Validation of the TASS/SMR-S Code for the PRHRS Condensation Heat Transfer Model

International Nuclear Information System (INIS)

Jun, In Sub; Yang, Soo Hyoung; Chung, Young Jong; Lee, Won Jae

2011-01-01

When some accidents or events are occurred in the SMART, the secondary system is used to remove the core decay heat for the long time such as a feedwater system. But if the feedwater system can't remove the residual core heat because of its malfunction, the core decay heat is removed using the Passive Residual Heat Removal System (PRHRS). The PRHRS is passive type safety system adopted to enhance the safety of the SMART. It can fundamentally eliminate the uncertainty of operator action. TASS/SMR-S (Transient And Setpoint Simulation/ System-integrated Modular Reactor-Safety) code has various heat transfer models reflecting the design features of the SMART. One of the heat transfer models is the PRHRS condensation heat transfer model. The role of this model is to calculate the heat transfer coefficient in the heat exchanger (H/X) tube side using the relevant heat transfer correlations for all of the heat transfer modes. In this paper, the validation of the condensation heat transfer model was carried out using the POSTECH H/X heat transfer test

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

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)

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

Verification of calculational models of heat exchange crisis and overcrisis heat transfer, used in the KORSAR code

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

Application of the TEMPEST computer code to canister-filling heat transfer problems

International Nuclear Information System (INIS)

Farnsworth, R.K.; Faletti, D.W.; Budden, M.J.

1988-03-01

Pacific Northwest Laboratory (PNL) researchers used the TEMPEST computer code to simulate thermal cooldown behavior of nuclear waste glass after it was poured into steel canisters for long-term storage. The objective of this work was to determine the accuracy and applicability of the TEMPEST code when used to compute canister thermal histories. First, experimental data were obtained to provide the basis for comparing TEMPEST-generated predictions. Five canisters were instrumented with appropriately located radial and axial thermocouples. The canister were filled using the pilot-scale ceramic melter (PSCM) at PNL. Each canister was filled in either a continous or a batch filling mode. One of the canisters was also filled within a turntable simulant (a group of cylindrical shells with heat transfer resistances similar to those in an actual melter turntable). This was necessary to provide a basis for assessing the ability of the TEMPEST code to also model the transient cooling of canisters in a melter turntable. The continous-fill model, Version M, was found to predict temperatures with more accuracy. The turntable simulant experiment demonstrated that TEMPEST can adequately model the asymmetric temperature field caused by the turntable geometry. Further, TEMPEST can acceptably predict the canister cooling history within a turntable, despite code limitations in computing simultaneous radiation and convection heat transfer between shells, along with uncertainty in stainless-steel surface emissivities. Based on the successful performance of TEMPEST Version M, development was initiated to incorporate 1) full viscous glass convection, 2) a dynamically adaptive grid that automatically follows the glass/air interface throughout the transient, and 3) a full enclosure radiation model to allow radiation heat transfer to non-nearest neighbor cells. 5 refs., 47 figs., 17 tabs

Turbomachinery Heat Transfer and Loss Modeling for 3D Navier-Stokes Codes

Science.gov (United States)

DeWitt, Kenneth; Ameri, Ali

2005-01-01

This report's contents focus on making use of NASA Glenn on-site computational facilities,to develop, validate, and apply models for use in advanced 3D Navier-Stokes Computational Fluid Dynamics (CFD) codes to enhance the capability to compute heat transfer and losses in turbomachiney.

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

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

Use of a general-purpose heat-transfer code for casting simulation

International Nuclear Information System (INIS)

Erickson, W.C.

1975-07-01

The practical use of numerical techniques in simulating casting solidification dictate that a general purpose heat transfer code be used and that results be obtained in an easy-to-analyze format. Color film plotting routines were developed for use with NASA's CINDA-3G heat transfer code; the combination of which meet the above criteria. The subroutine LQSLTR written for SINDA, the successor to CINDA-3G, was verified by comparing calculated results obtained using LQSLTR with those obtained using the specific heat method for handling the heat of fusion. Excellent agreement existed when similar data was used. When the more restrictive requirement of a 1 0 F melting range was used, comparable results were obtained. Uranium and lead rod castings were cast in instrumented graphite molds and the solidification sequence simulated using CINDA-3G. Discrepancies attributed to initial assumptions of instantaneous mold filling, uniform melt temperature, and intimate metal/mold contact were encountered. Further calculations using a model incorporating a gap between the mold and casting showed that the intimate contact assumption could not be used; a three-dimensional model also showed that the thermocouple assemblies used with the platinum--platinum-10 percent rhodium were a significant perturbation to the system. An L-shaped steel casting was simulated and the results compared to those reported in the literature. The experimental data for this casting were reproduced within the accuracy permitted by the thermal conductivity of the sand, thus demonstrating that agreement can be obtained when the mold material does not act as a chill. (U.S.)

Software requirements, design, and verification and validation for the FEHM application - a finite-element heat- and mass-transfer code

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

Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

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

Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

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

Development and preliminary assessment of the wall condensation heat transfer models for the SPACE code

International Nuclear Information System (INIS)

Park, Hyun Sik; Choi, Ki Yong; Moon, Sang Ki; Kim, Jung Woo; Kim, Kyung Doo

2009-01-01

The wall condensation heat transfer models are developed for the SPACE code and are assessed for various condensation conditions. Both default and alternative models were selected through an extensive literature survey. For a pure steam condensation, a maximum value among the Nusselt, Chato, and Shah's correlations is used in order to consider the geometric and turbulent effects. In the presence of non-condensable gases, the Colburn-Hougen's diffusion model was used as a default model and a non-iterative condensation model proposed by No and Park was selected as an alternative model. The wall condensation heat transfer models were assessed preliminarily by using arbitrary test conditions. Both wall condensation models could simulate the heat transfer coefficients and heat fluxes in the vertical, horizontal and turbulent conditions quite reasonably for a pure steam condensation. Both the default and alternative wall condensation models were also verified for the condensation heat transfer coefficient and heat flux in the presence of noncondensable gas. However, some improvements and further detailed verification are necessary for the condensation phenomena in the presence of noncondensable gas

An evaluation of analytical heat transfer area with various boiling heat transfer correlations in steam generator thermal sizing

International Nuclear Information System (INIS)

Jung, B. R.; Park, H. S.; Chung, D. M.; Baik, S. J.

1999-01-01

The computer program SAFE has been used to size and analyze the performance of a steam generator which has two types of heat transfer regions in Korean Standard Nuclear Power Plants (KSNP) and Korean Next Generation Reactor (KNGR) design. The SAFE code calculates the analytical boiling heat transfer area using the modified form of the saturated nucleate pool boiling correlation suggested by Rohsenow. The predicted heat transfer area in the boiling region is multiplied by a constant to obtain a final analytical heat transfer area. The inclusion of the multiplier in the analytical calculation has some disadvantage of loss of complete correlation by the governing heat transfer equation. Several comparative analyses have been performed quantitatively to evaluate the possibility of removing the multiplier in the analytical calculation in the SAFE code. The evaluation shows that the boiling correlation and multiplier used in predicting the boiling region heat transfer area can be replaced with other correlations predicting nearly the same heat transfer area. The removal of multiplier included in the analytical calculation will facilitate a direct use of a set of concerned analytical sizing values that can be exactly correlated by the governing heat transfer equation. In addition this will provide more reasonable basis for the steam generator thermal sizing calculation and enhance the code usability without loss of any validity of the current sizing procedure. (author)

Preparation of functions of computer code GENGTC and improvement for two-dimensional heat transfer calculations for irradiation capsules

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)

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.

Thermal-Flow Code for Modeling Gas Dynamics and Heat Transfer in Space Shuttle Solid Rocket Motor Joints

Science.gov (United States)

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.

An assessment of RELAP5 MOD3.1.1 condensation heat transfer modeling with GIRAFFE heat transfer tests

International Nuclear Information System (INIS)

Boyer, B.D.; Parlatan, Y.; Slovik, G.C.; Rohatgi, U.S.

1995-01-01

RELAP5 MOD3.1.1 is being used to simulate Loss of Coolant Accidents (LOCA) for the Simplified Boiling Water Reactor (SBWR) being proposed by General Electric (GE). One of the major components associated with the SBWR is the Passive Containment Cooling System (PCCS) which provides the long-term heat sink to reject decay heat. The RELAP5 MOD3.1.1 code is being assessed for its ability to represent accurately the PCCS. Data from the Phase 1, Step 1 Heat Transfer Tests performed at Toshiba's Gravity-Driven Integral Full-Height Test for Passive Heat Removal (GIRAFFE) facility will be used for assessing the ability of RELAP5 to model condensation in the presence of noncondensables. The RELAP5 MOD3.1.1 condensation model uses the University of California at Berkeley (UCB) correlation developed by Vierow and Schrock. The RELAP5 code uses this heat transfer coefficient with the gas velocity effect multiplier being limited to 2. This heat transfer option was used to analyze the condensation heat transfer in the GIRAFFE PCCS heat exchanger tubes in the Phase 1, Step 1 Heat Transfer Tests which were at a pressure of 3 bar and had a range of nitrogen partial pressure fractions from 0.0 to 0.10. The results of a set of RELAP5 calculations al these conditions were compared with the GIRAFFE data. The effects of PCCS cell nodings on the heat transfer process were also studied. The UCB correlation, as implemented in RELAP5, predicted the heat transfer to ±5% of the data with a three-node model. The three-node model has a large cell in the entrance region which smeared out the entrance effects on the heat transfer, which tend to overpredict the condensation. Hence, the UCB correlation predicts condensation heat transfer in the presence of noncondensable gases with only a coarse mesh. The cell length term in the condensation heat transfer correlation implemented in the code must be removed to allow for accurate calculations with smaller cell sizes

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

An assessment of RELAP5 MOD3.1.1 condensation heat transfer modeling with GIRAFFE heat transfer tests

International Nuclear Information System (INIS)

Boyer, B.D.; Parlatan, Y.; Slovik, G.C.

1995-01-01

RELAP5 MOD3.1.1 is being used to simulate Loss of Coolant Accidents (LOCA) for the Simplified Boiling Water Reactor (SBWR) being proposed by General Electric (GE). One of the major components associated with the SBWR is the Passive Containment Cooling System (PCCS) which provides the long-term heat sink to reject decay heat. The RELAP5 MOD3.1.1 code is being assessed for its ability to represent accurately the PCCS. Data from the Phase 1, Step 1 Heat Transfer Tests performed at Toshiba's Gravity-Driven Integral Full-Height Test for Passive Heat Removal (GIRAFFE) facility will be used for assessing the ability of RELAP5 to model condensation in the presence of noncondensables. The RELAP5 MOD3.1.1 condensation model uses the University of California at Berkeley (UCB) correlation developed by Vierow and Schrock. The RELAP5 code uses this heat transfer coefficient with the gas velocity effect multiplier being limited to 2. This heat transfer option was used to analyze the condensation heat transfer in the GIRAFFE PCCS heat exchanger tubes in the Phase 1, Step 1 Heat Transfer Tests which were at a pressure of 3 bar and had a range of nitrogen partial pressure fractions from 0.0 to 0.10. The results of a set of RELAP5 calculations at these conditions were compared with the GIRAFFE data. The effects of PCCS cell noding on the heat transfer process were also studied. The UCB correlation, as implemented in RELAP5, predicted the heat transfer to ±5% of the data with a three--node model. The three-node model has a large cell in the entrance region which smeared out the entrance effects on the heat transfer, which tend to overpredict the condensation. Hence, the UCB correlation predicts condensation heat transfer correlation implemented in the code must be removed to allow for accurate calculations with smaller cell sizes

An assessment of RELAP5 MOD3.1.1 condensation heat transfer modeling with GIRAFFE heat transfer tests

Energy Technology Data Exchange (ETDEWEB)

Boyer, B.D.; Parlatan, Y.; Slovik, G.C. [and others

1995-09-01

RELAP5 MOD3.1.1 is being used to simulate Loss of Coolant Accidents (LOCA) for the Simplified Boiling Water Reactor (SBWR) being proposed by General Electric (GE). One of the major components associated with the SBWR is the Passive Containment Cooling System (PCCS) which provides the long-term heat sink to reject decay heat. The RELAP5 MOD3.1.1 code is being assessed for its ability to represent accurately the PCCS. Data from the Phase 1, Step 1 Heat Transfer Tests performed at Toshiba`s Gravity-Driven Integral Full-Height Test for Passive Heat Removal (GIRAFFE) facility will be used for assessing the ability of RELAP5 to model condensation in the presence of noncondensables. The RELAP5 MOD3.1.1 condensation model uses the University of California at Berkeley (UCB) correlation developed by Vierow and Schrock. The RELAP5 code uses this heat transfer coefficient with the gas velocity effect multiplier being limited to 2. This heat transfer option was used to analyze the condensation heat transfer in the GIRAFFE PCCS heat exchanger tubes in the Phase 1, Step 1 Heat Transfer Tests which were at a pressure of 3 bar and had a range of nitrogen partial pressure fractions from 0.0 to 0.10. The results of a set of RELAP5 calculations at these conditions were compared with the GIRAFFE data. The effects of PCCS cell noding on the heat transfer process were also studied. The UCB correlation, as implemented in RELAP5, predicted the heat transfer to {plus_minus}5% of the data with a three--node model. The three-node model has a large cell in the entrance region which smeared out the entrance effects on the heat transfer, which tend to overpredict the condensation. Hence, the UCB correlation predicts condensation heat transfer correlation implemented in the code must be removed to allow for accurate calculations with smaller cell sizes.

