urge caution in attributing global significance to point heat-flow measurements on the Moon
Heat flow and heat generation in greenstone belts
Drury, M. J.
1986-01-01
Heat flow has been measured in Precambrian shields in both greenstone belts and crystalline terrains. Values are generally low, reflecting the great age and tectonic stability of the shields; they range typically between 30 and 50 mW/sq m, although extreme values of 18 and 79 mW/sq m have been reported. For large areas of the Earth's surface that are assumed to have been subjected to a common thermotectonic event, plots of heat flow against heat generation appear to be linear, although there may be considerable scatter in the data. The relationship is expressed as: Q = Q sub o + D A sub o in which Q is the observed heat flow, A sub o is the measured heat generation at the surface, Q sub o is the reduced heat flow from the lower crust and mantle, and D, which has the dimension of length, represents a scale depth for the distribution of radiogenic elements. Most authors have not used data from greenstone belts in attempting to define the relationship within shields, considering them unrepresentative and preferring to use data from relatively homogeneous crystalline rocks. A discussion follows.
Energy Technology Data Exchange (ETDEWEB)
Grazevicius, Audrius; Kaliatka, Algirdas [Lithuanian Energy Institute, Kaunas (Lithuania). Lab. of Nuclear Installation Safety
2017-07-15
The main functions of spent fuel pools are to remove the residual heat from spent fuel assemblies and to perform the function of biological shielding. In the case of loss of heat removal from spent fuel pool, the fuel rods and pool water temperatures would increase continuously. After the saturated temperature is reached, due to evaporation of water the pool water level would drop, eventually causing the uncover of spent fuel assemblies, fuel overheating and fuel rods failure. This paper presents an analysis of loss of heat removal accident in spent fuel pool of BWR 4 and a comparison of two different modelling approaches. The one-dimensional system thermal-hydraulic computer code RELAP5 and CFD tool ANSYS Fluent were used for the analysis. The results are similar, but the local effects cannot be simulated using a one-dimensional code. The ANSYS Fluent calculation demonstrated that this three-dimensional treatment allows to avoid the need for many one-dimensional modelling assumptions in the pool modelling and enables to reduce the uncertainties associated with natural circulation flow calculation.
Molecular dynamics study of solid-liquid heat transfer and passive liquid flow
Yesudasan Daisy, Sumith
equilibrium canonical ensemble (NVT) is simulated using thermostat algorithms. For research in heat transfer involving solid liquid interaction, we need to perform non equilibrium MD (NEMD) simulations. In such NEMD simulations, the methods used for simulating heating from a surface is very important and must capture proper physics and thermodynamic properties. Development of MD simulation techniques to simulate solid-liquid heating and the study of fundamental mechanism of passive flow is the main focus of this thesis. An accurate surface-heating algorithm was developed for water which can now allow the study of a whole new set of fundamental heat transfer problems at the nanoscale like surface heating/cooling of droplets, thin-films, etc. The developed algorithm is implemented in the in-house developed C++ MD code. A direct two dimensional local pressure estimation algorithm is also formulated and implemented in the code. With this algorithm, local pressure of argon and platinum interaction is studied. Also, the surface tension of platinum-argon (solid-liquid) was estimated directly from the MD simulations for the first time. Contact angle estimation studies of water on platinum, and argon on platinum were also performed. A thin film of argon is kept above platinum plate and heated in the middle region, leading to the evaporation and pressure reduction thus creating a strong passive flow in the near surface region. This observed passive liquid flow is characterized by estimating the pressure, density, velocity and surface tension using Eulerian mapping method. Using these simulation, we have demonstrated the fundamental nature and origin of surface-driven passive flow. Heat flux removed from the surface is also estimated from the results, which shows a significant improvement can be achieved in thermal management of electronic devices by taking advantage of surface-driven strong passive liquid flow. Further, the local pressure of water on silicon di-oxide surface is
Fluid flow and heat transfer modeling for castings
International Nuclear Information System (INIS)
Domanus, H.M.; Liu, Y.Y.; Sha, W.T.
1986-01-01
Casting is fundamental to manufacturing of many types of equipment and products. Although casting is a very old technology that has been in existence for hundreds of years, it remains a highly empirical technology, and production of new castings requires an expensive and time-consuming trial-and-error approach. In recent years, mathematical modeling of casting has received increasing attention; however, a majority of the modeling work has been in the area of heat transfer and solidification. Very little work has been done in modeling fluid flow of the liquid melt. This paper presents a model of fluid flow coupled with heat transfer of a liquid melt for casting processes. The model to be described in this paper is an extension of the COMMIX code and is capable of handling castings with any shape, size, and material. A feature of this model is the ability to track the liquid/gas interface and liquid/solid interface. The flow of liquid melt through the sprue and runners and into the mold cavity is calculated as well as three-dimensional temperature and velocity distributions of the liquid melt throughout the casting process. 14 refs., 13 figs
International Nuclear Information System (INIS)
Gosselin, Stephen R.; Cumblidge, Stephen E.; Anderson, Michael T.; Simonen, Fredric A.; Tinsley, G A.; Lydell, B.; Doctor, Steven R.
2005-01-01
Inservice inspection requirements for pressure retaining welds in the regenerative, letdown, and residual heat removal heat exchangers are prescribed in Section XI Articles IWB and IWC of the ASME Boiler and Pressure Vessel Code. Accordingly, volumetric and/or surface examinations are performed on heat exchanger shell, head, nozzle-to-head, and nozzle-to-shell welds. Inspection difficulties associated with the implementation of these Code-required examinations have forced operating nuclear power plants to seek relief from the U.S. Nuclear Regulatory Commission. The nature of these relief requests are generally concerned with metallurgical, geometry, accessibility, and radiation burden. Over 60% of licensee requests to the NRC identify significant radiation exposure burden as the principle reason for relief from the ASME Code examinations on regenerative heat exchangers. For the residual heat removal heat exchangers, 90% of the relief requests are associated with geometry and accessibility concerns. Pacific Northwest National Laboratory was funded by the NRC Office of Nuclear Regulatory Research to review current practice with regard to volumetric and/or surface examinations of shell welds of letdown heat exchangers regenerative heat exchangers and residual (decay) heat removal heat exchangers Design, operating, common preventative maintenance practices, and potential degradation mechanisms are reviewed. A detailed survey of domestic and international PWR-specific operating experience was performed to identify pressure boundary failures (or lack of failures) in each heat exchanger type and NSSS design. The service data survey was based on the PIPExp- database and covers PWR plants worldwide for the period 1970-2004. Finally a risk assessment of the current ASME Code inspection requirements for residual heat removal, letdown, and regenerative heat exchangers is performed. The results are then reviewed to discuss the examinations relative to plant safety and
ComboCoding: Combined intra-/inter-flow network coding for TCP over disruptive MANETs
Directory of Open Access Journals (Sweden)
Chien-Chia Chen
2011-07-01
Full Text Available TCP over wireless networks is challenging due to random losses and ACK interference. Although network coding schemes have been proposed to improve TCP robustness against extreme random losses, a critical problem still remains of DATA–ACK interference. To address this issue, we use inter-flow coding between DATA and ACK to reduce the number of transmissions among nodes. In addition, we also utilize a “pipeline” random linear coding scheme with adaptive redundancy to overcome high packet loss over unreliable links. The resulting coding scheme, ComboCoding, combines intra-flow and inter-flow coding to provide robust TCP transmission in disruptive wireless networks. The main contributions of our scheme are twofold; the efficient combination of random linear coding and XOR coding on bi-directional streams (DATA and ACK, and the novel redundancy control scheme that adapts to time-varying and space-varying link loss. The adaptive ComboCoding was tested on a variable hop string topology with unstable links and on a multipath MANET with dynamic topology. Simulation results show that TCP with ComboCoding delivers higher throughput than with other coding options in high loss and mobile scenarios, while introducing minimal overhead in normal operation.
Valentin Rodriguez, Francisco Ivan
dissipating capabilities of helium flow, due to natural circulation in the system at both high and low pressure, were also examined. These experimental results are useful for the development and validation of VHTR design and safety analysis codes. Numerical simulations were performed using a Multiphysics computer code, COMSOL, displaying less than 5% error between the measured graphite temperatures in both the heated and cooled channels. Finally, new correlations have been proposed describing the thermal-hydraulic phenomena in buoyancy driven flows in both heated and cooled channels.
FILM-30: A Heat Transfer Properties Code for Water Coolant
International Nuclear Information System (INIS)
MARSHALL, THERON D.
2001-01-01
A FORTRAN computer code has been written to calculate the heat transfer properties at the wetted perimeter of a coolant channel when provided the bulk water conditions. This computer code is titled FILM-30 and the code calculates its heat transfer properties by using the following correlations: (1) Sieder-Tate: forced convection, (2) Bergles-Rohsenow: onset to nucleate boiling, (3) Bergles-Rohsenow: partially developed nucleate boiling, (4) Araki: fully developed nucleate boiling, (5) Tong-75: critical heat flux (CHF), and (6) Marshall-98: transition boiling. FILM-30 produces output files that provide the heat flux and heat transfer coefficient at the wetted perimeter as a function of temperature. To validate FILM-30, the calculated heat transfer properties were used in finite element analyses to predict internal temperatures for a water-cooled copper mockup under one-sided heating from a rastered electron beam. These predicted temperatures were compared with the measured temperatures from the author's 1994 and 1998 heat transfer experiments. There was excellent agreement between the predicted and experimentally measured temperatures, which confirmed the accuracy of FILM-30 within the experimental range of the tests. FILM-30 can accurately predict the CHF and transition boiling regimes, which is an important advantage over current heat transfer codes. Consequently, FILM-30 is ideal for predicting heat transfer properties for applications that feature high heat fluxes produced by one-sided heating
Magnetohydrodynamic flow and heat transfer around a heated cylinder of arbitrary conductivity
Tassone, A.; Nobili, M.; Caruso, G.
2017-11-01
The interaction of the liquid metal with the plasma confinement magnetic field constitutes a challenge for the design of fusion reactor blankets, due to the arise of MHD effects: increased pressure drops, heat transfer suppression, etc. To overcome these issues, a dielectric fluid can be employed as coolant for the breeding zone. A typical configuration involves pipes transverse to the liquid metal flow direction. This numerical study is conducted to assess the influence of pipe conductivity on the MHD flow and heat transfer. The CFD code ANSYS CFX was employed for this purpose. The fluid is assumed to be bounded by rectangular walls with non-uniform thickness and subject to a skewed magnetic field with the main component aligned with the cylinder axis. The simulations were restricted to Re = (20; 40) and M = (10; 50). Three different scenarios for the obstacle were considered: perfectly insulating, finite conductivity and perfectly conducting. The electrical conductivity was found to affect the channel pressure penalty due to the obstacle insertion only for M = 10 and just for the two limiting cases. A general increment of the heat transfer with M was found due to the tendency of the magnetic field to equalize the flow rate between the sub-channels individuated by the pipe. The best results were obtained with the insulating pipe, due to the reduced electromagnetic drag. The generation of counter-rotating vortices close to the lateral duct walls was observed for M = 50 and perfectly conducting pipe as a result of the modified currents distribution.
SCDAP/RELAP5 modeling of heat transfer and flow losses in lower head porous debris. Rev. 1
International Nuclear Information System (INIS)
Siefken, L.J.; Coryell, E.W.; Paik, S.; Kuo, H.
1999-01-01
Designs are described for implementing models for calculating the heat transfer and flow losses in porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and nonporous debris that results from core material slumping into the lower head. Currently, the COUPLE model has the capability to model convective and radiative heat transfer from the surfaces of nonporous debris in a detailed manner and to model only in a simplistic manner the heat transfer from porous debris. In order to advance beyond the simplistic modeling for porous debris, designs are developed for detailed calculations of heat transfer and flow losses in porous debris. Correlations are identified for convective heat transfer in porous debris for the following modes of heat transfer; (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, and (5) film boiling. Interphase heat transfer is modeled in an approximate ma nner. Designs are described for models to calculate the flow losses and interphase drag of fluid flowing through the interstices of the porous debris, and to apply these variables in the momentum equations in the RELAP5 part of the code. Since the models for heat transfer and flow losses in porous debris in the lower head are designed for general application, a design is also described for implementation of these models to the analysis of porous debris in the core region. A test matrix is proposed for assessing the capability of the implemented models to calculate the heat transfer and flow losses in porous debris. The implementation of the models described in this report is expected to improve the COUPLE code calculation of the temperature distribution in porous debris and in the lower head that supports the debris. The implementation of these models is also expected to improve the calculation of the temperature and flow distribution in porous debris in the core region
MINET, Transient Fluid Flow and Heat Transfer Power Plant Network Analysis
International Nuclear Information System (INIS)
Van Tuyle, G.J.
2002-01-01
1 - Description of program or function: MINET (Momentum Integral Network) was developed for the transient analysis of intricate fluid flow and heat transfer networks, such as those found in the balance of plant in power generating facilities. It can be utilized as a stand-alone program or interfaced to another computer program for concurrent analysis. Through such coupling, a computer code limited by either the lack of required component models or large computational needs can be extended to more fully represent the thermal hydraulic system thereby reducing the need for estimating essential transient boundary conditions. The MINET representation of a system is one or more networks of volumes, segments, and boundaries linked together via heat exchangers only, i.e., heat can transfer between networks, but fluids cannot. Volumes are used to represent tanks or other volume components, as well as locations in the system where significant flow divisions or combinations occur. Segments are composed of one or more pipes, pumps, heat exchangers, turbines, and/or valves each represented by one or more nodes. Boundaries are simply points where the network interfaces with the user or another computer code. Several fluids can be simulated, including water, sodium, NaK, and air. 2 - Method of solution: MINET is based on a momentum integral network method. Calculations are performed at two levels, the network level (volumes) and the segment level. Equations conserving mass and energy are used to calculate pressure and enthalpy within volumes. An integral momentum equation is used to calculate the segment average flow rate. In-segment distributions of mass flow rate and enthalpy are calculated using local equations of mass and energy. The segment pressure is taken to be the linear average of the pressure at both ends. This method uses a two-plus equation representation of the thermal hydraulic behavior of a system of heat exchangers, pumps, pipes, valves, tanks, etc. With the
Saturated flow boiling heat transfer in water-heated vertical annulus
International Nuclear Information System (INIS)
Sun Licheng; Yan Changqi; Sun Zhonning
2005-01-01
This paper describes the saturated flow boiling heat transfer characteristics of water at 1 atm and low velocities in water-heated vertical annuli with equivalent diameters of 10 mm and 6 mm. Test section is consisted of two concentric circular tubes outer of which is made of quartz, so the whole test courses can be visualized. There are three main flow patterns of bubble flow, churn flow and churn-annular flow in the annuli, most important of which is churn flow. Flooding is the mechanism of churn flow and churn can enhance the heat transport between steam and water; Among the three factors of mass flux, inlet subcooling and annulus width, the last one has great effect on heat transport, moderately decreasing the annulus width can enhance the heat transfer; Combined annular flow model with theory of flooding and turbulent Prandtl Number, the numerical value of heat flux is given, the shape of test boiling curve and that of calculated by model is very alike, but there is large discrepancy between test data and calculated results, the most possible reason is that some parameters given by fluid flooding model are based on experimental data of common circular tubes, but not of annuli. Doing more research on flooding in annulus, particularly narrow annulus, is necessary for calculating the saturated boiling in annulus. (authors)
Investigation of Two-Phase Flow Regime Maps for Development of Thermal-Hydraulic Analysis Codes
International Nuclear Information System (INIS)
Kim, Kyung Doo; Kim, Byoung Jae; Lee, Seong Wook
2010-04-01
This reports is a literature survey on models and correlations for determining flow pattern that are used to simulate thermal-hydraulics in nuclear reactors. Determination of flow patterns are a basis for obtaining physical values of wall/interfacial friction, wall/interfacial heat transfer, and droplet entrainment/de-entrainment. Not only existing system codes, such as RELAP5-3D, TRAC-M, MARS, TRACE, CATHARE) but also up-to-date researches were reviewed to find models and correlations
Case-crossover analysis of heat-coded deaths and vulnerable subpopulations: Oklahoma, 1990-2011
Moore, Brianna F.; Brooke Anderson, G.; Johnson, Matthew G.; Brown, Sheryll; Bradley, Kristy K.; Magzamen, Sheryl
2017-11-01
The extent of the association between temperature and heat-coded deaths, for which heat is the primary cause of death, remains largely unknown. We explored the association between temperature and heat-coded deaths and potential interactions with various demographic and environmental factors. A total of 335 heat-coded deaths that occurred in Oklahoma from 1990 through 2011 were identified using heat-related International Classification of Diseases codes, cause-of-death nomenclature, and narrative descriptions. Conditional logistic regression models examined the association between temperature and heat index on heat-coded deaths. Interaction by demographic factors (age, sex, marital status, living alone, outdoor/heavy labor occupations) and environmental factors (ozone, PM10, PM2.5) was also explored. Temperatures ≥99 °F (the median value) were associated with approximately five times higher odds of a heat-coded death as compared to temperatures effect estimates were attenuated when exposure to heat was characterized by heat index. The interaction results suggest that effect of temperature on heat-coded deaths may depend on sex and occupation. For example, the odds of a heat-coded death among outdoor/heavy labor workers exposed to temperatures ≥99 °F was greater than expected based on the sum of the individual effects (observed OR = 14.0, 95% CI 2.7, 72.0; expected OR = 4.1 [2.8 + 2.3-1.0]). Our results highlight the extent of the association between temperature and heat-coded deaths and emphasize the need for a comprehensive, multisource definition of heat-coded deaths. Furthermore, based on the interaction results, we recommend that states implement or expand heat safety programs to protect vulnerable subpopulations, such as outdoor workers.
CFD investigation of flow and heat transfer of nanofluids in isoflux spirally fluted tubes
Salama, Amgad; Azamatov, Abdulaziz Irgashevich; El-Amin, Mohamed; Sun, Shuyu; Huang, Huancong
2012-01-01
In this work, the problem of flow and heat transfer of nanofluids in spirally fluted tubes is investigated numerically using the CFD code Fluent. The tube investigated in this work is characterized by the existence of helical ridging which
Experimental study on flow pattern and heat transfer of inverted annular flow
International Nuclear Information System (INIS)
Takenaka, Nobuyuki; Akagawa, Koji; Fujii, Terushige; Nishida, Koji
1990-01-01
Experimental results are presented on flow pattern and heat transfer in the regions from inverted annular flow to dispersed flow in a vertical tube using freon R-113 as a working fluid at atmospheric pressure to discuss the correspondence between them. Axial distributions of heat transfer coefficient are measured and flow patterns are observed. The heat transfer characteristics are divided into three regions and a heat transfer characteristics map is proposed. The flow pattern changes from inverted annular flow (IAF) to dispersed flow (DF) through inverted slug flow (ISF) for lower inlet velocities and through agitated inverted annular flow (AIAF) for higher inlet velocities. A flow pattern map is obtained which corresponds well with the heat transfer characteristic map. (orig.)
Energy Technology Data Exchange (ETDEWEB)
Moridis, George; Freeman, Craig
2013-09-30
We developed two new EOS additions to the TOUGH+ family of codes, the RealGasH2O and RealGas . The RealGasH2O EOS option describes the non-isothermal two-phase flow of water and a real gas mixture in gas reservoirs, with a particular focus in ultra-tight (such as tight-sand and shale gas) reservoirs. The gas mixture is treated as either a single-pseudo-component having a fixed composition, or as a multicomponent system composed of up to 9 individual real gases. The RealGas option has the same general capabilities, but does not include water, thus describing a single-phase, dry-gas system. In addition to the standard capabilities of all members of the TOUGH+ family of codes (fully-implicit, compositional simulators using both structured and unstructured grids), the capabilities of the two codes include: coupled flow and thermal effects in porous and/or fractured media, real gas behavior, inertial (Klinkenberg) effects, full micro-flow treatment, Darcy and non-Darcy flow through the matrix and fractures of fractured media, single- and multi-component gas sorption onto the grains of the porous media following several isotherm options, discrete and fracture representation, complex matrix-fracture relationships, and porosity-permeability dependence on pressure changes. The two options allow the study of flow and transport of fluids and heat over a wide range of time frames and spatial scales not only in gas reservoirs, but also in problems of geologic storage of greenhouse gas mixtures, and of geothermal reservoirs with multi-component condensable (H2O and CH4) and non-condensable gas mixtures. The codes are verified against available analytical and semi-analytical solutions. Their capabilities are demonstrated in a series of problems of increasing complexity, ranging from isothermal flow in simpler 1D and 2D conventional gas reservoirs, to non-isothermal gas flow in 3D fractured shale gas reservoirs involving 4 types of fractures, micro-flow, non-Darcy flow and gas
Mansoor, Mohammad M.
2012-02-01
A 3D-conjugate numerical investigation was conducted to predict heat transfer characteristics in a rectangular cross-sectional micro-channel employing simultaneously developing single-phase flows. The numerical code was validated by comparison with previous experimental and numerical results for the same micro-channel dimensions and classical correlations based on conventional sized channels. High heat fluxes up to 130W/cm 2 were applied to investigate micro-channel thermal characteristics. The entire computational domain was discretized using a 120×160×100 grid for the micro-channel with an aspect ratio of (α=4.56) and examined for Reynolds numbers in the laminar range (Re 500-2000) using FLUENT. De-ionized water served as the cooling fluid while the micro-channel substrate used was made of copper. Validation results were found to be in good agreement with previous experimental and numerical data [1] with an average deviation of less than 4.2%. As the applied heat flux increased, an increase in heat transfer coefficient values was observed. Also, the Reynolds number required for transition from single-phase fluid to two-phase was found to increase. A correlation is proposed for the results of average Nusselt numbers for the heat transfer characteristics in micro-channels with simultaneously developing, single-phase flows. © 2011 Elsevier Ltd.
International Nuclear Information System (INIS)
Giles, G.E.; DeVault, R.M.; Turner, W.D.; Becker, B.R.
1976-05-01
A description is given of the development and verification of a generalized coupled conduction-convection, multichannel heat transfer computer program to analyze specific safety questions involving high temperature gas-cooled reactors (HTGR). The HEXEREI code was designed to provide steady-state and transient heat transfer analysis of the HTGR active core using a basic hexagonal mesh and multichannel coolant flow. In addition, the core auxiliary cooling systems were included in the code to provide more complete analysis of the reactor system during accidents involving reactor trip and cooling down on the auxiliary systems. Included are brief descriptions of the components of the HEXEREI code and sample HEXEREI analyses compared with analytical solutions and other heat transfer codes
Chen, Yih-Kang
1992-01-01
Effect of flow field properties on the heating distribution over a 140 deg blunt cone was determined for a Martian atmosphere using Euler, Navier-Stokes (NS), viscous shock layer (VSL), and reacting boundary layer (BLIMPK) equations. The effect of gas kinetics on the flow field and the surface heating distribution were investigated. Gas models with nine species and nine reactions were implemented into the codes. Effects of surface catalysis on the heating distribution were studied using a surface kinetics model having five reactions.
Computational multi-fluid dynamics predictions of critical heat flux in boiling flow
Energy Technology Data Exchange (ETDEWEB)
Mimouni, S., E-mail: stephane.mimouni@edf.fr; Baudry, C.; Guingo, M.; Lavieville, J.; Merigoux, N.; Mechitoua, N.
2016-04-01
Highlights: • A new mechanistic model dedicated to DNB has been implemented in the Neptune-CFD code. • The model has been validated against 150 tests. • Neptune-CFD code is a CFD tool dedicated to boiling flows. - Abstract: Extensive efforts have been made in the last five decades to evaluate the boiling heat transfer coefficient and the critical heat flux in particular. Boiling crisis remains a major limiting phenomenon for the analysis of operation and safety of both nuclear reactors and conventional thermal power systems. As a consequence, models dedicated to boiling flows have being improved. For example, Reynolds Stress Transport Model, polydispersion and two-phase flow wall law have been recently implemented. In a previous work, we have evaluated computational fluid dynamics results against single-phase liquid water tests equipped with a mixing vane and against two-phase boiling cases. The objective of this paper is to propose a new mechanistic model in a computational multi-fluid dynamics tool leading to wall temperature excursion and onset of boiling crisis. Critical heat flux is calculated against 150 tests and the mean relative error between calculations and experimental values is equal to 8.3%. The model tested covers a large physics scope in terms of mass flux, pressure, quality and channel diameter. Water and R12 refrigerant fluid are considered. Furthermore, it was found that the sensitivity to the grid refinement was acceptable.
Computational multi-fluid dynamics predictions of critical heat flux in boiling flow
International Nuclear Information System (INIS)
Mimouni, S.; Baudry, C.; Guingo, M.; Lavieville, J.; Merigoux, N.; Mechitoua, N.
2016-01-01
Highlights: • A new mechanistic model dedicated to DNB has been implemented in the Neptune_CFD code. • The model has been validated against 150 tests. • Neptune_CFD code is a CFD tool dedicated to boiling flows. - Abstract: Extensive efforts have been made in the last five decades to evaluate the boiling heat transfer coefficient and the critical heat flux in particular. Boiling crisis remains a major limiting phenomenon for the analysis of operation and safety of both nuclear reactors and conventional thermal power systems. As a consequence, models dedicated to boiling flows have being improved. For example, Reynolds Stress Transport Model, polydispersion and two-phase flow wall law have been recently implemented. In a previous work, we have evaluated computational fluid dynamics results against single-phase liquid water tests equipped with a mixing vane and against two-phase boiling cases. The objective of this paper is to propose a new mechanistic model in a computational multi-fluid dynamics tool leading to wall temperature excursion and onset of boiling crisis. Critical heat flux is calculated against 150 tests and the mean relative error between calculations and experimental values is equal to 8.3%. The model tested covers a large physics scope in terms of mass flux, pressure, quality and channel diameter. Water and R12 refrigerant fluid are considered. Furthermore, it was found that the sensitivity to the grid refinement was acceptable.
RELAP5 and SIMMER-III code assessment on CIRCE decay heat removal experiments
International Nuclear Information System (INIS)
Bandini, Giacomino; Polidori, Massimiliano; Meloni, Paride; Tarantino, Mariano; Di Piazza, Ivan
2015-01-01
Highlights: • The CIRCE DHR experiments simulate LOHS+LOF transients in LFR systems. • Decay heat removal by natural circulation through immersed heat exchangers is investigated. • The RELAP5 simulation of DHR experiments is presented. • The SIMMER-III simulation of DHR experiments is presented. • The focus is on the transition from forced to natural convection and stratification in a large pool. - Abstract: In the frame of THINS Project of the 7th Framework EU Program on Nuclear Fission Safety, some experiments were carried out on the large scale LBE-cooled CIRCE facility at the ENEA/Brasimone Research Center to investigate relevant safety aspects associated with the removal of decay heat through heat exchangers (HXs) immersed in the primary circuit of a pool-type lead fast reactor (LFR), under loss of heat sink (LOHS) accidental conditions. The start-up and operation of this decay heat removal (DHR) system relies on natural convection on the primary side and then might be affected by coolant mixing and temperature stratification phenomena occurring in the LBE pool. The main objectives of the CIRCE experimental campaign were to verify the behavior of the DHR system under representative accidental conditions and provide a valuable database for the assessment of both CFD and system codes. The reproduced accidental conditions refer to a station blackout scenario, namely a protected LOHS and loss of flow (LOF) transient. In this paper the results of 1D RELAP5 and 2D SIMMER-III simulations are compared with the experimental data of more representative DHR transients T-4 and T-5 in order to verify the capability of these codes to reproduce both forced and natural convection conditions observed in the primary circuit and the right operation of the DHR system for decay heat removal. Both codes are able to reproduce the stationary conditions and with some uncertainties the transition to natural convection conditions until the end of the transient phase. The trend
International Nuclear Information System (INIS)
Vishnoi, A.K.; Dasgupta, A.; Chandraker, D.K.; Nayak, A.K.; Vijayan, P.K.
2015-01-01
Two-phase natural circulation loops have extensive applications in nuclear and process industries. One of the major concerns with natural circulation is the occurrence of the various types of flow instabilities, which can cause premature boiling crisis due to flow and power oscillations. In this work a transient computer code COPCOS (Code for Prediction of CHF under Oscillating flow and power condition) has been developed to predict the premature occurrence of CHF (critical heat flux) under oscillating flow and power. The code incorporates conduction equation of the fuel and coolant energy equation. For CHF prediction, CHF look-up table developed by Groeneveld is used. A facility named CHF and Instability Loop (CHIL) has been set up to study the effect of oscillatory flow on CHF. CHF and Instability Loop (CHIL) is a simple rectangular loop having a 10.5 mm ID and 1.2 m long test section. The flow through the test section is controlled by a canned motor pump using a Variable Frequency Drive (VFD). This leads to the ability of having a very precise control over flow oscillations which can be induced in the test section. The effect of frequency and amplitude of flow oscillation on occurrence of premature CHF has been investigated in this facility using COPCOS. Full paper covers details of COPCOS code, description of the facility and effect of frequency and the effect of oscillatory flow on CHF in the facility. (author)
International Nuclear Information System (INIS)
Kolev, N.I.
1991-12-01
The second part of the IVA3 code description contains the constitutive models used for the interfacial transport phenomena and the code validation results. First 20 flow patterns are defined and the transition criteria are discussed. The dynamic fragmentation and coalescence models used in IVA3 are documented. After the description of the models for predicting the flow patterns and flow structure sizes the models for the interfacial mechanical interaction are described. Finally the models for interfacial heat and mass transfer are given with emphasis on the time averaging of the heat and mass source terms. The code validation passes several stages from simple tests on well known benchmarks trough simulation of one-, two-, and three-phase flows in simple and complicated geometries. The gradually increase of the complexity and the successful comparison of the predictions with experimental data is the main characteristic of the verification procedure. It is demonstrated by several examples that IVA3 is a powerful tool for three-fluid modelling of complicated three-phase flows in complex geometry with strong thermal and mechanical interaction between the velocity fields. (orig.) [de
A study of the dispersed flow interfacial heat transfer model of RELAP5/MOD2.5 and RELAP5/MOD3
Energy Technology Data Exchange (ETDEWEB)
Andreani, M. [Swiss Federal Institute of Technology, Zurich (Switzerland); Analytis, G.T.; Aksan, S.N. [Paul Scherrer Institute, Villigen (Switzerland)
1995-09-01
The model of interfacial heat transfer for the dispersed flow regime used in the RELAP5 computer codes is investigated in the present paper. Short-transient calculations of two low flooding rate tube reflooding experiments have been performed, where the hydraulic conditions and the heat input to the vapour in the post-dryout region were controlled for the predetermined position of the quench front. Both RELAP5/MOD2.5 and RELAP5/MOD3 substantially underpredicted the exit vapour temperature. The mass flow rate and quality, however, were correct and the heat input to the vapour was larger than the actual one. As the vapour superheat at the tube exit depends on the balance between the heat input from the wall and the heat exchange with the droplets, the discrepancy between the calculated and the measured exit vapour temperature suggested that the inability of both codes to predict the vapour superheat in the dispersed flow region is due to the overprediction of the interfacial heat transfer rate.
CCC, Heat Flow and Mass Flow in Liquid Saturated Porous Media
International Nuclear Information System (INIS)
Mangold, D.C.; Lippmann, M.J.; Bodvarsson, G.S.
1982-01-01
1 - Description of problem or function: The numerical model CCC (conduction-convection-consolidation) solves the heat and mass flow equations for a fully, liquid-saturated, anisotropic porous medium and computes one-dimensional (vertical) consolidation of the simulated systems. The model has been applied to problems in the fields of geothermal reservoir engineering, aquifer thermal energy storage, well testing, radioactive waste isolation, and in situ coal combustion. The code has been validated against analytic solutions for fluid and heat flow, and against a field experiment for underground storage of hot water. 2 - Method of solution: The model employs the Integrated Finite Difference Method (IFDM) in discretizing the saturated porous medium and formulating the governing equations. The sets of equations are sol- ved by an iterative solution technique. The vertical deformation of the medium is calculated using the one-dimensional consolidation theory of Terzaghi. 3 - Restrictions on the complexity of the problem: Maximum of 12 materials. It is assumed that: (a) Darcy's law adequately describes fluid movement through fractured and porous media. (b) The rock and fluid are in thermal equilibrium at any given time. (c) Energy changes due to the fluid compressibility, acceleration and viscous dissipation are neglected. (d) One-dimensional consolidation theory adequately describes the vertical deformation of the medium
International Nuclear Information System (INIS)
Sirvydas, A.; Poskas, R.
2006-01-01
We present the results on numerical investigation of the local opposing mixed convection heat transfer in a vertical flat channel with symmetrical heating at low Reynolds numbers. Numerical two-dimensional simulation was performed for the same channel and for the same conditions as in the experiment using the FLUENT 6.1 code. The unsteady flow investigations were performed in airflow for the experimental conditions at the Reynolds number 2130 and Grashof number 6.2* 10 8 . Quasi-steady flow investigations were performed for two Reynolds numbers (2130 and 4310) and the Grashof number up to 3.1*10 9 in order to simulate the buoyancy effect on the flow structure. In both steady and quasi-steady modelling cases the results demonstrated that under the high buoyancy effect the chequerwise local circular flow took place near the heated walls. This made velocity profiles asymmetrical and caused pulsations of the wall temperature. Wall temperature had a pulsatory character, however, the resulting averaged values correlated rather well with experimental data for steady and quasi-steady cases for Re in = 2130. For Re in = 4310, the resulting averaged values for x/d e ≤25 correlated rather well with experimental data. When x/d e > 25, the difference between the experimental and the calculated wall temperature was increasing, increasing, possibly due to a steady flow and heat transfer modelling. (author)
Langseth, M. G.
1977-01-01
The principal components of the experiment were probes, each with twelve thermometers of exceptional accuracy and stability, that recorded temperature variations at the surface and in the regolith down to 2.5 m. The Apollo 15 experiment and the Apollo 17 probes recorded lunar surface and subsurface temperatures. These data provided a unique and valuable history of the interaction of solar energy with lunar surface and the effects of heat flowing from the deep interior out through the surface of the moon. The interpretation of these data resulted in a clearer definition of the thermal and mechanical properties of the upper two meters of lunar regolith, direct measurements of the gradient in mean temperature due to heat flow from the interior and a determination of the heat flow at the Apollo 15 and Apollo 17 sites.
Dunn, James C.; Hardee, Harry C.; Striker, Richard P.
1985-01-01
A convective heat flow probe device is provided which measures heat flow and fluid flow magnitude in the formation surrounding a borehole. The probe comprises an elongate housing adapted to be lowered down into the borehole; a plurality of heaters extending along the probe for heating the formation surrounding the borehole; a plurality of temperature sensors arranged around the periphery of the probe for measuring the temperature of the surrounding formation after heating thereof by the heater elements. The temperature sensors and heater elements are mounted in a plurality of separate heater pads which are supported by the housing and which are adapted to be radially expanded into firm engagement with the walls of the borehole. The heat supplied by the heater elements and the temperatures measured by the temperature sensors are monitored and used in providing the desired measurements. The outer peripheral surfaces of the heater pads are configured as segments of a cylinder and form a full cylinder when taken together. A plurality of temperature sensors are located on each pad so as to extend along the length and across the width thereof, with a heating element being located in each pad beneath the temperature sensors. An expansion mechanism driven by a clamping motor provides expansion and retraction of the heater pads and expandable packer-type seals are provided along the probe above and below the heater pads.
Flow and heat transfer in laminar–turbulent transitional flow regime under rolling motion
International Nuclear Information System (INIS)
Yuan, Hongsheng; Tan, Sichao; Zhuang, Nailiang; Lan, Shu
2016-01-01
Highlights: • Flow and heat transfer experiment in transitional flow regime under rolling motion. • Increases of average friction factor and Nu were found. • Periodic breakdown of laminar flow contributes to the increase. • Nonlinear variation of pressure drop or Nu with Re also contributes to the increase. • Effect of critical Reynolds number shift was discussed. - Abstract: Flow and heat transfer characteristics under rolling motion are extremely important to thermohydraulic analysis of offshore nuclear reactors. An experimental study was conducted in a heated rectangular channel to investigate flow and heat transfer in laminar–turbulent transitional flow regime under rolling motion. The results showed that the average friction factor and Nusselt number are higher than that of the corresponding steady flow as the flow rate fluctuates in transitional flow regime. Larger relative flow rate fluctuation was observed under larger rolling amplitude or higher rolling frequency. In the same manner, larger increases of average friction factor and Nusselt number were achieved under larger rolling amplitude or higher rolling frequency. The increases were mainly caused by the flow rate fluctuation through periodic breakdown of laminar flow and development of turbulence in laminar–turbulent transitional flow regime. First, turbulence, which enhances the rate of momentum and energy exchange, occurs near the crest of flow rate wave even the flow is still in laminar flow regime according to the average Reynolds number. Second, as a result of rapid increases of the friction and heat transfer with Reynolds number in transitional flow regime, the increases of the friction and the heat transfer near the crest of flow rate wave are larger than the decreases of them near the trough of flow rate wave, which also contributes to increases of average friction and heat transfer. Additionally, the effect of critical Reynolds number shift under unsteady flow and heating
International Nuclear Information System (INIS)
Shepherd, I.M.
1982-01-01
The computer code BUSH has been developed for the calculation of steady state heat transfer in a rod bundle. For a given power, flow and geometry it can calculate the temperatures in the rods, coolant and shroud assuming that at any axial level each rod can be described by one temperature and the coolant fluid is also radially uniform at this level. Heat transfer by convection and radiation are handled and the geometry is flexible enough to model nearly all types of envisaged shroud design for the SUPERSARA test series. The modular way in which BUSH has been written makes it suitable for future development, either within the present BUSH framework or as part of a more advanced code
Improved Flow Modeling in Transient Reactor Safety Analysis Computer Codes
International Nuclear Information System (INIS)
Holowach, M.J.; Hochreiter, L.E.; Cheung, F.B.
2002-01-01
A method of accounting for fluid-to-fluid shear in between calculational cells over a wide range of flow conditions envisioned in reactor safety studies has been developed such that it may be easily implemented into a computer code such as COBRA-TF for more detailed subchannel analysis. At a given nodal height in the calculational model, equivalent hydraulic diameters are determined for each specific calculational cell using either laminar or turbulent velocity profiles. The velocity profile may be determined from a separate CFD (Computational Fluid Dynamics) analysis, experimental data, or existing semi-empirical relationships. The equivalent hydraulic diameter is then applied to the wall drag force calculation so as to determine the appropriate equivalent fluid-to-fluid shear caused by the wall for each cell based on the input velocity profile. This means of assigning the shear to a specific cell is independent of the actual wetted perimeter and flow area for the calculational cell. The use of this equivalent hydraulic diameter for each cell within a calculational subchannel results in a representative velocity profile which can further increase the accuracy and detail of heat transfer and fluid flow modeling within the subchannel when utilizing a thermal hydraulics systems analysis computer code such as COBRA-TF. Utilizing COBRA-TF with the flow modeling enhancement results in increased accuracy for a coarse-mesh model without the significantly greater computational and time requirements of a full-scale 3D (three-dimensional) transient CFD calculation. (authors)
Heat transfer in a counterflow heat exchanger at low flow rates
International Nuclear Information System (INIS)
Hashimoto, A.; Hattori, N.; Naruke, K.
1995-01-01
A study was made of heat transfer in a double-tube heat exchanger at low flow rates of water. The temperatures of fluid and tube walls in the axial direction of tube were measured precisely at flow rate ratios of annulus to inner tube (or flow rate ratios of inner tube to annulus W i /W a , Re i approx. = 80 - 4000), W a /W i =0.1 - 1.1. In parallel with experiment, numerical calculation for forced-convection heat transfer was also carried out for laminar flows in the same tube configuration as experiment. Average over-all coefficients of heat transfer, obtained by experiments, indicate the same characteristics as numerical calculation in the examined range of flow rate ratio. Their experimental values, however, are somewhat larger than those of calculation at small values of flow rate ratio. (author)
Development, validation and application of NAFA 2D-CFD code
International Nuclear Information System (INIS)
Vaidya, A.M.; Maheshwari, N.K.; Vijayan, P.K.; Saha, D.
2010-01-01
A 2D axi-symmetric code named NAFA (Version 1.0) is developed for studying the pipe flow under various conditions. It can handle laminar/ turbulent flows, with or without heat transfer, under sub-critical/super-critical conditions. The code solves for momentum, energy equations with standard k-ε turbulence model (with standard wall functions). It solves pipe flow subjected to 'velocity inlet', 'wall', 'axis' and 'pressure outlet' boundary conditions. It is validated for several cases by comparing its results with experimental data/analytical solutions/correlations. The code has excellent convergence characteristics as verified from fall of equation residual in each case. It has proven capability of generating mesh independent results for laminar as well as turbulent flows. The code is applied to supercritical flows. For supercritical flows, the effect of mesh size on prediction of heat transfer coefficient is studied. With grid refinement, the Y + reduces and reaches the limiting value of 11.63. Hence the accuracy is found to increase with grid refinement. NAFA is able to qualitatively predict the effect of heat flux and operating pressure on heat transfer coefficient. The heat transfer coefficient matches well with experimental values under various conditions. (author)
Heat flow map of the Bohemian massif
Energy Technology Data Exchange (ETDEWEB)
Cermak, V [Geophys. Inst., CS Acad. of Sci.
1977-01-01
Forty seven heat flow values for the Bohemian massif were used to determine the heat flow pattern of the area. By including data from neighboring countries it was possible to draw an isothermal map outlining the geothermal activity. As a result, it is possible to closely correlate the heat flow and the tectonic structure. It is obvious that the areas of high geothermal activity correspond to zones of crustal weakness associated with two major faults bordering the rigid central section of the massif. The highest heat flow values coincide with the axis of the sedimentary basin. The development of these heat flow patterns should assist in the recognition of probable areas of geothermal resources and several promising sites are readily discernible.
Room Heat-Up Analysis with GOTHIC code
International Nuclear Information System (INIS)
Jimenez, G.; Olza, J. M.
2010-01-01
The GOTHIC T M computer code is a state-of-the art program for modeling multiphase, multicomponent fluid flow. GOTHIC is rapidly becoming the industry-standard code for performing both containment design basis accident (DBA) analyses and analyses to support equipment qualification. GOTHIC has a flexible nodding structure that allows both lumped parameter and 3-D modeling capabilities. Multidimensional analysis capabilities greatly enhance the study of noncondensable gases and stratification and permit the calculation of flow field details within any given volume.
International Nuclear Information System (INIS)
Nomura, Yasushi; Someya, Hiroyuki; Ito, Haruhiko.
1992-11-01
Capsules for irradiation tests in the JMTR (Japan Materials Testing Reactor), consist of irradiation specimens surrounded by a cladding tube, holders, an inner tube and a container tube (from 30mm to 65mm in diameter). And the annular gaps between these structural materials in the capsule are filled with liquids or gases. Cooling of the capsule is done by reactor primary coolant flowing down outside the capsule. Most of the heat generated by fission in fuel specimens and gamma absorption in structural materials is directed radially to the capsule container outer surface. In thermal performance calculations for capsule design, an one(r)-dimensional heat transfer computer code entitled (Generalyzed Gap Temperature Calculation), GENGTC, originally developed in Oak Ridge National Laboratory, U.S.A., has been frequently used. In designing a capsule, are needed many cases of parametric calculations with respect to changes materials and gap sizes. And in some cases, two(r,z)-dimensional heat transfer calculations are needed for irradiation test capsules with short length fuel rods. Recently the authors improved the original one-dimensional code GENGTC, (1) to simplify preparation of input data, (2) to perform automatic calculations for parametric survey based on design temperatures, ect. Moreover, the computer code has been improved to perform r-z two-dimensional heat transfer calculation. This report describes contents of the preparation of the one-dimensional code GENGTC and the improvement for the two-dimensional code GENGTC-2, together with their code manuals. (author)
Immanuel, Y.; Pullepu, Bapuji; Sambath, P.
2018-04-01
A two dimensional mathematical model is formulated for the transitive laminar free convective, incompressible viscous fluid flow over vertical cone with variable surface heat flux combined with the effects of heat generation and absorption is considered . using a powerful computational method based on thermoelectric analogy called Network Simulation Method (NSM0, the solutions of governing nondimensionl coupled, unsteady and nonlinear partial differential conservation equations of the flow that are obtained. The numerical technique is always stable and convergent which establish high efficiency and accuracy by employing network simulator computer code Pspice. The effects of velocity and temperature profiles have been analyzed for various factors, namely Prandtl number Pr, heat flux power law exponent n and heat generation/absorption parameter Δ are analyzed graphically.
Augmentation of forced flow boiling heat transfer by introducing air flow into subcooled water flow
International Nuclear Information System (INIS)
Koizumi, Y.; Ohtake, H.; Yuasa, T.; Matsushita, N.
2001-01-01
The effect of air injection into a subcooled water flow on boiling heat transfer and a critical heat flux (CHF) was examined experimentally. Experiments were conducted in the range of subcooling of 50 K, a superficial velocity of water and air Ul = 0.17 ∼ 3.4 and Ug = 0 ∼ 15 m/s, respectively. A test heat transfer surface was a 5 mm wide, 40 mm long and 0.5 mm thick stainless steel sheet embedded on the bottom wall of a 10 mm high and 20 mm wide rectangular flow channel. Nine times enhancement of the heat transfer coefficient in the non-boiling region was attained at the most by introducing an air flow into a water single-phase flow. The heat transfer improvement was prominent when the water flow rate was low and the air introduction was large. The present results of the non-boiling heat transfer were well correlated with the Lockhart-Martinelli parameter X tt ; h TP /h L0 = 5.0(1/ X tt ) 0.5 . The air introduction has some effect on the augmentation of heat transfer in the boiling region, however, the two-phase flow effect was little and the boiling was dominant in the fully developed boiling region. The CHF was improved a little by the air introduction in the high water flow region. However, that was rather greatly reduced in the low flow region. Even so, the general trend by the air introduction was that qCHF increased as the air introduction was increased. The heat transfer augmentation in the non-boiling region was attained by less power increase than that in the case that only the water flow rate was increased. From the aspect of the power consumption and the heat transfer enhancement, the small air introduction in the low water flow rate region seemed more profitable, although the air introduction in the high water flow rate region and also the large air introduction were still effective in the augmentation of the heat transfer in the non-boiling region. (author)
Valenzuela, Javier
2001-01-01
A radial flow heat exchanger (20) having a plurality of first passages (24) for transporting a first fluid (25) and a plurality of second passages (26) for transporting a second fluid (27). The first and second passages are arranged in stacked, alternating relationship, are separated from one another by relatively thin plates (30) and (32), and surround a central axis (22). The thickness of the first and second passages are selected so that the first and second fluids, respectively, are transported with laminar flow through the passages. To enhance thermal energy transfer between first and second passages, the latter are arranged so each first passage is in thermal communication with an associated second passage along substantially its entire length, and vice versa with respect to the second passages. The heat exchangers may be stacked to achieve a modular heat exchange assembly (300). Certain heat exchangers in the assembly may be designed slightly differently than other heat exchangers to address changes in fluid properties during transport through the heat exchanger, so as to enhance overall thermal effectiveness of the assembly.
Matson, D. L.; Ransford, G. A.; Johnson, T. V.
1981-01-01
The existing ground-based measurements of Io's thermal emission at infrared wavelengths of 8.4, 10.6, and 21 microns have been reexamined. Present in these data is the signature of hot spots, presumably similar to the hot spots seen by the IRIS experiment on Voyager. It is possible to extract from these data the total amount of power radiated. Since the hot spots are believed to be a result of deep-seated activity in Io and since the remainder of Io's surface is an extraordinarily poor thermal conductor, the power radiated by the hot spots is essentially the total heat flow. The analysis yields a heat flow of 2 + or - 1 W/sq m. This value is tremendously large in comparison to the average heat flow of the earth (0.06 W/sq m) and the moon (0.02 W/sq m), but is characteristic of active geothermal areas on the earth. A heat flow this large requires that the interior of Io be at least partially molten on a global scale.