Validation of Heat Transfer and Film Cooling Capabilities of the 3-D RANS Code TURBO

Science.gov (United States)

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.

Castor-1C spent fuel storage cask decay heat, heat transfer, and shielding analyses

International Nuclear Information System (INIS)

Rector, D.R.; McCann, R.A.; Jenquin, U.P.; Heeb, C.M.; Creer, J.M.; Wheeler, C.L.

1986-12-01

This report documents the decay heat, heat transfer, and shielding analyses of the Gesellschaft fuer Nuklear Services (GNS) CASTOR-1C cask used in a spent fuel storage demonstration performed at Preussen Elektra's Wurgassen nuclear power plant. The demonstration was performed between March 1982 and January 1984, and resulted in cask and fuel temperature data and cask exterior surface gamma-ray and neutron radiation dose rate measurements. The purpose of the analyses reported here was to evaluate decay heat, heat transfer, and shielding computer codes. The analyses consisted of (1) performing pre-look predictions (predictions performed before the analysts were provided the test data), (2) comparing ORIGEN2 (decay heat), COBRA-SFS and HYDRA (heat transfer), and QAD and DOT (shielding) results to data, and (3) performing post-test analyses if appropriate. Even though two heat transfer codes were used to predict CASTOR-1C cask test data, no attempt was made to compare the two codes. The codes are being evaluated with other test data (single-assembly data and other cask data), and to compare the codes based on one set of data may be premature and lead to erroneous conclusions

BUSH: A computer code for calculating steady state heat transfer in LWR rod bundles under accident conditions

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

Comparison of heat transfer models for reciprocating compressor

International Nuclear Information System (INIS)

Tuhovcak, J.; Hejcik, J.; Jicha, M.

2016-01-01

Highlights: • Comparison of integral heat transfer models. • Influence of heat transfer model on volumetric and isentropic efficiency. • Various gases used as working fluid. - Abstract: One of the main factors affecting the efficiency of reciprocating compressor is heat transfer inside the cylinder. An analysis of heat transfer could be done using numerical models or integral correlations developed mainly from approaches used in combustion engines; however their accuracy is not completely verified due to the complicated experimental set up. The goal of this paper is to analyse the effect of heat transfer on compressor efficiency. Various integral correlations were compared for different compressor settings and fluids. CoolProp library was used in the code to obtain the properties of common coolants and gases. A comparison was done using the in-house code developed in Matlab, based on 1st Law of Thermodynamics.

Summary of the models and methods 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 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

Assessment of heat transfer correlations for supercritical water in the frame of best-estimate code validation

International Nuclear Information System (INIS)

Jaeger, Wadim; Espinoza, Victor H. Sanchez; Schneider, Niko; Hurtado, Antonio

2009-01-01

Within the frame of the Generation IV international forum six innovative reactor concepts are the subject of comprehensive investigations. In some projects supercritical water will be considered as coolant, moderator (as for the High Performance Light Water Reactor) or secondary working fluid (one possible option for Liquid Metal-cooled Fast Reactors). Supercritical water is characterized by a pronounced change of the thermo-physical properties when crossing the pseudo-critical line, which goes hand in hand with a change in the heat transfer (HT) behavior. Hence, it is essential to estimate, in a proper way, the heat-transfer coefficient and subsequently the wall temperature. The scope of this paper is to present and discuss the activities at the Institute for Reactor Safety (IRS) related to the implementation of correlations for wall-to-fluid HT at supercritical conditions in Best-Estimate codes like TRACE as well as its validation. It is important to validate TRACE before applying it to safety analyses of HPLWR or of other reactor systems. In the past 3 decades various experiments have been performed all over the world to reveal the peculiarities of wall-to-fluid HT at supercritical conditions. Several different heat transfer phenomena such as HT enhancement (due to higher Prandtl numbers in the vicinity of the pseudo-critical point) or HT deterioration (due to strong property variations) were observed. Since TRACE is a component based system code with a finite volume method the resolution capabilities are limited and not all physical phenomena can be modeled properly. But Best -Estimate system codes are nowadays the preferred option for safety related investigations of full plants or other integral systems. Thus, the increase of the confidence in such codes is of high priority. In this paper, the post-test analysis of experiments with supercritical parameters will be presented. For that reason various correlations for the HT, which considers the characteristics

Analysis of turbulent natural convection heat transfer in a lower plenum during external cooling using the COSMO code

Energy Technology Data Exchange (ETDEWEB)

Noguchi, H. [Nuclear Power Engineering Corp., Tokyo (Japan); Sawatari, Y.; Imada, T. [Fuji Research Institute Corporation, Tokyo (Japan)

2000-11-01

The behavior of a large volumetrically heated melt pool is important to evaluate the feasibility of in-vessel retention by external flooding as an accident management. The COSMO (Coolability Simulation of Molten corium during severe accident) code has been developed at NUPEC to simulate turbulent natural convection heat transfer with internal heat source. The COSMO code solves thermal hydraulic conservation equations with turbulent model and can simulate melting and solidification process. The standard k-{epsilon} model has a limitation to describe the turbulent natural convection in the very high Rayleigh number condition (10{sup 16}-10{sup 17}) assumed to occur in a lower plenum of RPV during a severe accident. This limitation results from the assumption of an analogy of momentum and energy transfer phenomena in the standard model. In this paper the modified turbulent model in which the turbulent number is treated, as a function of the flux Richardson number derived from the experiment, has been incorporated and verified by using the BALI experiments. It was found that the prediction of averaged Nusselt number became better than that of the standard model. In order to extend the COSMO code to the actual scale analysis under the external flooding conditions, more realistic boundary condition derived from the experiments should be treated. In this work the CHF correlation from ULPU experiment or the heat transfer coefficient correlation from CYBL experiment have been applied. The preliminary analysis of an actual scale analysis has been carried out under the condition of the TMI-2 accident. (author)

Analysis of turbulent natural convection heat transfer in a lower plenum during external cooling using the COSMO code

International Nuclear Information System (INIS)

Noguchi, H.; Sawatari, Y.; Imada, T.

2000-01-01

The behavior of a large volumetrically heated melt pool is important to evaluate the feasibility of in-vessel retention by external flooding as an accident management. The COSMO (Coolability Simulation of Molten corium during severe accident) code has been developed at NUPEC to simulate turbulent natural convection heat transfer with internal heat source. The COSMO code solves thermal hydraulic conservation equations with turbulent model and can simulate melting and solidification process. The standard k-ε model has a limitation to describe the turbulent natural convection in the very high Rayleigh number condition (10 16 -10 17 ) assumed to occur in a lower plenum of RPV during a severe accident. This limitation results from the assumption of an analogy of momentum and energy transfer phenomena in the standard model. In this paper the modified turbulent model in which the turbulent number is treated, as a function of the flux Richardson number derived from the experiment, has been incorporated and verified by using the BALI experiments. It was found that the prediction of averaged Nusselt number became better than that of the standard model. In order to extend the COSMO code to the actual scale analysis under the external flooding conditions, more realistic boundary condition derived from the experiments should be treated. In this work the CHF correlation from ULPU experiment or the heat transfer coefficient correlation from CYBL experiment have been applied. The preliminary analysis of an actual scale analysis has been carried out under the condition of the TMI-2 accident. (author)

Heat transfer performance test of PDHRS heat exchangers of PGSFR using STELLA-1 facility

Energy Technology Data Exchange (ETDEWEB)

Hong, Jonggan, E-mail: hong@kaeri.re.kr; Yeom, Sujin; Eoh, Jae-Hyuk; Lee, Tae-Ho; Jeong, Ji-Young

2017-03-15

Highlights: • Heat transfer performance test of heat exchangers of PGSFR PDHRS is conducted using STELLA-1 facility. • Steady-state test results of DHX and AHX show good agreement with theoretical results of design codes. • Design codes for DHX and AHX are validated by STELLA-1 experimental results. • Heat transport capability of DHX and AHX is turned out to be satisfactory for reliable plant operation. - Abstract: The STELLA-1 facility was designed and constructed to carry out separate effect tests of the decay heat exchanger (DHX) and natural draft sodium-to-air heat exchanger (AHX), which are key components of the safety-grade decay heat removal system in PGSFR. The DHX is a sodium-to-sodium heat exchanger with a straight tube arrangement, and the AHX is a sodium-to-air heat exchanger with a helically coiled tube arrangement. The model heat exchangers in STELLA-1 have been designed to meet their own similitude conditions from the prototype ones, of which scale ratios were set to be unity in height (or length) and 1/2.5 in heat transfer rate. Consequently, the overall heat transfer coefficients and log-mean temperature differences of the prototypes have been preserved as well. The steady-state test results for each model heat exchanger obtained from STELLA-1 showed good agreement with the theoretical results of the computer design codes for thermal-sizing and a performance analysis of the DHX and AHX. In the DHX result comparison, the discrepancies in the heat transfer rate ranged from −4.4% to 2.0%, and in the AHX result comparison, they ranged from −11.1% to 12.6%. Therefore, the first step in thermal design codes validation for sodium heat exchangers, e.g., DHX and AHX, has been successfully completed with the experimental database obtained from STELLA-1. In addition, the heat transfer performance of the DHX and AHX was found to be satisfactory enough to secure a reliable decay heat removal performance.

A one-dimensional heat transfer model for parallel-plate thermoacoustic heat exchangers.

Science.gov (United States)

de Jong, J A; Wijnant, Y H; de Boer, A

2014-03-01

A one-dimensional (1D) laminar oscillating flow heat transfer model is derived and applied to parallel-plate thermoacoustic heat exchangers. The model can be used to estimate the heat transfer from the solid wall to the acoustic medium, which is required for the heat input/output of thermoacoustic systems. The model is implementable in existing (quasi-)1D thermoacoustic codes, such as DeltaEC. Examples of generated results show good agreement with literature results. The model allows for arbitrary wave phasing; however, it is shown that the wave phasing does not significantly influence the heat transfer.

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

Radiation heat transfer model for the SCDAP code

International Nuclear Information System (INIS)

Sohal, M.S.

1984-01-01

A radiation heat transfer model has been developed for severe fuel damage analysis which accounts for anisotropic effects of reflected radiation. The model simplifies the view factor calculation which results in significant savings in computational cost with little loss of accuracy. Radiation heat transfer rates calculated by the isotropic and anisotropic models compare reasonably well with those calculated by other models. The model is applied to an experimental nuclear rod bundle during a slow boiloff of the coolant liquid, a situation encountered during a loss of coolant accident with severe fuel damage. At lower temperatures and also lower temperature gradients in the core, the anisotropic effect was not found to be significant

Validation of Numerical Schemes in a Thermal-Hydraulic Analysis Code for a Natural Convection Heat Transfer of a Molten Pool

International Nuclear Information System (INIS)

Kim, Jong Tae; Ha, Kwang Soon; Kim, Hwan Yeol; Park, Rae Joon; Song, Jin Ho

2010-01-01

, unsteady turbulence models based on filtered or volume-averaged governing equations have been applied for the turbulent natural convection heat transfer. Tran et al. used large eddy simulation (LES) for the analysis of molten corium coolability. The numerical instability is related to a gravitational force of the molten corium. A staggered grid method on an orthogonal structured grid is used to prohibit a pressure oscillation in the numerical solution. But it is impractical to use the structured grid for a partially filled spherical pool, a cone-type pool or triangular pool. An unstructured grid is an alternative for the nonrectangular pools. In order to remove the checkerboard- like pressure oscillation on the unstructured grid, some special interpolation scheme is required. In order to evaluate in-vessel coolability of the molten corium for a pressurized water reactor (PWR), thermo-hydraulic analysis code LILAC had been developed. LILAC has a capability of multi-layered conjugate heat transfer with melt solidification. A solution domain can be 2-dimensional, axisymmetric, and 3-dimensional. LILAC is based on the unstructured mesh technology to discretized non-rectangular pool geometry. Because of too limited man-power to maintain the code, it becomes more and more difficult to implement new physical and numerical models in the code along with increased complication of the code. Recently, open source CFD code OpenFOAM has been released and applied to many academic and engineering areas. OpenFOAM is based on the very similar numerical schemes to the LILAC code. It has many physical and numerical models for multi-physics analysis. And because it is based on object-oriented programming, it is known that new models can be easily implemented and is very fast with a lower possibility of coding errors. This is a very attractive feature for the development, validation and maintenance of an analysis code. On the contrary to commercial CFD codes, it is possible to modify and add

Improvement and validation of the wall heat transfer package of RELAP5/MOD3.3

International Nuclear Information System (INIS)

Wu, Pan; Xiong, Xiaofei; Shan, Jianqiang; Gou, Junli; Zhang, Bin; Zhang, Bo

2016-01-01

Highlights: • A new heat transfer package has been developed. • It has been incorporated into RELAP5/MOD3.3 to verify its advantages. • The results of modified code were compared with available experimental data. • The results showed that higher prediction accuracy was achieved. - Abstract: The process of energy transfer from heat structure to control volume is determined by the wall-to-fluid heat transfer package, which is crucial for nuclear reactor safety analysis codes. The current logic for selection of heat transfer modes of RELAP5/MOD3.3 code is too complex and may result in incorrect heat transfer mode judgment. Also, the narrow application scope of film boiling heat transfer correlations may result in large errors in film boiling region which is of paramount importance for the predicted peak clad temperatures during hypothetical LB-LOCAs in PWRs. In this study, a new heat transfer package has been developed and incorporated into the RELAP5/MOD3.3 code. Differing from the original package, the modified one consists of twelve heat transfer modes and proposes a new logic for selection of heat transfer modes. For each mode, the models in the existing safety analysis codes and the leading models in literature have been reviewed in order to determine the best model which can easily be applicable to the RELAP5/MOD3.3 code. Particularly (1) a new package of heat transfer correlations are produced; (2) a new logic for selection of film boiling and transition boiling heat transfer modes is proposed which use minimum film boiling temperature and critical heat flux temperature as distinguished points. The modified code has been validated by comparing the analysis results with available experimental data from tube post dryout experiments and loss-of-fluid test (LOFT) facility. The calculation results showed that the improved package could better predict the experimental phenomena with higher prediction accuracy.

Improvement and validation of the wall heat transfer package of RELAP5/MOD3.3

Energy Technology Data Exchange (ETDEWEB)

Wu, Pan; Xiong, Xiaofei; Shan, Jianqiang, E-mail: jqshan@mail.xjtu.edu.cn; Gou, Junli; Zhang, Bin; Zhang, Bo

2016-12-15

Highlights: • A new heat transfer package has been developed. • It has been incorporated into RELAP5/MOD3.3 to verify its advantages. • The results of modified code were compared with available experimental data. • The results showed that higher prediction accuracy was achieved. - Abstract: The process of energy transfer from heat structure to control volume is determined by the wall-to-fluid heat transfer package, which is crucial for nuclear reactor safety analysis codes. The current logic for selection of heat transfer modes of RELAP5/MOD3.3 code is too complex and may result in incorrect heat transfer mode judgment. Also, the narrow application scope of film boiling heat transfer correlations may result in large errors in film boiling region which is of paramount importance for the predicted peak clad temperatures during hypothetical LB-LOCAs in PWRs. In this study, a new heat transfer package has been developed and incorporated into the RELAP5/MOD3.3 code. Differing from the original package, the modified one consists of twelve heat transfer modes and proposes a new logic for selection of heat transfer modes. For each mode, the models in the existing safety analysis codes and the leading models in literature have been reviewed in order to determine the best model which can easily be applicable to the RELAP5/MOD3.3 code. Particularly (1) a new package of heat transfer correlations are produced; (2) a new logic for selection of film boiling and transition boiling heat transfer modes is proposed which use minimum film boiling temperature and critical heat flux temperature as distinguished points. The modified code has been validated by comparing the analysis results with available experimental data from tube post dryout experiments and loss-of-fluid test (LOFT) facility. The calculation results showed that the improved package could better predict the experimental phenomena with higher prediction accuracy.

A comprehensive examination of heat transfer correlations suitable for reactor safety analysis

International Nuclear Information System (INIS)

Groeneveld, D.C.; Snoek, C.W.

1986-01-01

Due to the inadequate understanding of heat transfer mechanisms, an empirical approach is often necessary. This approach requires the derivation of empirical heat transfer correlations for each heat transfer configuration, resulting in numerous correlations for each heat transfer mode. A simplification that is frequently used is to combine these heat transfer correlations using some suitably defined local parameters to characterize the heat transfer process. These local condition correlations, usually encountered in reactor safety codes are discussed in this paper

Heat transfer characteristics around a single heated rod immersed in sodium pool with gas jet injection

International Nuclear Information System (INIS)

Hideto Niikura; Kazuo Soga; Ken-ichiro Sugiyama; Akira Yamaguchi

2005-01-01

In a steam generator using liquid sodium, water intensely reacts with sodium when it leaks out from a heat transfer tube. It is important to evaluate the influence of sodium-water reaction to surrounding tubes and the shell. Hence, it has been desired to develop the simulation code for the evaluation of sodium-water reaction. From this viewpoint, the Japan Nuclear Cycle is now developing the SERAPHIM code. We reported a preliminary study to establish an experimental method for a single heated rod immersed in sodium pool with steam jet impingement planned in the near future as well as to obtain a preliminary data to verify the adequacy of SERAPHIM code. We first measured local and mean heat transfer coefficients around a horizontal single heated rod immersed in a water pool and a sodium pool with a limited volume in the experimental apparatus. It was confirmed that the mean heat transfer coefficients fairly agreed with the existing data for natural convection in water and sodium. Secondary we measured local and mean heat transfer coefficients around a horizontal single heated rod with Ar gas jet impingement immersed in the limited water pool and in the limited sodium pool. It was clearly observed that the local heat transfer coefficients in the sodium pool keep almost the same values in every angle regardless of increase in Ar gas jet velocity varied from about 8.7m/s to about 78m/s. On the other hand, it was confirmed in the water pool that local heat transfer coefficients on the forward stagnation side exposed in the Ar gas jet impingement increase with increasing the jet velocity while the local heat transfer coefficients on the opposite surface keep almost same values regardless of increase in the velocity. (authors)

Heat transfer in plate heat exchanger channels: Experimental validation of selected correlation equations

Directory of Open Access Journals (Sweden)

Cieśliński Janusz T.

2016-09-01

Full Text Available This study is focused on experimental investigation of selected type of brazed plate heat exchanger (PHEx. The Wilson plot approach was applied in order to estimate heat transfer coefficients for the PHEx passages. The main aim of the paper was to experimentally check ability of several correlations published in the literature to predict heat transfer coefficients by comparison experimentally obtained data with appropriate predictions. The results obtained revealed that Hausen and Dittus-Boelter correlations underestimated heat transfer coefficient for the tested PHEx by an order of magnitude. The Aspen Plate code overestimated heat transfer coefficient by about 50%, while Muley-Manglik correlation overestimated it from 1% to 25%, dependent on the value of Reynolds number and hot or cold liquid side.

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)

Accounting for the Effect of Noncondensing Gases on Interphasic Heat and Mass Transfer in the Two-Fluid Model Used in the KORSAR Code

Science.gov (United States)

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

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.)

Heat transfer in a magnet C

International Nuclear Information System (INIS)

Sircilli Neto, F.; Passaro, A.; Borges, E.M.

1991-01-01

The cooling systems of nuclear reactors for spacial applications include direct current electromagnetic pumps, which are used to circulate the coolant fluid thru the reactor core. In this work, the transfer of the heat generated by the electrical current in a magnet C excitation coils, which is used in a prototype pump, was evaluated. Considering the processes of heat transfer by conduction, natural convection and radiation, the results of simulation with the codes HEATING5 and AUTHEATS indicate the utilization of the 180 sup(0)C thermal class conductor for a working Joule power of 4 10 sup(4) W/m sup(3) in each magnet coil. (author)

Heat transfer

International Nuclear Information System (INIS)

Saad, M.A.

1985-01-01

Heat transfer takes place between material systems as a result of a temperature difference. The transmission process involves energy conversions governed by the first and second laws of thermodynamics. The heat transfer proceeds from a high-temperature region to a low-temperature region, and because of the finite thermal potential, there is an increase in entropy. Thermodynamics, however, is concerned with equilibrium states, which includes thermal equilibrium, irrespective of the time necessary to attain these equilibrium states. But heat transfer is a result of thermal nonequilibrium conditions, therefore, the laws of thermodynamics alone cannot describe completely the heat transfer process. In practice, most engineering problems are concerned with the rate of heat transfer rather than the quantity of heat being transferred. Resort then is directed to the particular laws governing the transfer of heat. There are three distinct modes of heat transfer: conduction, convection, and radiation. Although these modes are discussed separately, all three types may occur simultaneously

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)

Improved inter-assembly heat transfer modeling under low flow conditions for the Super System Code (SSC)

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

Numerical assessment and comparison of heat transfer characteristics of supercritical water in bare tubes and tubes with heat transfer enhancing appendages

International Nuclear Information System (INIS)

Farah, Amjad; Harvel, Glenn; Pioro, Igor

2015-01-01

Computational Fluid Dynamics (CFD) is a numerical approach to model fluids in multidimensional space using the Navier-Stokes equations and databases of fluid properties to arrive at a full simulation of a fluid dynamics and heat transfer system. A numerical study on heat transfer to supercritical water (SCW) flowing in a vertical tube is carried out using the ANSYS FLUENT code and employing the SST k-ω turbulence model. The 3D mesh consists of a 1/8 section (45deg radially) of a bare tube. The numerical results on wall temperature distributions under normal and deteriorated heat transfer conditions are compared to experimental results. The same geometry is then simulated with an orifice to study the effect of geometrical perturbation on the flow and heat transfer characteristics of SCW. The orifice is placed areas to test the effect on normal, deteriorated and enhanced heat transfer regimes. The flow effects and heat transfer characteristics will be studied around the appendages to arrive at a fundamental understanding of the phenomena related to supercritical water turbulence. (author)

Heat transfer enhancement

International Nuclear Information System (INIS)

Hasatani, Masanobu; Itaya, Yoshinori

1985-01-01

In order to develop energy-saving techniques and new energy techniques, and also most advanced techniques by making industrial equipment with high performance, heat transfer performance frequently becomes an important problem. In addition, the improvement of conventional heat transfer techniques and the device of new heat transfer techniques are often required. It is most proper that chemical engineers engage in the research and development for enhancing heat transfer. The research and development for enhancing heat transfer are important to heighten heat exchange efficiency or to cool equipment for preventing overheat in high temperature heat transfer system. In this paper, the techniques of enhancing radiative heat transfer and the improvement of radiative heat transfer characteristics are reported. Radiative heat transfer is proportional to fourth power of absolute temperature, and it does not require any heat transfer medium, but efficient heat-radiation converters are necessary. As the techniques of enhancing radiative heat transfer, the increase of emission and absorption areas, the installation of emissive structures and the improvement of radiative characteristics are discussed. (Kako, I.)

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.