Model validation of GAMMA code with heat transfer experiment for KO TBM in ITER
International Nuclear Information System (INIS)
Yum, Soo Been; Lee, Eo Hwak; Lee, Dong Won; Park, Goon Cherl
2013-01-01
Highlights: ► In this study, helium supplying system was constructed. ► Preparation for heat transfer experiment in KO TBM condition using helium supplying system was progressed. ► To get more applicable results, test matrix was made to cover the condition for KO TBM. ► Using CFD code; CFX 11, validation and modification for system code GAMMA was performed. -- Abstract: By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) test blanket module (TBM) for testing in ITER. A performance analysis for the thermal–hydraulics and a safety analysis for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs multicomponent mixture analysis), which was developed by the gas cooled reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM first wall (FW) mock-up made from the same material as the KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 1.1, 1.9 and 2.9 MPa, and under various ranges of flow rate from 0.0105 to 0.0407 kg/s with a constant wall temperature condition. In the present study, a thermal–hydraulic test was performed with the newly constructed helium supplying system, in which the design pressure and temperature were 9 MPa and 500 °C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 3 MPa pressure, 30 °C inlet temperature and 70 m/s helium velocity, which are almost same conditions of the KO TBM FW. One side of the mock-up was heated with a constant heat flux of 0.3–0.5 MW/m 2 using a graphite heating system, KoHLT-2 (Korea heat load test facility-2). Because the comparison result between CFX 11 and GAMMA showed a difference tendency, the modification of heat transfer correlation included in GAMMA was performed. And the modified GAMMA showed the strong parity with CFX
Heat Source Models in Simulation of Heat Flow in Friction Stir Welding
DEFF Research Database (Denmark)
Schmidt, Henrik Nikolaj Blich; Hattel, Jesper
2004-01-01
The objective of the present paper is to investigate the effect of including the tool probe and the material flow in the numerical modelling of heat flow in Friction Stir Welding (FSW). The contact condition at the interface between the tool and workpiece controls the heat transfer mechanisms....... The convective heat transfer due to the material flow affects the temperature fields. Models presented previously in literature allow the heat to flow through the probe volume, and the majority of them neglect the influence of the contact condition as the sliding condition is assumed. In the present work......, a number of cases are established. Each case represents a combination of a contact condition, i.e. sliding and sticking, and a stage of refinement regarding the heat source distribution. In the most detailed models the heat flow is forced around the probe volume by prescribing a velocity field in shear...
Heat source models in simulation of heat flow in friction stir welding
DEFF Research Database (Denmark)
Schmidt, Henrik Nikolaj Blich; Hattel, Jesper
2004-01-01
The objective of the present paper is to investigate the effect of including the tool probe and the material flow in the numerical modelling of heat flow in friction stir welding (FSW). The contact condition at the interface between the tool and workpiece controls the heat transfer mechanisms....... The convective heat transfer due to the material flow affects the temperature fields. Models presented previously in the literature allow the heat to flow through the probe volume, and the majority neglects the influence of the contact condition as the sliding condition is assumed. In this work, a number...... of cases is established. Each case represents a combination of a contact condition, i.e. sliding and sticking, and a stage of refinement regarding the heat source distribution. In the most detailed models, the heat flow is forced around the probe volume by prescribing a velocity field in shear layers...
International Nuclear Information System (INIS)
Banerjee, S.; Hassan, Y.A.
1995-01-01
Condensation in the presence of noncondensible gases plays an important role in the nuclear industry. The RELAP5/MOD3 thermal hydraulic code was used to study the ability of the code to predict this phenomenon. Two separate effects experiments were simulated using this code. These were the Massachusetts Institute of Technology's (MIT) Pressurizer Experiment, the MIT Single Tube Experiment. A new iterative approach to calculate the interface temperature and the degraded heat transfer coefficient was developed and implemented in the RELAP5/MOD3 thermal hydraulic code. This model employs the heat transfer simultaneously. This model was found to perform much better than the reduction factor approach. The calculations using the new model were found to be in much better agreement with the experimental values
Energy Technology Data Exchange (ETDEWEB)
Banerjee, S.; Hassan, Y.A. [Texas A& M Univ., College Station, TX (United States)
1995-09-01
Condensation in the presence of noncondensible gases plays an important role in the nuclear industry. The RELAP5/MOD3 thermal hydraulic code was used to study the ability of the code to predict this phenomenon. Two separate effects experiments were simulated using this code. These were the Massachusetts Institute of Technology`s (MIT) Pressurizer Experiment, the MIT Single Tube Experiment. A new iterative approach to calculate the interface temperature and the degraded heat transfer coefficient was developed and implemented in the RELAP5/MOD3 thermal hydraulic code. This model employs the heat transfer simultaneously. This model was found to perform much better than the reduction factor approach. The calculations using the new model were found to be in much better agreement with the experimental values.
VanDresar, N. T.; Siegwarth, J. D.
2001-01-01
One reason for NASA's interest in cryogenic two-phase flow with low mass and heat flux is the need to design spacecraft heat exchangers used for vaporizing cryogenic propellants. The CD-ROM provides digitized movies of particular flow patterns observed in experimental work. The movies have been provided in (QuickTime9Trademark) format, encoded at 320w x 240h pixels, 15 fps, using the Sorenson(Trademark) Video Codec for compression. Experiments were conducted to obtain data on the two-phase (liquid and vapor) flow behavior of cryogenic nitrogen and hydrogen under low mass and heat flux conditions. Tests were performed in normal gravity with a 1.5 degree up flow configuration. View ports in the apparatus permitted visual observation of the two-phase flow patterns. Computer codes to predict flow patterns were developed from theoretical/empirical models reported in the literature. Predictions from the computer codes were compared with experimental flow pattern observations. Results are presented employing the traditional two-dimensional flow pattern map format using the liquid and gas superficial velocities as coordinates. In general, the agreement between the experimental results and the analytical predictive methods is reasonably good. Small regions of the flow pattern maps are identified where the models are deficient as a result of neglecting phase change phenomena. Certain regions of the maps were beyond the range of the experiments and could not be completely validated. Areas that could benefit from further work include modeling of the transition from separated flow, collection of additional data in the bubble and annular flow regimes, and collection of experimental data at other inclination angles, tube diameters and high heat flux.
Heat flow anomalies and their interpretation
Chapman, David S.; Rybach, Ladislaus
1985-12-01
More than 10,000 heat flow determinations exist for the earth and the data set is growing steadily at about 450 observations per year. If heat flow is considered as a surface expression of geothermal processes at depth, the analysis of the data set should reveal properties of those thermal processes. They do, but on a variety of scales. For this review heat flow maps are classified by 4 different horizontal scales of 10 n km (n = 1, 2, 3 and 4) and attention is focussed on the interpretation of anomalies which appear with characteristic dimensions of 10 (n - 1) km in the respective representations. The largest scale of 10 4 km encompasses heat flow on a global scale. Global heat loss is 4 × 10 13 W and the process of sea floor spreading is the principal agent in delivering much of this heat to the surface. Correspondingly, active ocean ridge systems produce the most prominent heat flow anomalies at this scale with characteristic widths of 10 3 km. Shields, with similar dimensions, exhibit negative anomalies. The scale of 10 3 km includes continent wide displays. Heat flow patterns at this scale mimic tectonic units which have dimensions of a few times 10 2 km, although the thermal boundaries between these units are sometimes sharp. Heat flow anomalies at this scale also result from plate tectonic processes, and are associated with arc volcanism, back arc basins, hot spot traces, and continental rifting. There are major controversies about the extent to which these surface thermal provinces reflect upper mantle thermal conditions, and also about the origin and evolution of the thermal state of continental lithosphere. Beginning with map dimensions of 10 2 km thermal anomalies of scale 10 1 km, which have a definite crustal origin, become apparent. The origin may be tectonic, geologic, or hydrologic. Ten kilometers is a common wavelength of topographic relief which drives many groundwater flow systems producing thermal anomalies. The largest recognized continental
Auxiliary Heat Exchanger Flow Distribution Test
International Nuclear Information System (INIS)
Kaufman, J.S.; Bressler, M.M.
1983-01-01
The Auxiliary Heat Exchanger Flow Distribution Test was the first part of a test program to develop a water-cooled (tube-side), compact heat exchanger for removing heat from the circulating gas in a high-temperature gas-cooled reactor (HTGR). Measurements of velocity and pressure were made with various shell side inlet and outlet configurations. A flow configuration was developed which provides acceptable velocity distribution throughout the heat exchanger without adding excessive pressure drop
Burnout heat flux in natural flow boiling
International Nuclear Information System (INIS)
Helal, M.M.; Darwish, M.A.; Mahmoud, S.I.
1978-01-01
Twenty runs of experiments were conducted to determine the critical heat flux for natural flow boiling with water flowing upwards through annuli of centrally heated stainless steel tube. The test section has concentric heated tube of 14mm diameter and heated lengthes of 15 and 25 cm. The outside surface of the annulus was formed by various glass tubes of 17.25, 20 and 25.9mm diameter. System pressure is atmospheric. Inlet subcooling varied from 18 to 5 0 C. Obtained critical heat flux varied from 24.46 to 62.9 watts/cm 2 . A number of parameters having dominant influence on the critical heat flux and hydrodynamic instability (flow and pressure oscillations) preceeding the burnout have been studied. These parameters are mass flow rate, mass velocity, throttling, channel geometry (diameters ratio, length to diameter ratio, and test section length), and inlet subcooling. Flow regimes before and at the moments of burnout were observed, discussed, and compared with the existing physical model of burnout
Analysis of two-phase flow and boiling heat transfer in inclined channel of core-catcher
International Nuclear Information System (INIS)
Tahara, M.; Suzuki, Y.; Abe, N.; Kurita, T.; Hamazaki, R.; Kojima, Y.
2008-01-01
Passive Corium Cooling System (CCS) provides a function of ex-vessel debris cooling and molten core stabilization during a severe accident. CCS features inclined cooling channels arranged axi-symmetrically below the core-catcher basin. In order to estimate the coolability of the inclined cooling channel, it is indispensable to identify the flow pattern of the two-phase flow in the cooling channel. Several former studies for the two-phase flow pattern in the inclined channel are referred. Taitel and Dukler (1976) developed a prediction method of the flow pattern transition in horizontal and near horizontal tubes. Barnea et al. (1980) showed the flow pattern map of upward flow with 10 degrees inclination. Sakaguti et al. (1996) observed the two-phase flow patterns in the horizontal pipe connected with slightly upward pipe, in which the flow pattern in the pipe with a bending part was expressed by the combination of a basic flow pattern and some auxiliary flow patterns. Then we investigated these studies In order to identify the flow patterns observed in the inclined cooling channel of CCS. Furthermore we experimentally observed the flow patterns in the inclined cooling channel with various inlet conditions. As a result of the investigation and observation, typical flow patterns in the inclined cooling channel were identified. Two typical flow patterns were observed depending on the steam flow rate, one of which is 'elongated bubble 'flow, and the other is 'churn with collapsing backward and upward slug 'flow The flow and heat transfer in the inclined channel of CCS is analyzed by using a two-phase analysis code employing two-fluid model in which the constitutive equations for the two-phase flow in inclined channels are incorporated. That is, drift flux parameter for each of the elongated bubble flow, and the churn with collapsing backward and upward slug flow are incorporated to the two-phase analysis code, which are based on the rising velocity of the long bubble in
Effect of the configuration of the corner in a narrow rectangular channel on flow and heat transfer
International Nuclear Information System (INIS)
Xu Jianjun; Chen Bingde; Wang Xiaojun
2009-01-01
In order to further understand the effect of the configuration of the corner in a narrow rectangular channel on flow and heat transfer, flow field and temperature field in a narrow rectangular channel were numerical simulated by using CFD code CFX10.0. The results show under the condition of equal quantity of heat of solid which is obtained by decreasing the solid of the corner, the distributions of inside wall temperature for the orthogonal and circular type configurations of the corner are almost the same as that of the archetypal configuration, and those can simulate heat transfer of the archetypal con- figuration. Under the condition of equal Re, secondary flow and friction pressure of the orthogonal type configuration are almost the same as those of the circular type configuration, which shows that the circular type configuration of the corner in a narrow channel can substituted for the archetypal configuration to simulate flow and heat transfer in a narrow rectangular channel. (authors)
Numerical fluid flow and heat transfer calculations on multiprocessor systems
Energy Technology Data Exchange (ETDEWEB)
Oehman, G.A.; Malen, T.E.; Kuusela, P.
1989-01-01
The first part of the report presents the basic principles of parallel processing, and factors influencing tbe efficiency of practical applications are discussed. In a multiprocessor computer, different parts of the program code are executed in parallel, i.e. simultaneous with respect to time, on different processors, and thus it becomes possible to decrease the overall computation time by a factor, which in the ideal case is equal to the number of processors. The application study starts from the numerical solution of the twodimesional Laplace equation, which describes the steady heat conduction in a solid plate and advances through the solution of the three dimensional Laplace equation to the case of study laminar fluid flow in a twodimensional box at Reynolds numbers up to 20. Hereby the stream function-vorticity method is first applied and the SIMPLER method. The conventional (sequential) numerical algoritms for these fluid flow and heat transfer problems are found not to be ideally suited for conversion to parallel computation, but sped-up ratios considerably above 50 % of the theoretical maximum are regularly achieved in the runs. The numerical procedures we coded in the OCCAM-2 language and the test runs were performed at who Akademi on the imperimental HATHI-computers containing 16 T4l4 and 100 INMOS T800 transputers respectively.
Numerical fluid flow and heat transfer calculations on multiprocessor systems
Energy Technology Data Exchange (ETDEWEB)
Oehman, G.A.; Malen, T.E.; Kuusela, P.
1989-12-31
The first part of the report presents the basic principles of parallel processing, and factors influencing tbe efficiency of practical applications are discussed. In a multiprocessor computer, different parts of the program code are executed in parallel, i.e. simultaneous with respect to time, on different processors, and thus it becomes possible to decrease the overall computation time by a factor, which in the ideal case is equal to the number of processors. The application study starts from the numerical solution of the twodimesional Laplace equation, which describes the steady heat conduction in a solid plate and advances through the solution of the three dimensional Laplace equation to the case of study laminar fluid flow in a twodimensional box at Reynolds numbers up to 20. Hereby the stream function-vorticity method is first applied and the SIMPLER method. The conventional (sequential) numerical algoritms for these fluid flow and heat transfer problems are found not to be ideally suited for conversion to parallel computation, but sped-up ratios considerably above 50 % of the theoretical maximum are regularly achieved in the runs. The numerical procedures we coded in the OCCAM-2 language and the test runs were performed at who Akademi on the imperimental HATHI-computers containing 16 T4l4 and 100 INMOS T800 transputers respectively.
Modeling the scooping phenomenon for the heat transfer in liquid–gas horizontal slug flows
International Nuclear Information System (INIS)
Bassani, Carlos L.; Pereira, Fernando H.G.; Barbuto, Fausto A.A.; Morales, Rigoberto E.M.
2016-01-01
Highlights: • A low computational tool for heat transfer prediction on slug flows is presented. • The scooping phenomenon is modeled on a stationary approach. • The scooping phenomenon improved in 8% the heat transfer results. - Abstract: The heat transfer between the deep sea waters and the oil and gas mixtures flowing through production lines is a common situation in the petroleum industry. The optimum prediction of the liquid–gas flow parameters along those lines, when the intermittent flow pattern known as slug flow is dominant, has extreme importance in facilities' design. The mixture temperature drop caused by the colder sea waters, which can be regarded as an infinite medium with constant temperature, directly affects physical properties of the fluids such as the viscosity and specific mass. Gas expansion may also occur due to pressure and temperature gradients, thus changing the flow hydrodynamics. Finally, the temperature gradient affects the thermodynamic equilibrium between the phases, favoring wax deposition and thus increasing pressure drops or even blocking the production line. With those issues in mind, the present work proposes a stationary model to predict the mixture temperature distribution and the two-phase flow heat transfer coefficient based on the mass, momentum and energy conservation equations applied to different unit cell regions. The main contribution of the present work is the modeling of the thermal scooping phenomenon, i.e., the heat transfer between two adjacent unit cells due to the mass flux known as scooping. The model was implemented as a structured Fortran95 code with an upwind difference scheme. The results were compared to experimental data and presented good agreement. The analysis showed that the inclusion of the scooping phenomenon into the model resulted in an averaged 8% improvement in the temperature gradient calculation and heat transfer coefficient prediction for the flowing mixture.
Heat flow in the north-central Colorado Plateau
International Nuclear Information System (INIS)
Bodell, J.M.; Chapman, D.S.
1982-01-01
We report new heat flow measurements at 25 evenly distributed sites in the north-central Colorado Plateau. Heat flow values computed for these new sites and one previously published site range from 43 to 116 mW m -2 but fall into the following district subsets related to physiographic and tectonic elements within the Plateau: (1) heat flow of 51 mW m -2 (12 sites; s.d. 6) in the San Rafael Swell and Green River Desert which constitute the core of the Colorado Plateau at this latitude, (2) heat flows of 69 mW m -2 (5 sites; s.d. 10) in successive parallel north-south bands approaching the Wasatch Plateau to the west but still 80 km east of the Basin and Range physiographic boundary, (3) heat flow of 64 mW m -2 (5 sites; s.d. 2) along the Salt Anticline trend which strikes northwest in the region of Moab, Utah. Heat flow results for the entire Colorado Plateau have been reexamined in view of our new results, and the overall pattern supports the concept of a low heat flow 'thermal interior' for the plateau surrounded by a periphery some 100 km wide having substantially higher heat flow. Average heat flow in the thermal interior is about 60 mW m -2 compared to 80--90 mW m -2 in the periphery. This regional heat flow pattern supports a model of tertiary lithospheric thinning under the Colorado Plateau whereby the plateau is still in transient thermal response and a 15--20 m.y. lag between uplift and corresponding surface heat flow anomaly is to be expected. The position of the heat flow transition between our interior and peripheral regions in the northwest plateau is roughly consistent with lateral warming and weakening of the Colorado Plateau lithosphere initiated at the Basin and Range boundary some 20 m.y. ago
International Nuclear Information System (INIS)
Banas, A.O.; Carver, M.B.; Leung, J.C.H.; Bromley, B.P.
1992-10-01
The general purpose computational fluid dynamics code, Harwell-FLOW3D, has been used to simulate the effects of fuel rod obstructions on pressure drop and heat transfer in single phase turbulent flows in a concentric annular channel. The results of two and three dimensional simulations are reported for obstructions approximating the geometry of bearing pads used in 37 element CANDU fuel bundles. Pressure drop penalty and augmentation of heat transfer have been quantified and correlated with the obstruction geometrical parameters and the dimensionless numbers representing operating conditions. The predicted effects on pressure drop have been compared with several experimental correlations, yielding good agreement. The methodology presented offers results that can be used directly as input into thermalhydraulic analyses in subchannel and system codes. (Author) (23 figs., 15 refs.)
Lunar ash flow with heat transfer.
Pai, S. I.; Hsieh, T.; O'Keefe, J. A.
1972-01-01
The most important heat-transfer process in the ash flow under consideration is heat convection. Besides the four important nondimensional parameters of isothermal ash flow (Pai et al., 1972), we have three additional important nondimensional parameters: the ratio of the specific heat of the gas, the ratio of the specific heat of the solid particles to that of gas, and the Prandtl number. We reexamine the one dimensional steady ash flow discussed by Pai et al. (1972) by including the effects of heat transfer. Numerical results for the pressure, temperature, density of the gas, velocities of gas and solid particles, and volume fraction of solid particles as function of altitude for various values of the Jeffreys number, initial velocity ratio, and two different gas species (steam and hydrogen) are presented.
TACO: a finite element heat transfer code
International Nuclear Information System (INIS)
Mason, W.E. Jr.
1980-02-01
TACO is a two-dimensional implicit finite element code for heat transfer analysis. It can perform both linear and nonlinear analyses and can be used to solve either transient or steady state problems. Either plane or axisymmetric geometries can be analyzed. TACO has the capability to handle time or temperature dependent material properties and materials may be either isotropic or orthotropic. A variety of time and temperature dependent loadings and boundary conditions are available including temperature, flux, convection, and radiation boundary conditions and internal heat generation. Additionally, TACO has some specialized features such as internal surface conditions (e.g., contact resistance), bulk nodes, enclosure radiation with view factor calculations, and chemical reactive kinetics. A user subprogram feature allows for any type of functional representation of any independent variable. A bandwidth and profile minimization option is also available in the code. Graphical representation of data generated by TACO is provided by a companion post-processor named POSTACO. The theory on which TACO is based is outlined, the capabilities of the code are explained, the input data required to perform an analysis with TACO are described. Some simple examples are provided to illustrate the use of the code
International Nuclear Information System (INIS)
Larrauri, D.; Briere, E.
1997-12-01
After different validation simulations of flows through cylindrical and annular channels, a subcooled boiling flow through a rod bundle has been simulated with ASTRID Steam-Water of software. The experiment simulated is called Poseidon. It is a vertical rectangular channel with three heating rods inside. The thermohydraulic conditions of the simulated flow were close to the DNB conditions. The simulation results were analysed and compared against the available measurements of liquid and wall temperatures. ASTRID Steam-Water produced satisfactory results. The wall and the liquid temperatures were well predicted in the different parts of the flow. The void fraction reached 40 % in the vicinity of the heating rods. The distribution of the different calculated variables showed that a three-dimensional simulation gives essential information for the analysis of the physical phenomena involved in this kind of flow. The good results obtained in Poseidon geometry will encourage future rod bundle flow simulations and analyses with ASTRID Steam-Water code. (author)
Benchmarking NNWSI flow and transport codes: COVE 1 results
International Nuclear Information System (INIS)
Hayden, N.K.
1985-06-01
The code verification (COVE) activity of the Nevada Nuclear Waste Storage Investigations (NNWSI) Project is the first step in certification of flow and transport codes used for NNWSI performance assessments of a geologic repository for disposing of high-level radioactive wastes. The goals of the COVE activity are (1) to demonstrate and compare the numerical accuracy and sensitivity of certain codes, (2) to identify and resolve problems in running typical NNWSI performance assessment calculations, and (3) to evaluate computer requirements for running the codes. This report describes the work done for COVE 1, the first step in benchmarking some of the codes. Isothermal calculations for the COVE 1 benchmarking have been completed using the hydrologic flow codes SAGUARO, TRUST, and GWVIP; the radionuclide transport codes FEMTRAN and TRUMP; and the coupled flow and transport code TRACR3D. This report presents the results of three cases of the benchmarking problem solved for COVE 1, a comparison of the results, questions raised regarding sensitivities to modeling techniques, and conclusions drawn regarding the status and numerical sensitivities of the codes. 30 refs
Flow and Heat Transfer Characteristics of Turbulent Gas Flow in Microtube with Constant Heat Flux
International Nuclear Information System (INIS)
Hong, Chungpyo; Matsushita, Shinichi; Ueno, Ichiro; Asako, Yutaka
2012-01-01
Local friction factors for turbulent gas flows in circular microtubes with constant wall heat flux were obtained numerically. The numerical methodology is based on arbitrary-Lagrangian-Eulerian method to solve two-dimensional compressible momentum and energy equations. The Lam-Bremhorst's Low-Reynolds number turbulence model was employed to calculate eddy viscosity coefficient and turbulence energy. The simulations were performed for a wide flow range of Reynolds numbers and Mach numbers with different constant wall heat fluxes. The stagnation pressure was chosen in such a way that the outlet Mach number ranged from 0.07 to 1.0. Both Darcy friction factor and Fanning friction factor were locally obtained. The result shows that the obtained both friction factors were evaluated as a function of Reynolds number on the Moody chart. The values of Darcy friction factor differ from Blasius correlation due to the compressibility effects but the values of Fanning friction factor almost coincide with Blasius correlation. The wall heat flux varied from 100 to 10000 W/m 2 . The wall and bulk temperatures with positive heat flux are compared with those of incompressible flow. The result shows that the Nusselt number of turbulent gas flow is different from that of incompressible flow.
International Nuclear Information System (INIS)
Siddiqui, Faisal A.; Dasgupta, Engr Sarbadaman; Fartaj, Amir
2012-01-01
Highlights: ► Air side heat transfer and flow characteristics of mesochannel cross-flow heat exchanger are studied experimentally. ► Hot ethylene glycol–water mixture (50:50) at constant mass flow rate is used against varying air flow. ► Air side heat transfer and fluid flow key parameters such as Nusselt number, Colburn factor, friction factor are obtained. ► General correlations are proposed for air side heat transfer and fluid flow parameters. - Abstract: Air side force convective heat transfer and flow characteristics of cross-flow mesochannel heat exchanger are investigated experimentally. A series of experiments representing 36 different operating conditions have been conducted on a finned mesochannel heat exchanger through the fully automated dynamic single-phase experimental facility which is capable of handling a wide variety of working fluids in air-to-liquid cross-flow orientation. The mesochannel heat exchanger is made of 15 aluminum slabs with arrays of wavy fins between slabs; 68 one millimeter circular diameter port located at each slab, and the air side frontal area of 304-mm × 304-mm. The ethylene glycol–water mixture as the working fluid in the liquid side was forced to flow through mesochannels maintaining constant inlet temperature and flow rate at 74 °C and 0.0345 kg/s respectively whereas the inlet flowing air into the arrays of wavy fins was changed at four different temperature levels from 28 °C to 43 °C. Frontal air velocity was altered in nine steps from 3 m/s to 11 m/s at each temperature level corresponding range of Reynolds number 752 a a ) and Colburn factor (j a ) were found higher in comparison with other studies.
Improvement of MARS code reflood model
International Nuclear Information System (INIS)
Hwang, Moonkyu; Chung, Bub-Dong
2011-01-01
A specifically designed heat transfer model for the reflood process which normally occurs at low flow and low pressure was originally incorporated in the MARS code. The model is essentially identical to that of the RELAP5/MOD3.3 code. The model, however, is known to have under-estimated the peak cladding temperature (PCT) with earlier turn-over. In this study, the original MARS code reflood model is improved. Based on the extensive sensitivity studies for both hydraulic and wall heat transfer models, it is found that the dispersed flow film boiling (DFFB) wall heat transfer is the most influential process determining the PCT, whereas the interfacial drag model most affects the quenching time through the liquid carryover phenomenon. The model proposed by Bajorek and Young is incorporated for the DFFB wall heat transfer. Both space grid and droplet enhancement models are incorporated. Inverted annular film boiling (IAFB) is modeled by using the original PSI model of the code. The flow transition between the DFFB and IABF, is modeled using the TRACE code interpolation. A gas velocity threshold is also added to limit the top-down quenching effect. Assessment calculations are performed for the original and modified MARS codes for the Flecht-Seaset test and RBHT test. Improvements are observed in terms of the PCT and quenching time predictions in the Flecht-Seaset assessment. In case of the RBHT assessment, the improvement over the original MARS code is found marginal. A space grid effect, however, is clearly seen from the modified version of the MARS code. (author)
Heating patterns during cancer heat therapy as a function of blood flow
International Nuclear Information System (INIS)
Mendecki, J.; Friedenthal, E.; Botstein, C.; Sterzer, F.; Paglione, R.W.
1984-01-01
Heating patterns as a function of regional blood flow were evaluated in healthy tissues with different vascular characteristics as well as in a variety of tumors submitted to microwave and RF-induced hyperthermia. Generally, faster heating and slower cooling was demonstrated for tumors. Definite correlation was found between the power needed to heat given tissue volume to a specific temperature and the ability of this tissue to dissipate heat via vascular flow. The measurements show that during the early phase of heating of tumors temperature rises slowly up to about 40 0 C. indicating good heat exchanges but that at this level rapid increase of temperature occurs for relatively small increments of power input. It is suggested that blood flow in malignant tissue remains competent and responsive to low grade heating, but that at higher temperature levels, in contrast to normal tissue, tumor blood flow rapidly decreases indicating compromised vascular system. Implication for treatment protocols are discussed
Numerical simulation of shell-side heat transfer and flow of natural circulation heat exchanger
International Nuclear Information System (INIS)
Xue Ruojun; Deng Chengcheng; Li Chaojun; Wang Mingyuan
2012-01-01
In order to analyze the influence on the heat transfer and flow characteristics of the heat exchanger model of different solving models and structures, a variety of transformation to the model equivalent for the heat exchanger was studied. In this paper, Fluent software was used to simulate the temperature-field and flow-field of the equivalent model, and investigate its heat-transferring and flow characteristics. Through comparative analysis of the distribution of temperature-field and flow-field for different models, the heat-transferring process and natural convection situation of heat exchanger were deeply understood. The results show that the temperature difference between the inside and outside of the natural circulation heat exchanger tubes is larger and the flow is more complex, so the turbulence model is the more reasonable choice. Asymmetry of tubes position makes the flow and heat transfer of the fluid on both sides to be dissymmetrical and makes the fluid interaction, and increases the role of natural convection. The complex structure of heat exchanger makes the flow and heat transfer of the fluid on both sides to be irregular to some extent when straight tubes into C-bent are transformed, and all these make the turbulence intensity increase and improve the effect of heat transfer. (authors)
Heat transfer in two-phase flow of helium
International Nuclear Information System (INIS)
Subbotin, V.I.; Deev, V.I.; Solodovnikov, V.V.; Arkhipov, V.V.
1986-01-01
The results of experimental study of heat transfer in two-phase helium flow are presented. The effect of operating parameters (pressure, mass velocity, heat flux and quality) on boiling heat transfer intensity was investigated. A significant influence of boiling process prehistory on heat transfer coefficients was demonstrated. On the basis of experimental data obtained three typical regimes of flow boiling heat transfer were found. Analogy of heat transfer in flow boiling and pool boiling of helium and noncryogenic liquids was established. Correlations were developed which are in close agreement with available heat transfer data
Study of two-phase flow redistribution between two passes of a heat exchanger
International Nuclear Information System (INIS)
Mendes de Moura, L.F.
1989-04-01
The object of the present thesis deals with the study of two-phase flow redistribution between two passes of a heat exchanger. Mass flow rate measurements of each component performed at each channel outlet of the second pass allowed us to determine the influence of mass flow, gas quality, flow direction (upward or downward) and common header geometry upon flow redistribution. Local void fraction inside common header was measured with an optical probe. A two-dimensional two-phase flow computational code was developed from a two-fluid model. Modelling of interfacial momentum transfer was used in order to take into account twp-phase flow patterns in common headers. Numerical simulation results show qualitative agreement with experimental results. Present theoretical model limitations are analysed and future improvements are proposed [fr
Modelling of steam condensation in the primary flow channel of a gas-heated steam generator
International Nuclear Information System (INIS)
Kawamura, H.; Meister, G.
1982-10-01
A new simulation code has been developed for the analysis of steam ingress accidents in high temperatures reactors which evaluates the heat transfer in a steam generator headed by a mixture of helium and water steam. Special emphasis is laid on the analysis of steam condensation in the primary circuit of the steam generator. The code takes wall and bulk condensation into account. A new method is proposed to describe the entrainment of water droplets in the primary gas flow. Some typical results are given. Steam condensation in the primary channel may have a significant effect on temperature distributions. The effect on the heat transferred by the steam generator, however, is found to be not so prominent as might be expected. The reason is discussed. A simplified code will also be described, which gives results with reasonable accuracy within much shorter execution times. This code may be used as a program module in a program simulating the total primary circuit of a high temperature reactor. (orig.) [de
Validation of a Computational Fluid Dynamics (CFD) Code for Supersonic Axisymmetric Base Flow
Tucker, P. Kevin
1993-01-01
The ability to accurately and efficiently calculate the flow structure in the base region of bodies of revolution in supersonic flight is a significant step in CFD code validation for applications ranging from base heating for rockets to drag for protectives. The FDNS code is used to compute such a flow and the results are compared to benchmark quality experimental data. Flowfield calculations are presented for a cylindrical afterbody at M = 2.46 and angle of attack a = O. Grid independent solutions are compared to mean velocity profiles in the separated wake area and downstream of the reattachment point. Additionally, quantities such as turbulent kinetic energy and shear layer growth rates are compared to the data. Finally, the computed base pressures are compared to the measured values. An effort is made to elucidate the role of turbulence models in the flowfield predictions. The level of turbulent eddy viscosity, and its origin, are used to contrast the various turbulence models and compare the results to the experimental data.
An analytical model for annular flow boiling heat transfer in microchannel heat sinks
International Nuclear Information System (INIS)
Megahed, A.; Hassan, I.
2009-01-01
An analytical model has been developed to predict flow boiling heat transfer coefficient in microchannel heat sinks. The new analytical model is proposed to predict the two-phase heat transfer coefficient during annular flow regime based on the separated model. Opposing to the majority of annular flow heat transfer models, the model is based on fundamental conservation principles. The model considers the characteristics of microchannel heat sink during annular flow and eliminates using any empirical closure relations. Comparison with limited experimental data was found to validate the usefulness of this analytical model. The model predicts the experimental data with a mean absolute error 8%. (author)
Flow regimes and mechanistic modeling of critical heat flux under subcooled flow boiling conditions
Le Corre, Jean-Marie
Thermal performance of heat flux controlled boiling heat exchangers are usually limited by the Critical Heat Flux (CHF) above which the heat transfer degrades quickly, possibly leading to heater overheating and destruction. In an effort to better understand the phenomena, a literature review of CHF experimental visualizations under subcooled flow boiling conditions was performed and systematically analyzed. Three major types of CHF flow regimes were identified (bubbly, vapor clot and slug flow regime) and a CHF flow regime map was developed, based on a dimensional analysis of the phenomena and available data. It was found that for similar geometric characteristics and pressure, a Weber number (We)/thermodynamic quality (x) map can be used to predict the CHF flow regime. Based on the experimental observations and the review of the available CHF mechanistic models under subcooled flow boiling conditions, hypothetical CHF mechanisms were selected for each CHF flow regime, all based on a concept of wall dry spot overheating, rewetting prevention and subsequent dry spot spreading. It is postulated that a high local wall superheat occurs locally in a dry area of the heated wall, due to a cyclical event inherent to the considered CHF two-phase flow regime, preventing rewetting (Leidenfrost effect). The selected modeling concept has the potential to span the CHF conditions from highly subcooled bubbly flow to early stage of annular flow. A numerical model using a two-dimensional transient thermal analysis of the heater undergoing nucleation was developed to mechanistically predict CHF in the case of a bubbly flow regime. In this type of CHF two-phase flow regime, the high local wall superheat occurs underneath a nucleating bubble at the time of bubble departure. The model simulates the spatial and temporal heater temperature variations during nucleation at the wall, accounting for the stochastic nature of the boiling phenomena. The model has also the potential to evaluate
Exhaust bypass flow control for exhaust heat recovery
Reynolds, Michael G.
2015-09-22
An exhaust system for an engine comprises an exhaust heat recovery apparatus configured to receive exhaust gas from the engine and comprises a first flow passage in fluid communication with the exhaust gas and a second flow passage in fluid communication with the exhaust gas. A heat exchanger/energy recovery unit is disposed in the second flow passage and has a working fluid circulating therethrough for exchange of heat from the exhaust gas to the working fluid. A control valve is disposed downstream of the first and the second flow passages in a low temperature region of the exhaust heat recovery apparatus to direct exhaust gas through the first flow passage or the second flow passage.
Flow and heat transfer in parallel channel attached with equally-spaced ribs, 2
International Nuclear Information System (INIS)
Kunugi, Tomoaki; Takizuka, Takakazu
1980-09-01
Using a computer code for the analysis of the flow and heat transfer in a parallel channel attached with equally-spaced ribs, calculations are performed when a pitch to rib-width ratio is 7 : 1, a rib-width to rib-height ratio is 2 : 1 and a channel-height to rib-height is 3 : 1. Assuming that the fluid properties and the heat-flux at the wall of this channel are constant, characteristics of the flow and heat transfer are analyzed in the range of Reynolds number from 10 to 250. The following results are obtained: (1) The separation region behind a rib grows downstream with the increase of Reynolds number. (2) The pressure drop of ribbed channel is greater than that of the smooth channel, and increases as Reynolds number increases. (3) The mean Nusselt number of ribbed channel is about 10 - 11 at the upper wall and about 7.5 at the lower wall in the range of Reynolds number from 10 to 250. (author)
Simulation of boiling flow in evaporator of separate type heat pipe with low heat flux
International Nuclear Information System (INIS)
Kuang, Y.W.; Wang, Wen; Zhuan, Rui; Yi, C.C.
2015-01-01
Highlights: • A boiling flow model in a separate type heat pipe with 65 mm diameter tube. • Nucleate boiling is the dominant mechanism in large pipes at low mass and heat flux. • The two-phase heat transfer coefficient is less sensitive to the total mass flux. - Abstract: The separate type heat pipe heat exchanger is considered to be a potential selection for developing passive cooling spent fuel pool – for the passive pressurized water reactor. This paper simulates the boiling flow behavior in the evaporator of separate type heat pipe, consisting of a bundle of tubes of inner diameter 65 mm. It displays two-phase characteristic in the evaporation section of the heat pipe working in low heat flux. In this study, the two-phase flow model in the evaporation section of the separate type heat pipe is presented. The volume of fluid (VOF) model is used to consider the interaction between the ammonia gas and liquid. The flow patterns and flow behaviors are studied and the agitated bubbly flow, churn bubbly flow are obtained, the slug bubble is likely to break into churn slug or churn froth flow. In addition, study on the heat transfer coefficients indicates that the nucleate boiling is the dominant mechanism in large pipes at low mass and heat flux, with the heat transfer coefficient being less sensitive to the total mass flux
Directory of Open Access Journals (Sweden)
Christophe Morel
2009-01-01
Full Text Available This paper describes the modeling of boiling multisize bubbly flows and its application to the simulation of the DEBORA experiment. We follow the method proposed originally by Kamp, assuming a given mathematical expression for the bubble diameter pdf. The original model is completed by the addition of some new terms for vapor compressibility and phase change. The liquid-to-interface heat transfer term, which essentially determines the bubbles condensation rate in the DEBORA experiment, is also modeled with care. First numerical results realized with the Neptune_CFD code are presented and discussed.
Papadimitriou, P.; Skorek, T.
THESUS is a thermohydraulic code for the calculation of steady state and transient processes of two-phase cryogenic flows. The physical model is based on four conservation equations with separate liquid and gas phase mass conservation equations. The thermohydraulic non-equilibrium is calculated by means of evaporation and condensation models. The mechanical non-equilibrium is modeled by a full-range drift-flux model. Also heat conduction in solid structures and heat exchange for the full spectrum of heat transfer regimes can be simulated. Test analyses of two-channel chilldown experiments and comparisons with the measured data have been performed.
Critical heat flux and exit film flow rate in a flow boiling system
International Nuclear Information System (INIS)
Ueda, Tatsuhiro; Isayama, Yasushi
1981-01-01
The critical heat flux in a flowing boiling system is an important problem in the evaporating tubes with high thermal load such as nuclear reactors and boilers, and gives the practical design limit. When the heat flux in uniformly heated evaporating tubes is gradually raised, the tube exit quality increases, and soon, the critical heat flux condition arises, and the wall temperature near tube exit rises rapidly. In the region of low exit quality, the critical heat flux condition is caused by the transition from nucleating boiling, and in the region of high exit quality, it is caused by dry-out. But the demarcation of both regions is not clear. In this study, for the purpose of obtaining the knowledge concerning the critical heat flux condition in a flowing boiling system, the relation between the critical heat flux and exit liquid film flow rate was examined. For the experiment, a uniformly heated vertical tube supplying R 113 liquid was used, and the measurement in the range of higher heating flux and mass velocity than the experiment by Ueda and Kin was carried out. The experimental setup and experimental method, the critical heat flux and exit quality, the liquid film flow rate at heating zone exit, and the relation between the critical heat flux and the liquid film flow rate at exit are described. (Kako, I.)
Sensitivity analysis of the Gupta and Park chemical models on the heat flux by DSMC and CFD codes
Morsa, Luigi; Festa, Giandomenico; Zuppardi, Gennaro
2012-11-01
The present study is the logical continuation of a former paper by the first author in which the influence of the chemical models by Gupta and by Park on the computation of heat flux on the Orion and EXPERT capsules was evaluated. Tests were carried out by the direct simulation Monte Carlo code DS2V and by the computational fluiddynamic (CFD) code H3NS. DS2V implements the Gupta model, while H3NS implements the Park model. In order to compare the effects of the chemical models, the Park model was implemented also in DS2V. The results showed that DS2V and H3NS compute a different composition both in the flow field and on the surface, even using the same chemical model (Park). Furthermore DS2V computes, by the two chemical models, different compositions in the flow field but the same composition on the surface, therefore the same heat flux. In the present study, in order to evaluate the influence of these chemical models also in a CFD code, the Gupta and the Park models have been implemented in FLUENT. Tests by DS2V and by FLUENT, have been carried out for the EXPERT capsule at the altitude of 70 km and with velocity of 5000 m/s. The capsule experiences a hypersonic, continuum low density regime. Due to the energy level of the flow, the vibration equation, lacking in the original version of FLUENT, has been implemented. The results of the heat flux computation verify that FLUENT is quite sensitive to the Gupta and to the Park chemical models. In fact, at the stagnation point, the percentage difference between the models is about 13%. On the opposite the DS2V results by the two models are practically equivalent.
Verification and Validation of Heat Transfer Model of AGREE Code
Energy Technology Data Exchange (ETDEWEB)
Tak, N. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Seker, V.; Drzewiecki, T. J.; Downar, T. J. [Department of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Michigan (United States); Kelly, J. M. [US Nuclear Regulatory Commission, Washington (United States)
2013-05-15
The AGREE code was originally developed as a multi physics simulation code to perform design and safety analysis of Pebble Bed Reactors (PBR). Currently, additional capability for the analysis of Prismatic Modular Reactor (PMR) core is in progress. Newly implemented fluid model for a PMR core is based on a subchannel approach which has been widely used in the analyses of light water reactor (LWR) cores. A hexagonal fuel (or graphite block) is discretized into triangular prism nodes having effective conductivities. Then, a meso-scale heat transfer model is applied to the unit cell geometry of a prismatic fuel block. Both unit cell geometries of multi-hole and pin-in-hole types of prismatic fuel blocks are considered in AGREE. The main objective of this work is to verify and validate the heat transfer model newly implemented for a PMR core in the AGREE code. The measured data in the HENDEL experiment were used for the validation of the heat transfer model for a pin-in-hole fuel block. However, the HENDEL tests were limited to only steady-state conditions of pin-in-hole fuel blocks. There exist no available experimental data regarding a heat transfer in multi-hole fuel blocks. Therefore, numerical benchmarks using conceptual problems are considered to verify the heat transfer model of AGREE for multi-hole fuel blocks as well as transient conditions. The CORONA and GAMMA+ codes were used to compare the numerical results. In this work, the verification and validation study were performed for the heat transfer model of the AGREE code using the HENDEL experiment and the numerical benchmarks of selected conceptual problems. The results of the present work show that the heat transfer model of AGREE is accurate and reliable for prismatic fuel blocks. Further validation of AGREE is in progress for a whole reactor problem using the HTTR safety test data such as control rod withdrawal tests and loss-of-forced convection tests.
SCDAP/RELAP5 Modeling of Heat Transfer and Flow Losses in Lower Head Porous Debris
International Nuclear Information System (INIS)
Coryell, E.W.; Siefken, L.J.; Paik, S.
1998-01-01
Designs are described for implementing models for calculating the heat transfer and flow losses in porous debris in the lower head of a reactor vessel. The COUPLE model in SCDAP/RELAP5 represents both the porous and non-porous debris that results from core material slumping into the lower head. Currently, the COUPLE model has the capability to model convective and radiative heat transfer from the surfaces of non-porous debris in a detailed manner and to model only in a simplistic manner the heat transfer from porous debris. In order to advance beyond the simplistic modeling for porous debris, designs are developed for detailed calculations of heat transfer and flow losses in porous debris. Correlations are identified for convective heat transfer in porous debris for the following modes of heat transfer; (1) forced convection to liquid, (2) forced convection to gas, (3) nucleate boiling, (4) transition boiling, and (5) film boiling. Interphase heat transfer is modeled in an approximate manner. A design is also described for implementing a model of heat transfer by radiation from debris to the interstitial fluid. A design is described for implementation of models for flow losses and interphase drag in porous debris. Since the models for heat transfer and flow losses in porous debris in the lower head are designed for general application, a design is also described for implementation of these models to the analysis of porous debris in the core region. A test matrix is proposed for assessing the capability of the implemented models to calculate the heat transfer and flow losses in porous debris. The implementation of the models described in this report is expected to improve the COUPLE code calculation of the temperature distribution in porous debris and in the lower head that supports the debris. The implementation of these models is also expected to improve the calculation of the temperature and flow distribution in porous debris in the core region
Two-phase flow heat transfer in nuclear reactor systems
International Nuclear Information System (INIS)
Koncar, Bostjan; Krepper, Eckhard; Bestion, Dominique; Song, Chul-Hwa; Hassan, Yassin A.
2013-01-01
occurring in at least two different spatial scales. Uncertainty in modelling of bubble departure diameter at boiling was studied by M. Matkovic and B. Koncar. In this article the propagation of input uncertainties for the simplified model of bubble departure size is evaluated. A methodology for estimating the prediction capability of a given correlation is provided taking into account its range of applicability. Aqueous nanofluids have a great potential for cooling applications, hence they have been studied in the article of P.N. Alekseev et al. as a possible coolant in pressurized water reactor (PWR). The theoretical study presents how a stable formation of nanoparticles in water solution can be established. Formation of fractal nanoparticles with a higher thermal conductivity than water can enhance the heat transfer of water used as a coolant in PWR. Apart from solid particles, also alternative formation of gaseous nanoparticles in density fluctuations of water is discussed. The article of R. Rzehak and E. Krepper provides a comprehensive overview of the state-of-the-art in the field of CFD modelling of subcooled flow boiling. The efficient predictive capability of current models requires calibration of model parameters over a wide range of measured data and operating conditions.The results presented in the article confirmed the great potential of the existing modelling approach for the 3D simulation of subcooled flow boiling in industrial applications but also highlight the need for specific model improvements to achieve highly accurate predictions. Two articles deal with one-dimensional analyses of two phase flows. In the article of O.Costa et al., a rapid depressurization in vertical heated pipe is simulated with the in-house 1D computer code WAHA, which was developed specifically for simulations of two-phase water hammer phenomena. The WAHA results were confronted with the simulations of the well-known system code RELAP5 on the same experimental data. The thermal
Inclusion of pressure and flow in the KITES MHD equilibrium code
International Nuclear Information System (INIS)
Raburn, Daniel; Fukuyama, Atsushi
2013-01-01
One of the simplest self-consistent models of a plasma is single-fluid magnetohydrodynamic (MHD) equilibrium with no bulk fluid flow under axisymmetry. However, both fluid flow and non-axisymmetric effects can significantly impact plasma equilibrium and confinement properties: in particular, fluid flow can produce profile pedestals, and non-axisymmetric effects can produce islands and stochastic regions. There exist a number of computational codes which are capable of calculating equilibria with arbitrary flow or with non-axisymmetric effects. Previously, a concept for a code to calculate MHD equilibria with flow in non-axisymmetric systems was presented, called the KITES (Kyoto ITerative Equilibrium Solver) code. Since then, many of the computational modules for the KITES code have been completed, and the work-in-progress KITES code has been used to calculate non-axisymmetric force-free equilibria. Additional computational modules are required to allow the KITES code to calculate equilibria with pressure and flow. Here, the authors report on the approaches used in developing these modules and provide a sample calculation with pressure. (author)
Nonsteady heat conduction code with radiation boundary conditions
International Nuclear Information System (INIS)
Fillo, J.A.; Benenati, R.; Powell, J.