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

Discussion on the Heat and Mass Transfer Model on the Pool Surface

Energy Technology Data Exchange (ETDEWEB)

Hong, Soon-Joon; Choo, Yeon-Jun [FNC Tech., Yongin (Korea, Republic of); Ha, Sang-Jun [KHNP Central Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Heat transfer on the pool surface involves the evaporation and condensation of steam in the presence of non-condensable gas. It is a kind of inter-phase heat transfer. This phenomenon has been regarded as less important on the thermal hydraulic behaviors such as pressure, temperature, hydrogen distribution, and so on in the nuclear reactor containment building. As a matter of fact, several RAIs (requests for additional information) during the licensing review of the developed CAP have been presented. And early in 2000s the steam condensation on the water surface of IRWST was a concern of APR1400 design. Such an increased concern is believed because it is a newly adopted system. This study discusses the pool surface heat transfer by reviewing the models of several well-known containment analysis codes, and conducting the sensitivities. This study discussed the pool surface heat transfer. The related models of CAP, GOTHIC, CONTEMPT-LT, and CONTEMPT4 were compared. The sensitivity of heat transfer coefficient for SKN3 and 4 using conventional code CONTEMPT-LT/028-A showed little effect. And the sensitivity of relative humidity and heat transfer area for latent heat transfer shows that CAP locates between GOTHIC and CONTEMPT4/MOD. The sensitivity for sensible heat transfer also shows similar trend. Conclusively, current CAP model of pool surface heat transfer has no fatal defect.

Discussion on the Heat and Mass Transfer Model on the Pool Surface

International Nuclear Information System (INIS)

Hong, Soon-Joon; Choo, Yeon-Jun; Ha, Sang-Jun

2016-01-01

Heat transfer on the pool surface involves the evaporation and condensation of steam in the presence of non-condensable gas. It is a kind of inter-phase heat transfer. This phenomenon has been regarded as less important on the thermal hydraulic behaviors such as pressure, temperature, hydrogen distribution, and so on in the nuclear reactor containment building. As a matter of fact, several RAIs (requests for additional information) during the licensing review of the developed CAP have been presented. And early in 2000s the steam condensation on the water surface of IRWST was a concern of APR1400 design. Such an increased concern is believed because it is a newly adopted system. This study discusses the pool surface heat transfer by reviewing the models of several well-known containment analysis codes, and conducting the sensitivities. This study discussed the pool surface heat transfer. The related models of CAP, GOTHIC, CONTEMPT-LT, and CONTEMPT4 were compared. The sensitivity of heat transfer coefficient for SKN3 and 4 using conventional code CONTEMPT-LT/028-A showed little effect. And the sensitivity of relative humidity and heat transfer area for latent heat transfer shows that CAP locates between GOTHIC and CONTEMPT4/MOD. The sensitivity for sensible heat transfer also shows similar trend. Conclusively, current CAP model of pool surface heat transfer has no fatal defect

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

Cryogenic heat transfer

CERN Document Server

Barron, Randall F

2016-01-01

Cryogenic Heat Transfer, Second Edition continues to address specific heat transfer problems that occur in the cryogenic temperature range where there are distinct differences from conventional heat transfer problems. This updated version examines the use of computer-aided design in cryogenic engineering and emphasizes commonly used computer programs to address modern cryogenic heat transfer problems. It introduces additional topics in cryogenic heat transfer that include latent heat expressions; lumped-capacity transient heat transfer; thermal stresses; Laplace transform solutions; oscillating flow heat transfer, and computer-aided heat exchanger design. It also includes new examples and homework problems throughout the book, and provides ample references for further study.

Incorporation of a Helical Tube Heat Transfer Model in the MARS Thermal Hydraulic Systems Analysis Code for the T/H Analyses of the SMART Reactor

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)

Implementation of a Transition Model in a NASA Code and Validation Using Heat Transfer Data on a Turbine Blade

Science.gov (United States)

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

Heat transfer

Indian Academy of Sciences (India)

First page Back Continue Last page Overview Graphics. Heat transfer. Heat conduction in solid slab. Convective heat transfer. Non-linear temperature. variation due to flow. HEAT FLUX AT SURFACE. conduction/diffusion.

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

Numerical investigation of two- and three-dimensional heat transfer in expander cycle engines

Science.gov (United States)

Burch, Robert L.; Cheung, Fan-Bill

1993-01-01

The concept of using tube canting for enhancing the hot-side convective heat transfer in a cross-stream tubular rocket combustion chamber is evaluated using a CFD technique in this study. The heat transfer at the combustor wall is determined from the flow field generated by a modified version of the PARC Navier-Stokes Code, using the actual dimensions, fluid properties, and design parameters of a split-expander demonstrator cycle engine. The effects of artificial dissipation on convergence and solution accuracy are investigated. Heat transfer results predicted by the code are presented. The use of CFD in heat transfer calculations is critically examined to demonstrate the care needed in the use of artificial dissipation for good convergence and accurate solutions.

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)

Convective heat transfer

CERN Document Server

Kakac, Sadik; Pramuanjaroenkij, Anchasa

2014-01-01

Intended for readers who have taken a basic heat transfer course and have a basic knowledge of thermodynamics, heat transfer, fluid mechanics, and differential equations, Convective Heat Transfer, Third Edition provides an overview of phenomenological convective heat transfer. This book combines applications of engineering with the basic concepts of convection. It offers a clear and balanced presentation of essential topics using both traditional and numerical methods. The text addresses emerging science and technology matters, and highlights biomedical applications and energy technologies. What’s New in the Third Edition: Includes updated chapters and two new chapters on heat transfer in microchannels and heat transfer with nanofluids Expands problem sets and introduces new correlations and solved examples Provides more coverage of numerical/computer methods The third edition details the new research areas of heat transfer in microchannels and the enhancement of convective heat transfer with nanofluids....

Modeling microscale heat transfer using Calore.

Energy Technology Data Exchange (ETDEWEB)

Gallis, Michail A.; Rader, Daniel John; Wong, Chung-Nin Channy; Bainbridge, Bruce L.; Torczynski, John Robert; Piekos, Edward Stanley

2005-09-01

Modeling microscale heat transfer with the computational-heat-transfer code Calore is discussed. Microscale heat transfer problems differ from their macroscopic counterparts in that conductive heat transfer in both solid and gaseous materials may have important noncontinuum effects. In a solid material, three noncontinuum effects are considered: ballistic transport of phonons across a thin film, scattering of phonons from surface roughness at a gas-solid interface, and scattering of phonons from grain boundaries within the solid material. These processes are modeled for polycrystalline silicon, and the thermal-conductivity values predicted by these models are compared to experimental data. In a gaseous material, two noncontinuum effects are considered: ballistic transport of gas molecules across a thin gap and accommodation of gas molecules to solid conditions when reflecting from a solid surface. These processes are modeled for arbitrary gases by allowing the gas and solid temperatures across a gas-solid interface to differ: a finite heat transfer coefficient (contact conductance) is imposed at the gas-solid interface so that the temperature difference is proportional to the normal heat flux. In this approach, the behavior of gas in the bulk is not changed from behavior observed under macroscopic conditions. These models are implemented in Calore as user subroutines. The user subroutines reside within Sandia's Source Forge server, where they undergo version control and regression testing and are available to analysts needing these capabilities. A Calore simulation is presented that exercises these models for a heated microbeam separated from an ambient-temperature substrate by a thin gas-filled gap. Failure to use the noncontinuum heat transfer models for the solid and the gas causes the maximum temperature of the microbeam to be significantly underpredicted.

First international workshop on fundamental aspects of post-dryout heat transfer: proceedings

International Nuclear Information System (INIS)

Lee, R.

1984-12-01

The purpose of the First International Workshop on Fundamental Aspects of Post-Dryout Heat Transfer was to review recent developments and the state of art in the field of post-dryout heat transfer. The workshop centered on interchanging ideas, reviewing current research results, and defining future research needs. The following five sessions dealing with the fundamental aspects of post-dryout heat transfer were held. A Computer Code Modeling and Flow Phenomena session was held dealing with flow rgimes, drop size, drop formation and behavior, interfacial area, interfacial drag, and computer modeling. A Quenching Phenomena session was held dealing with nature of rewetting, maximum wetting temperature, Leidenfrost phenomenon and heat transfer in the vicinity of quench front. A Low-Void Heat Transfer session was held dealing with inverted annular-flow heat transfer, inverted slug-flow heat transfer thermal non-equilibrium and computer modeling. A Dispersed-Flow Heat Transfer session was held dealing with drop interfacial heat transfer, vapor convection, thermal non-equilibrium and correlations and models

Analysis of natural convection heat transfer with crust formation in the molten metal pool using CONV-2 and 3D computer codes

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

Heat transfer: Pittsburgh 1987

International Nuclear Information System (INIS)

Lyczkowski, R.W.

1987-01-01

This book contains papers divided among the following sections: Process Heat Transfer; Thermal Hydraulics and Phase Change Phenomena; Analysis of Multicomponent Multiphase Flow and Heat Transfer; Heat Transfer in Advanced Reactors; General Heat Transfer in Solar Energy; Numerical Simulation of Multiphase Flow and Heat Transfer; High Temperature Heat Transfer; Heat Transfer Aspects of Severe Reactor Accidents; Hazardous Waste On-Site Disposal; and General Papers

CFD analysis of heat transfer in a vertical annular gas gap

International Nuclear Information System (INIS)

Borgohain, A.; Maheshwari, N.K.; Vijayan, P.K.

2011-01-01

Heat transfer analysis in a vertical annulus is carried out by using a CFD code TRIO-U. The results are used to understand heat transfer in the vertical annulus. An experimental study is taken from literature for the CFD analysis. The geometry of the test section of the experiment is simulated in TRIO-U. The operating conditions of the experiment are simulated by imposing appropriate boundary conditions. Three modes of the heat transfer, conduction, radiation and convection in the gas gap are considered in the analysis. From the analysis it is found that TRIO-U can successfully handle all modes heat transfer. The theoretical results for heat transfer have been compared with experimental data. This paper deals with the detailed CFD modelling and analysis. (author)

Development of tube rupture evaluation code for FBR steam generator (II). Modification of heat transfer model in sodium side

International Nuclear Information System (INIS)

Hamada, H.; Kurihara, A.

2003-05-01

The thermal effect of sodium-water reaction jet on neighboring heat transfer tubes was examined to rationally evaluate the structural integrity of the tube for overheating rupture under a water leak in an FBR steam generator. Then, the development of new heat transfer model and the application analysis were carried out. Main results in this paper are as follows. (1) The evaluation method of heat flux and heat transfer coefficient (HTC) on the tube exposed to reaction jet was developed. By using the method, it was confirmed that the heat flux could be realistically evaluated in comparison with the previous method. (2) The HTC between reaction jet and the tube was theoretically examined in the two-phase flow model, and new heat transfer model considering the effect of fluid temperature and cover gas pressure was developed. By applying the model, a tentative experimental correlation was conservatively obtained by using SWAT-1R test data. (3) The new model was incorporated to the Tube Rupture Evaluation Code (TRUE), and the conservatism of the model was confirmed by using sodium-water reaction data such as the SWAT-3 tests. (4) In the application analysis of the PFR large leak event, there was no significant difference of calculation results between the new model and previous one; the importance of depressurization in the tube was confirmed. (5) In the application analysis of the Monju evaporator, it was confirmed that the calculation result in the previous model would be more conservative than that in the new one and that the maximum cumulative damage of 25% could be reduced in the new model. (author)

Simulations and experiments of laminar heat transfer for Therminol heat transfer fluids in a rifled tube

International Nuclear Information System (INIS)

Xu, Weiguo; Ren, Depeng; Ye, Qing; Liu, Guodong; Lu, Huilin; Wang, Shuai

2016-01-01

Graphical abstract: Predicted laminar Nusselt number using regression correlation of Therminol-55 heat transfer fluid is in agreement with experiments in the rifled tube. - Highlights: • Heat transfer coefficient and friction factor are measured and predicted in the rifled tube. • Correlations for Nusselt number and friction factor are proposed. • The roughness height of 0.425 mm in transition SST model is suggested as an input parameter. • k–kl–ω transition and transition SST models are recommended for laminar–turbulent transition. • Thermal enhancement factor and synergy angle are predicted in the rifled tube. - Abstract: Simulations and experiments of flow and heat transfer behavior of Therminol-55 heat transfer fluid have been conducted in a horizontal rifled tube with outer diameter and inner diameter 25.0 and 20.0 mm, pitch and rib height of 12.0 and 1.0 mm, respectively. Numerical simulations of three-dimensional flow behavior of Therminol-55 heat transfer fluid are carried out using FLUENT code in the rifled tube. Experimental results show that the heat transfer and thermal performance of Therminol-55 heat transfer fluid in the rifled tube are considerably improved compared to those of the smooth tube. The Nusselt number increases with the increase of Reynolds number, and is from 3.5 to 5.1 times over the smooth tube. Also, the pressure drop results reveal that the average friction factor of the ribbed tube is in a range of 2.2 and 4.2 times over the smooth tube. Predictive Nusselt number and friction factor correlations have been presented. The numerical results show that the laminar flow model is valid only at lower Reynolds number in the developed laminar flow of rifled tube. The k–kl–ω transition model and transition SST model with roughness of 0.425 mm are recommended for the predictions of transition process from laminar to turbulent flow in the rifled tube.

Heat transfer system

Science.gov (United States)

Not Available

1980-03-07

A heat transfer system for a nuclear reactor is described. Heat transfer is accomplished within a sealed vapor chamber which is substantially evacuated prior to use. A heat transfer medium, which is liquid at the design operating temperatures, transfers heat from tubes interposed in the reactor primary loop to spaced tubes connected to a steam line for power generation purposes. Heat transfer is accomplished by a two-phase liquid-vapor-liquid process as used in heat pipes. Condensible gases are removed from the vapor chamber through a vertical extension in open communication with the chamber interior.

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)

Solving the heat transfer in the cold rain of a cross flow cooling tower. N3S code - cooling tower release

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

Influence on Heat Transfer Coefficient of Heat Exchanger by Velocity and Heat Transfer Temperature Difference

Directory of Open Access Journals (Sweden)

WANG Fang

2017-04-01

Full Text Available Aimed to insufficient heat transfer of heat exchanger, research the influence on the heat transfer coefficient impacted by velocity and heat transfer temperature difference of tube heat exchanger. According to the different heat transfer temperature difference and gas velocity，the experimental data were divided into group. Using the control variable method，the above two factors were analyzed separately. K一△T and k一:fitting curve were clone to obtain empirical function. The entire heat exchanger is as the study object，using numerical simulation methods，porous media，k一￡model，second order upwind mode，and pressure一velocity coupling with SIMPLE algorithm，the entire heat exchanger temperature field and the heat transfer coefficient distribution were given. Finally the trend of the heat transfer coefficient effected by the above two factors was gotten.

POST CRITICAL HEAT TRANSFER AND FUEL CLADDING OXIDATION

Directory of Open Access Journals (Sweden)

Vojtěch Caha

2016-12-01

Full Text Available The knowledge of heat transfer coefficient in the post critical heat flux region in nuclear reactor safety is very important. Although the nuclear reactors normally operate at conditions where critical heat flux (CHF is not reached, accidents where dryout occur are possible. Most serious postulated accidents are a loss of coolant accident or reactivity initiated accident which can lead to CHF or post CHF conditions and possible disruption of core integrity. Moreover, this is also influenced by an oxide layer on the cladding surface. The paper deals with the study of mathematical models and correlations used for heat transfer calculation, especially in post dryout region, and fuel cladding oxidation kinetics of currently operated nuclear reactors. The study is focused on increasing of accuracy and reliability of safety limit calculations (e.g. DNBR or fuel cladding temperature. The paper presents coupled code which was developed for the solution of forced convection flow in heated channel and oxidation of fuel cladding. The code is capable of calculating temperature distribution in the coolant, cladding and fuel and also the thickness of an oxide layer.

Heat transfer probe

Science.gov (United States)

Frank, Jeffrey I.; Rosengart, Axel J.; Kasza, Ken; Yu, Wenhua; Chien, Tai-Hsin; Franklin, Jeff

2006-10-10

Apparatuses, systems, methods, and computer code for, among other things, monitoring the health of samples such as the brain while providing local cooling or heating. A representative device is a heat transfer probe, which includes an inner channel, a tip, a concentric outer channel, a first temperature sensor, and a second temperature sensor. The inner channel is configured to transport working fluid from an inner inlet to an inner outlet. The tip is configured to receive at least a portion of the working fluid from the inner outlet. The concentric outer channel is configured to transport the working fluid from the inner outlet to an outer outlet. The first temperature sensor is coupled to the tip, and the second temperature sensor spaced apart from the first temperature sensor.

Basic heat transfer

CERN Document Server

Bacon, D H

2013-01-01

Basic Heat Transfer aims to help readers use a computer to solve heat transfer problems and to promote greater understanding by changing data values and observing the effects, which are necessary in design and optimization calculations.The book is concerned with applications including insulation and heating in buildings and pipes, temperature distributions in solids for steady state and transient conditions, the determination of surface heat transfer coefficients for convection in various situations, radiation heat transfer in grey body problems, the use of finned surfaces, and simple heat exc

Numerical Modeling of Transient Heat Transfer in Longitudinal Fin

Directory of Open Access Journals (Sweden)

Farshad Panahizadeh

2017-11-01

Full Text Available The main objective of the present numerical study is to investigate the transient heat transfer in one kind of all-purpose longitudinal fin with the triangular profile. The lateral surface of the concerned fin and the tip of it are subjected to general situations included heat flux at the base and insulation on the tip. For this study developed a one dimensional in house code written by Fortran 90 programming language by using finite difference method with an implicit scheme in unsteady state condition. Generally, the result of this study in time variation state after 700 seconds is steady. The results also show the fin efficiency by increasing the time of study decreases due to a reduction in the total heat transfer which is happened in the fin. The grid independence study shows that for the number of nodes greater than 20 the result will not be changed and same as before. Finally, the result of Fortran code verified by commercial CFD code which relies on finite difference method and it was shown have a consistent agreement

BWR spent fuel storage cask performance test. Volume 2. Pre- and post-test decay heat, heat transfer, and shielding analyses

International Nuclear Information System (INIS)

Wiles, L.E.; Lombardo, N.J.; Heeb, C.M.; Jenquin, U.P.; Michener, T.E.; Wheeler, C.L.; Creer, J.M.; McCann, R.A.

1986-06-01

This report describes the decay heat, heat transfer, and shielding analyses conducted in support of performance testing of a Ridhihalgh, Eggers and Associates REA 2033 boiling water reactor (BWR) spent fuel storage cask. The cask testing program was conducted for the US Department of Energy (DOE) Commercial Spent Fuel Management Program by the Pacific Northwest Laboratory (PNL) and by General Electric at the latters' Morris Operation (GE-MO) as reported in Volume I. The analyses effort consisted of performing pretest calculations to (1) select spent fuel for the test; (2) symmetrically load the spent fuel assemblies in the cask to ensure lateral symmetry of decay heat generation rates; (3) optimally locate temperature and dose rate instrumentation in the cask and spent fuel assemblies; and (4) evaluate the ORIGEN2 (decay heat), HYDRA and COBRA-SFS (heat transfer), and QAD and DOT (shielding) computer codes. The emphasis of this second volume is on the comparison of code predictions to experimental test data in support of the code evaluation process. Code evaluations were accomplished by comparing pretest (actually pre-look, since some predictions were not completed until testing was in progress) predictions with experimental cask testing data reported in Volume I. No attempt was made in this study to compare the two heat transfer codes because results of other evaluations have not been completed, and a comparison based on one data set may lead to erroneous conclusions

Methodology for verification of heat transfer crisis in the nuclear fuel assemblies

International Nuclear Information System (INIS)

Sharaevsky, I. G.; Sharaevskaya, E. I.; Domashev, E. D.; Arkhypov, A. P.; Kolochko, V. N.

2003-01-01

Reliable operation of water-water type nuclear energy units and design of new generation reactors are not to be provided with wide application of best estimate ThermalHydraulic (TH) codes. It is accepted to consider that up-to-date versions of the codes are featured not only by wide range of NPPs equipment modeling and high ergonomic characteristics of realized in the codes interfaces but comprehensive substantiation of its governing component viz correlations and closure relations systems The pointed correlations and closure relations provide mathematical restraint of the main differential equations system which are necessary for adequate description of the main classes of two-phase flow TH regimes. The principal fact is that without physically justificated correlations and adequate closure relations first of all concerning heat transfer crisis at boiling (DNB) the acceptable reliability of numerical solutions cannot be guaranteed by the codes. But the significant part of realized in the codes correlations mainly on heat transfer crisis are based on the experimental data obtained more than 30 years ago for cylindrical channels. It is known that for TH reliability calculations of the WWERs core with rod fuel elements, such correlations can be applied with caution as it give significantly conservative values of critical heat flux especially at under pressure accident regimes. Moreover because of irregularity of the flow TH parameters on fuel rod elements cross-section distribution the heat transfer crisis regimes are originated only in separate 'hot' cells. Additionally it should be underlined that realized in the codes correlations and closure relations do not consider possibility occurring in the steam generating channels high frequency oscillation instability which poses a threat to the reactor safety. The high frequency oscillations can bring to the fuel elements destruction at heat fluxes much less than the critical ones. Now this type of oscillation

Simulation of the VISTA SG heat transfer experiment using MIDAS/SMR

International Nuclear Information System (INIS)

Park, Jong Hwa; Kim, Dong Ha; Chung, Young Jong; Park, Sun Hee; Cho, Seong Won

2011-01-01

As the SMART plant was designed with the helical type tubes in the steam generators, the heat transfer model in that geometry has been implemented in the TASS/SMR-S code and used for the safety analysis. The same correlation was implemented in the MIDAS/SMR, which is being used for the severe accident analyses, to model heat transfer at the steam generators. In this study, the VISTA SG experiment with the helical steam generator tube was simulated with MIDAS/SMR to compare the heat transfer rates through the helical tube

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

Experimental heat transfer to supercritical carbon dioxide flowing upward vertical tube with highly conducting surroundings

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.

Numerical Modeling of Ablation Heat Transfer

Science.gov (United States)

Ewing, Mark E.; Laker, Travis S.; Walker, David T.

2013-01-01

A unique numerical method has been developed for solving one-dimensional ablation heat transfer problems. This paper provides a comprehensive description of the method, along with detailed derivations of the governing equations. This methodology supports solutions for traditional ablation modeling including such effects as heat transfer, material decomposition, pyrolysis gas permeation and heat exchange, and thermochemical surface erosion. The numerical scheme utilizes a control-volume approach with a variable grid to account for surface movement. This method directly supports implementation of nontraditional models such as material swelling and mechanical erosion, extending capabilities for modeling complex ablation phenomena. Verifications of the numerical implementation are provided using analytical solutions, code comparisons, and the method of manufactured solutions. These verifications are used to demonstrate solution accuracy and proper error convergence rates. A simple demonstration of a mechanical erosion (spallation) model is also provided to illustrate the unique capabilities of the method.

Heat Transfer Computations of Internal Duct Flows With Combined Hydraulic and Thermal Developing Length

Science.gov (United States)

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.

Inter-subchannel heat transfer modeling for a subchannel analysis of liquid metal-cooled reactors

International Nuclear Information System (INIS)

Hae-Yong, Jeong; Kwi-Seok, Ha; Young-Min, Kwon; Yong-Bum, Lee; Dohee, Hahn

2007-01-01

In a subchannel approach, the temperature, pressure and velocity in a subchannel are averaged, and one representative thermal-hydraulic condition specifies the state of a subchannel. To enhance the predictability of a subchannel analysis code, it is required to model the inter-subchannel heat transfer between the adjacent subchannels as accurately as possible. One of the critical parameters which determine the thermal-hydraulic behavior of the coolant in subchannels is the heat conduction between two neighboring sub-channels. This portion of a heat transfer becomes more important in the design of an LMR (Liquid Metal-cooled Reactor) because of the high heat capacity of the liquid metal coolant. The other important part of heat transfer is the mixing of flow as a form of cross flow. Especially, the turbulent mixing caused by the eddy motion of fluid across the gap between the subchannels enhances the exchange of the momentum and the energy through the gap with no net transport of the mass. Major results of recent efforts on these modeling have been implemented in a subchannel analysis code MATRA-LMR-FB. The analysis shows that the accuracy of a subchannel analysis code is improved by enhancing the models describing the conduction heat transfer and the cross-flow mixing, especially at low flow rate. (authors)

Analysis of Heat Transfer

International Nuclear Information System (INIS)

2003-08-01

This book deals with analysis of heat transfer which includes nonlinear analysis examples, radiation heat transfer, analysis of heat transfer in ANSYS, verification of analysis result, analysis of heat transfer of transition with automatic time stepping and open control, analysis of heat transfer using arrangement of ANSYS, resistance of thermal contact, coupled field analysis such as of thermal-structural interaction, cases of coupled field analysis, and phase change.