1975-01-01
A heat-transfer model for studying the temperature build-up in graphite blankets for fusion reactors is presented. In essence, the computer code developed is for two-dimensional, nonsteady heat conduction in heterogeneous, anisotropic solids with nonuniform internal heating. Thermal radiation as well as bremsstrahlung radiation boundary conditions are included. Numerical calculations are performed for two design options by varying the wall loading, bremsstrahlung, surface layer thickness and thermal conductivity, blanket dimensions, time step and grid size. (auth)
Phenomenological modeling of critical heat flux: The GRAMP code and its validation
International Nuclear Information System (INIS)
Ahmad, M.; Chandraker, D.K.; Hewitt, G.F.; Vijayan, P.K.; Walker, S.P.
2013-01-01
Highlights: ► Assessment of CHF limits is vital for LWR optimization and safety analysis. ► Phenomenological modeling is a valuable adjunct to pure empiricism. ► It is based on empirical representations of the (several, competing) phenomena. ► Phenomenological modeling codes making ‘aggregate’ predictions need careful assessment against experiments. ► The physical and mathematical basis of a phenomenological modeling code GRAMP is presented. ► The GRAMP code is assessed against measurements from BARC (India) and Harwell (UK), and the Look Up Tables. - Abstract: Reliable knowledge of the critical heat flux is vital for the design of light water reactors, for both safety and optimization. The use of wholly empirical correlations, or equivalently “Look Up Tables”, can be very effective, but is generally less so in more complex cases, and in particular cases where the heat flux is axially non-uniform. Phenomenological models are in principle more able to take into account of a wider range of conditions, with a less comprehensive coverage of experimental measurements. These models themselves are in part based upon empirical correlations, albeit of the more fundamental individual phenomena occurring, rather than the aggregate behaviour, and as such they too require experimental validation. In this paper we present the basis of a general-purpose phenomenological code, GRAMP, and then use two independent ‘direct’ sets of measurement, from BARC in India and from Harwell in the United Kingdom, and the large dataset embodied in the Look Up Tables, to perform a validation exercise on it. Very good agreement between predictions and experimental measurements is observed, adding to the confidence with which the phenomenological model can be used. Remaining important uncertainties in the phenomenological modeling of CHF, namely the importance of the initial entrained fraction on entry to annular flow, and the influence of the heat flux on entrainment rate
Investigation of Abnormal Heat Transfer and Flow in a VHTR Reactor Core
Energy Technology Data Exchange (ETDEWEB)
Kawaji, Masahiro [City College of New York, NY (United States); Valentin, Francisco I. [City College of New York, NY (United States); Artoun, Narbeh [City College of New York, NY (United States); Banerjee, Sanjoy [City College of New York, NY (United States); Sohal, Manohar [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schultz, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States); McEligot, Donald M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)
2015-12-21
The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spots in the VHTR core.
Investigation of Abnormal Heat Transfer and Flow in a VHTR Reactor Core
International Nuclear Information System (INIS)
Kawaji, Masahiro; Valentin, Francisco I.; Artoun, Narbeh; Banerjee, Sanjoy; Sohal, Manohar; Schultz, Richard; McEligot, Donald M.
2015-01-01
The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spots in the VHTR core.
Analysis of the flow structure and heat transfer in a vertical mantle heat exchanger
DEFF Research Database (Denmark)
Knudsen, Søren; Morrison, GL; Behnia, M
2005-01-01
initially mixed and initially stratified inner tank and mantle. The analysis of the heat transfer showed that the flow in the mantle near the inlet is mixed convection flow and that the heat transfer is dependent on the mantle inlet temperature relative to the core tank temperature at the mantle level. (C......The flow structure inside the inner tank and inside the mantle of a vertical mantle heat exchanger was investigated using a full-scale tank designed to facilitate flow visualisation. The flow structure and velocities in the inner tank and in the mantle were measured using a Particle Image...... Velocimetry (PIV) system. A Computational Fluid Dynamics (CFD) model of the vertical mantle heat exchanger was also developed for a detailed evaluation of the heat flux at the mantle wall and at the tank wall. The flow structure was evaluated for both high and low temperature incoming flows and for both...
Heat transfer investigation of molten salts under laminar and turbulent flow regimes
International Nuclear Information System (INIS)
Srivastava, A.K.; Vaidya, A.M.; Maheshwari, N.K.; Vijayan, P.K.
2014-01-01
High temperature reactor and solar thermal power plants use Molten Salt as a coolant, as it has low melting point and high boiling point, enabling us to operate the system at low pressure. Molten fluoride salt (eutectic mixture of LiF-NaF-KF) and molten nitrate salt (mixture of NaNO 3 and KNO 3 in 60:40 ratios by weight) are proposed as a candidate coolant for High Temperature Reactors (HTR) and solar power plant respectively. BARC is developing a 600 MWth pebble bed high temperature reactor, cooled by natural circulation of fluoride salt and capable of supplying process heat at 1000℃ to facilitate hydrogen production by splitting water. Beside this, BARC is also developing a 2MWe solar power tower system using molten nitrate salt as a primary coolant and storage medium. In order to design this, it is necessary to study the heat transfer characteristics of various molten salts. Most of the previous studies related to molten salts are based on the experimental works. These experiments essentially measured the physical properties of molten salts and their heat transfer characteristics. Ferri et al. introduced the property definitions for molten salts in the RELAP5 code to perform transient simulations at the ProvaCollettoriSolari (PCS) test facility. In this paper, a CFD analysis has been performed to study the heat transfer characteristics of molten fluoride salt and molten nitrate salt flowing in a circular pipe for various regimes of flow. Simulation is performed with the help of in-house developed CFD code, NAFA, acronym for Numerical Analysis of Flows in Axi-symmetric geometries. Uniform velocity and temperature distribution are set as the inlet boundary condition and pressure is employed at the outlet boundary condition. The inlet temperature for all simulation is set as 300℃ for nitrate salt and 500℃ for fluoride salt and the operating pressure is 1 atm in both the cases
Thaw flow control for liquid heat transport systems
Kirpich, Aaron S.
1989-01-01
In a liquid metal heat transport system including a source of thaw heat for use in a space reactor power system, the thaw flow throttle or control comprises a fluid passage having forward and reverse flow sections and a partition having a plurality of bleed holes therein to enable fluid flow between the forward and reverse sections. The flow throttle is positioned in the system relatively far from the source of thaw heat.
The FLUFF code for calculating finned surface heat transfer -description and user's guide
International Nuclear Information System (INIS)
Fry, C.J.
1985-08-01
FLUFF is a computer code for calculating heat transfer from finned surfaces by convection and radiation. It can also represent heat transfer by radiation to a partially emitting and absorbing medium within the fin cavity. The FLUFF code is useful not only for studying the behaviour of finned surfaces but also for deriving heat fluxes which can be applied as boundary conditions to other heat transfer codes. In this way models of bodies with finned surfaces may be greatly simplified since the fins need not be explicitly represented. (author)
International Nuclear Information System (INIS)
Bezrukov, Yu.A.; Shchekoldin, V.I.
2002-01-01
On the basis of the Gidropress OKB (Special Design Bureau) experimental data bank one verified the KORSAR code design models and correlations as to heat exchange crisis and overcrisis heat transfer as applied to the WWER reactor normal and emergency conditions. The VI.006.000 version of KORSAR code base calculations is shown to describe adequately the conducted experiments and to deviate insignificantly towards the conservative approach. So it may be considered as one of the codes ensuring more precise estimation [ru
Study on boiling heat transfer of subcooled flow under oscillatory flow condition
International Nuclear Information System (INIS)
Ohtake, Hiroyasu; Yamazaki, Satoshi; Koizumi, Yasuo
2004-01-01
The Onset of Nucleate Boiling, the point of Net Vapor Generation and Critical Heat Flux on subcooled flow boiling under oscillatory flow, focusing on liquid velocity, amplitude and frequency of oscillatory flow were investigated experimentally and analytically. Experiments were conducted using a copper thin-film and subcooled water in a range of the liquid velocity from 0.27 to 4.07 m/s at 0.10MPa. The liquid subcooling was 20K. Frequency of oscillatory flow was 2 and 4 Hz, respectively; amplitude of oscillatory flow was 25 and 50% in a ratio of main flow rate, respectively. Temperatures at Onset of Nuclear Boiling and Critical Heat Flux obtained in the experiments decreased with the oscillatory flow. The decrease of liquid velocity by oscillatory flow caused the ONB and the CHF to decrease. On the other hand, heat flux at Net Vapor Generation decreased with oscillatory flow; the increase of liquid velocity by oscillatory flow caused the NVG to decrease. (author)
Predictions of Critical Heat Flux Using the ASSERT-PV Subchannel Code for a CANFLEX Variant Bundle
International Nuclear Information System (INIS)
Onder, Ebru Nihan; Leung, Laurence; Kim, Hung; Rao, Yanfei
2009-01-01
The ASSERT-PV subchannel code developed by AECL has been applied as a design-assist tool to the advanced CANDU 1 reactor fuel bundle. Based primarily on the CANFLEX 2 fuel bundle, several geometry changes (such as element sizes and pitchcircle diameters of various element rings) were examined to optimize the dryout power and pressure-drop performances of the new fuel bundle. An experiment was performed to obtain dryout power measurements for verification of the ASSERT-PV code predictions. It was carried out using an electrically heated, Refrigerant-134a cooled, fuel bundle string simulator. The axial power profile of the simulator was uniform, while the radial power profile of the element rings was varied simulating profiles in bundles with various fuel compositions and burn-ups. Dryout power measurements are predicted closely using the ASSERT-PV code, particularly at low flows and low pressures, but are overpredicted at high flows and high pressures. The majority of data shows that dryout powers are underpredicted at low inlet-fluid temperatures but overpredicted at high inlet-fluid temperatures
Bridging Inter-flow and Intra-flow Network Coding for Video Applications
DEFF Research Database (Denmark)
Hansen, Jonas; Krigslund, Jeppe; Roetter, Daniel Enrique Lucani
2013-01-01
transmission approach to decide how much and when to send redundancy in the network, and a minimalistic feedback mechanism to guarantee delivery of generations of the different flows. Given the delay constraints of video applications, we proposed a simple yet effective coding mechanism, Block Coding On The Fly...
Directory of Open Access Journals (Sweden)
Ning Wang
2014-06-01
Full Text Available Supercritical convective heat transfer characteristics of hydrocarbon fuel play a fundamental role in the active cooling technology of scramjet. In this paper, a 2D-axisymmetric numerical study of supercritical heat transfer of RP3 flowing inside the cooling channels of scramjet has been conducted. The main thermophysical properties of RP3, including density, specific heat, and thermal conductivity, are obtained from experimental data, while viscosity is evaluated from a commercial code with a ten-species surrogate. Effects of heat flux, mass flow rate, and inlet temperature on supercritical heat transfer processes have been investigated. Results indicate that when the wall temperature rises above the pseudocritical temperature of RP3, heat transfer coefficient decreases as a result of drastic decrease of the specific heat. The conventional heat transfer correlations, that is, Gnielinski formula, are no longer proper for the supercritical heat transfer of RP3. The modified Jackson and Hall formula, which was proposed for supercritical CO2 and water, gives good prediction except when the wall temperature is near or higher than the pseudocritical temperature.
International Nuclear Information System (INIS)
Vitorello, I.
1978-01-01
Heat flow and heat production results are reported from nineteen widely spaced sites in eastern and central parts of Brazil. Three sites in the stable Sao Francisco Craton comprising rocks with Transamazonic ages (2600 to 1800 Ma) or older present an average heat flow of 41.8 +- 4.6 (standard error of the mean=sem) mW m -2 , typical of shield areas; eight sites located in the Late Precambrian Braziliane metamorphic belt have an average heat flow of 54.7 +- 3.8 (sem) mW m -2 ; and four sites in the Parana basin, locus of a Late Jurassic-Early Cretaceous basaltic volcanicity, have a mean heat flow of 70.1 +- 5.9 (sem) mW m -2 . Heat flow results from the Late Cretaceous-Early Tertiary alkalic intrusion of Pocos de Caldas have yielded a site mean of 55.3 mW m -2 . These results indicate a systematic decrease of heat flow with increasing age of the last tectonothermal event. As an explanation for this pattern, a model comprising three main heat flow components is advanced: radiogenic heat from the crust (40%), with the decrease of this contribution with time being achieved by erosional removal of radioactive material; a residual heat from a transient thermal perturbation associated with tectogenesis; and a uniform heat flow of about 28 mW m -2 from deeper sources. The Coastal Brazilian Shield is characterized by ordinary surface and reduced heat flow, but its heat production appears to be less concentrated near the surface, and distributed over a greater depth. Because of the variation in plate thickness, relative movements between the South American plate and the underlying mantle material are possibly constrained to depths exceeding 400 km
Heat transfer and fluid flow in minichannels and microchannels
Kandlikar, Satish; Li, Dongqing; Colin, Stephane; King, Michael R
2014-01-01
Heat exchangers with minichannel and microchannel flow passages are becoming increasingly popular due to their ability to remove large heat fluxes under single-phase and two-phase applications. Heat Transfer and Fluid Flow in Minichannels and Microchannels methodically covers gas, liquid, and electrokinetic flows, as well as flow boiling and condensation, in minichannel and microchannel applications. Examining biomedical applications as well, the book is an ideal reference for anyone involved in the design processes of microchannel flow passages in a heat exchanger. Each chapter is accompan
International Nuclear Information System (INIS)
Chen Yunmei
1994-01-01
In this paper we study the heat flow of harmonic maps between two compact Riemannian manifolds. The global existence of the regular solution and the weak solution, as well as the blow up of the weak solution are discussed. (author). 14 refs
Critical heat flux and flow pattern for water flow in annular geometry
International Nuclear Information System (INIS)
Park, Jae Wook; Baek, Won Pil; Chang, Soon Heung
1996-01-01
An experimental study on critical heat flux (CHF) and two-phase flow visualization has been performed for water flow in internally-heated, vertical, concentric annuli under near atmospheric pressure. Tests have been done under stable forced-circulation, upward and downward flow conditions with three test sections of relatively large gap widths (heated length = 0.6 m, inner diameter = 19 mm, outer diameter = 29, 35 and 51 mm). The outer wall of the test section was made up of the transparent Pyrex tube to allow the observation of flow patterns near the CHF occurrence. The CHF mechanism was changed in the order of flooding, churn-to-annular flow transition, and local dryout under a large bubble in churn flow as the flow rate was increased from zero to higher values. Observed parametric trends are consistent with the previous understanding except that the CHF for downward flow is considerably lower than that for upward flow
The improvement of the heat transfer model for sodium-water reaction jet code
International Nuclear Information System (INIS)
Hashiguchi, Yoshirou; Yamamoto, Hajime; Kamoshida, Norio; Murata, Shuuichi
2001-02-01
For confirming the reasonable DBL (Design Base Leak) on steam generator (SG), it is necessary to evaluate phenomena of sodium-water reaction (SWR) in an actual steam generator realistically. The improvement of a heat transfer model on sodium-water reaction (SWR) jet code (LEAP-JET ver.1.40) and application analysis to the water injection tests for confirmation of propriety for the code were performed. On the improvement of the code, the heat transfer model between a inside fluid and a tube wall was introduced instead of the prior model which was heat capacity model including both heat capacity of the tube wall and inside fluid. And it was considered that the fluid of inside the heat exchange tube was able to treat as water or sodium and typical heat transfer equations used in SG design were also introduced in the new heat transfer model. Further additional work was carried out in order to improve the stability of the calculation for long calculation time. The test calculation using the improved code (LEAP-JET ver.1.50) were carried out with conditions of the SWAT-IR·Run-HT-2 test. It was confirmed that the SWR jet behavior on the result and the influence to the result of the heat transfer model were reasonable. And also on the improved code (LEAP-JET ver.1.50), user's manual was revised with additional I/O manual and explanation of the heat transfer model and new variable name. (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)
Two phase nonequilibrium heat transfer in the TRAC-PD2 code
International Nuclear Information System (INIS)
Mandell, D.A.; Liles, D.R.
1980-01-01
TRAC is a best-estimate, multidimensional, nonequilibrium computer code intended for the analysis of loss-of-coolant accidents (LOCA's) in light water reactors. TRAC-PD2 is the third, detailed, pressurized water reactor version of the code. The TRAC code is modular both by components and by function. That is, vessels, pipes, pumps, etc. can be coupled together in any manner in order to simulate a reactor or a particular experimental facility. Individual physical phenomena are also coded in separate subroutines. This paper discusses the wall to fluid heat transfer coefficient correlations, the interfacial heat transfer models, and presents results for several experimental facilities
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)
Heat transfer of liquid-metal magnetohydrodynamic flow with internal heat generation
International Nuclear Information System (INIS)
Kumamaru, Hiroshige; Kurita, Kazuhisa; Kodama, Satoshi
2000-01-01
Numerical calculations on heat transfer of a magnetohydrodynamic (MHD) flow with internal heat generation in a rectangular channel have been performed for the cases of very-large Hartmann numbers, finite wall conductivities and small aspect ratio (i.e. small length ratios of the channel side perpendicular to the applied magnetic field and the side parallel to the field), simulating typical conditions for a fusion-reactor blanket. The Nusselt numbers of the MHD flow in rectangular channels with aspect ratios of 1/10 to 1/40 for Hartmann numbers of ∼5 x 10 5 become ∼10 times higher than those for the corresponding flow under no magnetic field. The Nusselt number becomes higher as the internal heat generation rate increases as far as the heat generation rates in a fusion reactor blanket are considered. (author)
Two-phase computer codes for zero-gravity applications
International Nuclear Information System (INIS)
Krotiuk, W.J.
1986-10-01
This paper discusses the problems existing in the development of computer codes which can analyze the thermal-hydraulic behavior of two-phase fluids especially in low gravity nuclear reactors. The important phenomenon affecting fluid flow and heat transfer in reduced gravity is discussed. The applicability of using existing computer codes for space applications is assessed. Recommendations regarding the use of existing earth based fluid flow and heat transfer correlations are made and deficiencies in these correlations are identified
Heat flow during sawtooth collapse in tokamak plasmas
International Nuclear Information System (INIS)
Hanada, Kazuaki
1994-01-01
Heat flow during sawtooth collapse was studied on the WT-3 tokamak by using temporal evolution of soft X-ray intensity profile in the poloidal cross section in a lower hybrid current driven plasma as well as an electron cyclotron heated plasma. Two phase in sawtooth collapses were observed. In the first phases, the hottest spot that is the peak of the soft X-ray distribution approaches the inversion surface and heat flows out through a narrow gate on the inversion surface. In the second phase, the hottest spot stays on the inversion surface, and heat flows out through the whole inversion surface. This suggests that magnetic reconnection as predicted by Kadomtsev's model occurs in the first phase, but in the second phase, a different mechanism dominates heat flow. (author)
Heat flow, heat transfer and lithosphere rheology in geothermal areas: Features and examples
Ranalli, G.; Rybach, L.
2005-10-01
Surface heat flow measurements over active geothermal systems indicate strongly positive thermal anomalies. Whereas in "normal" geothermal settings, the surface heat flow is usually below 100-120 mW m - 2 , in active geothermal areas heat flow values as high as several watts per meter squared can be found. Systematic interpretation of heat flow patterns sheds light on heat transfer mechanisms at depth on different lateral, depth and time scales. Borehole temperature profiles in active geothermal areas show various signs of subsurface fluid movement, depending on position in the active system. The heat transfer regime is dominated by heat advection (mainly free convection). The onset of free convection depends on various factors, such as permeability, temperature gradient and fluid properties. The features of heat transfer are different for single or two-phase flow. Characteristic heat flow and heat transfer features in active geothermal systems are demonstrated by examples from Iceland, Italy, New Zealand and the USA. Two main factors affect the rheology of the lithosphere in active geothermal areas: steep temperature gradients and high pore fluid pressures. Combined with lithology and structure, these factors result in a rheological zonation with important consequences both for geodynamic processes and for the exploitation of geothermal energy. As a consequence of anomalously high temperature, the mechanical lithosphere is thin and its total strength can be reduced by almost one order of magnitude with respect to the average strength of continental lithosphere of comparable age and thickness. The top of the brittle/ductile transition is located within the upper crust at depths less than 10 km, acts as the root zone of listric normal faults in extensional environments and, at least in some cases, is visible on seismic reflection lines. These structural and rheological features are well illustrated in the Larderello geothermal field in Tuscany.
International Nuclear Information System (INIS)
Park, R. J.; Kang, K. H.; Kim, S. B.; Kim, H. D.; Choi, S. M.
1998-01-01
Analytical studies have been performed on natural convection heat transfer with crust formation in a molten metal pool to validate and evaluate experimental data using the CONV-2 and 3D computer codes. Two types of steady state tests, a low and high geometric aspect ratio case in the molten metal pool, were performed to investigate crust thickness as a function of boundary conditions. The CONV-2 and 3D computer codes were developed under the OECD/NEA RASPLAV project to simulate two- and three-dimensional natural convection heat transfer with crust formation, respectively. The Rayleigh-Benard flow patterns in the molten metal pool contribute to the temperature distribution, which affects non-uniform crust formation. The CONV-2D results on crust thickness are a little higher than the experimental data because of heat loss during the test. In comparison of the CONV-3D results with the CONV-2D results on crust thickness, the three-dimensional results are higher than the two-dimensional results, because of three dimensional natural convection flow and wall effect
Simulation of thermal-hydraulic process in reactor of HTR-PM based on flow and heat transfer network
International Nuclear Information System (INIS)
Zhou Kefeng; Zhou Yangping; Sui Zhe; Ma Yuanle
2012-01-01
The development of HTR-PM full scale simulator (FSS) is an important part in the project. The simulation of thermal-hydraulic process in reactor is one of the key technologies in the development of FSS. The simulation of thermal-hydraulic process in reactor was studied. According to the geometry structures and the characteristics of thermal-hydraulic process in reactor, the model was setup in components construction way. Based on the established simulation method of flow and heat transfer network, a Fortran code was developed and the simulation of thermal-hydraulic process was achieved. The simulation results of 50% FP steady state, 100% FP steady state and control rod mistakenly ascension accidents were given. The verification of simulation results was carried out by comparing with the design and analysis code THERMIX. The results show that the method and model based on flow and heat transfer network can meet the requirements of FSS and reflect the features of thermal-hydraulic process in HTR-PM. (authors)
Assessment of the IVA3 code for multifield flow simulation. Formal report
Energy Technology Data Exchange (ETDEWEB)
Stewart, H.B.
1995-07-01
This report presents an assessment of the IVA3 computer code for multifield flow simulation, as applied to the premixing phase of a hypothetical steam explosion in a water-cooled power reactor. The first section of this report reviews the derivation of the basic partial differential equations of multifield modeling, with reference to standard practices in the multiphase flow literature. Basic underlying assumptions and approximations are highlighted, and comparison is made between IVA3 and other codes in current use. Although Kolev`s derivation of these equations is outside the mainstream of the multiphase literature, the basic partial differential equations are in fact nearly equivalent to those in other codes. In the second section, the assumptions and approximations required to pass from generic differential equations to a specific working form are detailed. Some modest improvements to the IVA3 model are suggested. In Section 3, the finite difference approximations to the differential equations are described. The discretization strategy is discussed with reference to numerical stability, accuracy, and the role of various physical phenomena - material convection, sonic propagation, viscous stress, and interfacial exchanges - in the choice of discrete approximations. There is also cause for concern about the approximations of time evolution in some heat transfer terms, which might be adversely affecting numerical accuracy. The fourth section documents the numerical solution method used in IVA3. An explanation for erratic behavior sometimes observed in the first outer iteration is suggested, along with possible remedies. Finally, six recommendations for future assessment and improvement of the IVA3 model and code are made.
New terrestrial heat flow measurements on the Nazca Plate
Energy Technology Data Exchange (ETDEWEB)
Anderson, R N [Columbia Univ., Palisades, NY; Langseth, M G; Vacquier, V; Francheteau, J
1976-03-01
Sixty-seven new heat flow measurements on the Nazca Plate are reported, and the thermal regimes of three specific areas on the plate are examined. The Nazca Ridge is an aseismic ridge which may have been generated as an ''island trail'' from the Easter Island ''hot spot'' and/or may be a fossil transform fault. The Nazca Ridge has lower heat flow than the surrounding sea floor implying that the ridge might have low ''effective'' thermal conductivity causing heat to preferentially flow or refract to surrounding ocean crust which has higher conductivity, or, the low heat flow values may be caused by hydrothermal circulation on the ridge. The Carnegie Plateau is an elevated region south of the Carnegie Ridge on the northeastern Nazca Plate with high heat flow and shallow topography consistent with an age of less than 20 m.y. B.P. The central Nazca Plate is an area of highly variable heat flow which is possibly related to thin sediment and to rough regional topography.
Coded Ultrasound for Blood Flow Estimation Using Subband Processing
DEFF Research Database (Denmark)
Gran, Fredrik; Udesen, Jesper; Nielsen, Michael Bachamnn
2008-01-01
the excitation signal is broadband and has good spatial resolution after pulse compression. This means that time can be saved by using the same data for B-mode imaging and blood flow estimation. Two different coding schemes are used in this paper, Barker codes and Golay codes. The performance of the codes......This paper investigates the use of coded excitation for blood flow estimation in medical ultrasound. Traditional autocorrelation estimators use narrow-band excitation signals to provide sufficient signal-to-noise-ratio (SNR) and velocity estimation performance. In this paper, broadband coded...... signals are used to increase SNR, followed by subband processing. The received broadband signal is filtered using a set of narrow-band filters. Estimating the velocity in each of the bands and averaging the results yields better performance compared with what would be possible when transmitting a narrow...
LavaSIM: the effect of heat transfer in 3D on lava flow characteristics (Invited)
Fujita, E.
2013-12-01
Characteristics of lava flow are governed by many parameters like lava viscosity, effusion rate, ground topography, etc. The accuracy and applicability of lava flow simulation code is evaluated whether the numerical simulation can reproduce these features quantitatively, which is important from both strategic and scientific points of views. Many lava flow simulation codes are so far proposed, and they are classified into two categories, i.e., the deterministic and the probabilistic models. LavaSIM is one of the former category models, and has a disadvantage of time consuming. But LavaSIM can solves the equations of continuity, motion, energy by step and has an advantage in the calculation of three-dimensional analysis with solid-liquid two phase flow, including the heat transfer between lava, solidified crust, air, water and ground, and three-dimensional convection in liquid lava. In other word, we can check the detailed structure of lava flow by LavaSIM. Therefore, this code can produce both channeled and fan-dispersive flows. The margin of the flow is solidified by cooling and these solidified crusts control the behavior of successive lava flow. In case of a channel flow, the solidified margin supports the stable central main flow and elongates the lava flow distance. The cross section of lava flow shows that the liquid lava flows between solidified crusts. As for the lava extrusion flow rate, LavaSIM can include the time function as well as the location of the vents. In some cases, some parts of the solidified wall may be broken by the pressure of successive flow and/or re-melting. These mechanisms could characterize complex features of the observed lava flows at many volcanoes in the world. To apply LavaSIM to the benchmark tests organized by V-hub is important to improve the lava flow evaluation technique.
2010-01-01
... 9 Animals and Animal Products 2 2010-01-01 2010-01-01 false Canning with heat processing and hermetically sealed containers; closures; code marking; heat processing; incubation. 355.25 Section 355.25... IDENTIFICATION AS TO CLASS, QUALITY, QUANTITY, AND CONDITION Inspection Procedure § 355.25 Canning with heat...
Applications of thermoelectric modules on heat flow detection.
Leephakpreeda, Thananchai
2012-03-01
This paper presents quantitative analysis and practical scenarios of implementation of the thermoelectric module for heat flow detection. Mathematical models of the thermoelectric effects are derived to describe the heat flow from/to the detected media. It is observed that the amount of the heat flow through the thermoelectric module proportionally induces the conduction heat owing to the temperature difference between the hot side and the cold side of the thermoelectric module. In turn, the Seebeck effect takes place in the thermoelectric module where the temperature difference is converted to the electric voltage. Hence, the heat flow from/to the detected media can be observed from both the amount and the polarity of the voltage across the thermoelectric module. Two experiments are demonstrated for viability of the proposed technique by the measurements of the heat flux through the building wall and thermal radiation from the outdoor environment during daytime. Copyright © 2011 ISA. Published by Elsevier Ltd. All rights reserved.
Ameri, Ali A.
2012-01-01
The purpose of this report is to summarize and document the work done to enable a NASA CFD code to model laminar-turbulent transition process on an isolated turbine blade. The ultimate purpose of the present work is to down-select a transition model that would allow the flow simulation of a variable speed power turbine to be accurately performed. The flow modeling in its final form will account for the blade row interactions and their effects on transition which would lead to accurate accounting for losses. The present work only concerns itself with steady flows of variable inlet turbulence. The low Reynolds number k- model of Wilcox and a modified version of the same model will be used for modeling of transition on experimentally measured blade pressure and heat transfer. It will be shown that the k- model and its modified variant fail to simulate the transition with any degree of accuracy. A case is thus made for the adoption of more accurate transition models. Three-equation models based on the work of Mayle on Laminar Kinetic Energy were explored. The three-equation model of Walters and Leylek was thought to be in a relatively mature state of development and was implemented in the Glenn-HT code. Two-dimensional heat transfer predictions of flat plate flow and two-dimensional and three-dimensional heat transfer predictions on a turbine blade were performed and reported herein. Surface heat transfer rate serves as sensitive indicator of transition. With the newly implemented model, it was shown that the simulation of transition process is much improved over the baseline k- model for the single Reynolds number and pressure ratio attempted; while agreement with heat transfer data became more satisfactory. Armed with the new transition model, total-pressure losses of computed three-dimensional flow of E3 tip section cascade were compared to the experimental data for a range of incidence angles. The results obtained, form a partial loss bucket for the chosen blade
International Nuclear Information System (INIS)
Yu, S.-O.; Kim, M.; Kim, H.-J.
2002-01-01
A CANDU reactor has the unique features and the intrinsic safety related characteristics that distinguish it from other water-cooled thermal reactors. If there is the loss of coolant accident (LOCA) and a coincident failure of the emergency coolant injection (ECI) system, the heavy water moderator is continuously cooled, providing a heat sink for decay heat produced in the fuel. Therefore, it is one of major concerns to estimate the local subcooling of moderator inside the calandria vessel under postulated accident in CANDU safety analyses. The Canadian Nuclear Safety Commission (CNSC), a regulatory body in Canada, categorized the integrity of moderator as a generic safety issue and recommended that a series of experimental works be performed to verify the safety evaluation codes for individual simulated condition of nuclear power plant, comparing with the results of three-dimensional experimental data. In this study, three-dimensional analyses of fluid flow and heat transfer have been performed to assess thermal-hydraulic characteristics for moderator simulation conducted by SPEL (Sheridan Park Experimental Laboratory) experimental facility. The parametric study has also carried out to investigate the effect of major parameters such as flowrate, temperature, and heat load generated from the heaters on the temperature and flow distribution inside the moderator. Three flow patterns have been identified in the moderator with flowrate, heat generation, or both. As the transition of fluid flow is progressed, it is found that the dimensionless numbers (Ar) and the ratio of buoyancy to inertia forces are constant. (author)
International Nuclear Information System (INIS)
Young Jin Lee; Bub Dong Chung; Jong Chull Jo; Hho Jung Kim; Un Chul Lee
2004-01-01
SMART is a medium sized integral type advanced pressurized water reactor currently under development at KAERI. The steam generators of SMART are designed with helically coiled tubes and these are designed to produce superheated steam. The helical shape of the tubes can induce strong centrifugal effect on the secondary coolant as it flows inside the tubes. The presence of centrifugal effect is expected to enhance the formation of cross-sectional circulation flows within the tubes that will increase the overall heat transfer. Furthermore, the centrifugal effect is expected to enhance the moisture separation and thus make it easier to produce superheated steam. MARS is a best-estimate thermal-hydraulic systems analysis code with multi-phase, multi-dimensional analysis capability. The MARS code was produced by restructuring and merging the RELAP5 and the COBRA-TF codes. However, MARS as well as most other best-estimate systems analysis codes in current use lack the detailed models needed to describe the thermal hydraulics of helically coiled tubes. In this study, the heat transfer characteristics and relevant correlations for both the tube and shell sides of helical tubes have been investigated, and the appropriate models have been incorporated into the MARS code. The newly incorporated helical tube heat transfer package is available to the MARS users via selection of the appropriate option in the input. A performance analysis on the steam generator of SMART under full power operation was carried out using the modified MARS code. The results of the analysis indicate that there is a significant improvement in the code predictability. (authors)
CFD investigation of flow and heat transfer of nanofluids in isoflux spirally fluted tubes
Salama, Amgad
2012-01-01
In this work, the problem of flow and heat transfer of nanofluids in spirally fluted tubes is investigated numerically using the CFD code Fluent. The tube investigated in this work is characterized by the existence of helical ridging which is usually obtained by embossing a smooth tube. A tube of diameter of 15 mm, 1.5 mm groove depth and a single helix with pitch of 64 mm is chosen for simulation. This geometry has been chosen for simulation because it has been investigated experimentally for pure fluids and would, therefore, provide a verification framework with our CFD model. The result of our CFD investigation compares very well with the experimental work conducted on this tube geometry. Interesting patterns are highlighted and investigated including the existence of flow swirl as a result of the existence of the spirally enhanced ridges. This swirl flow enhances heat transfer characteristics of this system as reported in the literatures. This study also shows that further enhancement is achieved if small amount of nanoparticles are introduced to the fluid. These nanoparticles (metallic-based nanoparticles) when introduced to the fluid enhances its heat transfer characteristics.
International Nuclear Information System (INIS)
Mohseni, Mahdi; Bazargan, Majid
2014-01-01
Highlights: • The entropy generation in supercritical fluid flows has been numerically investigated. • The mechanisms of entropy generation are different near and away from the walls. • In the near wall region, the energy dissipation is the deciding parameter. • Away from the wall, the heat transfer is the effective factor in entropy generation. • The bulk Be number is greater in the liquid-like region than in vapor-like region. - Abstract: In this study, a two dimensional CFD code has been developed to investigate entropy generation in turbulent mixed convection heat transfer flow of supercritical fluids. Since the fluid properties vary significantly under supercritical conditions, the changes of entropy generation are large. The contribution of each of the mechanisms of entropy production (heat transfer and energy dissipation) is compared in different regions of the flow. The results show that the mechanisms of entropy generation act differently in the near wall region within the viscous sub-layer and in the region away from the wall. The effects of the wall heat flux on the entropy generation are also investigated
Verification of combined thermal-hydraulic and heat conduction analysis code FLOWNET/TRUMP
International Nuclear Information System (INIS)
Maruyama, Soh; Fujimoto, Nozomu; Sudo, Yukio; Kiso, Yoshihiro; Murakami, Tomoyuki.
1988-09-01
This report presents the verification results of the combined thermal-hydraulic and heat conduction analysis code, FLOWNET/TRUMP which has been utilized for the core thermal hydraulic design, especially for the analysis of flow distribution among fuel block coolant channels, the determination of thermal boundary conditions for fuel block stress analysis and the estimation of fuel temperature in the case of fuel block coolant channel blockage accident in the design of the High Temperature Engineering Test Reactor(HTTR), which the Japan Atomic Energy Research Institute has been planning to construct in order to establish basic technologies for future advanced very high temperature gas-cooled reactors and to be served as an irradiation test reactor for promotion of innovative high temperature new frontier technologies. The verification of the code was done through the comparison between the analytical results and experimental results of the Helium Engineering Demonstration Loop Multi-channel Test Section(HENDEL T 1-M ) with simulated fuel rods and fuel blocks. (author)
Verification of combined thermal-hydraulic and heat conduction analysis code FLOWNET/TRUMP
Maruyama, Soh; Fujimoto, Nozomu; Kiso, Yoshihiro; Murakami, Tomoyuki; Sudo, Yukio
1988-09-01
This report presents the verification results of the combined thermal-hydraulic and heat conduction analysis code, FLOWNET/TRUMP which has been utilized for the core thermal hydraulic design, especially for the analysis of flow distribution among fuel block coolant channels, the determination of thermal boundary conditions for fuel block stress analysis and the estimation of fuel temperature in the case of fuel block coolant channel blockage accident in the design of the High Temperature Engineering Test Reactor(HTTR), which the Japan Atomic Energy Research Institute has been planning to construct in order to establish basic technologies for future advanced very high temperature gas-cooled reactors and to be served as an irradiation test reactor for promotion of innovative high temperature new frontier technologies. The verification of the code was done through the comparison between the analytical results and experimental results of the Helium Engineering Demonstration Loop Multi-channel Test Section(HENDEL T(sub 1-M)) with simulated fuel rods and fuel blocks.
Studies of thermal-hydrodynamic flow instability, 2
International Nuclear Information System (INIS)
Suzuoki, Akira
1977-01-01
For reliable prediction of flow stability in sodium-heated steam generators, a dynamic model was proposed for boiling flow oscillation in parallel channel systems, and an analysis code was developed. The model contains a description of a sodium flow exchanging heat with a water flow in counter-current fashion. The code was applied to three representative flow systems whose heating conditions differed from each other, whereby their flow stabilities were compared with a focus on the effects of heating condition. Eigenvalues and flow impedances of the oscillation determined for each system reveal that: (1) Two fundamental systems for the steam generator, parallel tube system in an evaporator and steam generator modules arranged in parallel, have different stabilities under low frequency oscillation. (2) Existing analysis model conditioned on constant heat flux gives different results on stability from those of either steam generator model under low frequency oscillation. (auth.)
Heat transfer and flow structure evaluation of a synthetic jet emanating from a planar heat sink
International Nuclear Information System (INIS)
Manning, Paul; Persoons, Tim; Murray, Darina
2014-01-01
Direct impinging synthetic jets are a proven method for heat transfer enhancement, and have been subject to extensive research. However, despite the vast amount of research into direct synthetic jet impingement, there has been little research investigating the effects of a synthetic jet emanating from a heated surface, this forms the basis of the current research investigation. Both single and multiple orifices are integrated into a planar heat sink forming a synthetic jet, thus allowing the heat transfer enhancement and flow structures to be assessed. The heat transfer analysis highlighted that the multiple orifice synthetic jet resulted in the greatest heat transfer enhancements. The flow structures responsible for these enhancements were identified using a combination of flow visualisation, thermal imaging and thermal boundary layer analysis. The flow structure analysis identified that the synthetic jets decreased the thermal boundary layer thickness resulting in a more effective convective heat transfer process. Flow visualisation revealed entrainment of local air adjacent to the heated surface; this occurred from vortex roll-up at the surface of the heat sink and from the highly sheared jet flow. Furthermore, a secondary entrainment was identified which created a surface impingement effect. It is proposed that all three flow features enhance the heat transfer characteristics of the system.
Development of a detailed core flow analysis code for prismatic fuel reactors
International Nuclear Information System (INIS)
Bennett, R.G.
1990-01-01
The detailed analysis of the core flow distribution in prismatic fuel reactors is of interest for modular high-temperature gas-cooled reactor (MHTGR) design and safety analyses. Such analyses involve the steady-state flow of helium through highly cross-connected flow paths in and around the prismatic fuel elements. Several computer codes have been developed for this purpose. However, since they are proprietary codes, they are not generally available for independent MHTGR design confirmation. The previously developed codes do not consider the exchange or diversion of flow between individual bypass gaps with much detail. Such a capability could be important in the analysis of potential fuel block motion, such as occurred in the Fort St. Vrain reactor, or for the analysis of the conditions around a flow blockage or misloaded fuel block. This work develops a computer code with fairly general-purpose capabilities for modeling the flow in regions of prismatic fuel cores. The code, called BYPASS solves a finite difference control volume formulation of the compressible, steady-state fluid flow in highly cross-connected flow paths typical of the MHTGR
Preparation and Interpretation of Heat Flow Map of Turkey
International Nuclear Information System (INIS)
Ozturk, S.; Karli, R.; Destur, M.
2007-01-01
There exist a lot of data indicating our country takes place on an impotrant Kown heat flow anomaly. The preparation of a detailed 'Heat Flow Map' as a result of rational studies and depending upon this the determination of the distribution of heat in litosphere, except from the scientific benefits; shall enlighten subjects such as oil basen analysis, prospection of hydrothermal ores and earthquakes and further shall increase the feasibility of planning geothermal energy research.In between years 1995- 2005; as a part of project of the Geophysical Department of MTA with the purpose of preperation of Heat Flow Maps of Turkey, the heat flow measurments had been carried on at the convenient cold water wells. Using the Thermic and Gamma-Ray measurments and calculated conductivity coefficients of the representative rock samples of formation, heat flow map had been prepared. A distance of 10-30 km had been kept carefully betwen the wells of interest a total of 80204 m Thermic and Gamma-Ray logs and 420 rock samples from 695 wells, had been used in the study. Then according to the Lambert Projection, using the Surfer 8.02 and Grapher4 programmes The Heat Flow Maps of Turkey of scale 1:1000000 had been obtained.Some regional researches indicate that Turkey takes place in a part of Europe of high heat flux. Unfortunately there exist no detailed heat flow map of our country up to now. This shows the importance of present project
Flow dynamics of volume-heated boiling pools
International Nuclear Information System (INIS)
Ginsberg, T.; Jones, O.C.; Chen, J.C.
1979-01-01
Safety analyses of fast breeder reactors require understanding of the two-phase fluid dynamic and heat transfer characteristics of volume-heated boiling pool systems. Design of direct contact three-phase boilers, of practical interest in the chemical industries also requires understanding of the fundamental two-phase flow and heat transfer behavior of volume boiling systems. Several experiments have been recently reported relevant to the boundary heat-loss mechanisms of boiling pool systems. Considerably less is known about the two-phase fluid dynamic behavior of such systems. This paper describes an experimental investigation of the steady-state flow dynamics of volume-heated boiling pool systems
Assessment of subchannel code ASSERT-PV for flow-distribution predictions
International Nuclear Information System (INIS)
Nava-Dominguez, A.; Rao, Y.F.; Waddington, G.M.
2014-01-01
Highlights: • Assessment of the subchannel code ASSERT-PV 3.2 for the prediction of flow distribution. • Open literature and in-house experimental data to quantify ASSERT-PV predictions. • Model changes assessed against vertical and horizontal flow experiments. • Improvement of flow-distribution predictions under CANDU-relevant conditions. - Abstract: This paper reports an assessment of the recently released subchannel code ASSERT-PV 3.2 for the prediction of flow-distribution in fuel bundles, including subchannel void fraction, quality and mass fluxes. Experimental data from open literature and from in-house tests are used to assess the flow-distribution models in ASSERT-PV 3.2. The prediction statistics using the recommended model set of ASSERT-PV 3.2 are compared to those from previous code versions. Separate-effects sensitivity studies are performed to quantify the contribution of each flow-distribution model change or enhancement to the improvement in flow-distribution prediction. The assessment demonstrates significant improvement in the prediction of flow-distribution in horizontal fuel channels containing CANDU bundles
Assessment of subchannel code ASSERT-PV for flow-distribution predictions
Energy Technology Data Exchange (ETDEWEB)
Nava-Dominguez, A., E-mail: navadoma@aecl.ca; Rao, Y.F., E-mail: raoy@aecl.ca; Waddington, G.M., E-mail: waddingg@aecl.ca
2014-08-15
Highlights: • Assessment of the subchannel code ASSERT-PV 3.2 for the prediction of flow distribution. • Open literature and in-house experimental data to quantify ASSERT-PV predictions. • Model changes assessed against vertical and horizontal flow experiments. • Improvement of flow-distribution predictions under CANDU-relevant conditions. - Abstract: This paper reports an assessment of the recently released subchannel code ASSERT-PV 3.2 for the prediction of flow-distribution in fuel bundles, including subchannel void fraction, quality and mass fluxes. Experimental data from open literature and from in-house tests are used to assess the flow-distribution models in ASSERT-PV 3.2. The prediction statistics using the recommended model set of ASSERT-PV 3.2 are compared to those from previous code versions. Separate-effects sensitivity studies are performed to quantify the contribution of each flow-distribution model change or enhancement to the improvement in flow-distribution prediction. The assessment demonstrates significant improvement in the prediction of flow-distribution in horizontal fuel channels containing CANDU bundles.
On the calculation of dynamic and heat loads on a three-dimensional body in a hypersonic flow
Bocharov, A. N.; Bityurin, V. A.; Evstigneev, N. M.; Fortov, V. E.; Golovin, N. N.; Petrovskiy, V. P.; Ryabkov, O. I.; Teplyakov, I. O.; Shustov, A. A.; Solomonov, Yu S.
2018-01-01
We consider a three-dimensional body in a hypersonic flow at zero angle of attack. Our aim is to estimate heat and aerodynamic loads on specific body elements. We are considering a previously developed code to solve coupled heat- and mass-transfer problem. The change of the surface shape is taken into account by formation of the iterative process for the wall material ablation. The solution is conducted on the multi-graphics-processing-unit (multi-GPU) cluster. Five Mach number points are considered, namely for M = 20-28. For each point we estimate body shape after surface ablation, heat loads on the surface and aerodynamic loads on the whole body and its elements. The latter is done using Gauss-type quadrature on the surface of the body. The comparison of the results for different Mach numbers is performed. We also estimate the efficiency of the Navier-Stokes code on multi-GPU and central processing unit architecture for the coupled heat and mass transfer problem.
International Nuclear Information System (INIS)
Andreani, M.; Yadigaroglu, G.; Paul Scherrer Inst.
1992-08-01
Dispersed Flow Film Boiling is the heat transfer regime that occurs at high void fractions in a heated channel. The way this heat transfer mode is modelled in the NRC computer codes (RELAP5 and TRAC) and the validity of the assumptions and empirical correlations used is discussed. An extensive review of the theoretical and experimental work related with heat transfer to highly dispersed mixtures reveals the basic deficiencies of these models: the investigation refers mostly to the typical conditions of low rate bottom reflooding, since the simulation of this physical situation by the computer codes has often showed poor results. The alternative models that are available in the literature are reviewed, and their merits and limits are highlighted. The modifications that could improve the physics of the models implemented in the codes are identified
Heat Flow, Regional Geophysics and Lithosphere Structure In The Czech Republic
Safanda, J.; Cermak, V.; Kresl, M.; Dedecek, P.
Paper summarises and critically revises heat flow data that have been collected in the Czech Republic to date. The regional heat flow density map was prepared in view of all existing heat flow data completed with the similar in the surrounding countries and taking into consideration also temperature measurements in deep boreholes. Crustal temperature profiles were calculated by using the available geological information, results of deep seismic sounding and the laboratory data on radiogenic heat produc- tion and thermal conductivity. Special attention was paid to numerous temperature logs in two sedimentary basins, namely in the Cheb and Ostrava-Karvina coal basins, for which detailed heat flow patterns were proposed. Relationships between heat flow distribution and the crustal/lithosphere evolution, between heat flow and the heat pro- duction of the crustal rocks, heat flow and crustal thickness and the steady-state vs. transient heat transport are discussed.