Numerical Modeling of Conjugate Heat Transfer in Fluid Network

Science.gov (United States)

Majumdar, Alok

2004-01-01

Fluid network modeling with conjugate heat transfer has many applications in Aerospace engineering. In modeling unsteady flow with heat transfer, it is important to know the variation of wall temperature in time and space to calculate heat transfer between solid to fluid. Since wall temperature is a function of flow, a coupled analysis of temperature of solid and fluid is necessary. In cryogenic applications, modeling of conjugate heat transfer is of great importance to correctly predict boil-off rate in propellant tanks and chill down of transfer lines. In TFAWS 2003, the present author delivered a paper to describe a general-purpose computer program, GFSSP (Generalized Fluid System Simulation Program). GFSSP calculates flow distribution in complex flow circuit for compressible/incompressible, with or without heat transfer or phase change in all real fluids or mixtures. The flow circuit constitutes of fluid nodes and branches. The mass, energy and specie conservation equations are solved at the nodes where as momentum conservation equations are solved at the branches. The proposed paper describes the extension of GFSSP to model conjugate heat transfer. The network also includes solid nodes and conductors in addition to fluid nodes and branches. The energy conservation equations for solid nodes solves to determine the temperatures of the solid nodes simultaneously with all conservation equations governing fluid flow. The numerical scheme accounts for conduction, convection and radiation heat transfer. The paper will also describe the applications of the code to predict chill down of cryogenic transfer line and boil-off rate of cryogenic propellant storage tank.

A study on gap heat transfer of LWR fuel rods under reactivity initiated accident conditions

International Nuclear Information System (INIS)

Fujishiro, Toshio

1984-03-01

Gap heat transfer between fuel pellet and cladding have a large influence on the LWR fuel behaviors under reactivity initiated accident (RIA) conditions. The objective of the present study is to investigate the effects of gap heat transfer on RIA fuel behaviors based on the results of the gap-gas parameter tests in NSRR and on their analysis with NSR-77 code. Through this study, transient variations of gap heat transfer, the effects of the gap heat transfer on fuel thermal behaviors and on fuel failure, effects of pellet-cladding sticking by eutectic formation, and the effects of cladding collapse under high external pressure have been clearified. The studies have also been performed on the applicability and its limit of modified Ross and Stoute equation which is extensively utilized to evaluate the gap heat transfer coefficient in the present fuel behavior codes. The method to evaluate the gap conductance to the conditions beyond the applicability limit of the Ross and Stoute equation has also been proposed. (author)

Comparison of the results of several heat transfer computer codes when applied to a hypothetical nuclear waste repository

International Nuclear Information System (INIS)

Claiborne, H.C.; Wagner, R.S.; Just, R.A.

1979-12-01

A direct comparison of transient thermal calculations was made with the heat transfer codes HEATING5, THAC-SIP-3D, ADINAT, SINDA, TRUMP, and TRANCO for a hypothetical nuclear waste repository. With the exception of TRUMP and SINDA (actually closer to the earlier CINDA3G version), the other codes agreed to within +-5% for the temperature rises as a function of time. The TRUMP results agreed within +-5% up to about 50 years, where the maximum temperature occurs, and then began an oscillary behavior with up to 25% deviations at longer times. This could have resulted from time steps that were too large or from some unknown system problems. The available version of the SINDA code was not compatible with the IBM compiler without using an alternative method for handling a variable thermal conductivity. The results were about 40% low, but a reasonable agreement was obtained by assuming a uniform thermal conductivity; however, a programming error was later discovered in the alternative method. Some work is required on the IBM version to make it compatible with the system and still use the recommended method of handling variable thermal conductivity. TRANCO can only be run as a 2-D model, and TRUMP and CINDA apparently required longer running times and did not agree in the 2-D case; therefore, only HEATING5, THAC-SIP-3D, and ADINAT were used for the 3-D model calculations. The codes agreed within +-5%; at distances of about 1 ft from the waste canister edge, temperature rises were also close to that predicted by the 3-D model

Wire-wrapped rod-bundle heat-transfer analysis for LMFBR

International Nuclear Information System (INIS)

Wong, C.N.C.; Todreas, N.E.

1982-07-01

Helical wire wraps are widely used in the LMFBR fuel and blanket assemblies to provide coolant mixing and maintain proper spacing between fuel pins. The presence of the helical wire, however, may possibly induce heat transfer problems, such as the uncertainty of the maximum clad temperature as a result of the contact between the wires and the pins. In this study, the detailed transient three dimensional velocity and temperature distributions for the coolant around the pin will be determined by solving the governing momentum and energy equation numerically. A computer code HEATRAN has been developed to perform this calculation. Before the computer code HEATRAN is applied to the wire wrapped rod bundle problem, it is used to analyze a wide range of fluid and heat transfer problem to verify its capabilities

Evaluation of upward heat flux in ex-vessel molten core heat transfer using MELCOR

International Nuclear Information System (INIS)

Park, S.Y.; Park, J.H.; Kim, S.D.; Kim, D.H.; Kim, H.D.

2000-01-01

The purpose of this study is to share experiences of MELCOR application to resolve the molten corium-concrete interaction (MCCI) issue in the Korea Next Generation Reactor (KNGR). In the evaluation of concrete erosion, the heat transfer modeling from the molten corium internal to the corium pool surface is very important and uncertain. MELCOR employs Kutateladze or Greene's bubble-enhanced heat transfer model for the internal heat transfer. The phenomenological uncertainty is so large that the model provides several model parameters in addition to the phenomenological model for user flexibility. However, the model parameters do not work on Kutateladze correlation at the top of the molten layer. From our experience, a code modification is suggested to match the upward heat flux with the experimental results. In this analysis, minor modification was carried out to calculate heat flux from the top molten layer to corium surface, and efforts were made to find out the best value of the model parameter based on upward heat flux of MACE test M1B. Discussion also includes its application to KNGR. (author)

Blowdown heat transfer surface in RELAP4/MOD6 and data comparisons

International Nuclear Information System (INIS)

Nelson, R.A.; Sullivan, L.H.

1978-01-01

RELAP4 is a thermal hydraulic analysis tool written to analyze transients in light water reactors (LWR). To date, most of the applications for RELAP4 have been to analyze postulated LOCA transients in LWR and the response of experimental systems to loss-of-coolant experiments. An important part of these analyses is the prediction of the fuel rod or heater surface temperature which involves the calculation of surface heat transfer coefficients. The paper describes the outcome of a significant blowdown heat transfer development effort which is incorporated in RELAP4/MOD6 (the current version of the code available to the United States public from the Argonne Code Center). The primary emphasis in the MOD6 development was on a PWR reflood capability. The best-estimate blowdown heat transfer correlation and logic were added to provide improved blowdown predictive capability

Integration of Heat Transfer, Stress, and Particle Trajectory Simulation

Energy Technology Data Exchange (ETDEWEB)

Thuc Bui; Michael Read; Lawrence ives

2012-05-17

Calabazas Creek Research, Inc. developed and currently markets Beam Optics Analyzer (BOA) in the United States and abroad. BOA is a 3D, charged particle optics code that solves the electric and magnetic fields with and without the presence of particles. It includes automatic and adaptive meshing to resolve spatial scales ranging from a few millimeters to meters. It is fully integrated with CAD packages, such as SolidWorks, allowing seamless geometry updates. The code includes iterative procedures for optimization, including a fully functional, graphical user interface. Recently, time dependent, particle in cell capability was added, pushing particles synchronically under quasistatic electromagnetic fields to obtain particle bunching under RF conditions. A heat transfer solver was added during this Phase I program. Completed tasks include: (1) Added a 3D finite element heat transfer solver with adaptivity; (2) Determined the accuracy of the linear heat transfer field solver to provide the basis for development of higher order solvers in Phase II; (3) Provided more accurate and smoother power density fields; and (4) Defined the geometry using the same CAD model, while maintaining different meshes, and interfacing the power density field between the particle simulator and heat transfer solvers. These objectives were achieved using modern programming techniques and algorithms. All programming was in C++ and parallelization in OpenMP, utilizing state-of-the-art multi-core technology. Both x86 and x64 versions are supported. The GUI design and implementation used Microsoft Foundation Class.

Validation of CFD/Heat Transfer Software for Turbine Blade Analysis

Science.gov (United States)

Kiefer, Walter D.

2004-01-01

I am an intern in the Turbine Branch of the Turbomachinery and Propulsion Systems Division. The division is primarily concerned with experimental and computational methods of calculating heat transfer effects of turbine blades during operation in jet engines and land-based power systems. These include modeling flow in internal cooling passages and film cooling, as well as calculating heat flux and peak temperatures to ensure safe and efficient operation. The branch is research-oriented, emphasizing the development of tools that may be used by gas turbine designers in industry. The branch has been developing a computational fluid dynamics (CFD) and heat transfer code called GlennHT to achieve the computational end of this analysis. The code was originally written in FORTRAN 77 and run on Silicon Graphics machines. However the code has been rewritten and compiled in FORTRAN 90 to take advantage of more modem computer memory systems. In addition the branch has made a switch in system architectures from SGI's to Linux PC's. The newly modified code therefore needs to be tested and validated. This is the primary goal of my internship. To validate the GlennHT code, it must be run using benchmark fluid mechanics and heat transfer test cases, for which there are either analytical solutions or widely accepted experimental data. From the solutions generated by the code, comparisons can be made to the correct solutions to establish the accuracy of the code. To design and create these test cases, there are many steps and programs that must be used. Before a test case can be run, pre-processing steps must be accomplished. These include generating a grid to describe the geometry, using a software package called GridPro. Also various files required by the GlennHT code must be created including a boundary condition file, a file for multi-processor computing, and a file to describe problem and algorithm parameters. A good deal of this internship will be to become familiar with these

Experiment on transient heat transfer in closed narrow channel

International Nuclear Information System (INIS)

Ochiai, Masaaki

1985-01-01

Heat transfer coefficients and transient pressures in closed narrow channels were obtained experimentally, in order to assess the gap heat transfer models in the computer code WTRLGD which were devised to analyze the internal pressure behavior of waterlogged fuel rods. Gap widths of channels are 0.1--0.5mm to simulate the gap region of waterlogged fuel rods, and test fluids are water (7--89.2 0 C) and Freon-113 (9.2 0 C). The results show that the heater temperature and the pressure measured in the experiments without the DNB occurrence are simulated fairly well by the calculational model of WTRLGD where the heat transfer in a closed narrow channel is evaluated with one-dimensional transient thermal conduction equation and Jens and Lottes' correlation for nucleate boiling. Consequently, it is also suggested that the above equations are available for evaluation of heat flux from fuel to internal water of waterlogged fuel rods. The film boiling heat transfer coefficient was in the same order of that evaluated by Bromley's correlation and the DNB heat flux was smaller than that obtained in quasi-steady experiments with ordinary systems, although the experimental data for them were not enough. (author)

Introduction to heat transfer

CERN Document Server

SUNDÉN, B

2012-01-01

Presenting the basic mechanisms for transfer of heat, Introduction to Heat Transfer gives a deeper and more comprehensive view than existing titles on the subject. Derivation and presentation of analytical and empirical methods are provided for calculation of heat transfer rates and temperature fields as well as pressure drop. The book covers thermal conduction, forced and natural laminar and turbulent convective heat transfer, thermal radiation including participating media, condensation, evaporation and heat exchangers.

Gas flow environmental and heat transfer nonrotating 3D program

Science.gov (United States)

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.

Conjugate Compressible Fluid Flow and Heat Transfer in Ducts

Science.gov (United States)

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.

Advanced Computational Methods for Thermal Radiative Heat Transfer

Energy Technology Data Exchange (ETDEWEB)

Tencer, John; Carlberg, Kevin Thomas; Larsen, Marvin E.; Hogan, Roy E.,

2016-10-01

Participating media radiation (PMR) in weapon safety calculations for abnormal thermal environments are too costly to do routinely. This cost may be s ubstantially reduced by applying reduced order modeling (ROM) techniques. The application of ROM to PMR is a new and unique approach for this class of problems. This approach was investigated by the authors and shown to provide significant reductions in the computational expense associated with typical PMR simulations. Once this technology is migrated into production heat transfer analysis codes this capability will enable the routine use of PMR heat transfer in higher - fidelity simulations of weapon resp onse in fire environments.

Heat transfer fluids containing nanoparticles

Science.gov (United States)

Singh, Dileep; Routbort, Jules; Routbort, A.J.; Yu, Wenhua; Timofeeva, Elena; Smith, David S.; France, David M.

2016-05-17

A nanofluid of a base heat transfer fluid and a plurality of ceramic nanoparticles suspended throughout the base heat transfer fluid applicable to commercial and industrial heat transfer applications. The nanofluid is stable, non-reactive and exhibits enhanced heat transfer properties relative to the base heat transfer fluid, with only minimal increases in pumping power required relative to the base heat transfer fluid. In a particular embodiment, the plurality of ceramic nanoparticles comprise silicon carbide and the base heat transfer fluid comprises water and water and ethylene glycol mixtures.

TRIP: a finite element computer program for the solution of convection heat transfer problems

International Nuclear Information System (INIS)

Slagter, W.; Roodbergen, H.A.

1976-01-01

The theory and use of the finite element code TRIP are described. The code calculates temperature distributions in three-dimensional continua subjected to convection heat transfer. A variational principle for transport phenomena is applied to solve the convection heat transfer problem with temperature and heat flux boundary conditions. The finite element discretization technique is used to reduce the continuous spatial solution into a finite number of unknowns. The method is developed in detail to determine temperature distributions in coolant passages of fuel rod bundles which are idealized by hexahedral elements. The development of the TRIP code is discussed and the listing of the program is given in FORTRAN IV. An example is given to illustrate the validity and practicality of the method

Investigation into the heat transfer performance of helically ribbed surfaces

International Nuclear Information System (INIS)

Firth, R.J.

1981-12-01

The first part of an investigation into flow and heat transfer in annular channels and seven pin clusters is described. One of the main aims of the project is to improve cluster heat transfer prediction codes for helically ribbed surfaces. A study is made of the heat transfer and flow characteristics of a helically ribbed pin in an annular channel. It is shown that the swirling flow, which is induced by the helical ribs, gives rise to substantially enhanced diffusivity levels. This phenomenon had not been taken into account by previous analysis techniques. The methods for analysing heat transfer and pressure drop data from annular channels which were originally developed for non-swirling flow are generalised to accommodate swirling flow. The new methods are shown to be consistent with empirical data. Roughness parameter data is presented for helically ribbed surfaces with an axial rib pitch into height ratio of about 7. (author)

Heat transfer direction dependence of heat transfer coefficients in annuli

Science.gov (United States)

Prinsloo, Francois P. A.; Dirker, Jaco; Meyer, Josua P.

2018-04-01

In this experimental study the heat transfer phenomena in concentric annuli in tube-in-tube heat exchangers at different annular Reynolds numbers, annular diameter ratios, and inlet fluid temperatures using water were considered. Turbulent flow with Reynolds numbers ranging from 15,000 to 45,000, based on the average bulk fluid temperature was tested at annular diameter ratios of 0.327, 0.386, 0.409 and 0.483 with hydraulic diameters of 17.00, 22.98, 20.20 and 26.18 mm respectively. Both heated and cooled annuli were investigated by conducting tests at a range of inlet temperatures between 10 °C to 30 °C for heating cases, and 30 °C to 50 °C for cooling cases. Of special interest was the direct measurement of local wall temperatures on the heat transfer surface, which is often difficult to obtain and evasive in data-sets. Continuous verification and re-evaluation of temperatures measurements were performed via in-situ calibration. It is shown that inlet fluid temperature and the heat transfer direction play significant roles on the magnitude of the heat transfer coefficient. A new adjusted Colburn j-factor definition is presented to describe the heating and cooling cases and is used to correlate the 894 test cases considered in this study.

Proposed heat transfer model for the gas-liquid heat transfer effects observed in the Stanford Research Institute scaled tests

International Nuclear Information System (INIS)

Corradini, M.; Sonin, A.A.; Todreas, N.

1976-12-01

In 1971-72, the Stanford Research Institute conducted a series of scaled experiments which simulated a sodium-vapor expansion in a hypothetical core disruptive accident (HCDA) for the Fast Flux Test Facility. A non-condensible explosive source was used to model the pressure-volume expansion characteristics of sodium vapor as predicted by computer code calculations. Rigid piston-cylinder experiments ( 1 / 10 and 1 / 30 scale) were undertaken to determine these expansion characteristics. The results showed that the pressure-volume characteristics depend significantly on the presence of water in the cylinder reducing the work output by about 50 percent when a sufficient water depth was present. The study presented proposes that the mechanism of heat transfer between the water and high temperature gas was due to area enhancement by Taylor instabilities at the gas-liquid interface. A simple heat transfer model is proposed which describes this energy transport process and agrees well with the experimental data from both scaled experiments. The consequences of this analysis suggest that an estimate of the heat transfer to the cold slug during a full-scale HCDA due to sodium vapor expansion and the accompanying reduction in mechanical work energy warrants further investigation. The implication of this analysis is that for either sodium or fuel vapor expansion in an HCDA, there is an inherent heat transfer mechanism which significantly reduces the work output of the expanding bubble

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)

Updated heat transfer correlations for supercritical water-cooled reactor applications

International Nuclear Information System (INIS)

Mokry, S.J.; Pioro, I.L.; Farah, A.; King, K.

2011-01-01

In support of the development of SuperCritical Water-cooled Reactors (SCWRs), research is currently being conducted for heat-transfer at supercritical conditions. Currently, there are no experimental datasets for heat transfer from power reactor fuel bundles to the fuel coolant (Water) available in open literature. Therefore, for preliminary calculations, heat-transfer correlations obtained with bare tube data can be used as a conservative approach. A large set of experimental data, for supercritical water was analyzed and an updated heat-transfer correlation for forced-convective heat-transfer, in the normal heat transfer regime, was developed. This experimental dataset was obtained within conditions similar to those for proposed SCWR concepts. Thus, this new correlation can be used for preliminary heat-transfer calculations in SCWR fuel channels. It has demonstrated a good fit for the analyzed dataset. Experiments with SuperCritical Water (SCW) are very expensive. Therefore, a number of experiments are performed in modeling fluids, such as carbon dioxide and refrigerants. However, there is no common opinion if SC modeling fluids' correlations can be applied to SCW and vice versa. Therefore, a correlation for supercritical carbon dioxide heat transfer was developed as a less expensive alternative to using supercritical water. The conducted analysis also meets the objective of improving our fundamental knowledge of the transport processes and handling of supercritical fluids. These correlations can be used for supercritical water heat exchangers linked to indirect-cycle concepts and the cogeneration of hydrogen, for future comparisons with other independent datasets, with bundle data, for the verification of computer codes for SCWR core thermalhydraulics and for the verification of scaling parameters between water and modeling fluids. (author)

Adiabatic partition effect on natural convection heat transfer inside a square cavity

DEFF Research Database (Denmark)

Mahmoudi Nezhad, Sajjad; Rezaniakolaei, Alireza; yousefi, Tooraj

2018-01-01

A steady state and two-dimensional laminar free convection heat transfer in a partitioned cavity with horizontal adiabatic and isothermal side walls is investigated using both experimental and numerical approaches. The experiments and numerical simulations are carried out using a Mach......-Zehnder interferometer and a finite volume code, respectively. A horizontal and adiabatic partition, with angle of θ is adjusted such that it separates the cavity into two identical parts. Effects of this angel as well as Rayleigh number on the heat transfer from the side-heated walls are investigated in this study...... partition angle, the results show that the average Nusselt number and consequently the heat transfer enhance as the Rayleigh number increases. However, for a given Rayleigh number the maximum and the minimum heat transfer occurs at θ = 45°and θ = 90°, respectively. Two responsible mechanisms...

aerodynamics and heat transfer

Directory of Open Access Journals (Sweden)

J. N. Rajadas

1998-01-01

Full Text Available A multidisciplinary optimization procedure, with the integration of aerodynamic and heat transfer criteria, has been developed for the design of gas turbine blades. Two different optimization formulations have been used. In the first formulation, the maximum temperature in the blade section is chosen as the objective function to be minimized. An upper bound constraint is imposed on the blade average temperature and a lower bound constraint is imposed on the blade tangential force coefficient. In the second formulation, the blade average and maximum temperatures are chosen as objective functions. In both formulations, bounds are imposed on the velocity gradients at several points along the surface of the airfoil to eliminate leading edge velocity spikes which deteriorate aerodynamic performance. Shape optimization is performed using the blade external and coolant path geometric parameters as design variables. Aerodynamic analysis is performed using a panel code. Heat transfer analysis is performed using the finite element method. A gradient based procedure in conjunction with an approximate analysis technique is used for optimization. The results obtained using both optimization techniques are compared with a reference geometry. Both techniques yield significant improvements with the multiobjective formulation resulting in slightly superior design.

Heat transfer characteristics of horizontal steam generators under natural circulation conditions

International Nuclear Information System (INIS)

Hyvaerinen, J.

1996-01-01

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

Radial heat transfer from fuel to moderator during LOCAs for CANDU PHW reactors

International Nuclear Information System (INIS)

Hildebrandt, J.G.; So, C.B.; Gillespie, G.E.; MacLean, G.

1983-01-01

In a postulated CANDU-PHW loss-of-coolant accident (LOCA) with coincident impaired emergency cooling, the axial transport of heat from the fuel by convection is reduced. This reduction in heat removal causes the fuel to heat up and the radial heat transfer to the moderator to become significant. This paper deals with two codes that predict the thermal response of fuel channels under LOCA conditions. New channel thermal radiation models in both RAMA, a thermalhydraulic code, and CHAN II, a fuel channel thermo-chemical code, are presented and their predictions are compared with the experimental results of an electrically heated bundle of 37 fuel pins. A second experiment, involving a single heated pin in a channel with flowing steam, is presented. The predictions of RAMA and CHAN II are compared with this experiment to verify the codes' thermo-chemical models. There is good agreement between the predictions of both codes and the experimental results

Radiative heat transfer

CERN Document Server

Modest, Michael F

2013-01-01

The third edition of Radiative Heat Transfer describes the basic physics of radiation heat transfer. The book provides models, methodologies, and calculations essential in solving research problems in a variety of industries, including solar and nuclear energy, nanotechnology, biomedical, and environmental. Every chapter of Radiative Heat Transfer offers uncluttered nomenclature, numerous worked examples, and a large number of problems-many based on real world situations-making it ideal for classroom use as well as for self-study. The book's 24 chapters cover the four major areas in the field: surface properties; surface transport; properties of participating media; and transfer through participating media. Within each chapter, all analytical methods are developed in substantial detail, and a number of examples show how the developed relations may be applied to practical problems. It is an extensive solution manual for adopting instructors. Features: most complete text in the field of radiative heat transfer;...

Simultaneous Heat and Mass Transfer Model for Convective Drying of Building Material

Science.gov (United States)

Upadhyay, Ashwani; Chandramohan, V. P.

2018-04-01

A mathematical model of simultaneous heat and moisture transfer is developed for convective drying of building material. A rectangular brick is considered for sample object. Finite-difference method with semi-implicit scheme is used for solving the transient governing heat and mass transfer equation. Convective boundary condition is used, as the product is exposed in hot air. The heat and mass transfer equations are coupled through diffusion coefficient which is assumed as the function of temperature of the product. Set of algebraic equations are generated through space and time discretization. The discretized algebraic equations are solved by Gauss-Siedel method via iteration. Grid and time independent studies are performed for finding the optimum number of nodal points and time steps respectively. A MATLAB computer code is developed to solve the heat and mass transfer equations simultaneously. Transient heat and mass transfer simulations are performed to find the temperature and moisture distribution inside the brick.

User's manual for the Heat Pipe Space Radiator design and analysis Code (HEPSPARC)

Science.gov (United States)

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.