Analysis on flow characteristic of nuclear heating reactor
International Nuclear Information System (INIS)
Jiang Shengyao; Wu Xinxin
1997-06-01
The experiment was carried out on the test loop HRTL-5, which simulates the geometry and system design of a 5 MW Nuclear heating reactor. The analysis was based on a one-dimensional two-phase flow drift model with conservation equations for mass, steam mass, energy and momentum. Clausius-Clapeyron equation was used for the calculation of flashing front in the riser. A set of ordinary equation, which describes the behavior of two-phase flow in the natural circulation system, was derived through integration of the above conservation equations in subcooled boiling region, bulk boiling region in the heated section and in the riser. The method of time-domain was used for the calculation. Both static and dynamic results are presented. System pressure, inlet subcooling and heat flux are varied as input parameters. The results show that, firstly, subcooled boiling in the heated section and void flashing in the riser have significant influence on the distribution of the void fraction, mass flow rate and stability of the system, especially at lower pressure, secondly, in a wide range of two-phase flow conditions, only subcooled boiling occurs in the heated section. For the designed two-phase regime operation of the 5 MW nuclear heating reactor, the temperature at the core exit has not reaches its saturation value. Thirdly, the mechanism of two-phase flow oscillation, namely, 'zero-pressure-drop', is described. In the wide range of inlet subcooling (0 K<ΔT<28 K) there exists three regions for system flow condition, namely, (1) stable two-phase flow, (2) bulk and subcooled boiling unstable flow, (3) subcooled boiling and single phase stable flow. The response of mass flow rate, after a small disturbance in the heat flux, is showed in the above inlet subcooling range, and based on it the instability map of the system is given through experiment and calculation. (3 refs., 9 figs.)
International Nuclear Information System (INIS)
Kurganov, V.A.; Gladuntsov, A.I.
1977-01-01
Analysed are the experimental data obtained for heat transfer to gaseous dissociating ammonium (NH 3 ) under heating in round pipes (steel Kh18N10T) at developed eddying input flow and marginal condition of heat supply gsub(c) approximately equal to const in the ranges of the following parameters: p=3-10 atm; Tsub(input)=310-720 K; Tsub(c) ( 3 ; gsub(c)/-anti rho W 8.8 kJ/kg; gsub(c)/(anti rho WCsub(p) sub(input)Tsub(input)) (<=) 0.0104; 1/d (<=) 150 (where Tsub(c) is the wall temperature, gsub(c) the heat flow density on wall, and anti rho W velocity). The discussion involves phenomena of worsened heat transfer at high heat loads. The authors show the basic relationship between these phenomena and laminarization of the near-wall flow at the input site of the pipe. The regularities of heat transfer were noted to undergo substantial transformation under laminarized flow
Analysis of the transient compressible vapor flow in heat pipes
Jang, J. H.; Faghri, A.; Chang, W. S.
1989-01-01
The transient compressible one-dimensional vapor flow dynamics in a heat pipe is modeled. The numerical results are obtained by using the implicit non-iterative Beam-Warming finite difference method. The model is tested for simulated heat pipe vapor flow and actual vapor flow in cylindrical heat pipes. A good comparison of the present transient results for the simulated heat pipe vapor flow with the previous results of a two-dimensional numerical model is achieved and the steady state results are in agreement with the existing experimental data. The transient behavior of the vapor flow under subsonic, sonic, and supersonic speeds and high mass flow rates are successfully predicted. The one-dimensional model also describes the vapor flow dynamics in cylindrical heat pipes at high temperatures.
Analysis of the transient compressible vapor flow in heat pipe
International Nuclear Information System (INIS)
Jang, J.H.; Faghri, A.; Chang, W.S.
1989-07-01
The transient compressible one-dimensional vapor flow dynamics in a heat pipe is modeled. The numerical results are obtained by using the implicit non-iterative Beam-Warming finite difference method. The model is tested for simulated heat pipe vapor flow and actual vapor flow in cylindrical heat pipes. A good comparison of the present transient results for the simulated heat pipe vapor flow with the previous results of a two-dimensional numerical model is achieved and the steady state results are in agreement with the existing experimental data. The transient behavior of the vapor flow under subsonic, sonic, and supersonic speeds and high mass flow rates are successfully predicted. The one-dimensional model also describes the vapor flow dynamics in cylindrical heat pipes at high temperatures
Analysis of the transient compressible vapor flow in heat pipe
Jang, Jong Hoon; Faghri, Amir; Chang, Won Soon
1989-01-01
The transient compressible one-dimensional vapor flow dynamics in a heat pipe is modeled. The numerical results are obtained by using the implicit non-iterative Beam-Warming finite difference method. The model is tested for simulated heat pipe vapor flow and actual flow in cylindrical heat pipes. A good comparison of the present transient results for the simulated heat pipe vapor flow with the previous results of a two-dimensional numerical model is achieved and the steady state results are in agreement with the existing experimental data. The transient behavior of the vapor flow under subsonic, sonic, and supersonic speeds and high mass flow rates are successfully predicted. The one-dimensional model also describes the vapor flow dynamics in cylindrical heat pipes at high temperatures.
International Nuclear Information System (INIS)
Cull, J.P.
1981-01-01
Secular and long-term periodic changes in surface temperature cause perturbations to the geothermal gradient which may be significant to depths of at least 1000 m, and major corrections are required to determine absolute values of heat flow from the Earth's interior. However, detailed climatic models remain contentious and estimates of error in geothermal gradients differ widely. Consequently, regions of anomalous heat flow which could contain geothermal resources may be more easily resolved by measuring relative values at a standard depth (e.g. 100 m) so that all data are subject to similar corrections. (orig./ME)
Decay Heat Calculations for Reactors: Development of a Computer Code ADWITA
International Nuclear Information System (INIS)
Raj, Devesh
2015-01-01
Estimation of release of energy (decay heat) over an extended period of time after termination of neutron induced fission is necessary for determining the heat removal requirements when the reactor is shutdown, and for fuel storage and transport facilities as well as for accident studies. A Fuel Cycle Analysis Code, ADWITA (Activation, Decay, Waste Incineration and Transmutation Analysis) which can generate inventory based on irradiation history and calculate radioactivity and decay heat for extended period of cooling, has been written. The method and data involved in Fuel Cycle Analysis Code ADWITA and some results obtained shall also be presented. (author)
CFD study of the heat transfer between a dilute gas particle suspension flow and an obstruction
International Nuclear Information System (INIS)
Nguyen, A.V.; Fletcher, C.A.J.
1999-01-01
The effect on heat transfer of solid particles suspended in a gas flow is of considerable importance in a number of industrial applications, ranging from coal combustion equipment and heat exchangers to catalytic reaction or cooling of nuclear reactors using gas graphite dust suspensions. Here, the heat transfer process between a dilute gas-particle suspension flow and an obstruction has been numerically investigated employing a novel Eulerian formulation for dilute gas particle suspension flows, which allows interaction of the key mechanisms to be quantified for the first time. As the particle reflection occurs around the obstruction, the heat transfer process has been modeled taking into account the incident and reflected particles explicitly. In the energy equations these particle families are treated separately. Only the effect on the gas convective heat transfer is expected to be of primary significance and investigated. The numerical computation is performed using the commercial computational fluid dynamics code, FLUENT, with the User Defined Subroutines. The authors study the heat transfer process between a dilute gas particle flow and an obstruction with simple geometries such as a 45 degree ramp and a cylindrical tube. The theoretical results for the latter case are compared with the available experimental data. The numerical simulation shows that both the particle size and the particle concentration (in the thermal boundary layer) affect the heat transfer process. Since both the particle incidence and reflection depend on the particle size and strongly influence the particle concentration distribution, they have to be physically correctly treated in the modeling of the heat transfer, as is demonstrated in the novel formulation. There is an optimum particle size for a maximum enhancement of the heat transfer. The particle concentration increases the efficiency of the heat transfer process expressed in terms of the local Nusselt numbers
Heat flow in a He II filled fin
International Nuclear Information System (INIS)
Warren, R.P.
1984-01-01
This chapter demonstrates the influence of diameter, length, Kapitza conductance and temperature on the heat carrying capacity of an externally cooled, circular He II filled channel with zero net mass flow and of negligible wall thermal resistance. Topics considered include the internal convection mechanism and the heat transfer model (boundary conditions, solution procedure). The large apparent thermal conductivity of He-II is explained by the two fluid model as an internal convection in which there is a counter flow of the normal and superfluids with no net mass flow. A separate bath is considered in which an He-IIp (pressurized superfluid helium) filled fin is immersed which extends from the heated reservoir. A single heat sink can serve multiple heat sources
Heat transfer to accelerating gas flows
International Nuclear Information System (INIS)
Kennedy, T.D.A.
1978-01-01
The development of fuels for gas-cooled reactors has resulted in a number of 'gas loop' experiments in materials-testing research reactors. In these experiments, efforts are made to reproduce the conditions expected in gas-cooled power reactors. Constant surface temperatures are sought over a short (300 mm) fuelled length, and because of entrance effects, an accelerating flow is required to increase the heat transfer down-stream from the entrance. Strong acceleration of a gas stream will laminarise the flow even at Reynolds Numbers up to 50000, far above values normally associated with laminar flow. A method of predicting heat transfer in this situation is presented here. An integral method is used to find the velocity profile; this profile is then used in an explicit finite-difference solution of the energy equation to give a temperature profile and resultant heat-transfer coefficient values. The Kline criterion, which compares viscous and disruptive forces, is used to predict whether the flow will be laminar. Experimental results are compared with predictions, and good agreement is found to exist. (author)
SCEPTIC, Pressure Drop, Flow Rate, Heat Transfer, Temperature in Reactor Heat Exchanger
International Nuclear Information System (INIS)
Kattchee, N.; Reynolds, W.C.
1975-01-01
1 - Nature of physical problem solved: SCEPTIC is a program for calculating pressure drop, flow rates, heat transfer rates, and temperature in heat exchangers such as fuel elements of typical gas or liquid cooled nuclear reactors. The effects of turbulent and heat interchange between flow passages are considered. 2 - Method of solution: The computation procedure amounts to a nodal of lumped parameter type of calculation. The axial mesh size is automatically selected to assure that a prescribed accuracy of results is obtained. 3 - Restrictions on the complexity of the problem: Maximum number of subchannels is 25, maximum number of heated surfaces is 46
DNS of turbulent channel flow with conjugate heat transfer at Prandtl number 0.01
Energy Technology Data Exchange (ETDEWEB)
Tiselj, Iztok, E-mail: iztok.tiselj@ijs.si [' Jozef Stefan' Institute, Jamova 39, SI-1000 Ljubljana (Slovenia); Cizelj, Leon, E-mail: leon.cizelj@ijs.si [' Jozef Stefan' Institute, Jamova 39, SI-1000 Ljubljana (Slovenia)
2012-12-15
Highlights: Black-Right-Pointing-Pointer DNS database for turbulent channel flow at Prandtl number 0.01 and various Re{sub {tau}}. Black-Right-Pointing-Pointer Two ideal boundary condition analyzed: non-fluctuating and fluctuating temperature. Black-Right-Pointing-Pointer DNS database with conjugate heat transfer for liquid sodium-steel contact. Black-Right-Pointing-Pointer Penetration of the turbulent temperature fluctuations into the solid wall analyzed. - Abstract: Direct Numerical Simulation (DNS) of the fully developed velocity and temperature fields in a turbulent channel flow coupled with the unsteady conduction in the heated walls was carried out. Simulations were performed with passive scalar approximation at Prandtl number 0.01, which roughly corresponds to the Prandtl number of liquid sodium. DNSs were performed at friction Reynolds numbers 180, 395 and 590. The obtained statistical quantities like mean temperatures, profiles of the root-mean-square (RMS) temperature fluctuations for various thermal properties of wall and fluid, and various wall thicknesses were obtained from a pseudo-spectral channel-flow code. Even for the highest implemented Reynolds number the temperature profile in the fluid does not exhibit log-law region and the near-wall RMS temperature fluctuations show Reynolds number dependence. Conjugate heat transfer simulations of liquid sodium-steel system point to a relatively intensive penetration of turbulent temperature fluctuations into the heated wall. Database containing the results is available in a digital form.
DNS of turbulent channel flow with conjugate heat transfer at Prandtl number 0.01
International Nuclear Information System (INIS)
Tiselj, Iztok; Cizelj, Leon
2012-01-01
Highlights: ► DNS database for turbulent channel flow at Prandtl number 0.01 and various Re τ . ► Two ideal boundary condition analyzed: non-fluctuating and fluctuating temperature. ► DNS database with conjugate heat transfer for liquid sodium–steel contact. ► Penetration of the turbulent temperature fluctuations into the solid wall analyzed. - Abstract: Direct Numerical Simulation (DNS) of the fully developed velocity and temperature fields in a turbulent channel flow coupled with the unsteady conduction in the heated walls was carried out. Simulations were performed with passive scalar approximation at Prandtl number 0.01, which roughly corresponds to the Prandtl number of liquid sodium. DNSs were performed at friction Reynolds numbers 180, 395 and 590. The obtained statistical quantities like mean temperatures, profiles of the root-mean-square (RMS) temperature fluctuations for various thermal properties of wall and fluid, and various wall thicknesses were obtained from a pseudo-spectral channel-flow code. Even for the highest implemented Reynolds number the temperature profile in the fluid does not exhibit log-law region and the near-wall RMS temperature fluctuations show Reynolds number dependence. Conjugate heat transfer simulations of liquid sodium–steel system point to a relatively intensive penetration of turbulent temperature fluctuations into the heated wall. Database containing the results is available in a digital form.
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
Analysis of unscrammed loss of flow and heat sink for PRISM with GEM
International Nuclear Information System (INIS)
Slovik, G.C.; Van Tuyle, G.J.; Kennett, R.J.
1991-01-01
The US Department of Energy is sponsoring an advanced liquid-metal reactor design by General Electric Company (GE) called PRISM. The intent is to design a reactor with passively safe responses to many operational and severe accidents. PRISM is under review at the US Nuclear Regulatory Commission for licensability with Brookhaven National Laboratory providing technical assistance. Recently, the PRISM design has been modified to include three gas expansion modules (GEMs) on its core periphery. The GEMs were added to quickly reduce the power (by inserting negative reactivity) during loss-of-flow events to curtail peak fuel and clad temperatures predicted in the previous design. The GEM prototypes have been tested at the Fast Flux Test Facility. The significance of the GEMs in PRISM is discussed in this paper through the evaluation of the unprotected loss of flow (ULOF) and loss of heat sink using the SSC code. It has been demonstrated in the past that SSC predicts results similar to GE and other liquid-metal reactor codes
Conjugate Heat Transfer Study in Hypersonic Flows
Sahoo, Niranjan; Kulkarni, Vinayak; Peetala, Ravi Kumar
2018-04-01
Coupled and decoupled conjugate heat transfer (CHT) studies are carried out to imitate experimental studies for heat transfer measurement in hypersonic flow regime. The finite volume based solvers are used for analyzing the heat interaction between fluid and solid domains. Temperature and surface heat flux signals are predicted by both coupled and decoupled CHT analysis techniques for hypersonic Mach numbers. These two methodologies are also used to study the effect of different wall materials on surface parameters. Effectiveness of these CHT solvers has been verified for the inverse problem of wall heat flux recovery using various techniques reported in the literature. Both coupled and decoupled CHT techniques are seen to be equally useful for prediction of local temperature and heat flux signals prior to the experiments in hypersonic flows.
Flow and heat transfer thermohydraulic modelisation during the reflooding phase of a P.W.R.'s core
International Nuclear Information System (INIS)
Raymond, Patrick
1978-04-01
Some generalities about L.O.C.A. are first recalled. The French experimental studies about Emergency Core Cooling System are briefly described. The different heat transfer mechanisms to take into account, according to the flow pattern in the dry zone, and the correlations or methods to calculate them, are defined. Then the Thermohydraulic code computer: FLIRA, which describe the reflooding phase, and a modelisation taking into account the different flow patterns are setted. A first interpretation of ERSEC experiments with a tubular test section shows that it is possible, with this modelisation and some classical heat transfer correlations, to describe the reflooding phase. [fr
Visualisation of heat transfer in laminar flows
Speetjens, M.F.M.; Steenhoven, van A.A.
2009-01-01
Heat transfer in fluid flows traditionally is examined in terms of temperature field and heat-transfer coefficients at non-adiabatic walls. However, heat transfer may alternatively be considered as the transport of thermal energy by the total convective-conductive heat flux in a way analogous to the
The effects of radiogenic heat on groundwater flow
International Nuclear Information System (INIS)
Beddoes, R.J.; Tammemagi, H.Y.
1986-03-01
The effects of radiogenic heat released by a nuclear waste repository on the groundwater flow in the neighbouring rock mass is reviewed. The report presents an overview of the hydrogeologic properties of crystalline rocks in the Canadian Shield and also describes the mathematical theory of groundwater flow and heat transfer in both porous media and fractured rock. Numerical methods for the solution of the governing equations are described. A number of case histories are described where analyses of flow systems have been performed both with and without radiogenic heat sources. A number of relevant topics are reviewed such as the role of the porous medium model, boundary conditions and, most importantly, the role of complex coupled processes where the effects of heat and water flow are intertwined with geochemical and mechanical processes. The implications to radioactive waste disposal are discussed
Calculation codes for radiant heat transfers; Les codes de calcul de rayonnement thermique
Energy Technology Data Exchange (ETDEWEB)
NONE
1997-12-31
This document reports on 12 papers about computerized simulation and modeling of radiant heat transfers and fluid flows in various industrial and domestic situations: space heating, metal industry (furnaces, boilers..), aerospace industry (turbojet engines, combustion chambers) etc.. This workshop was organized by the ``radiation`` section of the French society of thermal engineers. (J.S.)
Calculation codes for radiant heat transfers; Les codes de calcul de rayonnement thermique
Energy Technology Data Exchange (ETDEWEB)
NONE
1996-12-31
This document reports on 12 papers about computerized simulation and modeling of radiant heat transfers and fluid flows in various industrial and domestic situations: space heating, metal industry (furnaces, boilers..), aerospace industry (turbojet engines, combustion chambers) etc.. This workshop was organized by the ``radiation`` section of the French society of thermal engineers. (J.S.)
Energy Technology Data Exchange (ETDEWEB)
Yidong Xia; Mitch Plummer; Robert Podgorney; Ahmad Ghassemi
2016-02-01
Performance of heat production process over a 30-year period is assessed in a conceptual EGS model with a geothermal gradient of 65K per km depth in the reservoir. Water is circulated through a pair of parallel wells connected by a set of single large wing fractures. The results indicate that the desirable output electric power rate and lifespan could be obtained under suitable material properties and system parameters. A sensitivity analysis on some design constraints and operation parameters indicates that 1) the fracture horizontal spacing has profound effect on the long-term performance of heat production, 2) the downward deviation angle for the parallel doublet wells may help overcome the difficulty of vertical drilling to reach a favorable production temperature, and 3) the thermal energy production rate and lifespan has close dependence on water mass flow rate. The results also indicate that the heat production can be improved when the horizontal fracture spacing, well deviation angle, and production flow rate are under reasonable conditions. To conduct the reservoir modeling and simulations, an open-source, finite element based, fully implicit, fully coupled hydrothermal code, namely FALCON, has been developed and used in this work. Compared with most other existing codes that are either closed-source or commercially available in this area, this new open-source code has demonstrated a code development strategy that aims to provide an unparalleled easiness for user-customization and multi-physics coupling. Test results have shown that the FALCON code is able to complete the long-term tests efficiently and accurately, thanks to the state-of-the-art nonlinear and linear solver algorithms implemented in the code.
Uncertainty in unprotected loss-of-heat-sink, loss-of-flow, and transient-overpower accidents.
Energy Technology Data Exchange (ETDEWEB)
Morris, E. E.; Nuclear Engineering Division
2007-10-08
The sensitivities of various output parameters to selected input parameters in unprotected combined loss of heat-sink and loss-of-flow (ULOHS), loss-of-flow (ULOF), and transient-overpower (UTOP) accidents are explored in this report. This line of investigation was suggested by R. A. Wigeland. For an initial examination of potential sensitivities, the MATWS computer program has been compiled as part of a dynamic link library (DLL) so that uncertain input parameters can be sampled from their probability distributions using the GoldSim simulation software. The MATWS program combines the point-kinetics module from the SAS4A/SASSYS computer code with a simplified representation of the reactor heat removal system. Coupling with the GoldSim software by means of a DLL not only provides a convenient mechanism for sampling the stochastic input parameters but also allows the use of various tools that are available in GoldSim for analyzing the dependence of various MATWS outputs on these parameters. Should a decision be made to continue this investigation, the techniques used to couple MATWS and GoldSim could also be applied to couple the SAS4A/SASSYS computer code with GoldSim. The work described here illustrates the type of results that can be obtained from the stochastic analysis.
Uncertainty in unprotected loss-of-heat-sink, loss-of-flow, and transient-overpower accidents
International Nuclear Information System (INIS)
Morris, E.E.
2007-01-01
The sensitivities of various output parameters to selected input parameters in unprotected combined loss of heat-sink and loss-of-flow (ULOHS), loss-of-flow (ULOF), and transient-overpower (UTOP) accidents are explored in this report. This line of investigation was suggested by R. A. Wigeland. For an initial examination of potential sensitivities, the MATWS computer program has been compiled as part of a dynamic link library (DLL) so that uncertain input parameters can be sampled from their probability distributions using the GoldSim simulation software. The MATWS program combines the point-kinetics module from the SAS4A/SASSYS computer code with a simplified representation of the reactor heat removal system. Coupling with the GoldSim software by means of a DLL not only provides a convenient mechanism for sampling the stochastic input parameters but also allows the use of various tools that are available in GoldSim for analyzing the dependence of various MATWS outputs on these parameters. Should a decision be made to continue this investigation, the techniques used to couple MATWS and GoldSim could also be applied to couple the SAS4A/SASSYS computer code with GoldSim. The work described here illustrates the type of results that can be obtained from the stochastic analysis
Coupled equations for transient water flow, heat flow, and ...
Indian Academy of Sciences (India)
interacting processes, including flow of fluids, deformation of porous materials, chemical reactions, and transport of ... systems involving the flow of water, heat, and deformation. Such systems are ..... Defined thus, αI is independent of boundary con- ditions in an ... perature change with free deformation at constant total stress ...
Heat transfer in vapour-liquid flow of carbon dioxide
International Nuclear Information System (INIS)
Yagov, V.V.
2009-01-01
During the last decade a number of studies of boiling heat transfer in carbon dioxide notably increase. As a field of CO 2 practical using corresponds to high reduced pressures, and a majority of available experimental data on CO 2 flow boiling even in submillimetric channels relate to turbulent liquid flow regimes, a possibility arises to develop sufficiently general method for HTC predicting. Under the above conditions nucleate boiling occurs up to rather high flow quality, even in annular flow regime due to extremely small size of an equilibrium vapour bubble. This conclusion is in agreement with the available experimental data. The predicting equation for nucleate boiling heat transfer developed by the present author in 1988 is valid for any nonmetallic liquid. A contribution of forced convection in heat transfer is calculated according to the Petukhov et al. equation with correction factor, which accounted for an effect of velocity increase due to evaporation. This effect can be essential at relatively small heat fluxes and rather high mass flow rates. The Reynolds analogy and homogeneous model are used in order to account for the convective heat transfer augmentation in two-phase flow. Due to low ratio of liquid and vapour densities at high reduced pressures the homogeneous approximation of two-phase flow seems to be warranted. A total heat transfer coefficient is calculated as an interpolated value of boiling and convective HTCs. The experimental data on CO 2 flow boiling related to regimes before heated wall dryout incipience are in rather good agreement with the calculations. (author)
Dry-out heat fluxes of falling film and low-mass flux upward-flow in heated tubes
International Nuclear Information System (INIS)
Koizumi, Yasuo; Ueda, Tatsuhiro; Matsuo, Teruyuki; Miyota, Yukio
1998-01-01
Dry-out heat fluxes were investigated experimentally for a film flow falling down on the inner surface of vertical heated-tubes and for a low mass flux forced-upward flow in the tubes using R 113. This work followed the study on those for a two-phase natural circulation system. For the falling film boiling, flow state observation tests were also performed, where dry-patches appearing and disappearing repeatedly were observed near the exit end of the heated section at the dry-out heat flux conditions. Relation between the dry-out heat flux and the liquid film flow rate is analyzed. The dry-out heat fluxes of the low mass flux upflow are expressed well by the correlation proposed in the previous work. The relation for the falling film boiling shows a similar trend to that for the upflow boiling, however, the dry-out heat fluxes of the falling film are much lower, approximately one third, than those of the upward flow. (author)
International Nuclear Information System (INIS)
McHugh, P.R.; Ramshaw, J.D.
1991-11-01
MAGMA is a FORTRAN computer code designed to viscous flow in in situ vitrification melt pools. It models three-dimensional, incompressible, viscous flow and heat transfer. The momentum equation is coupled to the temperature field through the buoyancy force terms arising from the Boussinesq approximation. All fluid properties, except density, are assumed variable. Density is assumed constant except in the buoyancy force terms in the momentum equation. A simple melting model based on the enthalpy method allows the study of the melt front progression and latent heat effects. An indirect addressing scheme used in the numerical solution of the momentum equation voids unnecessary calculations in cells devoid of liquid. Two-dimensional calculations can be performed using either rectangular or cylindrical coordinates, while three-dimensional calculations use rectangular coordinates. All derivatives are approximated by finite differences. The incompressible Navier-Stokes equations are solved using a new fully implicit iterative technique, while the energy equation is differenced explicitly in time. Spatial derivatives are written in conservative form using a uniform, rectangular, staggered mesh based on the marker and cell placement of variables. Convective terms are differenced using a weighted average of centered and donor cell differencing to ensure numerical stability. Complete descriptions of MAGMA governing equations, numerics, code structure, and code verification are provided. 14 refs
Energy Technology Data Exchange (ETDEWEB)
McHugh, P.R.; Ramshaw, J.D.
1991-11-01
MAGMA is a FORTRAN computer code designed to viscous flow in in situ vitrification melt pools. It models three-dimensional, incompressible, viscous flow and heat transfer. The momentum equation is coupled to the temperature field through the buoyancy force terms arising from the Boussinesq approximation. All fluid properties, except density, are assumed variable. Density is assumed constant except in the buoyancy force terms in the momentum equation. A simple melting model based on the enthalpy method allows the study of the melt front progression and latent heat effects. An indirect addressing scheme used in the numerical solution of the momentum equation voids unnecessary calculations in cells devoid of liquid. Two-dimensional calculations can be performed using either rectangular or cylindrical coordinates, while three-dimensional calculations use rectangular coordinates. All derivatives are approximated by finite differences. The incompressible Navier-Stokes equations are solved using a new fully implicit iterative technique, while the energy equation is differenced explicitly in time. Spatial derivatives are written in conservative form using a uniform, rectangular, staggered mesh based on the marker and cell placement of variables. Convective terms are differenced using a weighted average of centered and donor cell differencing to ensure numerical stability. Complete descriptions of MAGMA governing equations, numerics, code structure, and code verification are provided. 14 refs.
Frictional strength and heat flow of southern San Andreas Fault
Zhu, P. P.
2016-01-01
Frictional strength and heat flow of faults are two related subjects in geophysics and seismology. To date, the investigation on regional frictional strength and heat flow still stays at the stage of qualitative estimation. This paper is concentrated on the regional frictional strength and heat flow of the southern San Andreas Fault (SAF). Based on the in situ borehole measured stress data, using the method of 3D dynamic faulting analysis, we quantitatively determine the regional normal stress, shear stress, and friction coefficient at various seismogenic depths. These new data indicate that the southern SAF is a weak fault within the depth of 15 km. As depth increases, all the regional normal and shear stresses and friction coefficient increase. The former two increase faster than the latter. Regional shear stress increment per kilometer equals 5.75 ± 0.05 MPa/km for depth ≤15 km; regional normal stress increment per kilometer is equal to 25.3 ± 0.1 MPa/km for depth ≤15 km. As depth increases, regional friction coefficient increment per kilometer decreases rapidly from 0.08 to 0.01/km at depths less than ~3 km. As depth increases from ~3 to ~5 km, it is 0.01/km and then from ~5 to 15 km, and it is 0.002/km. Previously, frictional strength could be qualitatively determined by heat flow measurements. It is difficult to obtain the quantitative heat flow data for the SAF because the measured heat flow data exhibit large scatter. However, our quantitative results of frictional strength can be employed to investigate the heat flow in the southern SAF. We use a physical quantity P f to describe heat flow. It represents the dissipative friction heat power per unit area generated by the relative motion of two tectonic plates accommodated by off-fault deformation. P f is called "fault friction heat." On the basis of our determined frictional strength data, utilizing the method of 3D dynamic faulting analysis, we quantitatively determine the regional long-term fault
Convective heat transfer in supercritical flows of CO_2 in tubes with and without flow obstacles
International Nuclear Information System (INIS)
Eter, Ahmad; Groeneveld, Dé; Tavoularis, Stavros
2017-01-01
Highlights: • Measurements of supercritical heat transfer in tubes equipped with obstacles were obtained and compared with results in base tubes. • In general, flow obstacles improve supercritical heat transfer, but under certain conditions have a negative effect on it. • New correlations describing obstacle-enhanced supercritical heat transfer in the liquid-like and gas-like regimes are fitted to the data. - Abstract: Heat transfer measurements to CO_2-cooled tubes with and without flow obstacles at supercritical pressures were obtained at the University of Ottawa’s supercritical pressure test facility. The effects of obstacle geometry (obstacle pitch, obstacle shape, flow blockage) on the wall temperature and heat transfer coefficient were investigated. Tests were performed for vertical upward flow in a directly heated 8 mm ID tube for a pressure range from 7.69 to 8.36 MPa, a mass flux range from 200 to 1184 kg/m"2 s, and a heat flux range from 1 to 175 kW/m"2. The results are presented graphically in plots of wall temperature and heat transfer coefficient vs. bulk specific enthalpy of the fluid. The effects of flow parameters and flow obstacle geometry on supercritical heat transfer for both normal and deteriorated heat transfer are discussed. A comparison of the measurements with leading prediction methods for supercritical heat transfer in bare tubes and for spacer effects is also presented. The optimum increase in heat transfer coefficient was found to be for blunt obstacles, having a large flow blockage, and a short obstacle pitch.
Heat Transfer treatment in computer codes for safety analysis
International Nuclear Information System (INIS)
Jerele, A.; Gregoric, M.
1984-01-01
Increased number of operating nuclear power plants has stressed importance of nuclear safety evaluation. For this reason, accordingly to regulatory commission request, safety analyses with computer codes are preformed. In this paper part of this thermohydraulic models dealing with wall-to-fluid heat transfer correlations in computer codes TRAC=PF1, RELAP4/MOD5, RELAP5/MOD1 and COBRA-IV is discussed. (author)
International Nuclear Information System (INIS)
Lin, Yueh-Hung; Li, Guang-Cheng; Yang, Chien-Yuh
2015-01-01
This study provides an Infrared Thermal Image observation on the evaporation heat transfer of refrigerant R-410A in plate heat exchanger with various flow arrangement and exit superheat conditions. An experimental method was derived for estimating the superheat region area of two-phase refrigerant evaporation in plate heat exchanger. The experimental results show that the superheat region area for parallel flow is much larger than that for counter flow as that estimated by Yang et al. [9]. There is an early superheated region at the central part of the plate heat exchanger for parallel flow arrangement. This effect is not significant for counter flow arrangement. The Yang et al. [9] method under estimated the superheat area approximately 40%–53% at various flow rates and degree of exit superheat. Even though the flow inside a plate heat exchanger is extremely turbulent because of the chevron flow passages, the assumption of uniform temperature distribution in the cross section normal to the bulk flow direction will cause significant uncertainties for estimating the superheat area for refrigerant evaporating in a plate heat exchanger
International Nuclear Information System (INIS)
Liu, Wei; Tamai, Hidesada; Yoshida, Hiroyuki; Takase, Kazuyuki; Hayafune, Hiroki; Futagami, Satoshi; Kisohara, Naoyuki
2008-01-01
For the Steam Generator (SG) with straight double-walled heat transfer tube that used in sodium cooled Faster Breeder Reactor, flow instability is one of the most important items need researching. As the first step of the research, thermal hydraulics experiments were performed under high pressure condition in JAEA with using a straight tube. Pressure drop, heat transfer coefficients and void fraction data were derived. This paper evaluates the pressure drop data with TRAC-BF1 code. The Pffan's correlation for single phase flow and the Martinelli-Nelson's two-phase flow multiplier are found can be well predicted the present pressure drop data under high pressure condition. (author)
Heat Transfer Enhancement in Separated and Vortex Flows
Energy Technology Data Exchange (ETDEWEB)
Richard J. Goldstein
2004-05-27
This document summarizes the research performance done at the Heat Transfer Laboratory of the University of Minnesota on heat transfer and energy separation in separated and vortex flow supported by DOE in the period September 1, 1998--August 31, 2003. Unsteady and complicated flow structures in separated or vortex flows are the main reason for a poor understanding of heat transfer under such conditions. The research from the University of Minnesota focused on the following important aspects of understanding such flows: (1) Heat/mass transfer from a circular cylinder; (2) study of energy separation and heat transfer in free jet flows and shear layers; and (3) study of energy separation on the surface and in the wake of a cylinder in crossflow. The current study used three different experimental setups to accomplish these goals. A wind tunnel and a liquid tunnel using water and mixtures of ethylene glycol and water, is used for the study of prandtl number effect with uniform heat flux from the circular cylinder. A high velocity air jet is used to study energy separation in free jets. A high speed wind tunnel, same as used for the first part, is utilized for energy separation effects on the surface and in the wake of the circular cylinder. The final outcome of this study is a substantial advancement in this research area.
Non-linear heat transfer computer code by finite element method
International Nuclear Information System (INIS)
Nagato, Kotaro; Takikawa, Noboru
1977-01-01
The computer code THETA-2D for the calculation of temperature distribution by the two-dimensional finite element method was made for the analysis of heat transfer in a high temperature structure. Numerical experiment was performed for the numerical integration of the differential equation of heat conduction. The Runge-Kutta method of the numerical experiment produced an unstable solution. A stable solution was obtained by the β method with the β value of 0.35. In high temperature structures, the radiative heat transfer can not be neglected. To introduce a term of the radiative heat transfer, a functional neglecting the radiative heat transfer was derived at first. Then, the radiative term was added after the discretion by variation method. Five model calculations were carried out by the computer code. Calculation of steady heat conduction was performed. When estimated initial temperature is 1,000 degree C, reasonable heat blance was obtained. In case of steady-unsteady temperature calculation, the time integral by THETA-2D turned out to be under-estimation for enthalpy change. With a one-dimensional model, the temperature distribution in a structure, in which heat conductivity is dependent on temperature, was calculated. Calculation with a model which has a void inside was performed. Finally, model calculation for a complex system was carried out. (Kato, T.)
International Nuclear Information System (INIS)
Li, Si-Ning; Zhang, Hong-Na; Li, Xiao-Bin; Li, Qian; Li, Feng-Chen; Qian, Shizhi; Joo, Sang Woo
2017-01-01
Highlights: • Heat transfer performance of non-Newtonian fluid flow in a MHS is studied. • Pseudo-plastic fluid flow can clearly promote the heat transfer efficiency in MMC. • Heat transfer enhancement is attributed to the emergence of secondary flow. • The heat transfer uniformity can also be improved by pseudo-plastic fluid flow. - Abstract: As the miniaturization and integration become the leading trend of the micro-electro-mechanical systems, it is of great significance to improve the microscaled heat transfer performance. This paper presents a three-dimensional (3D) numerical simulation on the flow characteristics and heat transfer performance of non-Newtonian fluid flow in a manifold microchannel (MMC) heat sink and traditional microchannel (TMC) heat sink. The non-Newtonian fluid was described by the power-law model. The analyses concentrated on the non-Newtonian fluid effect on the heat transfer performance, including the heat transfer efficiency and uniformity of temperature distribution, as well as the influence of inlet/outlet configurations on fluid flow and heat transfer. Comparing with Newtonian fluid flow, pseudo-plastic fluid could reduce the drag resistance in both MMC and TMC, while the dilatant fluid brought in quite larger drag resistance. For the heat transfer performance, the introduction of pseudo-plastic fluid flow greatly improved the heat transfer efficiency owing to the generation of secondary flow due to the shear-thinning property. Besides, the temperature distribution in MMC was more uniform by using pseudo-plastic fluid. Moreover, the inlet/outlet configuration was also important for the design and arrangement of microchannel heat sinks, since the present work showed that the maximum temperature was prone to locating in the corners near the inlet and outlet. This work provides guidance for optimal design of small-scale heat transfer devices in many cooling applications, such as biomedical chips, electronic systems, and
User's manual for the Heat Pipe Space Radiator design and analysis Code (HEPSPARC)
Hainley, Donald C.
1991-01-01
A heat pipe space radiatior code (HEPSPARC), was written for the NASA Lewis Research Center and is used for the design and analysis of a radiator that is constructed from a pumped fluid loop that transfers heat to the evaporative section of heat pipes. This manual is designed to familiarize the user with this new code and to serve as a reference for its use. This manual documents the completed work and is intended to be the first step towards verification of the HEPSPARC code. Details are furnished to provide a description of all the requirements and variables used in the design and analysis of a combined pumped loop/heat pipe radiator system. A description of the subroutines used in the program is furnished for those interested in understanding its detailed workings.
International Nuclear Information System (INIS)
Liu, Qiusheng; Fukuda, Katsuya
2003-01-01
The transient heat transfer coefficients for forced convection flow of helium gas over a horizontal cylinder were measured under wide experimental conditions. The platinum cylinder with a diameter of 1.0 mm was used as test heater and heated by electric current with an exponentially increasing heat input of Q 0 exp(t/τ). The gas flow velocities ranged from 5 to 35 m/s, the gas temperatures ranged from 25 to 80degC, and the periods of heat generation rate, τ, ranged from 40 ms to 20 s. The surface superheat and heat flux increase exponentially as the heat generation rate increases with the exponential function. It was clarified that the heat transfer coefficient approaches the quasi-steady-state one for the period τ longer than about 1 s, and it becomes higher for the period shorter than around 1 s. The transient heat transfer shows less dependence on the gas flowing velocity when the period becomes very shorter. The gas temperature in this study shows little influence on the heat transfer coefficient. Semi-empirical correlation for quasi-steady-state heat transfer was obtained based on the experimental data. The ratios of transient Nusselt number Nu tr to quasi-steady-state Nusselt number Nu st at various periods, flow velocities, and gas temperatures were obtained. The heat transfer shifts to the quasi-steady-state heat transfer for longer periods and shifts to the transient heat transfer for shorter periods at the same flow velocity. It also approaches the quasi-steady-state one for higher flow velocity at the same period. Empirical correlation for transient heat transfer was also obtained based on the experimental data. (author)
Contribution to the modelling of flows and heat transfers during the reflooding phase of a PWR core
International Nuclear Information System (INIS)
Colas, D.
1984-01-01
This thesis contributes to modelise thermohydraulic phenomena occuring in a pressurized water nuclear reactor core during the reflood phase of a LOCA. The reference accident and phenomena occuring during reflooding are described as well as flow regime and heat transfer proposed models. With these models, we developed a code to compute fluid conditions and fuel rods temperatures in a reactor core chanel. In order to test this code, results of computation are compared with experiments (FLECHT Skewed Tests) and a conclusion is drawn [fr
Flow visualization in heat-generating porous media
International Nuclear Information System (INIS)
Lee, D.O.; Nilson, R.H.
1977-11-01
The work reported is in support of the Sandia Post-Accident Heat Removal Program, in which simulated LMFBR beds will be subjected to in-pile heating in the ACPR (Annular Core Pulsed Reactor). Flow visualization experiments were performed to gain some insight into the flow patterns and temperature distributions in a fluid-saturated heat-generating porous medium. Although much of the information presented is of a qualitative nature, it is useful in the recognition of the controlling transport process and in the formulation of analytic and numerical models
Heat flow in Indian Gondwana basins and heat production of their basement rocks
Energy Technology Data Exchange (ETDEWEB)
Rao, G.V.; Rao, R.U.M.
1983-01-01
Temperatures have been measured in eight boreholes (ranging from 260 to 800 m in depth) in five Gondwana basins of the Damodar and Son valleys. With the aid of about 250 thermal conductivity determinations on core samples from these holes, heat flow has been evaluated. Measurements of radioactive heat generation have been made on samples of Precambrian gneisses constituting the basement for the Sonhat (Son valley) and Chintalapudi (Godavari valley) basins. Heat-flow values from all of the Damodar valley basins are within the narrow range of 69-79 mW/m exp(2). The value from the Sonhat basin (107 mW/m exp(2)) is significantly higher. The generally high heat flows observed in Gondwana basins of India cannot be attributed to the known tectonism or igneous activity associated with these basins. The plots of heat flow vs. heat generation for three Gondwana basins (Jharia, Sonhat and Chintalapudi) are on the same line as those of three regions in the exposed Precambrian crystalline terrains in the northern part of the Indian shield. This indicates that the crust under exposed regions of the Precambrian crystalline rocks as well as the Gondwana basins, form an integral unit as far as the present-day geothermal character is concerned. (5 figs., 14 refs., 4 tables).
Heat flow in Indian Gondwana basins and heat production of their basement rocks
Rao, G. V.; Rao, R. U. M.
1983-01-01
Temperatures have been measured in eight boreholes (ranging from 260 to 800 m in depth) in five Gondwana basins of the Damodar and Son valleys. With the aid of about 250 thermal conductivity determinations on core samples from these holes, heat flow has been evaluated. Measurements of radioactive heat generation have been made on samples of Precambrian gneisses constituting the basement for the Sonhat (Son valley) and Chintalapudi (Godavari valley) basins. Heat-flow values from all of the Damodar valley basins are within the narrow range of 69-79 mW/m 2. The value from the Sonhat basin (107 mW/m 2) is significantly higher. The generally high heat flows observed in Gondwana basins of India cannot be attributed to the known tectonism or igneous activity associated with these basins. The plots of heat flow vs. heat generation for three Gondwana basins (Jharia, Sonhat and Chintalapudi) are on the same line as those of three regions in the exposed Precambrian crystalline terrains in the northern part of the Indian shield. This indicates that the crust under exposed regions of the Precambrian crystalline rocks as well as the Gondwana basins, form an integral unit as far as the present-day geothermal character is concerned.
Urquiza, Eugenio
This work presents a comprehensive thermal hydraulic analysis of a compact heat exchanger using offset strip fins. The thermal hydraulics analysis in this work is followed by a finite element analysis (FEA) to predict the mechanical stresses experienced by an intermediate heat exchanger (IHX) during steady-state operation and selected flow transients. In particular, the scenario analyzed involves a gas-to-liquid IHX operating between high pressure helium and liquid or molten salt. In order to estimate the stresses in compact heat exchangers a comprehensive thermal and hydraulic analysis is needed. Compact heat exchangers require very small flow channels and fins to achieve high heat transfer rates and thermal effectiveness. However, studying such small features computationally contributes little to the understanding of component level phenomena and requires prohibitive computational effort using computational fluid dynamics (CFD). To address this issue, the analysis developed here uses an effective porous media (EPM) approach; this greatly reduces the computation time and produces results with the appropriate resolution [1]. This EPM fluid dynamics and heat transfer computational code has been named the Compact Heat Exchanger Explicit Thermal and Hydraulics (CHEETAH) code. CHEETAH solves for the two-dimensional steady-state and transient temperature and flow distributions in the IHX including the complicating effects of temperature-dependent fluid thermo-physical properties. Temperature- and pressure-dependent fluid properties are evaluated by CHEETAH and the thermal effectiveness of the IHX is also calculated. Furthermore, the temperature distribution can then be imported into a finite element analysis (FEA) code for mechanical stress analysis using the EPM methods developed earlier by the University of California, Berkeley, for global and local stress analysis [2]. These simulation tools will also allow the heat exchanger design to be improved through an
Critical heat flux and flow pattern for water flow in annular geometry
International Nuclear Information System (INIS)
Park, J.-W.; Baek, W.-P.; Chang, S.H.
1997-01-01
An experimental study on critical heat flux (CHF) and two-phase flow visualization has been performed for water flow in internally-heated, vertical, concentric annuli under near atmospheric pressure. Tests have been done under stable forced-circulation, upward and downward flow conditions with three test sections of relatively large gap widths (heated length = 0.6 m, inner diameter 19 mm, outer diameter = 29, 35 and 51 mm). The outer wall of the test section was made up of the transparent Pyrex tube to allow the observation of flow patterns near the CHF occurrence. The CHF mechanism was changed in the order of flooding, churn-to-annular flow transition and local dryout under a large bubble in churn flow as the flow rate was increased from zero to higher values. Observed parametric trends are consistent with the previous understanding except that the CHF for downward flow is considerably lower than that for the upward flow. In addition to the experiment, selected CHF correlations for annuli are assessed based on 1156 experimental data from various sources. The Doerffer et al. (1994); Barnett (1966); Jannsen and Kervinen (1963); Levitan and Lantsman (1977) correlations show reasonable predictions for wide parameter ranges, among which the Doerffer et al. (1994) correlation shows the widest parameter ranges and a possibility of further improvement. However, there is no correlation predicting the low-pressure, low-flow CHF satisfactorily. (orig.)
Entropy resistance analyses of a two-stream parallel flow heat exchanger with viscous heating
International Nuclear Information System (INIS)
Cheng Xue-Tao; Liang Xin-Gang
2013-01-01
Heat exchangers are widely used in industry, and analyses and optimizations of the performance of heat exchangers are important topics. In this paper, we define the concept of entropy resistance based on the entropy generation analyses of a one-dimensional heat transfer process. With this concept, a two-stream parallel flow heat exchanger with viscous heating is analyzed and discussed. It is found that the minimization of entropy resistance always leads to the maximum heat transfer rate for the discussed two-stream parallel flow heat exchanger, while the minimizations of entropy generation rate, entropy generation numbers, and revised entropy generation number do not always. (general)
Energy Technology Data Exchange (ETDEWEB)
Bayoumi, M.; Charlot, R.; Ricque, R.
1976-05-01
For analyzing, correlating and extrapolating experimental burn-out results obtained with LWR rod bundles, it is necessary to know the distributions of mass flow rate and quality between the subchannels. A description is presented of an experimental study in progress at the CEN-Grenoble for determining and adjusting the laws of mixing in the FLICA Code which is used to predict these distributions. The experiments are performed on the FRENESIE loop with Freon 12. The test section, in vertical position, consists of a four rod bundle in a channel with square section. The heat flux is axially uniform. The flow of each subchannel can be sampled in ''isokinetic conditions,'' at the end of the heating length. Thermodynamic quality and mass flow rate of the samplings are measured in steady state conditions by using respectively a calorimeter and a turbine flow meter. The test facility is described and experimental data are presented and discussed.
Heat transfer and fluid flow in regular rod arrays with opposing flow
International Nuclear Information System (INIS)
Yang, J.W.
1979-01-01
The heat transfer and fluid flow problem of opposing flow in the fully developed laminar region has been solved analytically for regular rod arrays. The problem is governed by two parameters: the pitch-to-diameter ratio and the Grashof-to-Reynolds number ratio. The critical Gr/Re ratios for flow separation caused by the upward buoyancy force on the downward flow were evaluated for a large range of P/D ratios of the triangular array. Numerical results reveal that both the heat transfer and pressure loss are reduced by the buoyancy force. Applications to nuclear reactors are discussed
Electron and ion cyclotron heating calculations in the tandem-mirror modeling code MERTH
International Nuclear Information System (INIS)
Smith, G.R.
1985-01-01
To better understand and predict tandem-mirror experiments, we are building a comprehensive Mirror Equilibrium Radial Transport and Heating (MERTH) code. In this paper we first describe our method for developing the code. Then we report our plans for the installation of physics packages for electron- and ion-cyclotron heating of the plasma
2.5 MWT Heat Exchanger Designs for Passive DHRS in PGSFR
Energy Technology Data Exchange (ETDEWEB)
Kim, Dehee; Eoh, Jaehyuk; Lee, Tae-Ho [KAERI, Daejeon (Korea, Republic of)
2016-05-15
Decay Heat Removal System (DHRS) of PGSFR consists of two passive DHRS (PDHRS) trains and two active DHRS (ADHRS) trains. Recently, total heat removal capacity of the DHRS in the PGSFR has increased to 10 MWT from 4 MWT reflecting safety analysis results. Consequently, DHRS components including heat exchangers, dampers, electro-magnetic pump, fan, piping, expansion tank and stack have been newly designed. In this work, physical models and correlations to design two main components of the PDHRS, decay heat exchanger (DHX) and natural-draft sodium-to-air heat exchanger (AHX), are introduced and designed data are presented. Physical models and correlations applied for heat exchangers in the PDHRS design were introduced and design works using the SHXSA and AHXSA codes has been completed for 2.5 MWT decay heat removal capability. DHX and AHX are designed utilizing SHXSA and AHXSA codes, respectively. Those design codes have capability of thermal sizing and performance analysis for the shell-and-tube type and counter-current flow heat exchanger unit. Since both SHXSA and AHXSA codes are similar, following description is focused on the SHXSA code. A single flow channel associated with an individual heat transfer tube is basically considered for thermal sizing and then the calculation results and design variables regarding heat transfer and pressure drop, etc. are extended to whole tubes. Various correlations of heat transfer and pressure loss for the shell- and tubeside flows were implemented in the computer codes. The analysis domain is discretized into several control volumes and heat transfer and pressure losses are calculated in each control volume.
A simplified method of calculating heat flow through a two-phase heat exchanger
Energy Technology Data Exchange (ETDEWEB)
Yohanis, Y.G. [Thermal Systems Engineering Group, Faculty of Engineering, University of Ulster, Newtownabbey, Co Antrim, BT37 0QB Northern Ireland (United Kingdom)]. E-mail: yg.yohanis@ulster.ac.uk; Popel, O.S. [Non-traditional Renewable Energy Sources, Institute for High Temperatures, Russian Academy of Sciences, 13/19 Izhorskaya str., IVTAN, Moscow 125412 (Russian Federation); Frid, S.E. [Non-traditional Renewable Energy Sources, Institute for High Temperatures, Russian Academy of Sciences, 13/19 Izhorskaya str., IVTAN, Moscow 125412 (Russian Federation)
2005-10-01
A simplified method of calculating the heat flow through a heat exchanger in which one or both heat carrying media are undergoing a phase change is proposed. It is based on enthalpies of the heat carrying media rather than their temperatures. The method enables the determination of the maximum rate of heat flow provided the thermodynamic properties of both heat-carrying media are known. There will be no requirement to separately simulate each part of the system or introduce boundaries within the heat exchanger if one or both heat-carrying media undergo a phase change. The model can be used at the pre-design stage, when the parameters of the heat exchangers may not be known, i.e., to carry out an assessment of a complex energy scheme such as a steam power plant. One such application of this model is in thermal simulation exercises within the TRNSYS modeling environment.
A simplified method of calculating heat flow through a two-phase heat exchanger
International Nuclear Information System (INIS)
Yohanis, Y.G.; Popel, O.S.; Frid, S.E.
2005-01-01
A simplified method of calculating the heat flow through a heat exchanger in which one or both heat carrying media are undergoing a phase change is proposed. It is based on enthalpies of the heat carrying media rather than their temperatures. The method enables the determination of the maximum rate of heat flow provided the thermodynamic properties of both heat-carrying media are known. There will be no requirement to separately simulate each part of the system or introduce boundaries within the heat exchanger if one or both heat-carrying media undergo a phase change. The model can be used at the pre-design stage, when the parameters of the heat exchangers may not be known, i.e., to carry out an assessment of a complex energy scheme such as a steam power plant. One such application of this model is in thermal simulation exercises within the TRNSYS modeling environment
Ghamlouch, T.; Roux, S.; Bailleul, J.-L.; Lefèvre, N.; Sobotka, V.
2017-10-01
Today's aerospace industrial first priority is the quality improvement of the composite material parts with the reduction of the manufacturing time in order to increase their quality/cost ratio. A fabrication method that could meet these specifications especially for large parts is the autoclave curing process. In fact the autoclave molding ensures the thermal control of the composite parts during the whole curing cycle. However the geometry of the tools as well as their positioning in the autoclave induce non uniform and complex flows around composite parts. This heterogeneity implies non-uniform heat transfers which can directly impact on part quality. One of the main challenges is therefore to describe the flow field inside an autoclave as well as the convective heat transfer from the heated pressurized gas to the composite part and the mold. For this purpose, and given the technical issues associated with instrumentation and measurements in actual autoclaves, an autoclave model was designed and then manufactured based on similarity laws. This tool allows the measurement of the flow field around representative real industrial molds using the PIV technique and the characterization of the heat transfer thanks to thermal instrumentation. The experimental results are then compared with those derived from numerical simulations using a commercial RANS CFD code. This study aims at developing a semi-empirical approach for the prediction of the heat transfer coefficient around the parts and therefore predicts its thermal history during the process with a view of optimization.
International Nuclear Information System (INIS)
Kaminaga, Masanori
1990-02-01
The COOLOD-N code provides a capability for the analysis of the steady-state thermal-hydraulics of research reactors in which plate-type fuel is employed. This code is revised version of the COOLOD code, and is applicable not only to a forced convection cooling mode, but also to a natural convection cooling mode. In the code, a function to calculate flow rate under a natural convection, and a heat transfer package which was a subroutine program to calculate heat transfer coefficient, ONB temperature and DNB heat flux, and was especially developed for the upgraded JRR-3, have been newly added to the COOLOD code. The COOLOD-N code also has a capability of calculating the heat flux at onset of flow instability as well as DNB heat flux. (author)
Analysis of slip flow heat transfer between two unsymmetrically
Indian Academy of Sciences (India)
This paper presents an analytical investigation to study the heat transfer and fluid flow characteristics in the slip flow region for hydrodynamically and thermally fully developed flow between parallel plates.Both upper and lower plates are subjected to asymmetric heat flux boundary conditions. The effect of first ordervelocity ...
Liquid metal heat transfer in heat exchangers under low flow rate conditions
International Nuclear Information System (INIS)
Mochizuki, Hiroyasu
2015-01-01
The present paper describes the liquid metal heat transfer in heat exchangers under low flow rate conditions. Measured data from some experiments indicate that heat transfer coefficients of liquid metals at very low Péclet number are much lower than what are predicted by the well-known empirical relations. The cause of this phenomenon was not fully understood for many years. In the present study, one countercurrent-type heat exchanger is analyzed using three, separated countercurrent heat exchanger models: one is a heat exchanger model in the tube bank region, while the upper and lower plena are modeled as two heat exchangers with a single heat transfer tube. In all three heat exchangers, the same empirical correlation is used in the heat transfer calculation on the tube and the shell sides. The Nusselt number, as a function of the Péclet number, calculated from measured temperature and flow rate data in a 50 MW experimental facility was correctly reproduced by the calculation result, when the calculated result is processed in the same way as the experiment. Finally, it is clarified that the deviation is a superficial phenomenon which is caused by the heat transfer in the plena of the heat exchanger. (author)
Temperature-gated thermal rectifier for active heat flow control.
Zhu, Jia; Hippalgaonkar, Kedar; Shen, Sheng; Wang, Kevin; Abate, Yohannes; Lee, Sangwook; Wu, Junqiao; Yin, Xiaobo; Majumdar, Arun; Zhang, Xiang
2014-08-13
Active heat flow control is essential for broad applications of heating, cooling, and energy conversion. Like electronic devices developed for the control of electric power, it is very desirable to develop advanced all-thermal solid-state devices that actively control heat flow without consuming other forms of energy. Here we demonstrate temperature-gated thermal rectification using vanadium dioxide beams in which the environmental temperature actively modulates asymmetric heat flow. In this three terminal device, there are two switchable states, which can be regulated by global heating. In the "Rectifier" state, we observe up to 28% thermal rectification. In the "Resistor" state, the thermal rectification is significantly suppressed (Rectifier state. This temperature-gated rectifier can have substantial implications ranging from autonomous thermal management of heating and cooling systems to efficient thermal energy conversion and storage.
Heat transfer characteristics of alkali metals flowing across tube banks
International Nuclear Information System (INIS)
Sugiyama, K.; Ishiguro, R.; Kojima, Y.; Kanaoka, H.
2004-01-01
For the purpose of getting heat transfer coefficients of alkali metals flowing across tube banks at an acceptable level, we propose to use an inviscid-irrotational flow model, which is based on our flow visualization experiment. We show that the heat transfer coefficients obtained for the condition where only the test rod is heated in tube banks considerably differ from those obtained for the condition where all the rods are heated, because of interference between thick thermal boundary layers of alkali metals. We also confirm that the analytical values obtained by this flow model are in a reasonable agreement with experimental values. (author)
Analysis of Tube Bank Heat Transfer In Downward Directed Foam Flow
Directory of Open Access Journals (Sweden)
Jonas Gylys
2004-06-01
Full Text Available Apparatus with the foam flow are suitable to use in different technologies like heat exchangers, food industry, chemical and oil processing industry. Statically stable liquid foam until now is used in technologic systems rather seldom. Although a usage of this type of foam as heat transfer agent in foam equipment has a number of advantages in comparison with one phase liquid equipment: small quantity of liquid is required, heat transfer rate is rather high, mass of equipment is much smaller, energy consumption for foam delivery into heat transfer zone is lower. The paper analyzes the peculiarities of heat transfer from distributed in staggered order and perpendicular to foam flow in channel of rectangular cross section tube bundle to the foam flow. It was estimated the dependence of mean gas velocity and volumetric void fraction of foam flow to heat transfer in downward foam flow. Significant difference of heat transfer intensity from front and back tubes of tube row in laminar foam flow was noticed. Dependence of heat transfer on flow velocity and volumetric void fraction of foam was confirmed and estimated by criterion equations.
Building codes as barriers to solar heating and cooling of buildings
Energy Technology Data Exchange (ETDEWEB)
Meeker, F.O. III
1978-04-01
The application of building codes to solar energy systems for heating and cooling of buildings is discussed, using as typical codes the three model building codes most widely adopted by states and localities. Some potential barriers to solar energy systems are found, federal and state programs to deal with these barriers are discussed, and alternatives are suggested. To remedy this, a federal program is needed to encourage state adoption of standards and acceptance of certification of solar systems for code approval, and to encourage revisions to codes based on model legislation prepared for the federal government by the model codes groups.
CONTAIN code analyses of direct containment heating experiments
International Nuclear Information System (INIS)
Williams, D.C.; Griffith, R.O.; Tadios, E.L.; Washington, K.E.
1995-01-01
In some nuclear reactor core-melt accidents, a potential exists for molten core-debris to be dispersed into the containment under high pressure. Resulting energy transfer to the containment atmosphere can pressurize the containment. This process, known as direct containment heating (DCH), has been the subject of extensive experimental and analytical programs sponsored by the U.S. Nuclear Regulatory Commission (NRC). The DCH modeling has been an important focus for the development of the CONTAIN code. Results of a detailed independent peer review of the CONTAIN code were published recently. This paper summarizes work performed in support of the peer review in which the CONTAIN code was applied to analyze DCH experiments. Goals of this work were comparison of calculated and experimental results, CONTAIN DCH model assessment, and development of guidance for code users, including development of a standardized input prescription for DCH analysis
Turbulent Heat Transfer in Curved Pipe Flow
Kang, Changwoo; Yang, Kyung-Soo
2013-11-01
In the present investigation, turbulent heat transfer in fully-developed curved pipe flow with axially uniform wall heat flux has been numerically studied. The Reynolds numbers under consideration are Reτ = 210 (DNS) and 1,000 (LES) based on the mean friction velocity and the pipe radius, and the Prandtl number (Pr) is 0.71. For Reτ = 210 , the pipe curvature (κ) was fixed as 1/18.2, whereas three cases of κ (0.01, 0.05, 0.1) were computed in the case of Reτ = 1,000. The mean velocity, turbulent intensities and heat transfer rates obtained from the present calculations are in good agreement with the previous numerical and experimental results. To elucidate the secondary flow structures due to the pipe curvature, the mean quantities and rms fluctuations of the flow and temperature fields are presented on the pipe cross-sections, and compared with those of the straight pipe flow. To study turbulence structures and their influence on turbulent heat transfer, turbulence statistics including but not limited to skewness and flatness of velocity fluctuations, cross-correlation coefficients, an Octant analysis, and turbulence budgets are presented and discussed. Based on our results, we attempt to clarify the effects of Reynolds number and the pipe curvature on turbulent heat transfer. This research was supported by Basic Science Research Program through the National Research Foundation of Korea (NRF) funded by the Ministry of Education, Science and Technology (2010-0008457).
Yudov, Yu. V.
2018-03-01
A model is presented of the interphasic heat and mass transfer in the presence of noncondensable gases for the KORSAR/GP design code. This code was developed by FGUP NITI and the special design bureau OKB Gidropress. It was certified by Rostekhnadzor in 2009 for numerical substantiation of the safety of reactor installations with VVER reactors. The model is based on the assumption that there are three types of interphasic heat and mass transfer of the vapor component: vapor condensation or evaporation on the interphase under any thermodynamic conditions of the phases, pool boiling of the liquid superheated above the saturation temperature at the total pressure, and spontaneous condensation in the volume of gas phase supercooled below the saturation temperature at the vapor partial pressure. Condensation and evaporation on the interphase continuously occur in a two-phase flow and control the time response of the interphase heat and mass transfer. Boiling and spontaneous condensation take place only at the metastable condition of the phases and run at a quite high speed. The procedure used for calculating condensation and evaporation on the interphase accounts for the combined diffusion and thermal resistance of mass transfer in all regimes of the two-phase flow. The proposed approach accounts for, in a natural manner, a decrease in the rate of steam condensation (or generation) in the presence of noncondensing components in the gas phase due to a decrease (or increase) in the interphase temperature relative to the saturation temperature at the vapor partial pressure. The model of the interphase heat transfer also accounts for the processes of dissolution or release of noncondensing components in or from the liquid. The gas concentration at the interphase and on the saturation curve is calculated by the Henry law. The mass transfer coefficient in gas dissolution is based on the heat and mass transfer analogy. Results are presented of the verification of the
Sandia reactor kinetics codes: SAK and PK1D
International Nuclear Information System (INIS)
Pickard, P.S.; Odom, J.P.
1978-01-01
The Sandia Kinetics code (SAK) is a one-dimensional coupled thermal-neutronics transient analysis code for use in simulation of reactor transients. The time-dependent cross section routines allow arbitrary time-dependent changes in material properties. The one-dimensional heat transfer routines are for cylindrical geometry and allow arbitrary mesh structure, temperature-dependent thermal properties, radiation treatment, and coolant flow and heat-transfer properties at the surface of a fuel element. The Point Kinetics 1 Dimensional Heat Transfer Code (PK1D) solves the point kinetics equations and has essentially the same heat-transfer treatment as SAK. PK1D can address extended reactor transients with minimal computer execution time
Energy Technology Data Exchange (ETDEWEB)
Eter, Ahmad, E-mail: eng.eter@yahoo.com; Groeneveld, Dé, E-mail: degroeneveld@gmail.com; Tavoularis, Stavros, E-mail: stavros.tavoularis@uottawa.ca
2017-03-15
Highlights: • Measurements of supercritical heat transfer in tubes equipped with obstacles were obtained and compared with results in base tubes. • In general, flow obstacles improve supercritical heat transfer, but under certain conditions have a negative effect on it. • New correlations describing obstacle-enhanced supercritical heat transfer in the liquid-like and gas-like regimes are fitted to the data. - Abstract: Heat transfer measurements to CO{sub 2}-cooled tubes with and without flow obstacles at supercritical pressures were obtained at the University of Ottawa’s supercritical pressure test facility. The effects of obstacle geometry (obstacle pitch, obstacle shape, flow blockage) on the wall temperature and heat transfer coefficient were investigated. Tests were performed for vertical upward flow in a directly heated 8 mm ID tube for a pressure range from 7.69 to 8.36 MPa, a mass flux range from 200 to 1184 kg/m{sup 2} s, and a heat flux range from 1 to 175 kW/m{sup 2}. The results are presented graphically in plots of wall temperature and heat transfer coefficient vs. bulk specific enthalpy of the fluid. The effects of flow parameters and flow obstacle geometry on supercritical heat transfer for both normal and deteriorated heat transfer are discussed. A comparison of the measurements with leading prediction methods for supercritical heat transfer in bare tubes and for spacer effects is also presented. The optimum increase in heat transfer coefficient was found to be for blunt obstacles, having a large flow blockage, and a short obstacle pitch.
ATES/heat pump simulations performed with ATESSS code
Vail, L. W.
1989-01-01
Modifications to the Aquifer Thermal Energy Storage System Simulator (ATESSS) allow simulation of aquifer thermal energy storage (ATES)/heat pump systems. The heat pump algorithm requires a coefficient of performance (COP) relationship of the form: COP = COP sub base + alpha (T sub ref minus T sub base). Initial applications of the modified ATES code to synthetic building load data for two sizes of buildings in two U.S. cities showed insignificant performance advantage of a series ATES heat pump system over a conventional groundwater heat pump system. The addition of algorithms for a cooling tower and solar array improved performance slightly. Small values of alpha in the COP relationship are the principal reason for the limited improvement in system performance. Future studies at Pacific Northwest Laboratory (PNL) are planned to investigate methods to increase system performance using alternative system configurations and operations scenarios.
Development of a detailed core flow analysis code for prismatic fuel reactors
International Nuclear Information System (INIS)
Bennett, R.G.
1990-01-01
The development of a computer code for the analysis of the detailed flow of helium in prismatic fuel reactors is reported. The code, called BYPASS, solves, a finite difference control volume formulation of the compressible, steady state fluid flow in highly cross-connected flow paths typical of the Modular High-Temperature Gas Cooled Reactor (MHTGR). The discretization of the flow in a core region typically considers the main coolant flow paths, the bypass gap flow paths, and the crossflow connections between them. 16 refs., 5 figs
Heat and mass transfer and hydrodynamics in swirling flows (review)
Leont'ev, A. I.; Kuzma-Kichta, Yu. A.; Popov, I. A.
2017-02-01
Research results of Russian and foreign scientists of heat and mass transfer in whirling flows, swirling effect, superficial vortex generators, thermodynamics and hydrodynamics at micro- and nanoscales, burning at swirl of the flow, and technologies and apparatuses with the use of whirling currents for industry and power generation were presented and discussed at the "Heat and Mass Transfer in Whirling Currents" 5th International Conference. The choice of rational forms of the equipment flow parts when using whirling and swirling flows to increase efficiency of the heat-power equipment and of flow regimes and burning on the basis of deep study of the flow and heat transfer local parameters was set as the main research prospect. In this regard, there is noticeable progress in research methods of whirling and swirling flows. The number of computational treatments of swirling flows' local parameters has been increased. Development and advancement of the up to date computing models and national productivity software are very important for this process. All experimental works are carried out with up to date research methods of the local thermoshydraulic parameters, which enable one to reveal physical mechanisms of processes: PIV and LIV visualization techniques, high-speed and infrared photography, high speed registration of parameters of high-speed processes, etc. There is a problem of improvement of researchers' professional skills in the field of fluid mechanics to set adequately mathematics and physics problems of aerohydrodynamics for whirling and swirling flows and numerical and pilot investigations. It has been pointed out that issues of improvement of the cooling system and thermal protection effectiveness of heat-power and heat-transfer equipment units are still actual. It can be solved successfully using whirling and swirling flows as simple low power consumption exposing on the flow method and heat transfer augmentation.
Thermodynamic efficiency of information and heat flow
International Nuclear Information System (INIS)
Allahverdyan, Armen E; Janzing, Dominik; Mahler, Guenter
2009-01-01
A basic task of information processing is information transfer (flow). Here we study a pair of Brownian particles each coupled to a thermal bath at temperatures T 1 and T 2 . The information flow in such a system is defined via the time-shifted mutual information. The information flow nullifies at equilibrium, and its efficiency is defined as the ratio of the flow to the total entropy production in the system. For a stationary state the information flows from higher to lower temperatures, and its efficiency is bounded from above by (max[T 1 ,T 2 ])/(|T 1 −T 2 |). This upper bound is imposed by the second law and it quantifies the thermodynamic cost for information flow in the present class of systems. It can be reached in the adiabatic situation, where the particles have widely different characteristic times. The efficiency of heat flow—defined as the heat flow over the total amount of dissipated heat—is limited from above by the same factor. There is a complementarity between heat and information flow: the set-up which is most efficient for the former is the least efficient for the latter and vice versa. The above bound for the efficiency can be (transiently) overcome in certain non-stationary situations, but the efficiency is still limited from above. We study yet another measure of information processing (transfer entropy) proposed in the literature. Though this measure does not require any thermodynamic cost, the information flow and transfer entropy are shown to be intimately related for stationary states
Heat transfer critical conditions in two-plase flow
International Nuclear Information System (INIS)
Assis, M.C.V. de.
1980-02-01
The critical heat flux for forced-convection flow of water inside an uniformly heated circular channel is analysed, taking into account several flow patterns usually met in this type of investigation. Comments about nomenclature, experimental methods and influence of operational parameters used in the description of this phenomenon are made. The experimental results from 187 tests of critical heat flux at low pressure are presented. One empirical correlation between the critical heat flux and the independent parameters, was developed. Some correlations developed in other laboratories in the same range of parameters are mentioned and compared with present one. (Author) [pt
Small scale changes of geochemistry and flow field due to transient heat storage in aquifers
Bauer, S.; Boockmeyer, A.; Li, D.; Beyer, C.
2013-12-01
Heat exchangers in the subsurface are increasingly installed for transient heat storage due to the need of heating or cooling of buildings as well as the interim storage of heat to compensate for the temporally fluctuating energy production by wind or solar energy. For heat storage to be efficient, high temperatures must be achieved in the subsurface. Significant temporal changes of the soil and groundwater temperatures however effect both the local flow field by temperature dependent fluid parameters as well as reactive mass transport through temperature dependent diffusion coefficients, geochemical reaction rates and mineral equilibria. As the use of heat storage will be concentrated in urban areas, the use of the subsurface for (drinking) water supply and heat storage will typically coincide and a reliable prognosis of the processes occurring is needed. In the present work, the effects of a temporal variation of the groundwater temperature, as induced by a local heat exchanger introduced into a groundwater aquifer, are studied. For this purpose, the coupled non-isothermal groundwater flow, heat transport and reactive mass transport is simulated in the near filed of such a heat exchanger. By explicitly discretizing and incorporating the borehole, the borehole cementation and the heat exchanger tubes, a realistic geometrical and process representation is obtained. The numerical simulation code OpenGeoSys is used in this work, which incorporates the required processes of coupled groundwater flow, heat and mass transport as well as temperature dependent geochemistry. Due to the use of a Finite Element Method, a close representation of the geometric effects can be achieved. Synthetic scenario simulations for typical settings of salt water formations in northern Germany are used to investigate the geochemical effects arising from a high temperature heat storage by quantifying changes in groundwater chemistry and overall reaction rates. This work presents the
Prediction of a Heat Transfer to CO{sub 2} Flowing in an Upward Path at a Supercritical Pressure
Energy Technology Data Exchange (ETDEWEB)
Cho, Bong Hyun; Kim, Young In; Bae, Yoon Yeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)
2009-09-15
This study was performed to evaluate the prediction capability of a commercial CFD code and to investigate the effects of different geometries such as a 4.4 mm tube and an 8/10 mm annular channel on the detailed flow structures. A numerical simulation was performed for the conditions, at which the experimental data was produced by the test facility SPHINX. A 2-dimensional axisymmetric steady flow was assumed for computational simplicity. The RNG k-epsilon turbulence model (RNG) with an enhanced wall treatment option, SST k-omega (SST) and low Reynolds Abid turbulence model (ABD) were employed and the numerical predictions were compared with the experimental data generated from the experiment. The effects of the geometry on heat transfer were investigated. The flow and temperature fields were also examined in order to investigate the mechanism of heat transfer near the wall. The local heat transfer coefficient predicted by the RNG model is very close to the measurement result for the tube. In contrast, the local heat transfer coefficient predicted by the SST and ABD models is closer to the measurement for the annular channel
Transient convective heat transfer to laminar flow from a flat plate with constant heat capacity
International Nuclear Information System (INIS)
Hanawa, Juichi
1980-01-01
Most basic transient heat transfer problem is the transient response characteristics of forced convection heat transfer in the flow along a flat plate or in a tube. In case of the laminar flow along a flat plate, the profile method using steady temperature distribution has been mostly adopted, but its propriety has not been clarified yet. About the unsteady heat transfer in the laminar flow along a flat plate, the analysis or experiment evaluating the heat capacity of the flat plate exactly was never carried out. The purpose of this study is to determine by numerical calculation the unsteady characteristics of the boundary layer in laminar flow and to confirm them by experiment concerning the unsteady heat transfer when a flat plate with a certain heat capacity is placed in parallel in uniform flow and given a certain quantity of heat generation suddenly. The basic equation and the solution are given, and the method of numerical calculation and the result are explained. The experimental setup and method, and the experimental results are shown. Both results were in good agreement, and the response of wall temperature, the response of Nusselt number and the change of temperature distribution in course of time were able to be determined by applying Laplace transformation and numerical Laplace inverse transformation to the equation. (Kako, I.)
Heat flow at the Platanares, Honduras, geothermal site
Meert, Joseph G.; Smith, Douglas L.
1991-03-01
Three boreholes, PLTG-1, PLTG-2 and PLTG-3, were drilled in the Platanares, Honduras geothermal system to evaluate the geothermal energy potential of the site. The maximum reservoir temperature was previously estimated at 225-240°C using various types of chemical and isotopic geothermometry. Geothermal gradients of 139-239°C/km, calculated from two segments of the temperature-depth profile for borehole PLTG-2, were used to project a minimum depth to the geothermal reservoir of 1.2-1.7 km. Borehole PLTG-1 exhibited an erratic temperature distribution attributed to fluid movement through a series of isolated horizontal and subhorizontal fractures. The maximum measured temperature in borehole PLTG-1 was 150.4°C, and in PLTG-2 the maximum measured temperature was 104.3°C. PLTG-3 was drilled after this study and the maximum recorded temperature of 165°C is similar to the temperature encountered in PLTG-1. Heat flow values of 392 mWm -2 and 266 mWm -2 represent the first directly-measured heat flow values for Honduras and northen Central America. Radioactive heat generation, based on gamma-ray analyses of uranium, thorium and potassium in five core samples, is less than 2.0 μWm -3 and does not appear to be a major source of the high heat flow. Several authors have proposed a variety of extensional tectonic environments for western Honduras and these heat flow values, along with published estimates of heat flow, are supportive of this type of tectonic regime.
Energy Technology Data Exchange (ETDEWEB)
Austegard, Anders
1997-12-31
This thesis consists of two parts. Part 1: Experimental and numerical study of jetfire stop, and Part 2: Experimental and numerical study of a new kind of shell and tube heat exchanger with helical flow on shell side. Part 1 describes the development of the model for simulation of the temperature development through Viking jetfirestop. A simulation program is developed that calculates the temperature development through Viking jetfirestop. In the development of the model, measurements of reaction energy, pyrolysis and heat conductivity at low temperatures are made. The conductivity at higher temperatures and when pyrolysis reactions are going on is estimated experimentally and by numerical calculations. Full-scale jet fire test and small-scale xenon lamp experiments are made to test the simulation model. Part 2 contains the development of a model that simulate the fluid flow and heat transfer in a helical flow shell and tube heat exchanger. It consists of the development of a porosity model and a model for pressure drop and heat transfer as well as experiments in non-standard tube layouts. Results from the simulation program are compared with experiments on a helical flow shell and tube heat exchanger. This is a separate appendix volume, including computer codes and simulated results. 316 figs., 11 tabs.
Heat transfer enhancement in cross-flow heat exchanger using vortex generator
International Nuclear Information System (INIS)
Yoo, S. Y.; Kwon, H. K.; Kim, B. C.; Park, D. S.; Lee, S. S.
2003-01-01
Fouling is very serious problem in heat exchanger because it rapidly deteriorates the performance of heat exchanger. Cross-flow heat exchanger with vortex generators is developed, which enhance heat transfer and reduce fouling. In the present heat exchanger, shell and baffle are removed from the conventional shell-and-tube heat exchanger. The naphthalene sublimation technique is employed to measure the local heat transfer coefficients. The experiments are performed for single circular tube, staggered array tube bank and in-line array tube bank with and without vortex generators. Local and average Nusselt numbers of single tube and tube bank with vortex generator are investigated and compared to those of without vortex generator
Thermal performance modeling of cross-flow heat exchangers
Cabezas-Gómez, Luben; Saíz-Jabardo, José Maria
2014-01-01
This monograph introduces a numerical computational methodology for thermal performance modeling of cross-flow heat exchangers, with applications in chemical, refrigeration and automobile industries. This methodology allows obtaining effectiveness-number of transfer units (e-NTU) data and has been used for simulating several standard and complex flow arrangements configurations of cross-flow heat exchangers. Simulated results have been validated through comparisons with results from available exact and approximate analytical solutions. Very accurate results have been obtained over wide ranges
Enhancing heat transfer in microchannel heat sinks using converging flow passages
International Nuclear Information System (INIS)
Dehghan, Maziar; Daneshipour, Mahdi; Valipour, Mohammad Sadegh; Rafee, Roohollah; Saedodin, Seyfolah
2015-01-01
Highlights: • The fluid flow and conjugate heat transfer in microchannel heat sinks are studied. • The Poiseuille and Nusselt numbers are presented for width-tapered MCHS. • Converging walls are found to enhance the thermal performance of MCHS. • The optimum performance of MCHS for fixed inlet and outlet pressures is discussed. • For the optimum configuration, the pumping power is reduced up to 75%. - Abstract: Constrained fluid flow and conjugate heat transfer in microchannel heat sinks (MCHS) with converging channels are investigated using the finite volume method (FVM) in the laminar regime. The maximum pressure of the MCHS loop is assumed to be limited due to constructional or operational conditions. Results show that the Poiseuille number increases with increased tapering, while the required pumping power decreases. Meanwhile, the Nusselt number increases with tapering as well as the convection heat transfer coefficient. The MCHS having the optimum heat transfer performance is found to have a width-tapered ratio equal to 0.5. For this tapering configuration and at the maximum pressure constraint of 3000 Pa, the pumping power reduces by a factor of 4 while the overall heat removal rate is kept fixed in comparison with a straight channel
An experimental investigation of turbulent flow heat transfer through ...
African Journals Online (AJOL)
An experimental investigation has been carried out to study the turbulent flow heat transfer and to determine the pressure drop characteristics of air, flowing through a tube with insert. An insert of special geometry is used inside the tube. The test section is electrically heated, and air is allowed to flow as the working fluid ...
Geothermal heat exchanger with coaxial flow of fluids
Directory of Open Access Journals (Sweden)
Pejić Dragan M.
2005-01-01
Full Text Available The paper deals with a heat exchanger with coaxial flow. Two coaxial pipes of the secondary part were placed directly into a geothermal boring in such a way that geothermal water flows around the outer pipe. Starting from the energy balance of the exchanger formed in this way and the assumption of a study-state operating regime, a mathematical model was formulated. On the basis of the model, the secondary circle output temperature was determined as a function of the exchanger geometry, the coefficient of heat passing through the heat exchange areas, the average mass isobaric specific heats of fluid and mass flows. The input temperature of the exchanger secondary circle and the temperature of the geothermal water at the exit of the boring were taken as known values. Also, an analysis of changes in certain factors influencing the secondary water temperature was carried out. The parameters (flow temperature of the deep boring B-4 in Sijarinska Spa, Serbia were used. The theoretical results obtained indicate the great potential of this boring and the possible application of such an exchanger.
Directory of Open Access Journals (Sweden)
Yasuhisa Shinmoto
2017-11-01
Full Text Available The use of immiscible liquids for cooling of surfaces with high heat generation density is proposed based on the experimental verification of its superior cooling characteristics in fundamental systems of pool boiling and flow boiling in a tube. For the purpose of practical applications, however, heat transfer characteristics due to flow boiling in narrow rectangular channels with different small gap sizes need to be investigated. The immiscible liquids employed here are FC72 and water, and the gap size is varied as 2, 1, and 0.5 mm between parallel rectangular plates of 30 mm × 175 mm, where one plate is heated. To evaluate the effect of gap size, the heat transfer characteristics are compared at the same inlet velocity. The generation of large flattened bubbles in a narrow gap results in two opposite trends of the heat transfer enhancement due to thin liquid film evaporation and of the deterioration due to the extension of dry patch in the liquid film. The situation is the same as that observed for pure liquids. The latter negative effect is emphasized for extremely small gap sizes if the flow rate ratio of more-volatile liquid to the total is not reduced. The addition of small flow rate of less-volatile liquid can increase the critical heat flux (CHF of pure more-volatile liquid, while the surface temperature increases at the same time and assume the values between those for more-volatile and less-volatile liquids. By the selection of small flow rate ratio of more-volatile liquid, the surface temperature of pure less-volatile liquid can be decreased without reducing high CHF inherent in the less-volatile liquid employed. The trend of heat transfer characteristics for flow boiling of immiscible mixtures in narrow channels is more sensitive to the composition compared to the flow boiling in a round tube.
Flow and Heat Transfer in Cooling Microchannels with Phase-Change
Energy Technology Data Exchange (ETDEWEB)
Peles, Y P; Yarin, L P; Hetsroni, G [Technion, Israel Institute of Technology, Haifa (Israel) Faculty of Engineering
1998-05-19
The subject of the present work is the parametrical investigation of hydrodynamic and thermal characteristics of laminar flow with phase-change in a heating microchannels. The study is based on the quasi-one-dimensional model of non-isothermal capillary flow. This model takes into account the evolution of flow, heating and evaporation of the liquid, as well as the influence of capillary, inertia, friction and gravity forces. The effect of various parameters (sizes of microchannel, initial temperature of cooling liquid, wall heat flux etc.) on hydrodynamic and thermal structures of the flow, the length of heating, evaporation and superheat regions is studied. Thc specific features of the phenomena is discussed.
Flow and Heat Transfer in Cooling Microchannels with Phase-Change
International Nuclear Information System (INIS)
Peles, Y.P.; Yarin, L.P.; Hetsroni, G.
1998-01-01
The subject of the present work is the parametrical investigation of hydrodynamic and thermal characteristics of laminar flow with phase-change in a heating microchannels. The study is based on the quasi-one-dimensional model of non-isothermal capillary flow. This model takes into account the evolution of flow, heating and evaporation of the liquid, as well as the influence of capillary, inertia, friction and gravity forces. The effect of various parameters (sizes of microchannel, initial temperature of cooling liquid, wall heat flux etc.) on hydrodynamic and thermal structures of the flow, the length of heating, evaporation and superheat regions is studied. Thc specific features of the phenomena is discussed
Towards Effective Intra-flow Network Coding in Software Defined Wireless Mesh Networks
Directory of Open Access Journals (Sweden)
Donghai Zhu
2016-01-01
Full Text Available Wireless Mesh Networks (WMNs have potential to provide convenient broadband wireless Internet access to mobile users.With the support of Software-Defined Networking (SDN paradigm that separates control plane and data plane, WMNs can be easily deployed and managed. In addition, by exploiting the broadcast nature of the wireless medium and the spatial diversity of multi-hop wireless networks, intra-flow network coding has shown a greater benefit in comparison with traditional routing paradigms in data transmission for WMNs. In this paper, we develop a novel OpenCoding protocol, which combines the SDN technique with intra-flow network coding for WMNs. Our developed protocol can simplify the deployment and management of the network and improve network performance. In OpenCoding, a controller that works on the control plane makes routing decisions for mesh routers and the hop-by-hop forwarding function is replaced by network coding functions in data plane. We analyze the overhead of OpenCoding. Through a simulation study, we show the effectiveness of the OpenCoding protocol in comparison with existing schemes. Our data shows that OpenCoding outperforms both traditional routing and intra-flow network coding schemes.
Multiphase integral reacting flow computer code (ICOMFLO): User`s guide
Energy Technology Data Exchange (ETDEWEB)
Chang, S.L.; Lottes, S.A.; Petrick, M.
1997-11-01
A copyrighted computational fluid dynamics computer code, ICOMFLO, has been developed for the simulation of multiphase reacting flows. The code solves conservation equations for gaseous species and droplets (or solid particles) of various sizes. General conservation laws, expressed by elliptic type partial differential equations, are used in conjunction with rate equations governing the mass, momentum, enthalpy, species, turbulent kinetic energy, and turbulent dissipation. Associated phenomenological submodels of the code include integral combustion, two parameter turbulence, particle evaporation, and interfacial submodels. A newly developed integral combustion submodel replacing an Arrhenius type differential reaction submodel has been implemented to improve numerical convergence and enhance numerical stability. A two parameter turbulence submodel is modified for both gas and solid phases. An evaporation submodel treats not only droplet evaporation but size dispersion. Interfacial submodels use correlations to model interfacial momentum and energy transfer. The ICOMFLO code solves the governing equations in three steps. First, a staggered grid system is constructed in the flow domain. The staggered grid system defines gas velocity components on the surfaces of a control volume, while the other flow properties are defined at the volume center. A blocked cell technique is used to handle complex geometry. Then, the partial differential equations are integrated over each control volume and transformed into discrete difference equations. Finally, the difference equations are solved iteratively by using a modified SIMPLER algorithm. The results of the solution include gas flow properties (pressure, temperature, density, species concentration, velocity, and turbulence parameters) and particle flow properties (number density, temperature, velocity, and void fraction). The code has been used in many engineering applications, such as coal-fired combustors, air
A novel compact heat exchanger using gap flow mechanism.
Liang, J S; Zhang, Y; Wang, D Z; Luo, T P; Ren, T Q
2015-02-01
A novel, compact gap-flow heat exchanger (GFHE) using heat-transfer fluid (HTF) was developed in this paper. The detail design of the GFHE coaxial structure which forms the annular gap passage for HTF is presented. Computational fluid dynamics simulations were introduced into the design to determine the impacts of the gap width and the HTF flow rate on the GFHE performance. A comparative study on the GFHE heating rate, with the gap widths ranged from 0.1 to 1.0 mm and the HTF flow rates ranged from 100 to 500 ml/min, was carried out. Results show that a narrower gap passage and a higher HTF flow rate can yield a higher average heating rate in GFHE. However, considering the compromise between the GFHE heating rate and the HTF pressure drop along the gap, a 0.4 mm gap width is preferred. A testing loop was also set up to experimentally evaluate the GFHE capability. The testing results show that, by using 0.4 mm gap width and 500 ml/min HTF flow rate, the maximum heating rate in the working chamber of the as-made GFHE can reach 18 °C/min, and the average temperature change rates in the heating and cooling processes of the thermal cycle test were recorded as 6.5 and 5.4 °C/min, respectively. These temperature change rates can well satisfy the standard of IEC 60068-2-14:2009 and show that the GFHE developed in this work has sufficient heat exchange capacity and can be used as an ideal compact heat exchanger in small volume desktop thermal fatigue test apparatus.
International Nuclear Information System (INIS)
Peyghambarzadeh, S.M.; Sarafraz, M.M.; Vaeli, N.; Ameri, E.; Vatani, A.; Jamialahmadi, M.
2013-01-01
Highlights: ► The cooling performance of water and n-heptane is compared during subcooled flow boiling. ► Although n-heptane leaves the heat exchanger warmer it has a lower heat transfer coefficient. ► Flow rate, heat flux and degree of subcooling have direct effect on heat transfer coefficient. ► The predictions of some correlations are evaluated against experimental data. - Abstract: In this research, subcooled flow boiling heat transfer coefficients of pure n-heptane and distilled water at different operating conditions have been experimentally measured and compared. The heat exchanger consisted of vertical annulus which is heated from the inner cylindrical heater with variable heat flux (less than 140 kW/m 2 ). Heat flux is varied so that two different flow regimes from single phase forced convection to nucleate boiling condition are created. Meanwhile, liquid flow rate is changed in the range of 2.5 × 10 −5 –5.8 × 10 −5 m 3 /s to create laminar up to transition flow regimes. Three subcooling levels including 10, 20 and 30 °C are also considered. Experimental results demonstrated that subcooled flow boiling heat transfer coefficient increases when higher heat flux, higher liquid flow rate and greater subcooling level are applied. Furthermore, influence of the operating conditions on the bubbles generation on the heat transfer surface is also discussed. It is also shown that water is better cooling fluid in comparison with n-heptane
Two-phase flow instabilities in a silicon microchannels heat sink
International Nuclear Information System (INIS)
Bogojevic, D.; Sefiane, K.; Walton, A.J.; Lin, H.; Cummins, G.
2009-01-01
Two-phase flow instabilities are highly undesirable in microchannels-based heat sinks as they can lead to temperature oscillations with high amplitudes, premature critical heat flux and mechanical vibrations. This work is an experimental study of boiling instabilities in a microchannel silicon heat sink with 40 parallel rectangular microchannels, having a length of 15 mm and a hydraulic diameter of 194 μm. A series of experiments have been carried out to investigate pressure and temperature oscillations during the flow boiling instabilities under uniform heating, using water as a cooling liquid. Thin nickel film thermometers, integrated on the back side of a heat sink with microchannels, were used in order to obtain a better insight related to temperature fluctuations caused by two-phase flow instabilities. Flow regime maps are presented for two inlet water temperatures, showing stable and unstable flow regimes. It was observed that boiling leads to asymmetrical flow distribution within microchannels that result in high temperature non-uniformity and the simultaneously existence of different flow regimes along the transverse direction. Two types of two-phase flow instabilities with appreciable pressure and temperature fluctuations were observed, that depended on the heat to mass flux ratio and inlet water temperature. These were high amplitude/low frequency and low amplitude/high frequency instabilities. High speed camera imaging, performed simultaneously with pressure and temperature measurements, showed that inlet/outlet pressure and the temperature fluctuations existed due to alternation between liquid/two-phase/vapour flows. It was also determined that the inlet water subcooling condition affects the magnitudes of the temperature oscillations in two-phase flow instabilities and flow distribution within the microchannels.
Micro-channel convective boiling heat transfer with flow instabilities
International Nuclear Information System (INIS)
Consolini, L.; Thome, J.R.
2009-01-01
Flow boiling heat transfer in micro-channels has attracted much interest in the past decade, and is currently a strong candidate for high performance compact heat sinks, such as those required in electronics systems, automobile air conditioning units, micro-reactors, fuel cells, etc. Currently the literature presents numerous experimental studies on two-phase heat transfer in micro-channels, providing an extensive database that covers many different fluids and operating conditions. Among the noteworthy elements that have been reported in previous studies, is the sensitivity of micro-channel evaporators to oscillatory two-phase instabilities. These periodic fluctuations in flow and pressure drop either result from the presence of upstream compressibility, or are simply due to the interaction among parallel channels in multi-port systems. An oscillating flow presents singular characteristics that are expected to produce an effect on the local heat transfer mechanisms, and thus on the estimation of the two-phase heat transfer coefficients. The present investigation illustrates results for flow boiling of refrigerants R-134a, R-236fa, and R-245fa in a 510 μm circular micro-channel, exposed to various degrees of oscillatory compressible volume instabilities. The data describe the main features of the fluctuations in the temperatures of the heated wall and fluid, and draw attention to the differences in the measured unstable time-averaged heat transfer coefficients with respect to those for stable flow boiling. (author)
Micro-channel convective boiling heat transfer with flow instabilities
Energy Technology Data Exchange (ETDEWEB)
Consolini, L.; Thome, J.R. [Ecole Polytechnique Federale de Lausanne (Switzerland). Lab. de Transfert de Chaleur et de Masse], e-mail: lorenzo.consolini@epfl.ch, e-mail: john.thome@epfl.ch
2009-07-01
Flow boiling heat transfer in micro-channels has attracted much interest in the past decade, and is currently a strong candidate for high performance compact heat sinks, such as those required in electronics systems, automobile air conditioning units, micro-reactors, fuel cells, etc. Currently the literature presents numerous experimental studies on two-phase heat transfer in micro-channels, providing an extensive database that covers many different fluids and operating conditions. Among the noteworthy elements that have been reported in previous studies, is the sensitivity of micro-channel evaporators to oscillatory two-phase instabilities. These periodic fluctuations in flow and pressure drop either result from the presence of upstream compressibility, or are simply due to the interaction among parallel channels in multi-port systems. An oscillating flow presents singular characteristics that are expected to produce an effect on the local heat transfer mechanisms, and thus on the estimation of the two-phase heat transfer coefficients. The present investigation illustrates results for flow boiling of refrigerants R-134a, R-236fa, and R-245fa in a 510 {mu}m circular micro-channel, exposed to various degrees of oscillatory compressible volume instabilities. The data describe the main features of the fluctuations in the temperatures of the heated wall and fluid, and draw attention to the differences in the measured unstable time-averaged heat transfer coefficients with respect to those for stable flow boiling. (author)
Visualisation of heat transfer in unsteady laminar flows
Speetjens, M.F.M.; Steenhoven, van A.A.
2011-01-01
Heat transfer in fluid flows traditionally is examined in terms of temperature fields and heat-transfer coefficients. However, heat transfer may alternatively be considered as the transport of thermal energy by the total convective-conductive heat flux in a way analogous to the transport of fluid by
Critical heat flux of forced flow boiling in a narrow one-side heated rectangular flow channel
Energy Technology Data Exchange (ETDEWEB)
Limin, Zheng [Shanghai Nuclear Engineering Research and Design Inst., SH (China); Iguchi, Tadashi; Kureta, Masatoshi; Akimoto, Hajime
1997-08-01
The present work deals with the critical heat flux (CHF) under subcooled flow boiling in a narrow one-side uniformly heated rectangular flow channel. The range of interest of parameters such as pressure, flow velocity and subcooling is around 0.1 MPa, 5-15 ms{sup -1} and 50degC, respectively. The rectangular flow channel used is 50 mm long, 12 mm in width and 0.2 to 3 mm in height. Test conditions were selected by combination of the following parameters: Gap=0.2-3.0 mm (D{sub hy}=0.3934-4.8 mm); flow length, 50.0 mm; water mass flux, 4.94-14.82 Mgm{sup -2}s{sup -1} (water flow velocity, 5-15 ms{sup -1}); exit pressure, 0.1 MPa; inlet temperature, 50degC, inlet coolant subcooling, 50degC. Over 40 CHF stable data points were obtained. CHF increased with the gap and flow velocity in a non-linear fashion. HTC increased with flow velocity and decreasing gap. Based on the experimental results, an empirical correlation was developed, indicating the dependence of CHF on the gap and flow velocity. All of data points predicted within {+-}18% error band for the present experimental data. On the other hand, another similitude-based correlation was also developed, indicating the dependence of Boiling number (Bo) on Reynolds number (Re) and the variable of Gap/La, where La is a characteristic length known as Laplace capillary constant. For the limited present experimental data, all of data points were predicted within {+-}16%. (author)
Critical heat flux of forced flow boiling in a narrow one-side heated rectangular flow channel
International Nuclear Information System (INIS)
Zheng Limin; Iguchi, Tadashi; Kureta, Masatoshi; Akimoto, Hajime.
1997-08-01
The present work deals with the critical heat flux (CHF) under subcooled flow boiling in a narrow one-side uniformly heated rectangular flow channel. The range of interest of parameters such as pressure, flow velocity and subcooling is around 0.1 MPa, 5-15 ms -1 and 50degC, respectively. The rectangular flow channel used is 50 mm long, 12 mm in width and 0.2 to 3 mm in height. Test conditions were selected by combination of the following parameters: Gap=0.2-3.0 mm (D hy =0.3934-4.8 mm); flow length, 50.0 mm; water mass flux, 4.94-14.82 Mgm -2 s -1 (water flow velocity, 5-15 ms -1 ); exit pressure, 0.1 MPa; inlet temperature, 50degC, inlet coolant subcooling, 50degC. Over 40 CHF stable data points were obtained. CHF increased with the gap and flow velocity in a non-linear fashion. HTC increased with flow velocity and decreasing gap. Based on the experimental results, an empirical correlation was developed, indicating the dependence of CHF on the gap and flow velocity. All of data points predicted within ±18% error band for the present experimental data. On the other hand, another similitude-based correlation was also developed, indicating the dependence of Boiling number (Bo) on Reynolds number (Re) and the variable of Gap/La, where La is a characteristic length known as Laplace capillary constant. For the limited present experimental data, all of data points were predicted within ±16%. (author)
A Prototype Flux-Plate Heat-Flow Sensor for Venus Surface Heat-Flow Determinations
Morgan, Paul; Reyes, Celso; Smrekar, Suzanne E.
2005-01-01
Venus is the most Earth-like planet in the Solar System in terms of size, and the densities of the two planets are almost identical when selfcompression of the two planets is taken into account. Venus is the closest planet to Earth, and the simplest interpretation of their similar densities is that their bulk compositions are almost identical. Models of the thermal evolution of Venus predict interior temperatures very similar to those indicated for the regions of Earth subject to solid-state convection, but even global analyses of the coarse Pioneer Venus elevation data suggest Venus does not lose heat by the same primary heat loss mechanism as Earth, i.e., seafloor spreading. The comparative paucity of impact craters on Venus has been interpreted as evidence for relatively recent resurfacing of the planet associated with widespread volcanic and tectonic activity. The difference in the gross tectonic styles of Venus and Earth, and the origins of some of the enigmatic volcano-tectonic features on Venus, such as the coronae, appear to be intrinsically related to Venus heat loss mechanism(s). An important parameter in understanding Venus geological evolution, therefore, is its present surface heat flow. Before the complications of survival in the hostile Venus surface environment were tackled, a prototype fluxplate heat-flow sensor was built and tested for use under synthetic stable terrestrial surface conditions. The design parameters for this prototype were that it should operate on a conforming (sand) surface, with a small, self-contained power and recording system, capable of operating without servicing for at least several days. The precision and accuracy of the system should be < 5 mW/sq m. Additional information is included in the original extended abstract.
Numerical Simulation of a Coolant Flow and Heat Transfer in a Pebble Bed Reactor
International Nuclear Information System (INIS)
In, Wang-Kee; Kim, Min-Hwan; Lee, Won-Jae
2008-01-01
Pebble Bed Reactor(PBR) is one of the very high temperature gas cooled reactors(VHTR) which have been reviewed in the Generation IV International Forum as potential sources for future energy needs, particularly for a hydrogen production. The pebble bed modular reactor(PBMR) exhibits inherent safety features due to the low power density and the large amount of graphite present in the core. PBR uses coated fuel particles(TRISO) embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the PBR core during a reactor operation and the coolant flows around randomly distributed spheres. For the reliable operation and the safety of the PBR, it is important to understand the coolant flow structure and the fuel pebble temperature in the PBR core. There have been few experimental and numerical studies to investigate the fluid and heat transfer phenomena in the PBR core. The objective of this paper is to predict the fluid and heat transfer in the PBR core. The computational fluid dynamics (CFD) code, STAR-CCM+(V2.08) is used to perform the CFD analysis using the design data for the PBMR400
Nuclear research reactor IEA-R1 heat exchanger inlet nozzle flow - a preliminary study
International Nuclear Information System (INIS)
Angelo, Gabriel; Andrade, Delvonei Alves de; Fainer, Gerson; Angelo, Edvaldo
2009-01-01
As a computational fluid mechanics training task, a preliminary model was developed. ANSYS-CFX R code was used in order to study the flow at the inlet nozzle of the heat exchanger of the primary circuit of the nuclear research reactor IEA-R1. The geometry of the inlet nozzle is basically compounded by a cylinder and two radial rings which are welded on the shell. When doing so there is an offset between the holes through the shell and the inlet nozzle. Since it is not standardized by TEMA, the inlet nozzle was chosen for a preliminary study of the flow. Results for the proposed model are presented and discussed. (author)
Validation of a CFD code for Unsteady Flows with cyclic boundary Conditions
International Nuclear Information System (INIS)
Kim, Jong-Tae; Kim, Sang-Baik; Lee, Won-Jae
2006-01-01
Currently Lilac code is under development to analyze thermo-hydraulics of a high-temperature gas-cooled reactor (GCR). Interesting thermo-hydraulic phenomena in a nuclear reactor are usually unsteady and turbulent. The analysis of the unsteady flows by using a three dimension CFD code is time-consuming if the flow domain is very large. Hopefully, flow domains commonly encountered in the nuclear thermo-hydraulics is periodic. So it is better to use the geometrical characteristics in order to reduce the computational resources. To get the benefits from reducing the computation domains especially for the calculations of unsteady flows, the cyclic boundary conditions are implemented in the parallelized CFD code LILAC. In this study, the parallelized cyclic boundary conditions are validated by solving unsteady laminar and turbulent flows past a circular cylinder
Energy Technology Data Exchange (ETDEWEB)
Kasinathan, N.; Rajakumar, A.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)
1995-09-01
Post shutdown decay heat removal is an important safety requirement in any nuclear system. In order to improve the reliability of this function, Liquid metal (sodium) cooled fast breeder reactors (LMFBR) are equipped with redundant hot pool dipped immersion coolers connected to natural draught air cooled heat exchangers through intermediate sodium circuits. During decay heat removal, flow through the core, immersion cooler primary side and in the intermediate sodium circuits are also through natural convection. In order to establish the viability and validate computer codes used in making predictions, a 1:20 scale experimental model called RAMONA with water as coolant has been built and experimental simulation of decay heat removal situation has been performed at KfK Karlsruhe. Results of two such experiments have been compiled and published as benchmarks. This paper brings out the results of the numerical simulation of one of the benchmark case through a 1D/2D coupled code system, DHDYN-1D/THYC-2D and the salient features of the comparisons. Brief description of the formulations of the codes are also included.
International Nuclear Information System (INIS)
Peletier, Mark A.; Redig, Frank; Vafayi, Kiamars
2014-01-01
We consider three one-dimensional continuous-time Markov processes on a lattice, each of which models the conduction of heat: the family of Brownian Energy Processes with parameter m (BEP(m)), a Generalized Brownian Energy Process, and the Kipnis-Marchioro-Presutti (KMP) process. The hydrodynamic limit of each of these three processes is a parabolic equation, the linear heat equation in the case of the BEP(m) and the KMP, and a nonlinear heat equation for the Generalized Brownian Energy Process with parameter a (GBEP(a)). We prove the hydrodynamic limit rigorously for the BEP(m), and give a formal derivation for the GBEP(a). We then formally derive the pathwise large-deviation rate functional for the empirical measure of the three processes. These rate functionals imply gradient-flow structures for the limiting linear and nonlinear heat equations. We contrast these gradient-flow structures with those for processes describing the diffusion of mass, most importantly the class of Wasserstein gradient-flow systems. The linear and nonlinear heat-equation gradient-flow structures are each driven by entropy terms of the form −log ρ; they involve dissipation or mobility terms of order ρ 2 for the linear heat equation, and a nonlinear function of ρ for the nonlinear heat equation
Flow film boiling heat transfer in water and Freon-113
International Nuclear Information System (INIS)
Liu, Qiusheng; Shiotsu, Masahiro; Sakurai, Akira
2002-01-01
Experimental apparatus and method for film boiling heat transfer measurement on a horizontal cylinder in forced flow of water and Freon-113 under pressurized and subcooled conditions were developed. The experiments of film boiling heat transfer from single horizontal cylinders with diameters ranging from 0.7 to 5 mm in saturated and subcooled water and Freon-113 flowing upward perpendicular to the cylinders were carried out for the flow velocities ranging from 0 to 1 m/s under system pressures ranging from 100 to 500 kPa. Liquid subcoolings ranged from 0 to 50 K, and the cylinder surface superheats were raised up to 800 K for water and 400 K for Freon-113. The film boiling heat transfer coefficients obtained were depended on surface superheats, flow velocities, liquid subcoolings, system pressures and cylinder diameters. The effects of these parameters were systematically investigated under wider ranges of experimental conditions. It was found that the heat transfer coefficients are higher for higher flow velocities, subcoolings, system pressures, and for smaller cylinder diameters. The observation results of film boiling phenomena were obtained by a high-speed video camera. A new correlation for subcooled flow film boiling heat transfer was derived by modifying authors' correlation for saturated flow film boiling heat transfer with authors' experimental data under wide subcooled conditions. (author)
Development and Verification of a Pilot Code based on Two-fluid Three-field Model
Energy Technology Data Exchange (ETDEWEB)
Hwang, Moon Kyu; Bae, S. W.; Lee, Y. J.; Chung, B. D.; Jeong, J. J.; Ha, K. S.; Kang, D. H
2006-09-15
In this study, a semi-implicit pilot code is developed for a one-dimensional channel flow as three-fields. The three fields are comprised of a gas, continuous liquid and entrained liquid fields. All the three fields are allowed to have their own velocities. The temperatures of the continuous liquid and the entrained liquid are, however, assumed to be equilibrium. The interphase phenomena include heat and mass transfer, as well as momentum transfer. The fluid/structure interaction, generally, include both heat and momentum transfer. Assuming adiabatic system, only momentum transfer is considered in this study, leaving the wall heat transfer for the future study. Using 10 conceptual problems, the basic pilot code has been verified. The results of the verification are summarized below: It was confirmed that the basic pilot code can simulate various flow conditions (such as single-phase liquid flow, bubbly flow, slug/churn turbulent flow, annular-mist flow, and single-phase vapor flow) and transitions of the flow conditions. The pilot code was programmed so that the source terms of the governing equations and numerical solution schemes can be easily tested. The mass and energy conservation was confirmed for single-phase liquid and single-phase vapor flows. It was confirmed that the inlet pressure and velocity boundary conditions work properly. It was confirmed that, for single- and two-phase flows, the velocity and temperature of non-existing phase are calculated as intended. Complete phase depletion which might occur during a phase change was found to adversely affect the code stability. A further study would be required to enhance code capability in this regard.
Program Computes Flows Of Fluids And Heat
Cullimore, Brent; Ring, Steven; Welch, Mark
1993-01-01
SINDA'85/FLUINT incorporates lumped-parameter-network and one-dimensional-flow mathematical models. System enables analysis of mutual influences of thermal and flow phenomena. Offers two finite-difference numerical solution techniques: forward-difference explicit approximation and Crank-Nicholson approximation. Enables simulation of nonuniform heating and facilitates mathematical modeling of thin-walled heat exchangers. Ability to model nonequilibrium behavior within two-phase volumes included. Recent changes in program improve modeling of real evaporator pumps and other capillary-assist evaporators. Written in FORTRAN 77.
Critical heat flux in flow boiling in microchannels
Saha, Sujoy Kumar
2015-01-01
This Brief concerns the important problem of critical heat flux in flow boiling in microchannels. A companion edition in the SpringerBrief Subseries on Thermal Engineering and Applied Science to “Heat Transfer and Pressure Drop in Flow Boiling in Microchannels,” by the same author team, this volume is idea for professionals, researchers, and graduate students concerned with electronic cooling.
Heat transfer measurements of the 1983 kilauea lava flow.
Hardee, H C
1983-10-07
Convective heat flow measurements of a basaltic lava flow were made during the 1983 eruption of Kilauea volcano in Hawaii. Eight field measurements of induced natural convection were made, giving heat flux values that ranged from 1.78 to 8.09 kilowatts per square meter at lava temperatures of 1088 and 1128 degrees Celsius, respectively. These field measurements of convective heat flux at subliquidus temperatures agree with previous laboratory measurements in furnace-melted samples of molten lava, and are useful for predicting heat transfer in magma bodies and for estimating heat extraction rates for magma energy.
International Nuclear Information System (INIS)
Laurien, E.
2012-01-01
Within the Generation IV International Forum the Supercritical Water Reactor is investigated. For its core design and safety analysis the efficient prediction of flow and heat transfer parameters such as the wall-shear stress and the heat-transfer coefficient for pipe and channel flows is needed. For circular pipe flows a numerical model based on the one-dimensional conservation equations of mass, momentum end energy in the radial direction is presented, referred to as a 'semi-analytical' method. An accurate, high-order numerical method is employed to evaluate previously derived analytical solutions of the governing equations. Flow turbulence is modeled using the algebraic approach of Prandtl/van-Karman, including a model for the buffer layer. The influence of wall roughness is taken into account by a new modified numerical damping function of the turbulence model. The thermo-hydraulic properties of water are implemented according to the international standard of 1997. This method has the potential to be used within a sub-channel analysis code and as wall-functions for CFD codes to predict the wall shear stress and the wall temperature. The present study presents a validation of the method with comparison of model results with experiments and multi-dimensional computational (CFD) studies in a wide range of flow parameters. The focus is laid on forced convection flows related to reactor design and near-design conditions. It is found, that the method can accurately predict the wall temperature even under deterioration conditions as they occur in the selected experiments (Yamagata el al. 1972 at 24.5 MPa, Ornatski et al. 1971 at 25.5 and Swenson et al. 1963 at 22.75 MPa). Comparison of the friction coefficient under high heat flux conditions including significant viscosity and density reductions near the wall with various correlations for the hydraulic resistance will be presented; the best agreement is achieve with the correlation of Pioro et al. 2004. It is
Flow boiling heat transfer at low liquid Reynolds number
International Nuclear Information System (INIS)
Weizhong Zhang; Takashi Hibiki; Kaichiro Mishima
2005-01-01
Full text of publication follows: In view of the significance of a heat transfer correlation of flow boiling at conditions of low liquid Reynolds number or liquid laminar flow, and very few existing correlations in principle suitable for such flow conditions, this study is aiming at developing a heat transfer correlation of flow boiling at low liquid Reynolds number conditions. The obtained results are as follows: 1. A new heat transfer correlation has been developed for saturated flow boiling at low liquid Reynolds number conditions based on superimposition of two boiling mechanisms, namely convective boiling and nucleate boiling. In the new correlation, two terms corresponding to the mechanisms of nucleate boiling and convective boiling are obtained from the pool boiling correlation by Forster and Zuber and the analytical annular flow model by Hewitt and Hall-Taylor, respectively. 2. An extensive database was collected for saturated flow boiling heat transfer at low liquid Reynolds number conditions, including data for different channels geometries (circular and rectangular), flow orientations (vertical and horizontal), and working fluids (water, R11, R12, R113). 3. An extensive comparison of the new correlation with the collected database shows that the new correlation works satisfactorily with the mean deviation of 16.6% for saturated flow boiling at low liquid Reynolds number conditions. 4. The detailed discussion reveals the similarity of the newly developed correlation for flow boiling at low liquid Reynolds number to the Chen correlation for flow boiling at high liquid Reynolds number. The Reynolds number factor F can be analytically deduced in this study. (authors)
Systematic heat flow measurements across the Wagner Basin, northern Gulf of California
Neumann, Florian; Negrete-Aranda, Raquel; Harris, Robert N.; Contreras, Juan; Sclater, John G.; González-Fernández, Antonio
2017-12-01
A primary control on the geodynamics of rifting is the thermal regime. To better understand the geodynamics of rifting in the northern Gulf of California we systematically measured heat-flow across the Wagner Basin, a tectonically active basin that lies near the southern terminus of the Cerro Prieto fault. The heat flow profile is 40 km long, has a nominal measurement spacing of ∼1 km, and is collocated with a seismic reflection profile. Heat flow measurements were made with a 6.5-m violin-bow probe. Although heat flow data were collected in shallow water, where there are significant temporal variations in bottom water temperature, we use CTD data collected over many years to correct our measurements to yield accurate values of heat flow. After correction for bottom water temperature, the mean and standard deviation of heat flow across the western, central, and eastern parts of the basin are 220 ± 60, 99 ± 14, 889 ± 419 mW m-2, respectively. Corrections for sedimentation would increase measured heat flow across the central part of basin by 40 to 60%. We interpret the relatively high heat flow and large variability on the western and eastern flanks in terms of upward fluid flow at depth below the seafloor, whereas the lower and more consistent values across the central part of the basin are suggestive of conductive heat transfer. Moreover, heat flow across the central basin is consistent with gabbroic underplating at a depth of 15 km and suggests that continental rupture here has not gone to completion.
International Nuclear Information System (INIS)
Borges, Eduardo M.; Sabundjian, Gaiane
2017-01-01
The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2. gamma code in Angra-2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 250cm 2 of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of Angra-2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of Angra-2 during the postulated accident. The results obtained for Angra-2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core. (author)
Gravity-driven flow and heat transfer in a spent nuclear fuel storage pool
International Nuclear Information System (INIS)
Gay, R.R.
1983-01-01
The GFLOW code analyzes a three-dimensional rectangular porous medium by dividing the porous medium into a number of nodes or cells specified by the user. The finite difference form of the fluid conservation equations is solved for each node by application of a modified ''marker and cell'' numerical technique. The existence of spent nuclear fuel in any node is modeled by using a porosity value less than unity in that node and by including a surface heat transfer term in the fluid energy equation. In addition, local pressure losses due to grid spaces or other planar flow obstructions can be modeled by local loss coefficients. Heat conduction in the fuel is simulated by a fast running implicit finite difference model of the fuel, gap, and clad regions of the fuel rod
Code requirements document: MODFLOW 2.1: A program for predicting moderator flow patterns
International Nuclear Information System (INIS)
Peterson, P.F.
1992-03-01
Sudden changes in the temperature of flowing liquids can result in transient buoyancy forces which strongly impact the flow hydrodynamics via flow stratification. These effects have been studied for the case of potential flow of stratified liquids to line sinks, but not for moderator flow in SRS reactors. Standard codes, such as TRAC and COMMIX, do not have the capability to capture the stratification effect, due to strong numerical diffusion which smears away the hot/cold fluid interface. A related problem with standard codes is the inability to track plumes injected into the liquid flow, again due to numerical diffusion. The combined effects of buoyant stratification and plume dispersion have been identified as being important in operation of the Supplementary Safety System which injects neutron-poison ink into SRS reactors to provide safe shutdown in the event of safety rod failure. The MODFLOW code discussed here provides transient moderator flow pattern information with stratification effects, and tracks the location of ink plumes in the reactor. The code, written in Fortran, is compiled for Macintosh II computers, and includes subroutines for interactive control and graphical output. Removing the graphics capabilities, the code can also be compiled on other computers. With graphics, in addition to the capability to perform safety related computations, MODFLOW also provides an easy tool for becoming familiar with flow distributions in SRS reactors
Miniaturized heat flux sensor for high enthalpy plasma flow characterization
International Nuclear Information System (INIS)
Gardarein, Jean-Laurent; Battaglia, Jean-Luc; Lohlec, Stefan; Jullien, Pierre; Van Ootegemd, Bruno; Couzie, Jacques; Lasserre, Jean-Pierre
2013-01-01
An improved miniaturized heat flux sensor is presented aiming at measuring extreme heat fluxes of plasma wind tunnel flows. The sensor concept is based on an in-depth thermocouple measurement with a miniaturized design and an advanced calibration approach. Moreover, a better spatial estimation of the heat flux profile along the flow cross section is realized with this improved small sensor design. Based on the linearity assumption, the heat flux is determined using the impulse response of the sensor relating the heat flux to the temperature of the embedded thermocouple. The non-integer system identification (NISI) procedure is applied that allows a calculation of the impulse response from transient calibration measurements with a known heat flux of a laser source. The results show that the new sensor leads to radially highly resolved heat flux measurement for a flow with only a few centimetres in diameter, the so far not understood non-symmetric heat flux profiles do not occur with the new sensor design. It is shown that this former effect is not a physical effect of the flow, but a drawback of the classical sensor design. (authors)
International Nuclear Information System (INIS)
Hussein, H.M.S.
2007-01-01
In this work, a wickless heat pipes flat plate solar collector with a cross flow heat exchanger was investigated theoretically and experimentally under the meteorological conditions of Cairo, Egypt. The author's earlier simulation program of wickless heat pipes flat plate solar water heaters was modified to be valid for the present type of wickless heat pipes solar collector by including the solution of the dimensionless governing equations of the present analysis. For verifying the modified simulation program, a wickless heat pipes flat plate solar collector with a cross flow heat exchanger was designed, constructed, and tested at different meteorological conditions and operating parameters. These parameters include different cooling water mass flow rates and different inlet cooling water temperatures. The comparison between the experimental results and their corresponding simulated ones showed considerable agreement. Under different climatic conditions, the experimental and theoretical results showed that the optimal mass flow rate is very close to the ASHRAE standard mass flow rate for testing conventional flat plate solar collectors. Also, the experimental and theoretical results indicated that the number of wickless heat pipes has a significant effect on the collector efficiency
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
Post-Dryout Heat Transfer to a Refrigerant Flowing in Horizontal Evaporator Tubes
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.
Tests of the TRAC code against known analytical solutions for stratified flow
International Nuclear Information System (INIS)
Black, P.S.; Leslie, D.C.; Hewitt, G.F.
1987-01-01
The area averaged equations for gas-liquid flow are briefly summarized and related, for the specific case of stratified flow, to the shallow water equations commonly used in hydraulics. These equations are then compared to the equations used in TRAC-PF/MOD1 and are shown to differ in their treatment of the gravity head terms. A modification of the TRAC code is therefore necessary to bring it into line with established shallow water theory. The corrected form of the code was compared with a number of specific cases, each of which throws further light on the code behavior. The following areas are discussed in the paper: (1) the dam break problem; (2) Kelvin-Helmholtz instability; (3) counter-current flow; and (4) slug flow. It is concluded that detailed comparisons of the code with known analytic solutions and with a number of the more complex phenomenological experiments can give useful insights into its behavior
Critical review of conservation equations for two-phase flow in the U.S. NRC TRACE code
International Nuclear Information System (INIS)
Wulff, Wolfgang
2011-01-01
Research highlights: → Field equations as implemented in TRACE are incorrect. → Boundary conditions needed for cooling of nuclear fuel elements are wrong. → The two-fluid model in TRACE is not closed. → Three-dimensional flow modeling in TRACE has no basis. - Abstract: The field equations for two-phase flow in the computer code TRAC/RELAP Advanced Computational Engine or TRACE are examined to determine their validity, their capabilities and limitations in resolving nuclear reactor safety issues. TRACE was developed for the NRC to predict thermohydraulic phenomena in nuclear power plants during operational transients and postulated accidents. TRACE is based on the rigorously derived and well-established two-fluid field equations for 1-D and 3-D two-phase flow. It is shown that: (1)The two-fluid field equations for mass conservation as implemented in TRACE are wrong because local mass balances in TRACE are in conflict with mass conservation for the whole reactor system, as shown in Section . (2)Wrong equations of motion are used in TRACE in place of momentum balances, compromising at branch points the prediction of momentum transfer between, and the coupling of, loops in hydraulic networks by impedance (form loss and wall shear) and by inertia and thereby the simulation of reactor component interactions. (3)Most seriously, TRACE calculation of heat transfer from fuel elements is incorrect for single and two-phase flows, because Eq. of the TRACE Manual is wrong (see Section ). (4)Boundary conditions for momentum and energy balances in TRACE are restricted to flow regimes with single-phase wall contact because TRACE lacks constitutive relations for solid-fluid exchange of momentum and heat in prevailing flow regimes. Without a quantified assessment of consequences from (3) to (4), predictions of phasic fluid velocities, fuel temperatures and important safety parameters, e.g., peak clad temperature, are questionable. Moreover, TRACE cannot predict 3-D single- or
International Nuclear Information System (INIS)
Yang Ruichang; Liu Ruolei; Zhong Yong; Liu Tao
2006-01-01
This paper reports on an experimental study on transitional heat transfer of water flow in a heated vertical tube under natural circulation conditions. In the experiments the local and average heat transfer coefficients were obtained. The experimental data were compared with the predictions by a forced flow correlation available in the literature. The comparisons show that the Nusselt number value in the fully developed region is about 30% lower than the predictions by the forced flow correlation due to flow laminarization in the layer induced by co-current bulk natural circulation and free convection. By using the Rayleigh number Ra to represent the influence of free convection on heat transfer, the empirical correlations for the calculation of local and average heat transfer behavior in the tube at natural circulation have been developed. The empirical correlations are in good agreement with the experimental data. Based on the experimental results, the effect of the thermal entry-length behavior on heat transfer design in the tube under natural circulation was evaluated
Comparison of computer code calculations with FEBA test data
International Nuclear Information System (INIS)
Zhu, Y.M.
1988-06-01
The FEBA forced feed reflood experiments included base line tests with unblocked geometry. The experiments consisted of separate effect tests on a full-length 5x5 rod bundle. Experimental cladding temperatures and heat transfer coefficients of FEBA test No. 216 are compared with the analytical data postcalculated utilizing the SSYST-3 computer code. The comparison indicates a satisfactory matching of the peak cladding temperatures, quench times and heat transfer coefficients for nearly all axial positions. This agreement was made possible by the use of an artificially adjusted value of the empirical code input parameter in the heat transfer for the dispersed flow regime. A limited comparison of test data and calculations using the RELAP4/MOD6 transient analysis code are also included. In this case the input data for the water entrainment fraction and the liquid weighting factor in the heat transfer for the dispersed flow regime were adjusted to match the experimental data. On the other hand, no fitting of the input parameters was made for the COBRA-TF calculations which are included in the data comparison. (orig.) [de
SPLOSH III. A code for calculating reactivity and flow transients in CSGHWR
International Nuclear Information System (INIS)
Halsall, M.J.; Course, A.F.; Sidell, J.
1979-09-01
SPLOSH is a time dependent, one dimensional, finite difference (in time and space) coupled neutron kinetics and thermal hydraulics code for studying pressurised faults and control transients in water reactor systems. An axial single channel model with equally spaced mesh intervals is used to represent the neutronics of the reactor core. A radial finite difference model is used for heat conduction through the fuel pin, gas gap and can. Appropriate convective, boiling or post-dryout heat transfer correlations are used at the can-coolant interface. The hydraulics model includes the important features of the SGHWR primary loop including 'slave' channels in parallel with the 'mean' channel. Standard mass, energy and momentum equations are solved explicitly. Circuit features modelled include pumps, spray cooling and the SGHWR steam drum. Perturbations to almost any feature of the circuit model may be specified by the user although blowdown calculations resulting in critical or reversed flows are not permitted. Automatic reactor trips may be defined and the ensuing actions of moderator dumping and rod firing can be specified. (UK)
Two-dimensional heat flow apparatus
McDougall, Patrick; Ayars, Eric
2014-06-01
We have created an apparatus to quantitatively measure two-dimensional heat flow in a metal plate using a grid of temperature sensors read by a microcontroller. Real-time temperature data are collected from the microcontroller by a computer for comparison with a computational model of the heat equation. The microcontroller-based sensor array allows previously unavailable levels of precision at very low cost, and the combination of measurement and modeling makes for an excellent apparatus for the advanced undergraduate laboratory course.
International Nuclear Information System (INIS)
Viswanathan, H.S.
1995-01-01
The finite element code FEHMN is a three-dimensional finite element heat and mass transport simulator that can handle complex stratigraphy and nonlinear processes such as vadose zone flow, heat flow and solute transport. Scientists at LANL have been developed hydrologic flow and transport models of the Yucca Mountain site using FEHMN. Previous FEHMN simulations have used an equivalent K d model to model solute transport. In this thesis, FEHMN is modified making it possible to simulate the transport of a species with a rigorous chemical model. Including the rigorous chemical equations into FEHMN simulations should provide for more representative transport models for highly reactive chemical species. A fully kinetic formulation is chosen for the FEHMN reactive transport model. Several methods are available to computationally implement a fully kinetic formulation. Different numerical algorithms are investigated in order to optimize computational efficiency and memory requirements of the reactive transport model. The best algorithm of those investigated is then incorporated into FEHMN. The algorithm chosen requires for the user to place strongly coupled species into groups which are then solved for simultaneously using FEHMN. The complete reactive transport model is verified over a wide variety of problems and is shown to be working properly. The simulations demonstrate that gas flow and carbonate chemistry can significantly affect 14 C transport at Yucca Mountain. The simulations also provide that the new capabilities of FEHMN can be used to refine and buttress already existing Yucca Mountain radionuclide transport studies
THEORETICAL AND EXPERIMENTAL ANALYSIS OF A CROSS-FLOW HEAT EXCHANGER
Directory of Open Access Journals (Sweden)
R. Tuğrul OĞULATA
1996-03-01
Full Text Available In this study, cross-flow plate type heat exchanger has been investigated because of its effective use in waste heat recovery systems. For this purpose, a heat regain system has been investigated and manufactured in laboratory conditions. Manufactured heat exchanger has been tested with an applicable experimental set up and temperatures, velocity of the air and the pressure losses occuring in the system have been measured and the efficiency of the system has been determined. The irreversibility of heat exchanger has been taken into consideration while the design of heat exchanger is being performed. So minimum entropy generation number has been analysied with respect to second law of thermodynamics in cross-flow heat exchanger. The minimum entropy generation number depends on parameters called optimum flow path length, dimensionless mass velocity and dimensionless heat transfer area. Variations of entropy generation number with these parameters have been analysied and introduced their graphics with their comments.
Heat transfer and flow characteristics on a gas turbine shroud.
Obata, M; Kumada, M; Ijichi, N
2001-05-01
The work described in this paper is an experimental investigation of the heat transfer from the main flow to a turbine shroud surface, which may be applicable to ceramic gas turbines. Three kinds of turbine shrouds are considered with a flat surface, a taper surface and a spiral groove surface opposite to the blades in an axial flow turbine of actual turbo-charger. Heat transfer measurements were performed for the experimental conditions of a uniform heat flux or a uniform wall temperature. The effects of the inlet flow angle, rotational speed, and tip clearance on the heat transfer coefficient were clarified under on- and off-design flow conditions. The mean heat transfer coefficient was correlated to the blade Reynolds number and tip clearance, and compared with an experimental correlation and measurements of a flat surface. A comparison was also made for the measurement of static pressure distributions.
Non-standard model for electron heat transport for multidimensional hydrodynamic codes
Energy Technology Data Exchange (ETDEWEB)
Nicolai, Ph.; Busquet, M.; Schurtz, G. [CEA/DAM-Ile de France, 91 - Bruyeres Le Chatel (France)
2000-07-01
In simulations of laser-produced plasma, modeling of heat transport requires an artificial limitation of standard Spitzer-Haerm fluxes. To improve heat conduction processing, we have developed a multidimensional model which accounts for non-local features of heat transport and effects of self-generated magnetic fields. This consistent treatment of both mechanisms has been implemented in a two-dimensional radiation-hydrodynamic code. First results indicate good agreements between simulations and experimental data. (authors)
Non-standard model for electron heat transport for multidimensional hydrodynamic codes
International Nuclear Information System (INIS)
Nicolai, Ph.; Busquet, M.; Schurtz, G.
2000-01-01
In simulations of laser-produced plasma, modeling of heat transport requires an artificial limitation of standard Spitzer-Haerm fluxes. To improve heat conduction processing, we have developed a multidimensional model which accounts for non-local features of heat transport and effects of self-generated magnetic fields. This consistent treatment of both mechanisms has been implemented in a two-dimensional radiation-hydrodynamic code. First results indicate good agreements between simulations and experimental data. (authors)
Experimental investigation on Heat Transfer Performance of Annular Flow Path Heat Pipe
International Nuclear Information System (INIS)
Kim, In Guk; Kim, Kyung Mo; Jeong, Yeong Shin; Bang, In Cheol
2015-01-01
Mochizuki et al. was suggested the passive cooling system to spent nuclear fuel pool. Detail analysis of various heat pipe design cases was studied to determine the heat pipes cooling performance. Wang et al. suggested the concept PRHRS of MSR using sodium heat pipes, and the transient performance of high temperature sodium heat pipe was numerically simulated in the case of MSR accident. The meltdown at the Fukushima Daiichi nuclear power plants alarmed to the dangers of station blackout (SBO) accident. After the SBO accident, passive decay heat removal systems have been investigated to prevent the severe accidents. Mochizuki et al. suggested the heat pipes cooling system using loop heat pipes for decay heat removal cooling and analysis of heat pipe thermal resistance for boiling water reactor (BWR). The decay heat removal systems for pressurized water reactor (PWR) were suggested using natural convection mechanisms and modification of PWR design. Our group suggested the concept of a hybrid heat pipe with control rod as Passive IN-core Cooling System (PINCs) for decay heat removal for advanced nuclear power plant. Hybrid heat pipe is the combination of the heat pipe and control rod. In the present research, the main objective is to investigate the effect of the inner structure to the heat transfer performance of heat pipe containing neutron absorber material, B 4 C. The main objective is to investigate the effect of the inner structure in heat pipe to the heat transfer performance with annular flow path. ABS pellet was used instead of B 4 C pellet as cylindrical structures. The thermal performances of each heat pipes were measured experimentally. Among them, concentric heat pipe showed the best performance compared with others. 1. Annular evaporation section heat pipe and annular flow path heat pipe showed heat transfer degradation. 2. AHP also had annular vapor space and contact cooling surface per unit volume of vapor was increased. Heat transfer coefficient of
Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes
International Nuclear Information System (INIS)
Baratta, A.J.
1997-01-01
To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together
Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes
Energy Technology Data Exchange (ETDEWEB)
Baratta, A.J.
1997-07-01
To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts and engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.
ASME code considerations for the compact heat exchanger
Energy Technology Data Exchange (ETDEWEB)
Nestell, James [MPR Associates Inc., Alexandria, VA (United States); Sham, Sam [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)
2015-08-31
robustness. Classic shell and tube designs will be large and costly, and may only be appropriate in steam generator service in the SHX where boiling inside the tubes occurs. For other energy conversion systems, all of these features can be met in a compact heat exchanger design. This report will examine some of the ASME Code issues that will need to be addressed to allow use of a Code-qualified compact heat exchanger in IHX or SHX nuclear service. Most effort will focus on the IHX, since the safety-related (Class A) design rules are more extensive than those for important-to-safety (Class B) or commercial rules that are relevant to the SHX.
Improved choked flow model for MARS code
International Nuclear Information System (INIS)
Chung, Moon Sun; Lee, Won Jae; Ha, Kwi Seok; Hwang, Moon Kyu
2002-01-01
Choked flow calculation is improved by using a new sound speed criterion for bubbly flow that is derived by the characteristic analysis of hyperbolic two-fluid model. This model was based on the notion of surface tension for the interfacial pressure jump terms in the momentum equations. Real eigenvalues obtained as the closed-form solution of characteristic polynomial represent the sound speed in the bubbly flow regime that agrees well with the existing experimental data. The present sound speed shows more reasonable result in the extreme case than the Nguyens did. The present choked flow criterion derived by the present sound speed is employed in the MARS code and assessed by using the Marviken choked flow tests. The assessment results without any adjustment made by some discharge coefficients demonstrate more accurate predictions of choked flow rate in the bubbly flow regime than those of the earlier choked flow calculations. By calculating the Typical PWR (SBLOCA) problem, we make sure that the present model can reproduce the reasonable transients of integral reactor system
Implementation of an implicit method into heat conduction calculation of TRAC-PF1/MOD2 code
International Nuclear Information System (INIS)
Akimoto, Hajime; Abe, Yutaka; Ohnuki, Akira; Murao, Yoshio
1990-08-01
A two-dimensional unsteady heat conduction equation is solved in the TRAC-PF/MOD2 code to calculate temperature transients in fuel rod. A large CPU time is often required to get stable solution of temperature transients in the TRAC calculation with a small axial node size (less than 1.0 mm), because the heat conduction equation is discretized explicitly. To eliminate the restriction of the maximum time step size by the heat conduction calculation, an implicit method for solving the heat condition equation was developed and implemented into the TRAC code. Several assessment calculations were performed with the original and modified TRAC codes. It is confirmed that the implicit method is reliable and is successfully implemented into the TRAC code through comparison with theoretical solutions and assessment calculation results. It is demonstrated that the implicit method makes the heat conduction calculation practical even for the analyses of temperature transients with the axial node size less than 0.1 mm. (author)
Hendricks, R. C.; Braun, M. J.; Mullen, R. L.
1986-01-01
In systems where the design inlet and outlet pressure P sub amb are maintained above the thermodynamic critical pressure P sub c, it is often assumed that heat and mass transfer are governed by single-phase relations and that two-phase flows cannot occur. This simple rule of thumb is adequate in many low-power designs but is inadequate for high-performance turbomachines, boilers, and other systems where two-phase regions can exist even though P sub amb P sub c. Heat and mass transfer and rotordynamic-fluid-mechanic restoring forces depend on momentum differences, and those for a two-phase zone can differ significantly from those for a single-phase zone. By using a laminar, variable-property bearing code and a rotating boiler code, pressure and temperature surfaces were determined that illustrate nesting of a two-phase region within a supercritical pressure region. The method of corresponding states is applied to bearings with reasonable rapport.
Energy Technology Data Exchange (ETDEWEB)
Papukchiev, A., E-mail: angel.papukchiev@grs.de; Buchholz, S.
2017-02-15
Highlights: • ANSYS CFX is validated for gas and liquid metal flows. • L-STAR and TALL-3D experiments are simulated. • Complex flow and heat transfer phenomena are modelled. • Conjugate heat transfer has to be considered in CFD analyses. - Abstract: Within the FP7 European project THINS (Thermal Hydraulics of Innovative Nuclear Systems), numerical tools for the simulation of the thermal-hydraulics of next generation rector systems were developed, applied and validated for innovative coolants. The Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH participated in THINS with activities related to the development and validation of computational fluid dynamics (CFD) and coupled System Thermal Hydraulics (STH) – CFD codes. High quality measurements from the L-STAR and TALL-3D experiments were used to assess the numerical results. Two-equation eddy viscosity and scale resolving turbulence models were used in the validation process of ANSYS CFX for gas and liquid metal flows with conjugate heat transfer. This paper provides a brief overview on the main results achieved at GRS within the project.
Energy Technology Data Exchange (ETDEWEB)
Luo, Yang; Wen, Meimei [Department of Mechanical Engineering, Graduate School, Kyung Hee University, Yong-in, Kyunggi-do, 446-701 (Korea, Republic of); Kim, Chang Nyung, E-mail: cnkim@khu.ac.kr [Department of Mechanical Engineering, College of Engineering, Kyung Hee University, Yong-in, Kyunggi-do, 446-701 (Korea, Republic of); Yang, Shangjing [Department of Mechanical Engineering, Graduate School, Kyung Hee University, Yong-in, Kyunggi-do, 446-701 (Korea, Republic of)
2017-04-15
In this study, the characteristics of liquid metal (LM) magnetohydrodynamic (MHD) flow and convective heat transfer in a manifold with three sub-channels having locally different electric conductivity are investigated with the use of commercial code CFX, allowing an imbalance in flow rate among the sub-channels, which can be used for intensive cooling of the region with higher heat load in the blanket. In a manifold with co-flow multiple sub-channels, the electrical current can cross the fluid regions and channel walls, thus influencing the flow distribution in each sub-channel. In the present study, cases with various arrangements of the electric conductivity in different parts of the channel walls are investigated, yielding different distributions of the current and fluid flow in different cases. Here, the mechanism governing the imbalance in mass flow rate among the sub-channels is discussed. The interdependency of the fluid velocity, current and electric potential of LM MHD flows in the three sub-channels are analyzed in detail. The results show that, in the sub-channel surrounded by the walls with lower electric conductivity, higher axial velocity and superior heat extraction can be obtained, with an effective cooling associated with higher velocity, where the higher velocity is closely related to the distribution of the electromotive component of the current in the flow field.
International Nuclear Information System (INIS)
Luo, Yang; Wen, Meimei; Kim, Chang Nyung; Yang, Shangjing
2017-01-01
In this study, the characteristics of liquid metal (LM) magnetohydrodynamic (MHD) flow and convective heat transfer in a manifold with three sub-channels having locally different electric conductivity are investigated with the use of commercial code CFX, allowing an imbalance in flow rate among the sub-channels, which can be used for intensive cooling of the region with higher heat load in the blanket. In a manifold with co-flow multiple sub-channels, the electrical current can cross the fluid regions and channel walls, thus influencing the flow distribution in each sub-channel. In the present study, cases with various arrangements of the electric conductivity in different parts of the channel walls are investigated, yielding different distributions of the current and fluid flow in different cases. Here, the mechanism governing the imbalance in mass flow rate among the sub-channels is discussed. The interdependency of the fluid velocity, current and electric potential of LM MHD flows in the three sub-channels are analyzed in detail. The results show that, in the sub-channel surrounded by the walls with lower electric conductivity, higher axial velocity and superior heat extraction can be obtained, with an effective cooling associated with higher velocity, where the higher velocity is closely related to the distribution of the electromotive component of the current in the flow field.
Prediction of strongly-heated internal gas flows
International Nuclear Information System (INIS)
McEligot, D.M.; Shehata, A.M.; Kunugi, Tomoaki
1997-01-01
The purposes of the present article are to remind practitioners why the usual textbook approaches may not be appropriate for treating gas flows heated from the surface with large heat fluxes and to review the successes of some recent applications of turbulence models to this case. Simulations from various turbulence models have been assessed by comparison to the measurements of internal mean velocity and temperature distributions by Shehata for turbulent, laminarizing and intermediate flows with significant gas property variation. Of about fifteen models considered, five were judged to provide adequate predictions
NSPEC - A neutron spectrum code for beam-heated fusion plasmas
International Nuclear Information System (INIS)
Scheffel, J.
1983-06-01
A 3-dimensional computer code is described, which computes neutron spectra due to beam heating of fusion plasmas. Three types of interactions are considered; thermonuclear of plasma-plasma, beam-plasma and beam-beam interactions. Beam deposition is modelled by the NFREYA code. The applied steady state beam distribution as a function of pitch angle and velocity contains the effects of energy diffusion, friction, angular scattering, charge exchange, electric field and source pitch angle distribution. The neutron spectra, generated by Monte-Carlo methods, are computed with respect to given lines of sight. This enables the code to be used for neutron diagnostics. (author)
DEFF Research Database (Denmark)
Rakêt, Lars Lau; Søgaard, Jacob; Salmistraro, Matteo
2012-01-01
We consider Distributed Video Coding (DVC) in presence of communication errors. First, we present DVC side information generation based on a new method of optical flow driven frame interpolation, where a highly optimized TV-L1 algorithm is used for the flow calculations and combine three flows....... Thereafter methods for exploiting the error-correcting capabilities of the LDPCA code in DVC are investigated. The proposed frame interpolation includes a symmetric flow constraint to the standard forward-backward frame interpolation scheme, which improves quality and handling of large motion. The three...... flows are combined in one solution. The proposed frame interpolation method consistently outperforms an overlapped block motion compensation scheme and a previous TV-L1 optical flow frame interpolation method with an average PSNR improvement of 1.3 dB and 2.3 dB respectively. For a GOP size of 2...
Unsteady Flow in a Supersonic Turbine with Variable Specific Heats
Dorney, Daniel J.; Griffin, Lisa W.; Huber, Frank; Sondak, Douglas L.; Turner, James (Technical Monitor)
2001-01-01
Modern high-work turbines can be compact, transonic, supersonic, counter-rotating, or use a dense drive gas. The vast majority of modern rocket turbine designs fall into these Categories. These turbines usually have large temperature variations across a given stage, and are characterized by large amounts of flow unsteadiness. The flow unsteadiness can have a major impact on the turbine performance and durability. For example, the Space Transportation Main Engine (STME) fuel turbine, a high work, transonic design, was found to have an unsteady inter-row shock which reduced efficiency by 2 points and increased dynamic loading by 24 percent. The Revolutionary Reusable Technology Turbopump (RRTT), which uses full flow oxygen for its drive gas, was found to shed vortices with such energy as to raise serious blade durability concerns. In both cases, the sources of the problems were uncovered (before turbopump testing) with the application of validated, unsteady computational fluid dynamics (CFD) to the designs. In the case of the RRTT and the Alternate Turbopump Development (ATD) turbines, the unsteady CFD codes have been used not just to identify problems, but to guide designs which mitigate problems due to unsteadiness. Using unsteady flow analyses as a part of the design process has led to turbine designs with higher performance (which affects temperature and mass flow rate) and fewer dynamics problems. One of the many assumptions made during the design and analysis of supersonic turbine stages is that the values of the specific heats are constant. In some analyses the value is based on an average of the expected upstream and downstream temperatures. In stages where the temperature can vary by 300 to 500 K, however, the assumption of constant fluid properties may lead to erroneous performance and durability predictions. In this study the suitability of assuming constant specific heats has been investigated by performing three-dimensional unsteady Navier
Local heat transfer where heated rods touch in axially flowing water
International Nuclear Information System (INIS)
Kast, S.J.
1983-05-01
An anlaytic model is developed to predict the azimuthal width of a stablesteam blanket region near the line of contact between two heated rods cooled by axially flowing water at high pressure. The model is intended to aid analysis of reduced surface heat transfer capability for the abnormal configuration of nuclear fuel rods bowed into contact in the core of a pressurized water nuclear reactor. The analytic model predicts the azimuthal width of the steam blanket zone having reduced surface heat transfer as a function of rod average heat flux, subchannel coolant conditions and rod dimensions. The analytic model is developed from a heat balance between the heat generated in the wall of a heated empty tube and the heat transported away by transverse mixing and axial convection in the coolant subchannel. The model is developed for seveal geometries including heated rods in line contact, a heated rod touching a short insulating plane and a heated rod touching the inside of a metal guide tube
Assessment of critical flow models of RELAP5-MOD2 and CATHARE codes
International Nuclear Information System (INIS)
Hao Laomi; Zhu Zhanchuan
1992-01-01
The critical flow tests for the long and short nozzles conducted on the SUPER MOBY-DICK facility were analyzed using the RELAP5-MOD2 and CATHARE 1.3 codes to assess the critical flow models of two codes. The critical mass flux calculated for two nozzles are given. The CATHARE code has used the thermodynamic nonequilibrium sound velocity of the two-phase fluid as the critical flow criterion, and has the better interphase transfer models and calculates the critical flow velocities with the completely implicit solution. Therefore, it can well calculate the critical flowrate and can describe the effect of the geometry L/D on the critical flowrate
Wang, C. R.; Hingst, W. R.; Porro, A. R.
1991-01-01
The properties of 2-D shock wave/turbulent boundary layer interaction flows were calculated by using a compressible turbulent Navier-Stokes numerical computational code. Interaction flows caused by oblique shock wave impingement on the turbulent boundary layer flow were considered. The oblique shock waves were induced with shock generators at angles of attack less than 10 degs in supersonic flows. The surface temperatures were kept at near-adiabatic (ratio of wall static temperature to free stream total temperature) and cold wall (ratio of wall static temperature to free stream total temperature) conditions. The computational results were studied for the surface heat transfer, velocity temperature correlation, and turbulent shear stress in the interaction flow fields. Comparisons of the computational results with existing measurements indicated that (1) the surface heat transfer rates and surface pressures could be correlated with Holden's relationship, (2) the mean flow streamwise velocity components and static temperatures could be correlated with Crocco's relationship if flow separation did not occur, and (3) the Baldwin-Lomax turbulence model should be modified for turbulent shear stress computations in the interaction flows.
MHD and heat transfer benchmark problems for liquid metal flow in rectangular ducts
International Nuclear Information System (INIS)
Sidorenkov, S.I.; Hua, T.Q.; Araseki, H.
1994-01-01
Liquid metal cooling systems of a self-cooled blanket in a tokamak reactor will likely include channels of rectangular cross section where liquid metal is circulated in the presence of strong magnetic fields. MHD pressure drop, velocity distribution and heat transfer characteristics are important issues in the engineering design considerations. Computer codes for the reliable solution of three-dimensional MHD flow problems are needed for fusion relevant conditions. Argonne National Laboratory and The Efremov Institute have jointly defined several benchmark problems for code validation. The problems, described in this paper, are based on two series of rectangular duct experiments conducted at ANL; one of the series is a joint ANL/Efremov experiment. The geometries consist of variation of aspect ratio and wall thickness (thus wall conductance ratio). The transverse magnetic fields are uniform and nonuniform in the axial direction
Heat transfer in intermediate heat exchanger under low flow rate conditions
International Nuclear Information System (INIS)
Mochizuki, H.
2008-01-01
The present paper describes the heat transfer in intermediate heat exchangers (IHXs) of liquid metal cooled fast reactors when flow rate is low such as a natural circulation condition. Although empirical correlations of heat transfer coefficients for IHX were derived using test data at the fast reactor 'Monju' and 'Joyo' and also at the 50 MW steam generator facility, the heat transfer coefficient was very low compared to the well known correlation for liquid metals proposed by Seban-Shimazaki. The heat conduction in IHX was discussed as a possible cause of the low Nusselt number. As a result, the heat conduction is not significant under the natural circulation condition, and the heat conduction term in the energy equation can be neglected in the one-dimensional plant dynamics calculation. (authors)
Development of an advanced fluid-dynamic analysis code: α-flow
International Nuclear Information System (INIS)
Akiyama, Mamoru
1990-01-01
A Project for development of large scale three-dimensional fluid-dynamic analysis code, α-FLOW, coping with the recent advancement of supercomputers and workstations, has been in progress. This project is called the α-Project, which has been promoted by the Association for Large Scale Fluid Dynamics Analysis Code comprising private companies and research institutions such as universities. The developmental period for the α-FLOW is four years, March 1989 to March 1992. To date, the major portions of basic design and program preparation have been completed and the project is in the stage of testing each module. In this paper, the present status of the α-Project, design policy and outline of α-FLOW are described. (author)
Lunar Heat Flow Probe, Phase I
National Aeronautics and Space Administration — To accurately determine endogenic heat flow, both thermal gradient and thermal conductivity measurements are needed. The thermal gradient measurement can be achieved...
International Nuclear Information System (INIS)
Valls, E.Mas de les; Batet, L.; Medina, V. de; Fradera, J.; Sedano, L.
2011-01-01
Highlights: → 3D transient CFD code based on OpenFOAM toolbox and accounting for MHD and thermal et al. effects. → Hydrodynamic instabilities caused by the jet (generated at the gap narrowing) are found at Reynolds 480. → Hartmann 1740 is able to stabilise the flow. → A heat deposition corresponding to Gr = 5.21 x 10 9 is sufficient for buoyancy to be predominant at the bend region. Flow becomes unstable. → Tritium permeation ratio cannot be accurately predicted due to major uncertainties in Sievert's coefficient. - Abstract: Under fusion reactor operational conditions, heat deposition might cause a complex buoyant liquid metal flow in the HCLL blanket, what has a direct influence on tritium permeation ratio. In order to characterise the nature of this flow, a simplified HCLL channel, including the U-bend near the reactor first wall, is analysed using a finite volume CFD code, based on OpenFOAM toolbox, following an electric potential based formulation. Code validation results for developed MHD flow and magneto-convective flow are exposed. The influence of the HCLL U-bend on the flow pattern is studied with the validated code, covering the range of possible Reynolds numbers in HCLL-ITER blanket, and considering either electrically insulating or perfectly conducting walls. It can be stated that, despite the very low velocities and the high Hartmann number, flow pattern is complex and unsteady vortices are formed by the action of buoyancy forces together with the influence of the U-bend. Through the analysis, the flow physics is decoupled in order to identify the exact origin of vortex formation. A simplified tritium transport analysis, considering tritium as a passive scalar, has been carried out including a study on boundary conditions influence and a sensitivity analysis of tritium permeation fluxes to diffusivity and solubility parameters. Results show the relevance of Sievert's coefficient uncertainties, which alters the permeation ratio by an order of
Energy Technology Data Exchange (ETDEWEB)
Valls, E.Mas de les, E-mail: elisabet.masdelesvalls@gits.ws [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Technology for Fusion (T4F) Research Group, GREENER, Dept. of Heat Engines (UPC) (Spain); Batet, L. [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Technology for Fusion (T4F) Research Group, GREENER, Dept. of Physics and Nuclear Engineering (UPC) (Spain); Medina, V. de [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Sediment Transport Research Group, Dept. of Engineering Hydraulic, Marine and Environmental Engineering (UPC) (Spain); Fradera, J. [Technical University of Catalonia (UPC), Jordi Girona 1-3, 08034 Barcelona (Spain); Technology for Fusion (T4F) Research Group, GREENER, Dept. of Physics and Nuclear Engineering (UPC) (Spain); Sedano, L. [EURATOM-CIEMAT Fusion Association, Av. Complutense 22, 28040 Madrid (Spain)
2011-06-15
Highlights: > 3D transient CFD code based on OpenFOAM toolbox and accounting for MHD and thermal et al. effects. > Hydrodynamic instabilities caused by the jet (generated at the gap narrowing) are found at Reynolds 480. > Hartmann 1740 is able to stabilise the flow. > A heat deposition corresponding to Gr = 5.21 x 10{sup 9} is sufficient for buoyancy to be predominant at the bend region. Flow becomes unstable. > Tritium permeation ratio cannot be accurately predicted due to major uncertainties in Sievert's coefficient. - Abstract: Under fusion reactor operational conditions, heat deposition might cause a complex buoyant liquid metal flow in the HCLL blanket, what has a direct influence on tritium permeation ratio. In order to characterise the nature of this flow, a simplified HCLL channel, including the U-bend near the reactor first wall, is analysed using a finite volume CFD code, based on OpenFOAM toolbox, following an electric potential based formulation. Code validation results for developed MHD flow and magneto-convective flow are exposed. The influence of the HCLL U-bend on the flow pattern is studied with the validated code, covering the range of possible Reynolds numbers in HCLL-ITER blanket, and considering either electrically insulating or perfectly conducting walls. It can be stated that, despite the very low velocities and the high Hartmann number, flow pattern is complex and unsteady vortices are formed by the action of buoyancy forces together with the influence of the U-bend. Through the analysis, the flow physics is decoupled in order to identify the exact origin of vortex formation. A simplified tritium transport analysis, considering tritium as a passive scalar, has been carried out including a study on boundary conditions influence and a sensitivity analysis of tritium permeation fluxes to diffusivity and solubility parameters. Results show the relevance of Sievert's coefficient uncertainties, which alters the permeation ratio by an
Analysis of the one-dimensional transient compressible vapor flow in heat pipes
Jang, Jong H.; Faghri, Amir; Chang, Won S.
1991-01-01
The transient compressible one-dimensional vapor flow dynamics in a heat pipe is modeled. The numerical results are obtained by using the implicit non-iterative Beam-Warming finite difference method. The model is tested for simulated heat pipe vapor flow and actual vapor flow in cylindrical heat pipes. A good comparison of the present transient results for the simulated heat pipe vapor flow with the previous results of a two-dimensional numerical model is achieved and the steady state results are in agreement with the existing experimental data. The transient behavior of the vapor flow under subsonic, sonic, and supersonic speeds as well as high mass flow rates are successfully predicted.
Heat transfer and fluid flow in biological processes advances and applications
Becker, Sid
2015-01-01
Heat Transfer and Fluid Flow in Biological Processes covers emerging areas in fluid flow and heat transfer relevant to biosystems and medical technology. This book uses an interdisciplinary approach to provide a comprehensive prospective on biofluid mechanics and heat transfer advances and includes reviews of the most recent methods in modeling of flows in biological media, such as CFD. Written by internationally recognized researchers in the field, each chapter provides a strong introductory section that is useful to both readers currently in the field and readers interested in learning more about these areas. Heat Transfer and Fluid Flow in Biological Processes is an indispensable reference for professors, graduate students, professionals, and clinical researchers in the fields of biology, biomedical engineering, chemistry and medicine working on applications of fluid flow, heat transfer, and transport phenomena in biomedical technology. Provides a wide range of biological and clinical applications of fluid...
Study on Gas-liquid Falling Film Flow in Internal Heat Integrated Distillation Column
Liu, Chong
2017-10-01
Gas-liquid internally heat integrated distillation column falling film flow with nonlinear characteristics, study on gas liquid falling film flow regulation control law, can reduce emissions of the distillation column, and it can improve the quality of products. According to the distribution of gas-liquid mass balance internally heat integrated distillation column independent region, distribution model of heat transfer coefficient of building internal heat integrated distillation tower is obtained liquid distillation falling film flow in the saturated vapour pressure of liquid water balance, using heat transfer equation and energy equation to balance the relationship between the circulating iterative gas-liquid falling film flow area, flow parameter information, at a given temperature, pressure conditions, gas-liquid flow falling film theory makes the optimal parameters to achieve the best fitting value with the measured values. The results show that the geometric gas-liquid internally heat integrated distillation column falling film flow heat exchange area and import column thermostat, the average temperature has significant. The positive correlation between the heat exchanger tube entrance due to temperature difference between inside and outside, the heat flux is larger, with the increase of internal heat integrated distillation column temperature, the slope decreases its temperature rise, which accurately describes the internal gas-liquid heat integrated distillation tower falling film flow regularity, take appropriate measures to promote the enhancement of heat transfer. It can enhance the overall efficiency of the heat exchanger.
Direct containment heating models in the CONTAIN code
International Nuclear Information System (INIS)
Washington, K.E.; Williams, D.C.
1995-08-01
The potential exists in a nuclear reactor core melt severe accident for molten core debris to be dispersed under high pressure into the containment building. If this occurs, the set of phenomena that result in the transfer of energy to the containment atmosphere and its surroundings is referred to as direct containment heating (DCH). Because of the potential for DCH to lead to early containment failure, the U.S. Nuclear Regulatory Commission (USNRC) has sponsored an extensive research program consisting of experimental, analytical, and risk integration components. An important element of the analytical research has been the development and assessment of direct containment heating models in the CONTAIN code. This report documents the DCH models in the CONTAIN code. DCH models in CONTAIN for representing debris transport, trapping, chemical reactions, and heat transfer from debris to the containment atmosphere and surroundings are described. The descriptions include the governing equations and input instructions in CONTAIN unique to performing DCH calculations. Modifications made to the combustion models in CONTAIN for representing the combustion of DCH-produced and pre-existing hydrogen under DCH conditions are also described. Input table options for representing the discharge of debris from the RPV and the entrainment phase of the DCH process are also described. A sample calculation is presented to demonstrate the functionality of the models. The results show that reasonable behavior is obtained when the models are used to predict the sixth Zion geometry integral effects test at 1/10th scale
Direct containment heating models in the CONTAIN code
Energy Technology Data Exchange (ETDEWEB)
Washington, K.E.; Williams, D.C.
1995-08-01
The potential exists in a nuclear reactor core melt severe accident for molten core debris to be dispersed under high pressure into the containment building. If this occurs, the set of phenomena that result in the transfer of energy to the containment atmosphere and its surroundings is referred to as direct containment heating (DCH). Because of the potential for DCH to lead to early containment failure, the U.S. Nuclear Regulatory Commission (USNRC) has sponsored an extensive research program consisting of experimental, analytical, and risk integration components. An important element of the analytical research has been the development and assessment of direct containment heating models in the CONTAIN code. This report documents the DCH models in the CONTAIN code. DCH models in CONTAIN for representing debris transport, trapping, chemical reactions, and heat transfer from debris to the containment atmosphere and surroundings are described. The descriptions include the governing equations and input instructions in CONTAIN unique to performing DCH calculations. Modifications made to the combustion models in CONTAIN for representing the combustion of DCH-produced and pre-existing hydrogen under DCH conditions are also described. Input table options for representing the discharge of debris from the RPV and the entrainment phase of the DCH process are also described. A sample calculation is presented to demonstrate the functionality of the models. The results show that reasonable behavior is obtained when the models are used to predict the sixth Zion geometry integral effects test at 1/10th scale.
Benchmark calculation of subchannel analysis codes
International Nuclear Information System (INIS)
1996-02-01
In order to evaluate the analysis capabilities of various subchannel codes used in thermal-hydraulic design of light water reactors, benchmark calculations were performed. The selected benchmark problems and major findings obtained by the calculations were as follows: (1)As for single-phase flow mixing experiments between two channels, the calculated results of water temperature distribution along the flow direction were agreed with experimental results by tuning turbulent mixing coefficients properly. However, the effect of gap width observed in the experiments could not be predicted by the subchannel codes. (2)As for two-phase flow mixing experiments between two channels, in high water flow rate cases, the calculated distributions of air and water flows in each channel were well agreed with the experimental results. In low water flow cases, on the other hand, the air mixing rates were underestimated. (3)As for two-phase flow mixing experiments among multi-channels, the calculated mass velocities at channel exit under steady-state condition were agreed with experimental values within about 10%. However, the predictive errors of exit qualities were as high as 30%. (4)As for critical heat flux(CHF) experiments, two different results were obtained. A code indicated that the calculated CHF's using KfK or EPRI correlations were well agreed with the experimental results, while another code suggested that the CHF's were well predicted by using WSC-2 correlation or Weisman-Pei mechanistic model. (5)As for droplets entrainment and deposition experiments, it was indicated that the predictive capability was significantly increased by improving correlations. On the other hand, a remarkable discrepancy between codes was observed. That is, a code underestimated the droplet flow rate and overestimated the liquid film flow rate in high quality cases, while another code overestimated the droplet flow rate and underestimated the liquid film flow rate in low quality cases. (J.P.N.)
Modeling of a District Heating System and Optimal Heat-Power Flow
Directory of Open Access Journals (Sweden)
Wentao Yang
2018-04-01
Full Text Available With ever-growing interconnections of various kinds of energy sources, the coupling between a power distribution system (PDS and a district heating system (DHS has been progressively intensified. Thus, it is becoming more and more important to take the PDS and the DHS as a whole in energy flow analysis. Given this background, a steady state model of DHS is first presented with hydraulic and thermal sub-models included. Structurally, the presented DHS model is composed of three major parts, i.e., the straight pipe, four kinds of local pipes, and the radiator. The impacts of pipeline parameters and the environment temperature on heat losses and pressure losses are then examined. The term “heat-power flow” is next defined, and the optimal heat-power flow (OHPF model formulated as a quadratic planning problem, in which the objective is to minimize energy losses, including the heat losses and active power losses, and both the operational constraints of PDS and DHS are respected. The developed OHPF model is solved by the well-established IPOPT (Interior Point OPTimizer commercial solver, which is based on the YALMIP/MATLAB toolbox. Finally, two sample systems are served for demonstrating the characteristics of the proposed models.
Thermal heat-balance mode flow-to-frequency converter
Pawlowski, Eligiusz
2016-11-01
This paper presents new type of thermal flow converter with the pulse frequency output. The integrating properties of the temperature sensor have been used, which allowed for realization of pulse frequency modulator with thermal feedback loop, stabilizing temperature of sensor placed in the flowing medium. The system assures balancing of heat amount supplied in impulses to the sensor and heat given up by the sensor in a continuous way to the flowing medium. Therefore the frequency of output impulses is proportional to the heat transfer coefficient from sensor to environment. According to the King's law, the frequency of those impulses is a function of medium flow velocity around the sensor. The special feature of presented solution is total integration of thermal sensor with the measurement signal conditioning system. Sensor and conditioning system are not the separate elements of the measurement circuit, but constitute a whole in form of thermal heat-balance mode flow-to-frequency converter. The advantage of such system is easiness of converting the frequency signal to the digital form, without using any additional analogue-to-digital converters. The frequency signal from the converter may be directly connected to the microprocessor input, which with use of standard built-in counters may convert the frequency into numerical value of high precision. Moreover, the frequency signal has higher resistance to interference than the voltage signal and may be transmitted to remote locations without the information loss.
Directory of Open Access Journals (Sweden)
Sabanskis A.
2016-04-01
Full Text Available Monitoring of temperature, humidity and air flow velocity is performed in 5 experimental buildings with the inner size of 3×3×3 m3 located in Riga, Latvia. The buildings are equipped with different heating systems, such as an air-air heat pump, air-water heat pump, capillary heating mat on the ceiling and electric heater. Numerical simulation of air flow and heat transfer by convection, conduction and radiation is carried out using OpenFOAM software and compared with experimental data. Results are analysed regarding the temperature and air flow distribution as well as thermal comfort.
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
Two-phase interfacial area and flow regime modeling in FLOWTRAN-TF code
International Nuclear Information System (INIS)
Smith, F.G. III; Lee, S.Y.; Flach, G.P.; Hamm, L.L.
1992-01-01
FLOWTRAN-TF is a new two-component, two-phase thermal-hydraulics code to capture the detailed assembly behavior associated with loss-of-coolant accident analyses in multichannel assemblies of the SRS reactors. The local interfacial area of the two-phase mixture is computed by summing the interfacial areas contributed by each of three flow regimes. For smooth flow regime transitions, the code uses an interpolation technique in terms of component void fraction for each basic flow regime
Ocular blood flow decreases during passive heat stress in resting humans.
Ikemura, Tsukasa; Miyaji, Akane; Kashima, Hideaki; Yamaguchi, Yuji; Hayashi, Naoyuki
2013-12-06
Heat stress induces various physiological changes and so could influence ocular circulation. This study examined the effect of heat stress on ocular blood flow. Ocular blood flow, end-tidal carbon dioxide (P(ET)CO2) and blood pressure were measured for 12 healthy subjects wearing water-perfused tube-lined suits under two conditions of water circulation: (1) at 35 °C (normothermia) for 30 min and (2) at 50 °C for 90 min (passive heat stress). The blood-flow velocities in the superior temporal retinal arteriole (STRA), superior nasal retinal arteriole (SNRA), and the retinal and choroidal vessels (RCV) were measured using laser-speckle flowgraphy. Blood flow in the STRA and SNRA was calculated from the integral of a cross-sectional map of blood velocity. PETCO2 was clamped at the normothermia level by adding 5% CO2 to the inspired gas. Passive heat stress had no effect on the subjects' blood pressures. The blood-flow velocity in the RCV was significantly lower after 30, 60 and 90 min of passive heat stress than the normothermic level, with a peak decrease of 18 ± 3% (mean ± SE) at 90 min. Blood flow in the STRA and SNRA decreased significantly after 90 min of passive heat stress conditions, with peak decreases of 14 ± 3% and 14 ± 4%, respectively. The findings of this study suggest that passive heat stress decreases ocular blood flow irrespective of the blood pressure or arterial partial pressure of CO2.
Heat flow in the Izu-Ogasawara (Bonin)-Mariana arc
Energy Technology Data Exchange (ETDEWEB)
Yamazaki, T. (Geological Survey of Japan, Tsukuba (Japan))
1992-04-27
New heat flow data of 218 points obtained from 1984 to 1989 in the Izu-Ogasawara(Bonin)-Mariana Arc and their interpretation are presented. The purpose of the measurements is to evaluate the potential of hydrothermal activity and to increase an understanding of the tectonics in this region. The Sumisu Rift shows high and variable heat flow values, which suggests the existence of hydrothermal circulation. The Nishinoshima Trough is bounded on the west by a steep fault scarp, and heat flow is higher in the trough than to the west of the scarp. This contrast was considered to reflect difference in crustal thickness caused by old rifting. Local high heat flow associated with an intrusive body is found in the trough. The Mariana Trough north of 22{degree}N, which is currently in a rifting stage, shows thermal asymmetry. In the Mariana Trough south of 22{degree}N, which has developed to a spreading stage, anomalous temperature profiles suggesting hydrothermal circulation are found near the spreading centers. 44 refs., 11 figs., 1 tab.
The effect of buoyancy on flow and heat transfer in curved pipes
Mochizuki, Munekazu; Ishigaki, Hiroshi; 望月 宗和; 石垣 博
1994-01-01
Fully developed laminar flow in a heated horizontal curved pipe is studied numerically. The thermal boundary conditions at the wall are uniform wall heat flux axially and uniform wall temperature peripherally. Flow and heat transfer are governed by Dean number, Prandtl number and buoyancy number. Detailed prediction of the friction factor, average heat transfer rate, velocity profile, temperature profile and secondary-flow streamlines are given.
Heat transfer to a particle exposed to a rarefield ionized-gas flow
International Nuclear Information System (INIS)
Chen, X.; He, P.
1986-01-01
Analytical results are presented concerning the heat transfer to a spherical particle exposed to a high temperature, ionized- gas flow for the extreme case of free-molecule flow regime. It has been shown that the presence of relative velocity between the particle and the ionized gas reduces the floating potential on the particle, enhances the heat flux and causes appreciably non-uniform distribution of the local heat flux. Pronounced difference is found between metallic and non-metallic particles in the floating potential and the local heat flux distributions, in particular for the case with high gas-flow temperature. Relative contribution of atoms to the total heat flux is dominant for the case of low gas-flow temperature, while the heat flux is mainly caused by ions and electrons for the case of high gas-flow temperature
International Nuclear Information System (INIS)
Hata, K.; Fukuda, K.; Masuzaki, S.
2014-01-01
The subcooled boiling heat transfer and the steady state critical heat flux (CHF) in a vertical circular tube for the flow velocities (u=3.95 to 30.80 m/s) are systematically measured by the experimental water loop comprised of a multistage canned-type circulation pump with high pump head. The SUS304 test tube of inner diameter (d=6 mm) and heated length (L=59.5 mm) is used in this work. The outer surface temperatures of the SUS304 test tube with heating are observed by an infrared thermal imaging camera and a video camera. The subcooled boiling heat transfers for SUS304 test tube are compared with the values calculated by other workers' correlations for the subcooled boiling heat transfer. The influence of flow velocity on the subcooled boiling heat transfer and the CHF is investigated into details based on the experimental data. Nucleate boiling surface superheats at the CHF are close to the lower limit of the heterogeneous spontaneous nucleation temperature and the homogeneous spontaneous nucleation temperature. The dominant mechanism of the subcooled flow boiling CHF on the SUS304 circular tube is discussed at high liquid Reynolds number. On the other hand, theoretical equations for k-ε turbulence model in a circular tube of a 3 mm in diameter and a 526 mm long are numerically solved for heating of water on heated section of a 3 mm in diameter and a 67 mm long with various thicknesses of conductive sub-layer by using PHOENICS code under the same conditions as the experimental ones previously obtained considering the temperature dependence of thermo-physical properties concerned. The Platinum (Pt) test tube of inner diameter (d=3 mm) and heated length (L=66.5 mm) was used in this experiment. The thicknesses of conductive sub-layer from non-boiling regime to CHF are clarified. The thicknesses of conductive sub-layer at the CHF point are evaluated for various flow velocities. The experimental values of the CHF are also compared with the corresponding
Auxiliary plasma heating and fueling models for use in particle simulation codes
International Nuclear Information System (INIS)
Procassini, R.J.; Cohen, B.I.
1989-01-01
Computational models of a radiofrequency (RF) heating system and neutral-beam injector are presented. These physics packages, when incorporated into a particle simulation code allow one to simulate the auxiliary heating and fueling of fusion plasmas. The RF-heating package is based upon a quasilinear diffusion equation which describes the slow evolution of the heated particle distribution. The neutral-beam injector package models the charge exchange and impact ionization processes which transfer energy and particles from the beam to the background plasma. Particle simulations of an RF-heated and a neutral-beam-heated simple-mirror plasma are presented. 8 refs., 5 figs
An experimental and analytical study of fluid flow and critical heat flux in PWR fuel elements
International Nuclear Information System (INIS)
Bowditch, F.H.; Mogford, D.J.
1987-02-01
This report describes experiments that have been carried out at the Winfrith Establishment of the United Kingdom Atomic Energy Authority to determine the critical heat flux characteristics of pressurized water reactor fuel elements over an unusually wide range of coolant flow conditions that are relevant to both normal and fault conditions of reactor operation. The experiments were carried out in the TITAN loop using an electrically heated bundle of 25 rods of 9.5 mm diameter on a 12.7 mm pitch fitted with plain grids in order to provide a generic base for code validation. The fully tabulated experimental data for critical heat flux, pressure drop and sub-channel mixing are encompassed by ranges of pressure between 20 and 160 Bar, coolant flow between 150 and 3600 Kg/m 2 s, and coolant inlet temperature between 150 and 320 0 C. The results of the experiments are compared with predicted data based upon several established critical heat flux correlations. It is concluded that the extrapolation of some correlations to conditions beyond their intended range of application can lead to dangerous over estimates of critical heat flux, but the Winfrith WSC-2 and the EPRI NP-2609 correlations perform well over the whole data range and correlate all data with RMS errors of 9% and 6% respectively. (author)
An experimental study on micro-scale flow boiling heat transfer
International Nuclear Information System (INIS)
Tibirica, Cristiano Bigonha; Ribatski, Gherhardt
2009-01-01
In this paper, new experimental flow boiling heat transfer results in micro-scale tubes are presented. The experimental data were obtained in a horizontal 2.32 mm I.D. stainless steel tube with heating length of 464 mm, R134a as working fluid, mass velocities ranging from 50 to 600 kg/m 2 s, heat flux from 5 to 55 kW/m 2 , exit saturation temperatures of 22, 31 and 41 deg C, and vapor qualities from 0.05 to 0.98. Flow pattern characterization was also performed from images obtained by high speed filming. Heat transfer coefficient results from 2 to 14 kW/m 2 K were measured. It was found that the heat transfer coefficient is a strong function of the saturation pressure, heat flux, mass velocity and vapor quality. The experimental data were compared against the following micro-scale flow boiling predictive methods from the literature: Saitoh et al., Kandlikar, Zhang et al. and Thome et al. Comparisons against these methods based on the data segregated according to flow patterns were also performed. Though not satisfactory, Saitoh et al. worked the best and was able of capturing most of the experimental heat transfer trends. (author)
An experimental study on micro-scale flow boiling heat transfer
Energy Technology Data Exchange (ETDEWEB)
Tibirica, Cristiano Bigonha; Ribatski, Gherhardt [Universidade de Sao Paulo (USP), Sao Carlos, SP (Brazil). Escola de Engenharia. Dept. de Engenharia Mecanica
2009-07-01
In this paper, new experimental flow boiling heat transfer results in micro-scale tubes are presented. The experimental data were obtained in a horizontal 2.32 mm I.D. stainless steel tube with heating length of 464 mm, R134a as working fluid, mass velocities ranging from 50 to 600 kg/m{sup 2}s, heat flux from 5 to 55 kW/m{sup 2}, exit saturation temperatures of 22, 31 and 41 deg C, and vapor qualities from 0.05 to 0.98. Flow pattern characterization was also performed from images obtained by high speed filming. Heat transfer coefficient results from 2 to 14 kW/m{sup 2}K were measured. It was found that the heat transfer coefficient is a strong function of the saturation pressure, heat flux, mass velocity and vapor quality. The experimental data were compared against the following micro-scale flow boiling predictive methods from the literature: Saitoh et al., Kandlikar, Zhang et al. and Thome et al. Comparisons against these methods based on the data segregated according to flow patterns were also performed. Though not satisfactory, Saitoh et al. worked the best and was able of capturing most of the experimental heat transfer trends. (author)
Stretched flow of Oldroyd-B fluid with Cattaneo-Christov heat flux
Directory of Open Access Journals (Sweden)
T. Hayat
Full Text Available The objective of present attempt is to analyse the flow and heat transfer in the flow of an Oldroyd-B fluid over a non-linear stretching sheet having variable thickness. Characteristics of heat transfer are analyzed with temperature dependent thermal conductivity and heat source/sink. Cattaneo-Christov heat flux model is considered rather than Fourier’s law of heat conduction in the present flow analysis. Thermal conductivity varies with temperature. Resulting partial differential equations through laws of conservation of mass, linear momentum and energy are converted into ordinary differential equations by suitable transformations. Convergent series solutions for the velocity and temperature distributions are developed and discussed. Keywords: Oldroyd-B fluid, Variable sheet thickness, Cattaneo-Christov heat flux model, Heat source/sink, Temperature dependent thermal conductivity
Decay heat experiment and validation of calculation code systems for fusion reactor
Energy Technology Data Exchange (ETDEWEB)
Maekawa, Fujio; Ikeda, Yujiro [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Wada, Masayuki
1999-10-01
Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of {+-}10%. (author)
Decay heat experiment and validation of calculation code systems for fusion reactor
International Nuclear Information System (INIS)
Maekawa, Fujio; Ikeda, Yujiro; Wada, Masayuki
1999-10-01
Although accurate estimation of decay heat value is essential for safety analyses of fusion reactors against loss of coolant accidents and so on, no experimental work has been devoted to validating the estimation. Hence, a decay heat measurement experiment was performed as a task (T-339) of ITER/EDA. A new detector, the Whole Energy Absorption Spectrometer (WEAS), was developed for accurate and efficient measurements of decay heat. Decay heat produced in the thirty-two sample materials which were irradiated by 14-MeV neutrons at FNS/JAERI were measured with WEAS for a wide cooling time period from 1 min to 400 days. The data presently obtained were the first experimental decay heat data in the field of fusion. Validity of decay heat calculation codes of ACT4 and CINAC-V4, activation cross section libraries of FENDL/A-2.0 and JENDL Activation File, and decay data was investigated through analyses of the experiment. As a result, several points that should be modified were found in the codes and data. After solving the problems, it was demonstrated that decay heat valued calculated for most of samples were in good agreement with the experimental data. Especially for stainless steel 316 and copper, which were important materials for ITER, decay heat could be predicted with accuracy of ±10%. (author)
Fluid flow and heat transfer in rotating porous media
Vadasz, Peter
2016-01-01
This Book concentrates the available knowledge on rotating fluid flow and heat transfer in porous media in one single reference. Dr. Vadasz develops the fundamental theory of rotating flow and heat transfer in porous media and introduces systematic classification and identification of the relevant problems. An initial distinction between rotating flows in isothermal heterogeneous porous systems and natural convection in homogeneous non-‐isothermal porous systems provides the two major classes of problems to be considered. A few examples of solutions to selected problems are presented, highlighting the significant impact of rotation on the flow in porous media.
International Nuclear Information System (INIS)
Wang Junfeng; Huang Yanping; Wang Yanlin; Song Mingliang
2012-01-01
Highlights: ► Flow regimes were visually investigated in a heated narrow rectangular channel. ► Bubbly, churn, and annular flow were observed. Slug flow was never observed. ► Flow regime transition boundary could be predicted by existing criteria. ► Churn zone in present flow regime maps were poorly predicted by existing criteria. - Abstract: Flow regimes are very important in understanding two-phase flow resistance and heat transfer characteristics. In present work, two-phase flow regimes for steam–water flows in a single-side heated narrow rectangular channel, having a width of 40 mm and a gap of 3 mm, were visually studied at relatively low pressure and low mass flux condition. The flow regimes observed in this experiment could be classified into bubbly, churn and annular flow. Slug flow was never observed at any of the conditions in our experiment. Flow regime maps at the pressure of 0.7 MPa and 1.0 MPa were developed, and then the pressure effect on flow regime transition was analyzed. Based on the experimental results, the comparisons with some existing flow regime maps and transition criteria were conducted. The comparison results show that the bubbly transition boundary and annular formation boundary of heated steam–water flow were consistent with that of adiabatic air–water flow. However, the intermediate flow pattern between bubbly and annular flow was different. Hibiki and Mishima criteria could predict the bubbly transition boundary and annular formation boundary satisfactorily, but it poorly predicted churn zone in present experimental data.
Turbulence modeling and surface heat transfer in a stagnation flow region
Wang, C. R.; Yeh, F. C.
1987-01-01
Analysis for the turbulent flow field and the effect of freestream turbulence on the surface heat transfer rate of a stagnation flow is presented. The emphasis is on modeling and its augmentation of surface heat transfer rate. The flow field considered is the region near the forward stagnation point of a circular cylinder in a uniform turbulent mean flow.
Abnormal high surface heat flow caused by the Emeishan mantle plume
Jiang, Qiang; Qiu, Nansheng; Zhu, Chuanqing
2016-04-01
It is commonly believed that increase of heat flow caused by a mantle plume is small and transient. Seafloor heat flow data near the Hawaiian hotspot and the Iceland are comparable to that for oceanic lithosphere elsewhere. Numerical modeling of the thermal effect of the Parana large igneous province shows that the added heat flow at the surface caused by the magmatic underplating is less than 5mW/m2. However, the thermal effect of Emeishan mantle plume (EMP) may cause the surface hear-flow abnormally high. The Middle-Late Emeishan mantle plume is located in the western Yangtze Craton. The Sichuan basin, to the northeast of the EMP, is a superimposed basin composed of Paleozoic marine carbonate rocks and Mesozoic-Cenozoic terrestrial clastic rocks. The vitrinite reflectance (Ro) data as a paleogeothermal indicator records an apparent change of thermal regime of the Sichuan basin. The Ro profiles from boreholes and outcrops which are close to the center of the basalt province exhibit a 'dog-leg' style at the unconformity between the Middle and Upper Permian, and they show significantly higher gradients in the lower subsection (pre-Middle Permian) than the Upper subsection (Upper Permian to Mesozoic). Thermal history inversion based on these Ro data shows that the lower subsection experienced a heat flow peak much higher than that of the upper subsection. The abnormal heat flow in the Sichuan basin is consistent with the EMP in temporal and spatial distribution. The high-temperature magmas from deep mantle brought heat to the base of the lithosphere, and then large amount of heat was conducted upwards, resulting in the abnormal high surface heat flow.
Experimental study of supercritical water flow and heat transfer in vertical tube
International Nuclear Information System (INIS)
Li Hongbo; Yang Jue; Lu Donghua; Gu Hanyang; Zhao Meng
2012-01-01
The experiment of flow and heat transfer of supercritical water has been performed on the supercritical water multipurpose test loop co-constructed by China Guangdong Nuclear Power Group and Shanghai Jiao Tong University with a 7.6 mm vertical tube. Heat transfer experimental data is obtained. The results of experimental research of thermal-hydraulic parameters on flow and heat transfer of supercritical water show that: (1) Heat transfer enhancement occurs when the bulk temperature reaches pseudo-critical point with low mass flow velocity; (2) The heat transfer co- efficient and Nusselt number are decreased with the increasing of heat flux; (3) The wall temperature is decreased, but the heat transfer coefficient and Nusselt number are increased with the increasing of mass flow velocity; (4) The wall temperature is increased, but the heat transfer coefficient and Nusselt number are decreased with the increasing of sys- tem pressure. (authors)
Responding to the Effects of Extreme Heat: Baltimore City's Code Red Program.
Martin, Jennifer L
2016-01-01
Heat response plans are becoming increasingly more common as US cities prepare for heat waves and other effects of climate change. Standard elements of heat response plans exist, but plans vary depending on geographic location and distribution of vulnerable populations. Because heat events vary over time and affect populations differently based on vulnerability, it is difficult to compare heat response plans and evaluate responses to heat events. This article provides an overview of the Baltimore City heat response plan, the Code Red program, and discusses the city's response to the 2012 Ohio Valley/Mid Atlantic Derecho, a complex heat event. Challenges with and strategies for evaluating the program are reviewed and shared.
Experimental determination of heat transfer in a Poiseuille-Rayleigh-Bénard flow
Taher, R.; Abid, C.
2018-05-01
This paper deals with an experimental study of heat transfer in a Poiseuille-Rayleigh-Bénard flow. This situation corresponds to a mixed convection phenomenon in a horizontal rectangular channel uniformly heated from below. Flow visualisation and temperature measurements were achieved in order to describe the flow regimes and heat transfer behaviour. The classical measurement techniques such employing thermocouples give local measurement on one hand and on other hand they often disturb the flow. As the flow is three-dimensional, these techniques are not efficient. In order to not disturb the flow, a non-intrusive method is used for thermal measurement. The Planar laser Induced Fluorescence (PLIF) was implemented to determine thermal fields in the fluid. Experiments conducted for various Reynolds and Rayleigh numbers allow to determine the heat transfer and thus to propose correlation for Nusselt number for a mixed convection flow in Poiseuille-Rayleigh-Bénard configuration. First a description of the use of this technique in water flow is presented and then the obtained results for various Reynolds and Rayleigh numbers allow to propose a correlation for the Nusselt number for such configuration of mixed convection. The comparison between the obtained heat transfer and the pure forced convection one confirms the well-known result that the convective heat transfer is greatly enhanced in mixed convection. Indeed, secondary flow induced by buoyant forces contributes to the refreshment of thermal boundary layers and so acts like mixers, which significantly enhances heat transfer.
Energy Technology Data Exchange (ETDEWEB)
Ranganayakulu, C. [Aeronautical Development Agency, Bangalore (India); Seetharamu, K.N. [School of Mechanical Engineering, Univ. of Southern Malaysia (KCP), Tronoh (Malaysia)
2000-05-01
An analysis of a crossflow plate-fin compact heat exchanger, accounting for the combined effect of two-dimensional longitudinal heat conduction through the exchanger wall and nonuniform inlet fluid flow distribution on both hot and cold fluid sides is carried out using a finite element method. Using the fluid flow maldistribution models, the exchanger effectiveness and its deterioration due to the combined effects of longitudinal heat conduction and flow nonuniformity are calculated for various design and operating conditions of the exchanger. It was found that the performance deteriorations are quite significant in some typical applications due to the combined effects of wall longitudinal heat conduction and inlet fluid flow nonuniformity on crossflow plate-fin heat exchanger. (orig.)
Laminar fluid flow and heat transfer in a fin-tube heat exchanger with vortex generators
Energy Technology Data Exchange (ETDEWEB)
Yanagihara, J.I.; Rodriques, R. Jr. [Polytechnic School of Univ. of Sao Paolo, Sao Paolo (Brazil). Dept. of Mechanical Engineering
1996-12-31
Development of heat transfer enhancement techniques for fin-tube heat exchangers has great importance in industry. In recent years, heat transfer augmentation by vortex generators has been considered for use in plate fin-tube heat exchangers. The present work describes a numerical investigation about the influence of delta winglet pairs of vortex generators on the flow structure and heat transfer of a plate fin-tube channel. The Navier-Stokes and Energy equations are solved by the finite volume method using a boundary-fitted coordinate system. The influence of vortex generators parameters such as position, angle of attack and aspect ratio were investigated. Local and global influences of vortex generators in heat transfer and flow losses were analyzed by comparison with a model using smooth fin. The results indicate great advantages of this type of geometry for application in plate fin-tube heat exchangers, in terms of large heat transfer enhancement and small pressure loss penalty. (author)
Laminar fluid flow and heat transfer in a fin-tube heat exchanger with vortex generators
Energy Technology Data Exchange (ETDEWEB)
Yanagihara, J I; Rodriques, R Jr [Polytechnic School of Univ. of Sao Paolo, Sao Paolo (Brazil). Dept. of Mechanical Engineering
1997-12-31
Development of heat transfer enhancement techniques for fin-tube heat exchangers has great importance in industry. In recent years, heat transfer augmentation by vortex generators has been considered for use in plate fin-tube heat exchangers. The present work describes a numerical investigation about the influence of delta winglet pairs of vortex generators on the flow structure and heat transfer of a plate fin-tube channel. The Navier-Stokes and Energy equations are solved by the finite volume method using a boundary-fitted coordinate system. The influence of vortex generators parameters such as position, angle of attack and aspect ratio were investigated. Local and global influences of vortex generators in heat transfer and flow losses were analyzed by comparison with a model using smooth fin. The results indicate great advantages of this type of geometry for application in plate fin-tube heat exchangers, in terms of large heat transfer enhancement and small pressure loss penalty. (author)
Ice slurry flow and heat transfer during flow through tubes of rectangular and slit cross-sections
Directory of Open Access Journals (Sweden)
Niezgoda-Żelasko Beata
2014-09-01
Full Text Available The paper presents the results of experimental research of pressure drop and heat transfer coefficients of ice slurry during its flow through tubes of rectangular and slit cross-sections. Moreover, the work discusses the influence of solid particles, type of motion and cross-section on the changes in the pressure drop and heat transfer coefficient. The analysis presented in the paper allows for identification of the criterial relations used to calculate the Fanning factor and the Nusselt number for laminar and turbulent flow, taking into account elements such as phase change, which accompanies the heat transfer process. Ice slurry flow is treated as a generalized flow of a non-Newtonian fluid.
Heat transfer to air-water two-phase flow in slug/churn region
International Nuclear Information System (INIS)
Wadekar, V.V.; Tuzla, K.; Chen, J.C.
1996-01-01
Measured heat transfer data for air-water two-phase flow in the slug/churn flow region are reported. The measurements were obtained from a 1.3 m tall, 15.7 mm diameter vertical tube test-section. It is observed that the data exhibit different heat transfer characteristics to those predicted by the standard correlations for the convective component of flow boiling heat transfer. Comparison with the predictions of a slug flow model for evaporation shows a significant overprediction of the data. The reason for the overprediction is attributed to the sensible heating requirement of the gas phase. The slug flow model is therefore suitably modified for non-evaporating two-phase flow. This specially adapted model is found to give reasonably good predictions of the measured data
CFD Code Validation against Stratified Air-Water Flow Experimental Data
International Nuclear Information System (INIS)
Terzuoli, F.; Galassi, M.C.; Mazzini, D.; D'Auria, F.
2008-01-01
Pressurized thermal shock (PTS) modelling has been identified as one of the most important industrial needs related to nuclear reactor safety. A severe PTS scenario limiting the reactor pressure vessel (RPV) lifetime is the cold water emergency core cooling (ECC) injection into the cold leg during a loss of coolant accident (LOCA). Since it represents a big challenge for numerical simulations, this scenario was selected within the European Platform for Nuclear Reactor Simulations (NURESIM) Integrated Project as a reference two-phase problem for computational fluid dynamics (CFDs) code validation. This paper presents a CFD analysis of a stratified air-water flow experimental investigation performed at the Institut de Mecanique des Fluides de Toulouse in 1985, which shares some common physical features with the ECC injection in PWR cold leg. Numerical simulations have been carried out with two commercial codes (Fluent and Ansys CFX), and a research code (NEPTUNE CFD). The aim of this work, carried out at the University of Pisa within the NURESIM IP, is to validate the free surface flow model implemented in the codes against experimental data, and to perform code-to-code benchmarking. Obtained results suggest the relevance of three-dimensional effects and stress the importance of a suitable interface drag modelling
CFD Code Validation against Stratified Air-Water Flow Experimental Data
Directory of Open Access Journals (Sweden)
F. Terzuoli
2008-01-01
Full Text Available Pressurized thermal shock (PTS modelling has been identified as one of the most important industrial needs related to nuclear reactor safety. A severe PTS scenario limiting the reactor pressure vessel (RPV lifetime is the cold water emergency core cooling (ECC injection into the cold leg during a loss of coolant accident (LOCA. Since it represents a big challenge for numerical simulations, this scenario was selected within the European Platform for Nuclear Reactor Simulations (NURESIM Integrated Project as a reference two-phase problem for computational fluid dynamics (CFDs code validation. This paper presents a CFD analysis of a stratified air-water flow experimental investigation performed at the Institut de Mécanique des Fluides de Toulouse in 1985, which shares some common physical features with the ECC injection in PWR cold leg. Numerical simulations have been carried out with two commercial codes (Fluent and Ansys CFX, and a research code (NEPTUNE CFD. The aim of this work, carried out at the University of Pisa within the NURESIM IP, is to validate the free surface flow model implemented in the codes against experimental data, and to perform code-to-code benchmarking. Obtained results suggest the relevance of three-dimensional effects and stress the importance of a suitable interface drag modelling.
Energy Technology Data Exchange (ETDEWEB)
Borges, Eduardo M.; Sabundjian, Gaiane, E-mail: borges.em@hotmail.com, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)
2017-07-01
The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2. gamma code in Angra-2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 250cm{sup 2} of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of Angra-2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of Angra-2 during the postulated accident. The results obtained for Angra-2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core. (author)
Analysis of counter flow of corona wind for heat transfer enhancement
Shin, Dong Ho; Baek, Soo Hong; Ko, Han Seo
2018-03-01
A heat sink for cooling devices using the counter flow of a corona wind was developed in this study. Detailed information about the numerical investigations of forced convection using the corona wind was presented. The fins of the heat sink using the counter flow of a corona wind were also investigated. The corona wind generator with a wire-to-plate electrode arrangement was used for generating the counter flow to the fin. The compact and simple geometric characteristics of the corona wind generator facilitate the application of the heat sink using the counter flow, demonstrating the heat sink is effective for cooling electronic devices. Parametric studies were performed to analyze the effect of the counter flow on the fins. Also, the velocity and temperature were measured experimentally for the test mock-up of the heat sink with the corona wind generator to verify the numerical results. From a numerical study, the type of fin and its optimal height, length, and pitch were suggested for various heat fluxes. In addition, the correlations to calculate the mass of the developed heat sink and its cooling performance in terms of the heat transfer coefficient were derived. Finally, the cooling efficiencies corresponding to the mass, applied power, total size, and noise of the devices were compared with the existing commercial central processing unit (CPU) cooling devices with rotor fans. As a result, it was confirmed that the heat sink using the counter flow of the corona wind showed appropriate efficiencies for cooling electronic devices, and is a suitable replacement for the existing cooling device for high power electronics.
Studies of heat transport to forced-flow He II
International Nuclear Information System (INIS)
Dresner, L.; Kashani, A.; Van Sciver, S.W.
1985-01-01
Analytical and experimental studies of heat transport to forced-flow He II are reported. The work is pertinent to the transfer of He II in space. An analytical model has been developed that establishes a condition for two-phase flow to occur in the transfer line. This condition sets an allowable limit to the heat leak into the transfer line. Experimental measurements of pressure drop and flow meter performances indicate that turbulent He II can be analyzed in terms of classical pressure drop correlations
CFD simulation on critical heat flux of flow boiling in IVR-ERVC of a nuclear reactor
Energy Technology Data Exchange (ETDEWEB)
Zhang, Xiang, E-mail: zhangxiang3@snptc.com.cn [State Nuclear Power Technology Research & Development Center, South Area, Future Science and Technology Park, Chang Ping District, Beijing 102209 (China); Hu, Teng [State Nuclear Power Technology Research & Development Center, South Area, Future Science and Technology Park, Chang Ping District, Beijing 102209 (China); Chen, Deqi, E-mail: chendeqi@cqu.edu.cn [Key Laboratory of Low-grade Energy Utilization Technologies and Systems, Chongqing University, 400044 (China); Zhong, Yunke; Gao, Hong [Key Laboratory of Low-grade Energy Utilization Technologies and Systems, Chongqing University, 400044 (China)
2016-08-01
Highlights: • CFD simulation on CHF of boiling two-phase flow in ERVC is proposed. • CFD simulation result of CHF agrees well with that of experimental result. • The characteristics of boiling two-phase flow and boiling crisis are analyzed. - Abstract: The effectiveness of in-vessel retention (IVR) by external reactor vessel cooling (ERVC) strongly depends on the critical heat flux (CHF). As long as the local CHF does not exceed the local heat flux, the lower head of the pressure vessel can be cooled sufficiently to prevent from failure. In this paper, a CFD simulation is carried out to investigate the CHF of ERVC. This simulation is performed by a CFD code fluent couple with a boiling model by UDF (User-Defined Function). The experimental CHF of ERVC obtained by State Nuclear Power Technology Research and Development Center (SNPTRD) is used to validate this CFD simulation, and it is found that the simulation result agrees well with the experimental result. Based on the CFD simulation, detailed analysis focusing on the pressure distribution, velocity distribution, void fraction distribution, heating wall temperature distribution are proposed in this paper.
Two-phase flow instability in a liquid nitrogen heat exchanger, 2
International Nuclear Information System (INIS)
Kondoh, Tetsuya; Fukuda, Kenji; Hasegawa, Shu; Yamada, Hidetomo; Ryu, Hiroyuki.
1988-01-01
Experimental and analytical investigations are conducted on flow instability in a vertically installed liquid nitrogen shell and tube type heat exchanger. The experiments are carried out by making use of water steam as a secondary fluid and it is observed that flow instability occurs in the range of small inlet flow rate. Mode analysis of the flow instability oscillation reveals that there exists a fundamental mode and its higher harmonics up to the fourth. As the period of the fundamental mode is nearly equal to the transit time for a fluid particle to travel through the heated tube, it is suggested that this flow instability is of the density wave type. It is shown that the amount of exchanged heat, as well as the pressure drop, decrease when unstable flow oscillation occurs. An analysis of the static heat transfer and pressure drop characteristics can simulate the experimental results in the stable region. Linear stability analysis is also carried out to yield the stability map as well as the period of flow oscillation, which proved to agree with the experimental data qualitatively. (author)
Heat flow at the proposed Appalachian Ultradeep Core Hole (ADCOH) Site: Tectonic implications
Costain, John K.; Decker, Edward R.
The heat flow in northwestern South Carolina at the Appalachian Ultradeep Core Hole (ADCOH) site area is approximately 55 mW/m². This data supplements other data to the east in the Piedmont and Atlantic Coastal Plain provinces where heat flows > 55 mW/m² are characteristic of post- and late-synmetamorphic granitoids. Piedmont heat flow and heat generation data for granites, metagranites, and one Slate Belt site, in a zone approximately parallel to major structural Appalachian trends, define a linear relation. Tectonic truncation of heat-producing crust at a depth of about 8 km (a depth equal to the slope of the heat flow-heat production line) is proposed to explain the linear relation. Using the value of reduced heat flow estimated from this empirical relation, and assuming thicknesses of heat-producing crust defined by new ADCOH seismic data, the heat flow and heat production at the ADCOH site are consistent with a depth to the base of the Inner Piedmont crystalline allochthon of about 5.5 km. Seismic data at the ADCOH site confirm that the Inner Piedmont is tectonically truncated at about 5.5 km by the Blue Ridge master decollement. Temperatures at 10 km at the ADCOH site are predicted to be less than 200 °C.
Heat transfer and pressure drop in flow boiling in microchannels
Saha, Sujoy Kumar
2016-01-01
This Brief addresses the phenomena of heat transfer and pressure drop in flow boiling in micro channels occurring in high heat flux electronic cooling. A companion edition in the Springer Brief Subseries on Thermal Engineering and Applied Science to “Critical Heat Flux in Flow Boiling in Micro channels,” by the same author team, this volume is idea for professionals, researchers and graduate students concerned with electronic cooling.
Effect of regional heating on the liver blood flow in rats
International Nuclear Information System (INIS)
Nakajima, T.; Song, C.W.; Osborn, J.L.; Rhee, J.G.; Levitt, S.H.
1987-01-01
The authors measured the blood flow in the liver of rats heated with a radio frequency capacitive heating device. The blood flow through the hepatic artery, as measured with the radioactive microsphere method, was 0.21 ml/min/gm; it increased by 13% and 16% when heated for 15 minutes at 41 0 C and 43 0 C, respectively. The portal vein blood flow was 1.09 ml/min/gm and decreased by 12% and 20% on heating for 15 minutes at 41 0 C and 43 0 C, respectively. The total liver blood flow, therefore, decreased by 11% at 41 0 C and by 14% at 43 0 C from the control value of 1.30 ml/min/gm
Development of a computer code for Dalat research reactor transient analysis
International Nuclear Information System (INIS)
Le Vinh Vinh; Nguyen Thai Sinh; Huynh Ton Nghiem; Luong Ba Vien; Pham Van Lam; Nguyen Kien Cuong
2003-01-01
DRSIM (Dalat Reactor SIMulation) computer code has been developed for Dalat reactor transient analysis. It is basically a coupled neutronics-hydrodynamics-heat transfer code employing point kinetics, one dimensional hydrodynamics and one dimensional heat transfer. The work was financed by VAEC and DNRI in the framework of institutional R and D programme. Some transient problems related to reactivity and loss of coolant flow was carried out by DRSIM using temperature and void coefficients calculated by WIMS and HEXNOD2D codes. (author)
Development of flow network analysis code for block type VHTR core by linear theory method
International Nuclear Information System (INIS)
Lee, J. H.; Yoon, S. J.; Park, J. W.; Park, G. C.
2012-01-01
VHTR (Very High Temperature Reactor) is high-efficiency nuclear reactor which is capable of generating hydrogen with high temperature of coolant. PMR (Prismatic Modular Reactor) type reactor consists of hexagonal prismatic fuel blocks and reflector blocks. The flow paths in the prismatic VHTR core consist of coolant holes, bypass gaps and cross gaps. Complicated flow paths are formed in the core since the coolant holes and bypass gap are connected by the cross gap. Distributed coolant was mixed in the core through the cross gap so that the flow characteristics could not be modeled as a simple parallel pipe system. It requires lot of effort and takes very long time to analyze the core flow with CFD analysis. Hence, it is important to develop the code for VHTR core flow which can predict the core flow distribution fast and accurate. In this study, steady state flow network analysis code is developed using flow network algorithm. Developed flow network analysis code was named as FLASH code and it was validated with the experimental data and CFD simulation results. (authors)
International Nuclear Information System (INIS)
Yasuo Koizumi; Hiroyasu Ohtake; Tomoyuki Suzuki
2002-01-01
The influence of particle introduction into a subcooled water flow on boiling heat transfer and critical heat flux (CHF) was examined. When the water velocity was low, the particles crowded on the bottom wall of the flow channel and flowed just like sliding on the wall. When the water velocity was high, the particles were well dispersed in the water flow. In the non-boiling region, the heat transfer was augmented by the introduction of the particles into the water flow. As the introduction of the particles were increased, the augmentation was also increased in the high water flow rate region. However, it was independent upon the particle introduction rate in the low water flow rate region. The onset of boiling was delayed by the particle inclusion. The boiling heat transfer was enhanced by the particles. However, it was rather decreased in the high heat flux fully-developed-boiling region. The CHF was decreased by the particle inclusion in the low water flow region and was not affected in the high water flow region. (authors)
Heat transfer of pulsating laminar flow in pipes with wall thermal inertia
International Nuclear Information System (INIS)
Yuan, Hongsheng; Tan, Sichao; Wen, Jing; Zhuang, Nailiang
2016-01-01
The effects of wall thermal inertia on heat transfer of pulsating laminar flow with constant power density within the pipe wall are investigated theoretically. The energy equation of the fully developed flow and heat transfer is solved by separation of variables and Green's function. The effects of the pulsation amplitude and frequency, the Prandtl number and the wall heat capacity on heat transfer features characterized by temperature, heat flux and Nusselt number are analyzed. The results show that the oscillation of wall heat flux increases along with the wall thermal inertia, while the oscillation of temperature and Nusselt number is suppressed by the wall thermal inertia. The influence of pulsation on the average Nusselt number is also obtained. The pulsating laminar flow can reduce the average Nusselt number. The Nusselt number reduction of pipe flow are a little more remarkable than that of flow between parallel plates, which is mainly caused by differences in hydraulic and thermal performances of the channels. (authors)
A critical flow model for the Cathena thermalhydraulic code
International Nuclear Information System (INIS)
Popov, N.K.; Hanna, B.N.
1990-01-01
The calculation of critical flow rate, e.g., of choked flow through a break, is required for simulating a loss of coolant transient in a reactor or reactor-like experimental facility. A model was developed to calculate the flow rate through the break for given geometrical parameters near the break and fluid parameters upstream of the break for ordinary water, as well as heavy water, with or without non- condensible gases. This model has been incorporated in the CATHENA, one-dimensional, two-fluid thermalhydraulic code. In the CATHENA code a standard staggered-mesh, finite-difference representation is used to solve the thermalhydraulic equations. This model compares the fluid mixture velocity, calculated using the CATHENA momentum equations, with a critical velocity. When the mixture velocity is smaller than the critical velocity, the flow is assumed to be subcritical, and the model remains passive. When the fluid mixture velocity is higher than the critical velocity, the model sets the fluid mixture velocity equal to the critical velocity. In this paper the critical velocity at a link (momentum cell) is first estimated separately for single-phase liquid, two- phase, or single-phase gas flow condition at the upstream node (mass/energy cell). In all three regimes non-condensible gas can be present in the flow. For single-phase liquid flow, the critical velocity is estimated using a Bernoulli- type of equation, the pressure at the link is estimated by the pressure undershoot method
Characterizing the Heat Flow from Between Enceladus' Tiger Stripes
Howett, C.; Spencer, J. R.; Verbiscer, A.
2017-12-01
Enceladus' heat flow provides a fundamental constraint on its tidal dissipation mechanisms, orbital evolution, and the physical processes that generate the plumes. Determining the total amount of emission is proving difficult, as different techniques produce differing constraints. For example, an initial estimate of this value, 5.8±1.3 GW, was made by Spencer et al. (2006) using Cassini Composite Infrared Spectrometer (CIRS) 600 to 1100 cm-1 observations, which was refined using 10 to 600 cm-1 CIRS observations to 15.8±3.1 GW by Howett et al. (2011). However, recent reanalysis of high-spatial resolution 10 to 1100 cm-1 CIRS observations of Enceladus' active south polar region conducted by Spencer and Howett gives a heat flow of 4.64±0.23 GW. Whilst all of these heat flow estimates are much larger than those expected in a steady state, 1.1 GW (Meyer and Wisdom, 2007), their obvious discrepancy is a puzzle. In this work we seek to help understand these discrepancies by determining how much endogenic heat flow is coming from the funiscular terrain between Enceladus active tiger stripes.
Dynamic modelling for two-phase flow systems
International Nuclear Information System (INIS)
Guerra, M.A.
1991-06-01
Several models for two-phase flow have been studied, developing a thermal-hydraulic analysis code with one of these models. The program calculates, for one-dimensional cases with variable flow area, the transient behaviour of system process variables, when the boundary conditions (heat flux, flow rate, enthalpy and pressure) are functions of time. The modular structure of the code, eases the program growth. In fact, the present work is the basis for a general purpose accident and transient analysis code in nuclear reactors. Code verification has been made against RETRAN-02 results. Satisfactory results have been achieved with the present version of the code. (Author) [es
Comparative simulation of Stirling and Sibling cycle cryocoolers with two codes
International Nuclear Information System (INIS)
Mitchell, M.P.; Wilson, K.J.; Bauwens, L.
1989-01-01
The authors present a comparative analysis of Stirling and Sibling Cycle cryocoolers conducted with two different computer simulation codes. One code (CRYOWEISS) performs an initial analysis on the assumption of isothermal conditions in the machines and adjusts that result with decoupled loss calculations. The other code (MS*2) models fluid flows and heat transfers more realistically but ignores significant loss mechanisms, including flow friction and heat conduction through the metal of the machines. Surprisingly, MS*2 is less optimistic about performance of all machines even though it ignores losses that are modelled by CRYOWEISS. Comparison between constant-bore Stirling and Sibling machines shows that their performance is generally comparable over a range of temperatures, pressures and operating speeds. No machine was consistently superior or inferior according to both codes over the whole range of conditions studied
Numerical prediction of flow, heat transfer, turbulence and combustion
Spalding, D Brian; Pollard, Andrew; Singhal, Ashok K
1983-01-01
Numerical Prediction of Flow, Heat Transfer, Turbulence and Combustion: Selected Works of Professor D. Brian Spalding focuses on the many contributions of Professor Spalding on thermodynamics. This compilation of his works is done to honor the professor on the occasion of his 60th birthday. Relatively, the works contained in this book are selected to highlight the genius of Professor Spalding in this field of interest. The book presents various research on combustion, heat transfer, turbulence, and flows. His thinking on separated flows paved the way for the multi-dimensional modeling of turbu
Simulation of heat transfer in the unsaturated zone
International Nuclear Information System (INIS)
Zyvoloski, G.
1990-01-01
Heat transfer can play an important role in fluid flow near the emplacement site of high-level nuclear waste. The effects on far- field flow can be important in understanding net moisture fluxes above the repository zone. The convection in the unsaturated zone at the Yucca Mountain site was responsible for this movement. If this is so, then the convection could provide a mechanism for drying the rock above the repository zone and thus provide a buffer for heavy rainfall events. In addition, the convection would increase the movement of gaseous radionuclides such as 14 CO 2 , tritiated water vapor, and 129 I (Weeks, 1987). Because of the complexity of the problem, numerical models were required to calculate gas flow and vapor transport at the site. Kipp previously modeled this problem using the code HST3D. This code represents the flow of a single-phase fluid with both heat- and mass-transfer effects included. Water density and partial pressure effects are accounted for by the virtual temperature method. In this paper, the problem was simulated using the code FEHMN, a finite-element heat- and mass-transfer code being developed for the Yucca Mountain Project. The work described in this paper was done in preparation of the upcoming problem to be formulated for the Performance Assessment Calculation Exercise. 5 refs., 9 figs., 1 tab
Visualisation of heat transfer in 3D unsteady flows
Speetjens, M.F.M.; Steenhoven, van A.A.
2010-01-01
Heat transfer in fluid flows traditionally is examined in terms oftemperature field and heat-transfer coefficients at non-adiabaticwalls. However, heat transfer may alternatively be considered asthe transport of thermal energy by the total convective-conductiveheat flux in a way analogous to the
A review of the heat flow data of NE Morocco
Chiozzi, Paolo; Barkaoui, Alae-Eddine; Rimi, Abdelkrim; Verdoya, Massimo; Zarhloule, Yassine
2016-04-01
The Atlas chain is characterised by a SW-NE trending volcanic belt roughly extending from the Atlantic to the Mediterranean Sea and showing activity that spans in age mainly from Middle Miocene to Quaternary (14.6-0.3 Ma). The geochemical features of volcanism are mostly intraplate and alkaline with the exception of the northeastern termination of the belt where calc-alkaline series crop out. Lithospheric thermal and density models so far proposed, constrained by heat flow, gravity anomalies, geoid, and topography data, show that the Atlas chain is not supported isostatically by a thickened crust and a thin, hot and low-density lithosphere explains the high topography. One of the possible explanations for lithospheric mantle thinning, possibly in relation with the observed alkaline volcanism, is thermal erosion produced by either small-scale convection or activation of a small mantle plume, forming part of a hot and deep mantle reservoir system extending from the Canary Islands. This paper focuses on the several geothermal data available in the northeastern sector of the volcanic belt. The occurrence of an extensive, often artesian, carbonatic reservoir hosting moderately hot groundwater might boost the temperature gradient in the overlying impermeable cover, and consequently mask the deep thermal regime. We therefore revised the available dataset and investigated the contribution of advection. Temperature data available from water and oil wells were reprocessed and analysed in combination with thermal conductivity measurements on a wide set of lithotypes. Data were filtered according to rigid selection criteria, and, in the deeper boreholes, the heat flow was inferred by taking into account the porosity variation with depth and the temperature effect on the matrix and pore-filling fluid conductivity. Moreover, the possible effect of advection was evaluated with simple analytical models which envisage the carbonatic layers as confined aquifers heated by the
Heating limits of boiling downward two-phase flow in parallel channels
International Nuclear Information System (INIS)
Fukuda, Kenji; Kondoh, Tetsuya; Hasegawa, Shu; Sakai, Takaaki.
1989-01-01
Flow characteristics and heating limits of downward two-phase flow in single or parallel multi-channels are investigated experimentally and analytically. The heating section used is made of glass tube, in which the heater tube is inserted, and the flow regime inside it is observed. In single channel experiments with low flow rate conditions, it is found that, initially, gas phase which flows upward against the downward liquid phase flow condenses and diminishes as it flows up being cooled by inflowing liquid. However, as the heating power is increased, some portion of the gas phase reaches the top and accumulates to form an liquid level, which eventually causes the dryout. On the other hand, for high flow rate condition, the flooding at the bottom of the heated section is the cause of the dryout. In parallel multi-channels experiments, reversed (upward) flow which leads to the dryout is observed in some of these channels for low flow rate conditions, while the situation is the same to the single channel case for high flow rate conditions. Analyses are carried out to predict the onset of dryout in single channel using the drift flux model as well as the Wallis' flooding correlation. Above-mentioned two types of the dryout and their boundary are predicted which agree well with the experimental results. (author)
Flow visualization study of inverted annular flow of post dryout heat transfer region
International Nuclear Information System (INIS)
Ishii, M.; De Jarlais, G.
1985-01-01
The inverted annular flow is important in the area of LWR accident analysis in terms of the maximum cladding temperature and effectiveness of the emergency core cooling. However, the inverted annular flow thermal-hydraulics is not well understood due to its special heat transfer condition of film boiling. In view of this, the inverted flow is studied in detail experimentally. A new experimental apparatus has been constructed in which film boiling heat transfer can be established in a transparent test section. Data on liquid core stability, core break-up mechanism, and dispersed-core liquid slug and droplet sizes are obtained using F 113 as a test fluid. Both high speed movies and flash photographs are used
Large scale experiments with a 5 MW sodium/air heat exchanger for decay heat removal
International Nuclear Information System (INIS)
Stehle, H.; Damm, G.; Jansing, W.
1994-01-01
Sodium experiments in the large scale test facility ILONA were performed to demonstrate proper operation of a passive decay heat removal system for LMFBRs based on pure natural convection flow. Temperature and flow distributions on the sodium and the air side of a 5 MW sodium/air heat exchanger in a natural draught stack were measured during steady state and transient operation in good agreement with calculations using a two dimensional computer code ATTICA/DIANA. (orig.)
A Study of Heat Transfer and Flow Characteristics of Rising Taylor Bubbles
Scammell, Alexander David
2016-01-01
Practical application of flow boiling to ground- and space-based thermal management systems hinges on the ability to predict the systems heat removal capabilities under expected operating conditions. Research in this field has shown that the heat transfer coefficient within two-phase heat exchangers can be largely dependent on the experienced flow regime. This finding has inspired an effort to develop mechanistic heat transfer models for each flow pattern which are likely to outperform traditional empirical correlations. As a contribution to the effort, this work aimed to identify the heat transfer mechanisms for the slug flow regime through analysis of individual Taylor bubbles.An experimental apparatus was developed to inject single vapor Taylor bubbles into co-currently flowing liquid HFE 7100. The heat transfer was measured as the bubble rose through a 6 mm inner diameter heated tube using an infrared thermography technique. High-speed flow visualization was obtained and the bubble film thickness measured in an adiabatic section. Experiments were conducted at various liquid mass fluxes (43-200 kgm2s) and gravity levels (0.01g-1.8g) to characterize the effect of bubble drift velocityon the heat transfer mechanisms. Variable gravity testing was conducted during a NASA parabolic flight campaign.Results from the experiments showed that the drift velocity strongly affects the hydrodynamics and heat transfer of single elongated bubbles. At low gravity levels, bubbles exhibited shapes characteristic of capillary flows and the heat transfer enhancement due to the bubble was dominated by conduction through the thin film. At moderate to high gravity, traditional Taylor bubbles provided small values of enhancement within the film, but large peaks in the wake heat transfer occurred due to turbulent vortices induced by the film plunging into the trailing liquid slug. Characteristics of the wake heat transfer profiles were analyzed and related to the predicted velocity field
GPU accelerated study of heat transfer and fluid flow by lattice Boltzmann method on CUDA
Ren, Qinlong
,000 US dollars. The release of NVIDIA's CUDA architecture which includes both hardware and programming environment in 2007 makes GPU computing attractive. Due to its highly parallel nature, lattice Boltzmann method is successfully ported into GPU with a performance benefit during the recent years. In the current work, LBM CUDA code is developed for different fluid flow and heat transfer problems. In this dissertation, lattice Boltzmann method and immersed boundary method are used to study natural convection in an enclosure with an array of conduting obstacles, double-diffusive convection in a vertical cavity with Soret and Dufour effects, PCM melting process in a latent heat thermal energy storage system with internal fins, mixed convection in a lid-driven cavity with a sinusoidal cylinder, and AC electrothermal pumping in microfluidic systems on a CUDA computational platform. It is demonstrated that LBM is an efficient method to simulate complex heat transfer problems using GPU on CUDA.
UNSAT-H, an unsaturated soil water flow code for use at the Hanford site: code documentation
International Nuclear Information System (INIS)
Fayer, M.J.; Gee, G.W.
1985-10-01
The unsaturated soil moisture flow code, UNSAT-H, which was developed at Pacific Northwest Laboratory for assessing water movement at waste sites on the Hanford site, is documented in this report. This code is used in simulating the water dynamics of arid sites under consideration for waste disposal. The results of an example simulation of constant infiltration show excellent agreement with an analytical solution and another numerical solution, thus providing some verification of the UNSAT-H code. Areas of the code are identified for future work and include runoff, snowmelt, long-term climate and plant models, and parameter measurement. 29 refs., 7 figs., 2 tabs
Verification of the MOTIF code version 3.0
International Nuclear Information System (INIS)
Chan, T.; Guvanasen, V.; Nakka, B.W.; Reid, J.A.K.; Scheier, N.W.; Stanchell, F.W.
1996-12-01
As part of the Canadian Nuclear Fuel Waste Management Program (CNFWMP), AECL has developed a three-dimensional finite-element code, MOTIF (Model Of Transport In Fractured/ porous media), for detailed modelling of groundwater flow, heat transport and solute transport in a fractured rock mass. The code solves the transient and steady-state equations of groundwater flow, solute (including one-species radionuclide) transport, and heat transport in variably saturated fractured/porous media. The initial development was completed in 1985 (Guvanasen 1985) and version 3.0 was completed in 1986. This version is documented in detail in Guvanasen and Chan (in preparation). This report describes a series of fourteen verification cases which has been used to test the numerical solution techniques and coding of MOTIF, as well as demonstrate some of the MOTIF analysis capabilities. For each case the MOTIF solution has been compared with a corresponding analytical or independently developed alternate numerical solution. Several of the verification cases were included in Level 1 of the International Hydrologic Code Intercomparison Project (HYDROCOIN). The MOTIF results for these cases were also described in the HYDROCOIN Secretariat's compilation and comparison of results submitted by the various project teams (Swedish Nuclear Power Inspectorate 1988). It is evident from the graphical comparisons presented that the MOTIF solutions for the fourteen verification cases are generally in excellent agreement with known analytical or numerical solutions obtained from independent sources. This series of verification studies has established the ability of the MOTIF finite-element code to accurately model the groundwater flow and solute and heat transport phenomena for which it is intended. (author). 20 refs., 14 tabs., 32 figs
International Nuclear Information System (INIS)
Baek, Seong Gu; Park, Seung O.
2003-01-01
This paper provides the assessment of prediction performance of explicit algebraic stress and heat-flux models under conditions of mixed convective gas flows in a strongly-heated vertical tube. Two explicit algebraic stress models and four algebraic heat-flux models are selected for assessment. Eight combinations of explicit algebraic stress and heat-flux models are used in predicting the flows experimentally studied by Shehata and McEligot (IJHMT 41(1998) p.4333) in which property variation was significant. Among the various model combinations, the Wallin and Johansson (JFM 403(2000) p. 89) explicit algebraic stress model-Abe, Kondo, and Nagano (IJHFF 17(1996) p. 228) algebraic heat-flux model combination is found to perform best. We also found that the dimensionless wall distance y + should be calculated based on the local property rather than the property at the wall for property-variation flows. When the buoyancy or the property variation effects are so strong that the flow may relaminarize, the choice of the basic platform two-equation model is a most important factor in improving the predictions
International Nuclear Information System (INIS)
Wong, C.N.C.; Cheng, S.K.; Todreas, N.E.
1982-01-01
This report provides the HEATRAN user with programming and input information. HEATRAN is a computer program which is written to analyze the transient three dimensional single phase incompressible fluid flow and heat transfer problem. In this report, the programming information is given first. This information includes details concerning the code and structure. The description of the required input variables is presented next. Following the input description, the sample problems are described and HEATRAN's results are presented
Application of two-equation turbulence models to turbulent gas flow heated by a high heat flux
International Nuclear Information System (INIS)
Kawamura, Hiroshi
1978-01-01
Heat transfer in heated turbulent gas flow is analyzed using two-equation turbulence models. Four kinds of two-equation models are examined; that is, k-epsilon model by Jones-Launder, k-w model by Wilcox-Traci, k-kL model by Rotta, k-ω model by Saffman-Wilcox. The results are compared with more than ten experiments by seven authors. The k-kL model proposed originally by Rotta and modified by the present author is found to give relatively the best results. It well predicts the decrease in the heat transfer coefficient found in the heated turbulent gas flow; however, it fails to predict the laminarization due to a strong heating. (author)
Thermally determining flow and/or heat load distribution in parallel paths
Chainer, Timothy J.; Iyengar, Madhusudan K.; Parida, Pritish R.
2016-12-13
A method including obtaining calibration data for at least one sub-component in a heat transfer assembly, wherein the calibration data comprises at least one indication of coolant flow rate through the sub-component for a given surface temperature delta of the sub-component and a given heat load into said sub-component, determining a measured heat load into the sub-component, determining a measured surface temperature delta of the sub-component, and determining a coolant flow distribution in a first flow path comprising the sub-component from the calibration data according to the measured heat load and the measured surface temperature delta of the sub-component.
Heat and mass transfer from the mantle: heat flow and He-isotope constraints
Directory of Open Access Journals (Sweden)
B. G. Polyak
2005-06-01
Full Text Available Terrestrial heat flow density, q, is inversely correlated with the age, t, of tectono-magmatic activity in the Earth's crust (Polyak and Smirnov, 1966; etc.. «Heat flow-age dependence» indicates unknown temporal heat sources in the interior considered a priori as the mantle-derived diapirs. The validity of this hypothesis is demonstrated by studying the helium isotope ratio, 3He/4He = R, in subsurface fluids. This study discovered the positive correlation between the regionally averaged (background estimations of R- and q-values (Polyak et al., 1979a. Such a correlation manifests itself in both pan-regional scales (Norhtern Eurasia and separate regions, e.g., Japan (Sano et al., 1982, Eger Graben (Polyak et al., 1985 Eastern China rifts (Du, 1992, Southern Italy (Italiano et al., 2000, and elsewhere. The R-q relation indicates a coupled heat and mass transfer from the mantle into the crust. From considerations of heat-mass budget this transfer can be provided by the flux consisting of silicate matter rather than He or other volatiles. This conclusion is confirmed by the correlation between 3He/ 4He and 87Sr/86Sr ratios in the products of the volcanic and hydrothermal activity in Italy (Polyak et al., 1979b; Parello et al., 2000 and other places. Migration of any substance through geotemperature field transports thermal energy accumulated within this substance, i.e. represents heat and mass transfer. Therefore, only the coupled analysis of both material and energy aspects of this transfer makes it possible to characterise the process adequately and to decipher an origin of terrestrial heat flow observed in upper parts of the earth crust. An attempt of such kind is made in this paper.
Heat transfer to a dispersed two-phase flow and detailed quench front velocity research
International Nuclear Information System (INIS)
Boer, T.C. de; Molen, S.B. van der
1985-01-01
A programme to obtain a data base for 'Boildown and Reflood' computer code development and to obtain information on the influence of non-uniform temperature and/or power profile on the quench front velocity and prequench heat transfer, including unheated wall and grid effects, has been undertaken. It is in two parts. In the first (for the tube, annulus and a 4-rod bundle) an early wetting of the unheated shroud is shown. This leads to an increase in quench front velocity and in liquid transport downstream from the quench front. For the inverted annular flow regime the extended Bromley correlation gives good agreement with the experimental data. In the second part (36-rod bundle reflood test programme) the wall-temperature differences in the radial direction gives rise to heat transfer processes which are described and explained. (U.K.)
Wadhams, T.P.; MacLean, M.; Holden, M.S.; Cassady, A.M.
2009-01-01
An experimental program has been completed by CUBRC exploring laminar, transitional, and turbulent flows over a 7.0% scale model of the Project ORION CEV geometry. This program was executed primarily to answer questions concerning the increase in heat transfer on the windward, or "hot shoulder" of the CEV heat shield from laminar to turbulent flow. To answer these questions CUBRC constructed and instrumented a 14.0 inch diameter Project ORION CEV model and ran a range of Reynolds numbers based on diameter from 1.0 to over 40 million at a Mach number of 8.0. These Reynolds numbers were selected to cover laminar to turbulent heating data on the "hot shoulder". Data obtained during these runs will be used to guide design decisions as they apply to heat shield thickness and extent. Several experiments at higher enthalpies were achieved to obtain data for code validation with real gas effects and transition. CUBRC also performed computation studies of these experiments to aid in the data reduction process and study turbulence modeling.
Flow and heat transfer in a curved channel
Brinich, P. F.; Graham, R. W.
1977-01-01
Flow and heat transfer in a curved channel of aspect ratio 6 and inner- to outer-wall radius ratio 0.96 were studied. Secondary currents and large longitudinal vortices were found. The heat-transfer rates of the outer and inner walls were independently controlled to maintain a constant wall temperature. Heating the inner wall increased the pressure drop along the channel length, whereas heating the outer wall had little effect. Outer-wall heat transfer was as much as 40 percent greater than the straight-channel correlation, and inner-wall heat transfer was 22 percent greater than the straight-channel correlation.
Three-dimensional numerical study of heat transfer enhancement in separated flows
Kumar, Saurav; Vengadesan, S.
2017-11-01
The flow separation appears in a wide range of heat transfer applications and causes poor heat transfer performance. It motivates the study of heat transfer enhancement in laminar as well as turbulent flows over a backward facing step by means of an adiabatic fin mounted on the top wall. Recently, we have studied steady, 2-D numerical simulations in laminar flow and investigated the effect of fin length, location, and orientation. It revealed that the addition of fin causes enhancement of heat transfer and it is very effective to control the flow and thermal behavior. The fin is most effective and sensitive when it is placed exactly above the step. A slight displacement of the fin in upstream of the step causes the complete change of flow and thermal behavior. Based on the obtained 2-D results it is interesting to investigate the side wall effect in three-dimensional simulations. The comparison of two-dimensional and three-dimensional numerical simulations with the available experimental results will be presented. Special attention has to be given to capture unsteadiness in the flow and thermal field.
Models for fluid flows with heat transfer in mixed convection
International Nuclear Information System (INIS)
Mompean Munhoz da Cruz, G.
1989-06-01
Second order models were studied in order to predict turbulent flows with heat transfer. The equations used correspond to the characteristic scale of turbulent flows. The order of magnitude of the terms of the equation is analyzed by using Reynolds and Peclet numbers. The two-equation model (K-ε) is applied in the hydrodynamic study. Two models are developed for the heat transfer analysis: the Prt + teta 2 and the complete model. In the first model, the turbulent thermal diffusivity is calculated by using the Prandtl number for turbulent flow and an equation for the variance of the temperature fluctuation. The second model consists of three equations concerning: the turbulent heat flow, the variance of the temperature fluctuation and its dissipation ratio. The equations were validated by four experiments, which were characterized by the analysis of: the air flow after passing through a grid of constant average temperature and with temperature gradient, an axysymmetric air jet submitted to high and low heating temperature, the mixing (cold-hot) of two coaxial jets of sodium at high Peclet number. The complete model is shown to be the most suitable for the investigations presented [fr
A review of surface heat-flow data of the northern Middle Atlas (Morocco)
Chiozzi, Paolo; Barkaoui, Alae-Eddine; Rimi, Abdelkrim; Verdoya, Massimo; Zarhloule, Yassine
2017-12-01
We revised thermal data available from water and oil wells in the northern sector of the Middle Atlas region. To avoid biased estimation of surface heat flow caused by advection likely occurring in shallow aquifers, temperature measurements in water boreholes were carefully inspected and selected. The heat flow in the oil wells was inferred by taking into account the porosity variation with depth, the temperature effect on thermal conductivity of the matrix and the pore fluid, together with the contribution of the radiogenic heat production. Moreover, the possible bias in heat flow caused by convection occurring in confined carbonate aquifers was evaluated. The results of heat flow slightly modify the picture reported in previous investigations. The heat flow value over the investigated region is rather uniform (about 80 mW m-2) and is similar in oil wells and in water boreholes. Geothermal calculations indicate that such a surface heat flow is compatible with a ∼70 km thick thermal lithosphere and normal thermal conditions in the asthenospheric mantle.
Hybrid code simulation on mode conversion in the second harmonic ICRF heating
International Nuclear Information System (INIS)
Sakai, K.; Takeuchi, S.; Matsumoto, M.; Sugihara, R.
1985-01-01
ICRF second harmonic heating of a single-species plasma is studied by using a 1-1/2 dimensional quasi-neutral hybrid code. Mode conversion, transmission and reflection of the magnetosonic waves are confirmed, both for the high- and low-field-side excitations. The ion heating by waves propagating perpendicularly to the static magnetic field is also observed
Ocular blood flow decreases during passive heat stress in resting humans
Ikemura, Tsukasa; Miyaji, Akane; Kashima, Hideaki; Yamaguchi, Yuji; Hayashi, Naoyuki
2013-01-01
Background Heat stress induces various physiological changes and so could influence ocular circulation. This study examined the effect of heat stress on ocular blood flow. Findings Ocular blood flow, end-tidal carbon dioxide (P ETCO2) and blood pressure were measured for 12 healthy subjects wearing water-perfused tube-lined suits under two conditions of water circulation: (1) at 35°C (normothermia) for 30 min and (2) at 50°C for 90 min (passive heat stress). The blood-flow velocities in the s...
Enhanced two phase flow in heat transfer systems
Tegrotenhuis, Ward E; Humble, Paul H; Lavender, Curt A; Caldwell, Dustin D
2013-12-03
A family of structures and designs for use in devices such as heat exchangers so as to allow for enhanced performance in heat exchangers smaller and lighter weight than other existing devices. These structures provide flow paths for liquid and vapor and are generally open. In some embodiments of the invention, these structures can also provide secondary heat transfer as well. In an evaporate heat exchanger, the inclusion of these structures and devices enhance the heat transfer coefficient of the evaporation phase change process with comparable or lower pressure drop.
Natural convection heat transfer between vertical channel with flow resistance at the lower end
International Nuclear Information System (INIS)
Iwamoto, S.; Nishimura, S.; Ishihara, I.
2003-01-01
For natural convection in the geometrically complicated channel, the convection flow is suppressed by flow resistance due to such channel itself and the lopsided flow may take place. This could result in serious influences on the heat transfer in the channel. In order to investigate fundamentally the natural convection flow and heat transfer in such the channel, the vertical channel in which wall was heated with uniform heat flux and the flow resistance was given by small clearance between the lower end of channel and a wide horizontal floor. Flow pattern was observed by illuminating smoke filled in the channel and heat transfer rate was measured. (author)
Preliminary heat flow map of Europe. Explanatory text
Energy Technology Data Exchange (ETDEWEB)
Cermak, V.; Hurtig, E.
1977-08-08
A preliminary heat flow map of Europe was prepared, based on data contained in 401 references. The map was prepared on a scale of 1:5,000,000 and shows broad-scale geological structure (e.g., platforms, shields, foredeeps) and specialized rock suites (ophiolites, volcanites). Primary faults and thrust faults are indicated, and contours showing the depth of crystalline basement are given. Heat flow is plotted using 10.0 mW/m/sup 2/ isotherms. The accompanying explanatory text describes data acquisition and techniques of correction, and discusses some implications of the results.
Chaotic advection and heat transfer enhancement in Stokes flows
International Nuclear Information System (INIS)
Lefevre, A.; Mota, J.P.B.; Rodrigo, A.J.S.; Saatdjian, E.
2003-01-01
The heat transfer rate from a solid boundary to a highly viscous fluid can be enhanced significantly by a phenomenon which is called chaotic advection or Lagrangian turbulence. Although the flow is laminar and dominated by viscous forces, some fluid particle trajectories are chaotic due either to a suitable boundary displacement protocol or to a change in geometry. As in turbulent flow, the heat transfer rate enhancement between the boundary and the fluid is intimately linked to the mixing of fluid in the system. Chaotic advection in real Stokes flows, i.e. flows governed by viscous forces and that can be constructed experimentally, is reviewed in this paper. An emphasis is made on recent new results on 3-D time-periodic open flows which are particularly important in industry
Editorial to "Heat flow: recent advances"
Czech Academy of Sciences Publication Activity Database
Čermák, Vladimír; Huang, S.; Ravat, D.; Verdoya, M.
2018-01-01
Roč. 107, č. 1 (2018), s. 1-3 ISSN 1437-3254 Institutional support: RVO:67985530 Keywords : geothermics * climate change * terrestrial heat flow Subject RIV: DC - Siesmology, Volcanology, Earth Structure OBOR OECD: Volcanology Impact factor: 2.283, year: 2016
Directory of Open Access Journals (Sweden)
Mikielewicz Dariusz
2014-09-01
Full Text Available In the paper a method developed earlier by authors is applied to calculations of pressure drop and heat transfer coefficient for flow boiling and also flow condensation for some recent data collected from literature for such fluids as R404a, R600a, R290, R32,R134a, R1234yf and other. The modification of interface shear stresses between flow boiling and flow condensation in annular flow structure are considered through incorporation of the so called blowing parameter. The shear stress between vapor phase and liquid phase is generally a function of nonisothermal effects. The mechanism of modification of shear stresses at the vapor-liquid interface has been presented in detail. In case of annular flow it contributes to thickening and thinning of the liquid film, which corresponds to condensation and boiling respectively. There is also a different influence of heat flux on the modification of shear stress in the bubbly flow structure, where it affects bubble nucleation. In that case the effect of applied heat flux is considered. As a result a modified form of the two-phase flow multiplier is obtained, in which the nonadiabatic effect is clearly pronounced.
Heat transfer in flow past a continuously moving porous flat plate with heat flux
Digital Repository Service at National Institute of Oceanography (India)
Murty, T.V.R.; Sarma, Y.V.B.
The analysis of the heat transfer in flow past a continuously moving semi-infinite plate in the presence of suction/ injection with heat flux has been presented. Similarity solutions have been derived and the resulting equations are integrated...
International Nuclear Information System (INIS)
Takahashi, Yuya; Chen, Lin; Okajima, Junnosuke; Iga, Yuka; Komiya, Atsuki; Maruyama, Shigenao
2016-01-01
Highlights: • Effective cooling design by super-/sub-sonic air flow in microchannels is proposed. • Microscale supersonic flows is successfully generated and examined. • Microchannel flow density field were visualized quantitatively by interferometer. • The bump design shows great potential of heat transfer enhancement in microscale. - Abstract: With the fast development of electronic systems and the ever-increasing demand of thermally “smart” design in space and aeronautic engineering, the heat transfer innovations and high heat flux challenges have become a hot topic for decades. This study is aimed at the effective cooling heat transfer design by super-/sub-sonic air flow in microscale channels for high heat flux devices. The design is based on the low temperature flows with supersonic expansion in microscale, which yields a compact and simple design. By careful microelectromechanical process, microscale straight and bumped channels (with simple arc curve) are fabricated and experimentally tested in this study. The microscale flow field and density distributions under new designs are visualized quantitatively by an advanced phase-shifting interferometer system, which results are then compared carefully with numerical simulations. In this study, large differences between the two designs in density distribution and temperature changes (around 50 K) are found. The high heat flux potential for supersonic microchannel flows is realized and discussion into detail. It is confirmed that the bump design contributes significantly to the heat transfer enhancement, which shows potential for future application in novel system designs.
Free convection flow and heat transfer in pipe exposed to cooling
Energy Technology Data Exchange (ETDEWEB)
Mme, Uduak Akpan
2010-10-15
One of the challenges with thermal insulation design in subsea equipment is to minimize the heat loss through cold spots during production shut down. Cold spots are system components where insulation is difficult to implement, resulting in an insulation discontinuity which creates by nature a thermal bridge. It is difficult to avoid cold spots or thermal bridges in items like sensors, valves, connectors and supporting structures. These areas of reduced or no insulation are referred to as cold spots. Heat is drained faster through these spots, resulting in an increased local fluid density resulting in an internal fluid flow due to gravity and accelerated cool- down. This natural convection flow is important for both heat loss and internal distribution of the temperature. This thesis is presenting both experimental work and modelling work. A series of cool down tests and Computational Fluid Dynamics (CFD) simulations of these tests are presented. These tests and simulations were carried out in order to understand the flow physics involved in heat exchange processes caused by free convection flow in pipe exposed to cooling. Inclination of the pipe relative to the direction of gravity and temperature difference between cooling water and internal pipe water are the two main parameters investigated in this study. The experimental heat removal and temperature field is discussed and further interpreted by means of computational fluid dynamics. For prediction of the evolvement of the local temperature and heat flow, selection of an appropriate turbulence model is critical. Hence, different models and wall functions are investigated. The predicted temperature profiles and heat extraction rates are compered to the experiments for the selected turbulence models. Our main conclusions, supported by our experimental and CFD results, include: (i) Heat transfer from a localized cold spot in an inclined pipe is most efficient when the pipe orientation is close to horizontal. As the
Transient heat transfer for forced convection flow of helium gas
International Nuclear Information System (INIS)
Liu, Qiusheng; Fukuda, Katsuya; Sasaki, Kenji; Yamamoto, Manabu
1999-01-01
Transient heat transfer coefficients for forced convection flow of helium gas over a horizontal cylinder were measured using a forced convection test loop. The platinum heater with a diameter of 1.0 mm was heated by electric current with an exponential increase of Q 0 exp(t/τ). It was clarified that the heat transfer coefficient approaches the steady-state one for the period τ over 1 s, and it becomes higher for the period of τ shorter than 1 s. The transient heat transfer shows less dependent on the gas flowing velocity when the period becomes very shorter. Semi-empirical correlations for steady-state and transient heat transfer were developed based on the experimental data. (author)
Surface roughness effects on heat transfer in Couette flow
International Nuclear Information System (INIS)
Elia, G.G.
1981-01-01
A cell theory for viscous flow with rough surfaces is applied to two basic illustrative heat transfer problems which occur in Couette flow. Couette flow between one adiabatic surface and one isothermal surface exhibits roughness effects on the adiabatic wall temperature. Two types of rough cell adiabatic surfaces are studied: (1) perfectly insulating (the temperature gradient vanishes at the boundary of each cell); (2) average insulating (each cell may gain or lose heat but the total heat flow at the wall is zero). The results for the roughness on a surface in motion are postulated to occur because of fluid entrainment in the asperities on the moving surface. The symmetry of the roughness effects on thermal-viscous dissipation is discussed in detail. Explicit effects of the roughness on each surface, including combinations of roughness values, are presented to enable the case where the two surfaces may be from different materials to be studied. The fluid bulk temperature rise is also calculated for Couette flow with two ideal adiabatic surfaces. The effect of roughness on thermal-viscous dissipation concurs with the viscous hydrodynamic effect. The results are illustrated by an application to lubrication. (Auth.)
MINET [momentum integral network] code documentation
International Nuclear Information System (INIS)
Van Tuyle, G.J.; Nepsee, T.C.; Guppy, J.G.
1989-12-01
The MINET computer code, developed for the transient analysis of fluid flow and heat transfer, is documented in this four-part reference. In Part 1, the MINET models, which are based on a momentum integral network method, are described. The various aspects of utilizing the MINET code are discussed in Part 2, The User's Manual. The third part is a code description, detailing the basic code structure and the various subroutines and functions that make up MINET. In Part 4, example input decks, as well as recent validation studies and applications of MINET are summarized. 32 refs., 36 figs., 47 tabs
Noxious heat and scratching decrease histamine-induced itch and skin blood flow.
Yosipovitch, Gil; Fast, Katharine; Bernhard, Jeffrey D
2005-12-01
The aim of this study was to assess the effect of thermal stimuli or distal scratching on skin blood flow and histamine-induced itch in healthy volunteers. Twenty-one healthy volunteers participated in the study. Baseline measurements of skin blood flow were obtained on the flexor aspect of the forearm. These measurements were compared with skin blood flow after various stimuli: heating the skin, cooling the skin, noxious cold 2 degrees C, noxious heat 49 degrees C, and scratching via a brush with controlled pressure. Afterwards histamine iontophoresis was performed and skin blood flow and itch intensity were measured immediately after the above-mentioned stimuli. Scratching reduced mean histamine-induced skin blood flow and itch intensity. Noxious heat pain increased basal skin blood flow but reduced histamine-induced maximal skin blood flow and itch intensity. Cold pain and cooling reduced itch intensity, but neither affected histamine-induced skin blood flow. Sub-noxious warming the skin did not affect the skin blood flow or itch intensity. These findings suggest that heat pain and scratching may inhibit itch through a neurogenic mechanism that also affects skin blood flow.
A high performance cocurrent-flow heat pipe for heat recovery applications
Saaski, E. W.; Hartl, J. C.
1980-01-01
By the introduction of a plate-and-tube separator assembly into a heat pipe vapor core, it has been demonstrated that axial transport capacity in reflux mode can be improved by up to a factor of 10. This improvement is largely the result of eliminating the countercurrent shear that commonly limits reflux heat pipe axial capacity. With benzene, axial heat fluxes up to 1800 W/sq cm were obtained in the temperature range 40 to 80 C, while heat flux densities up to 3000 W/sq cm were obtained with R-11 over the temperature range 40 to 80 C. These very high axial capacities compare favorably with liquid metal limits; the sonic limit for liquid sodium, for example, is 3000 W/sq cm at 657 C. Computational models developed for these cocurrent flow heat pipes agreed with experimental data within + or - 25%.
International Nuclear Information System (INIS)
Moon, S.K.; Chun, S.Y.; Choi, K.Y.; Yang, S.K.
2001-01-01
An experimental study on transient critical heat flux (CHF) under flow coast-down has been performed for water flow in a non-uniformly heated vertical annulus under low flow and a wide range of pressure conditions. The objectives of this study are to systematically investigate the effect of the flow transient on the CHF and to compare the transient CHF with steady state CHF. The transient CHF experiments have been performed for three kinds of flow transient modes based on the coast-down data of the Kori 3/4 nuclear power plant reactor coolant pump. Most of the CHFs occurred in the annular-mist flow regime. Thus, it means that the possible CHF mechanism might be the liquid film dryout in the annular-mist flow regime. For flow transient mode with the smallest flow reduction rate, the time-to-CHF is the largest. At the same inlet subcooling, system pressure and heat flux, the effect of the initial mass flux on the critical mass flux can be negligible. However, the effect of the initial mass flux on the time-to-CHF becomes large as the heat flux decreases. Usually, the critical mass flux is large for slow flow reduction. There is a pressure effect on the ratio of the transient CHF data to steady state CHF data. Some conventional correlations show relatively better CHF prediction results for high system pressure, high quality and slow transient modes than for low system pressure, low quality and fast transient modes. (author)
Flow analysis of an innovative compact heat exchanger channel geometry
International Nuclear Information System (INIS)
Vitillo, F.; Cachon, L.; Reulet, F.; Millan, P.
2016-01-01
Highlights: • An innovative compact heat transfer technology is proposed. • Experimental measurements are shown to validate the CFD model. • CFD simulations s