Match properties of heat transfer and coupled heat and mass transfer processes in air-conditioning system

International Nuclear Information System (INIS)

Zhang Tao; Liu Xiaohua; Zhang Lun; Jiang Yi

2012-01-01

Highlights: ► Investigates match properties of heat or mass transfer processes in HVAC system. ► Losses are caused by limited transfer ability, flow and parameter mismatching. ► Condition of flow matching is the same heat capacity of the fluids. ► Parameter matching is only reached along the saturation line in air–water system. ► Analytical solutions of heat and mass transfer resistance are derived. - Abstract: Sensible heat exchangers and coupled heat and mass transfer devices between humid air and water/desiccant are commonly used devices in air-conditioning systems. This paper focuses on the match properties of sensible heat transfer processes and coupled heat and mass transfer processes in an effort to understand the reasons for performance limitations in order to optimize system performance. Limited heat transfer capability and flow mismatching resulted in heat resistance of the sensible heat transfer process. Losses occurred during the heat and mass transfer processes due to limited transfer capability, flow mismatching, and parameter mismatching. Flow matching was achieved when the heat capacities of the fluids were identical, and parameter matching could only be reached along the saturation line in air–water systems or the iso-concentration line in air–desiccant systems. Analytical solutions of heat transfer resistance and mass transfer resistance were then derived. The heat and mass transfer process close to the saturation line is recommended, and heating sprayed water resulted in better humidification performance than heating inlet air in the air humidifier.

Development and adaptation of conduction and radiation heat-transfer computer codes for the CFTL

International Nuclear Information System (INIS)

Conklin, J.C.

1981-08-01

RODCON and HOTTEL are two computational methods used to calculate thermal and radiation heat transfer for the Core Flow Test Loop (CFTL) analysis efforts. RODCON was developed at ORNL to calculate the internal temperature distribution of the fuel rod simulator (FRS) for the CFTL. RODCON solves the time-dependent heat transfer equation in two-dimensional (R angle) cylindrical coordinates at an axial plane with user-specified radial material zones and time- and position-variant surface conditions at the FRS periphery. Symmetry of the FRS periphery boundary conditions is not necessary. The governing elliptic, partial differential heat equation is cast into a fully implicit, finite-difference form by approximating the derivatives with a forward-differencing scheme with variable mesh spacing. The heat conduction path is circumferentially complete, and the potential mathematical problem at the rod center can be effectively ignored. HOTTEL is a revision of an algorithm developed by C.B. Baxi at the General Atomic Company (GAC) to be used in calculating radiation heat transfer in a rod bundle enclosed in a hexagonal duct. HOTTEL uses geometric view factors, surface emissivities, and surface areas to calculate the gray-body or composite view factors in an enclosure having multiple reflections in a nonparticipating medium

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.

Analytical modeling of heat transfer during the reflooding phase of the LOCA: the UCFLOOD code

International Nuclear Information System (INIS)

Yadigaroglu, G.; Arrieta, L.A.

1980-01-01

A mechanistic model of bottom-reflooding heat transfer is described. 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. The onset of liquid entrainment is determined using a criterion based on the instability of the liquid core in the inverted-annular flow regime. For heat transfer calculations, the channel is also divided into a number of regions. The heat transfer coefficients are functions of the local flow conditions. Quench front propagation is treated separately by a model including the effects of axial conduction. Good agreement of calculated and experimental results has been obtained

Separate effects tests for GOTHIC condensation and evaporative heat transfer models

International Nuclear Information System (INIS)

George, T.L.; Singh, A.

1994-01-01

The GOTHIC computer program, under development at EPRI/NAI, is a general purpose thermal hydraulics computer program for design, licensing, safety and operating analysis of nuclear containments and other confinement buildings. The code solves a nine equation model for three dimensional multiphase flow with separate mass, momentum and energy equations for vapor, liquid and drop phases. The vapor phase can be a gas mixture of steam and non-condensing gases. The phase balance equations are coupled by mechanistic and empirical models for interface mass, energy and momentum transfer that cover the entire flow regime from bubbly flow to film/drop flow. A variety of heat transfer correlations are available to model the fluid coupling to active and passive solid conductors. This paper focuses on the application of GOTHIC to two separate effects tests; condensation heat transfer on a vertical flat plate with varying bulk velocity, steam concentration and temperature, and evaporative heat transfer from a hot pool to a dry (superheated) atmosphere. Comparisons with experimental data is included for both tests. Results show the validity of two condensation heat transfer correlations as incorporated into GOTHIC and the interfacial heat and mass transfer models for the range of the experimental test conditions. Comparisons are also made for lumped versus multidimensional modeling for buoyancy controlled flow with evaporative heat transfer. (author). 13 refs., 1 tab., 10 figs

Separate effects tests for GOTHIC condensation and evaporative heat transfer models

International Nuclear Information System (INIS)

George, T.L.; Singh, A.

1996-01-01

The GOTHIC computer program, under development at NAI for EPRI, is a general purpose thermal hydraulics computer program for design, licensing, safety and operating analysis of nuclear containments and other confinement buildings. The code solves a nine-equation model for three-dimensional multiphase flow with separate mass, momentum and energy equations for vapor, liquid and drop phases. The vapor phase can be a gas mixture of steam and non-condensing gases. The phase balance equations are coupled by mechanistic and empirical models for interface mass, energy and momentum transfer that cover the entire flow regime from bubbly flow to film-drop flow. A variety of heat transfer correlations are available to model the fluid coupling to active and passive solid conductors. This paper focuses on the application of GOTHIC to two separate effects tests: condensation heat transfer on a vertical flat plate with varying bulk velocity, steam concentration and temperature, and evaporative heat transfer from a hot pool to a dry (superheated) atmosphere. Comparisons with experimental data are included for both tests. Results show the validity of two condensation heat transfer correlations as incorporated into GOTHIC and the interfacial heat and mass transfer models for the range of the experimental test conditions. Comparisons are also made for lumped vs. multidimensional modeling for buoyancy-controlled flow with evaporative heat transfer. (orig.)

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

The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation's energy, environmental, and energy security needs. Advanced high temperature reactor systems such as sodium fast reactors and high and very high temperature gas-cooled reactors are being considered for the next generation of nuclear reactor plant designs. The coolants for these high temperature reactor systems include liquid sodium and helium gas. Supercritical carbon dioxide (sCO₂), a fluid at a temperature and pressure above the supercritical point of CO₂, is currently being investigated by DOE as a working fluid for a nuclear or fossil-heated recompression closed Brayton cycle energy conversion system that operates at 550°C (1022°F) at 200 bar (2900 psi). Higher operating temperatures are envisioned in future developments. All of these design concepts require a highly effective heat exchanger that transfers heat from the nuclear or chemical reactor to the chemical process fluid or the to the power cycle. In the nuclear designs described above, heat is transferred from the primary to the secondary loop via an intermediate heat exchanger (IHX) and then from the intermediate loop to either a working process or a power cycle via a secondary heat exchanger (SHX). The IHX is a component in the primary coolant loop which will be classified as "safety related." The intermediate loop will likely be classified as "not safety related but important to safety." These safety classifications have a direct bearing on heat exchanger design approaches for the IHX and SHX. The very high temperatures being considered for the VHTR will require the use of very high temperature alloys for the IHX and SHX. Material cost considerations alone will dictate that the IHX and SHX be highly effective; that is, provide high heat transfer area in a small volume. This feature must be accompanied by low pressure drop and mechanical reliability and

Heat transfer modeling an inductive approach

CERN Document Server

Sidebotham, George

2015-01-01

This innovative text emphasizes a "less-is-more" approach to modeling complicated systems such as heat transfer by treating them first as "1-node lumped models" that yield simple closed-form solutions. The author develops numerical techniques for students to obtain more detail, but also trains them to use the techniques only when simpler approaches fail. Covering all essential methods offered in traditional texts, but with a different order, Professor Sidebotham stresses inductive thinking and problem solving as well as a constructive understanding of modern, computer-based practice. Readers learn to develop their own code in the context of the material, rather than just how to use packaged software, offering a deeper, intrinsic grasp behind models of heat transfer. Developed from over twenty-five years of lecture notes to teach students of mechanical and chemical engineering at The Cooper Union for the Advancement of Science and Art, the book is ideal for students and practitioners across engineering discipl...

Impacts of transient heat transfer modeling on prediction of advanced cladding fracture during LWR LBLOCA

Energy Technology Data Exchange (ETDEWEB)

Lee, Youho, E-mail: euo@kaist.ac.kr; Lee, Jeong Ik, E-mail: jeongiklee@kaist.ac.kr; NO, Hee Cheon, E-mail: hcno@kaist.ac.kr

2016-03-15

Highlights: • Use of constant heat transfer coefficient for fracture analysis is not sound. • On-time heat transfer coefficient should be used for thermal fracture prediction. • ∼90% of the actual fracture stresses were predicted with the on-time transient h. • Thermal-hydraulic codes can be used to better predict brittle cladding fracture. • Effects of surface oxides on thermal shock fracture should be accounted by h. - Abstract: This study presents the importance of coherency in modeling thermal-hydraulics and mechanical behavior of a solid for an advanced prediction of cladding thermal shock fracture. In water quenching, a solid experiences dynamic heat transfer rate evolutions with phase changes of the fluid over a short quenching period. Yet, such a dynamic change of heat transfer rates has been overlooked in the analysis of thermal shock fracture. In this study, we are presenting quantitative evidence against the prevailing use of a constant heat transfer coefficient for thermal shock fracture analysis in water. We conclude that no single constant heat transfer could suffice to depict the actual stress evolution subject to dynamic fluid phase changes. Use of the surface temperature dependent heat transfer coefficient will remarkably increase predictability of thermal shock fracture of brittle materials. The presented results show a remarkable stress prediction improvement up to 80–90% of the actual stress with the use of the surface temperature dependent heat transfer coefficient. For thermal shock fracture analysis of brittle fuel cladding such as oxidized zirconium-based alloy or silicon carbide during LWR reflood, transient subchannel heat transfer coefficients obtained from a thermal-hydraulics code should be used as input for stress analysis. Such efforts will lead to a fundamental improvement in thermal shock fracture predictability over the current experimental empiricism for cladding fracture analysis during reflood.

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

Investigation of Heat Transfer in Supercritical Fluids for Application to the Generation IV

Energy Technology Data Exchange (ETDEWEB)

Bae, Y. Y.; Kim, H. D.; Song, J. H.; Kim, H. Y.; Cho, B. H.; Kim, H.; Kang, D. J.

2007-08-15

Using a facility named SPHINX, which can accommodate a heat transfer test with CO{sub 2} at supercritical pressure, a series of tests was performed. The test geometries include tubes with the inner diameter of 4.4, 6.32 and 9 mm. a concentric annular passages with 8 x 10 mm, and an eccentric annular passages with 9.5 x 12.5 mm. Based on the test results, heat transfer correlations were developed and compared with the existing correlations. The heat transfer deterioration which may occur at certain conditions of heat and mass flux, were carefully studied and the published criteria were reviewed against our test results. Numerical calculation by using commercial CFD code, Fluent, were performed in order to provide the pre-test information for the heat transfer tests. Various turbulence models were evaluated and reliable models were suggested for each case

About calculation results of heat transfer in the fuel assembly clusters cooled by water with supercritical parameters

International Nuclear Information System (INIS)

Grabezhnaya, V.A.

2008-01-01

Paper reviews the numerical investigation into the heat transfer in the supercritical water cooled fuel assemblies on the basis of the various commercial codes. The turbulence available models specified in the codes describe adequately the experimental data in tubes within the range of flow temperatures away from the pseudocritical point, as well as under high mass velocities. There are k-ε type turbulence models that show qualitatively the local acceleration (slowdown) of the heat transfer in tubes, but they fail to describe the mentioned phenomena quantitatively. To determine the effect of grid spacers on the suppression of the heat transfer local slowdown and on the heat transfer acceleration in fuel assemblies and to ensure more accurate calculation of the fuel element cladding maximum temperature one should perform a number of the experiments making use of the fuel assembly models [ru

Review and proposal for heat transfer predictions at supercritical water conditions using existing correlations and experiments

Energy Technology Data Exchange (ETDEWEB)

Jaeger, Wadim, E-mail: wadim.jaeger@kit.edu [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, DE-76344 Eggenstein-Leopoldshafen (Germany); Sanchez Espinoza, Victor Hugo [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, DE-76344 Eggenstein-Leopoldshafen (Germany); Hurtado, Antonio [Technical University of Dresden, Institute of Power Engineering, DE-01062 Dresden (Germany)

2011-06-15

Highlights: > Implementation of heat transfer correlations for supercritical water into TRACE. > Simulation of several heat transfer experiments with modified TRACE version. > Most correlations are not able to reproduce the experimental results. > Bishop, Sandberg and Tong correlation is most suitable for TRACE applications. - Abstract: This paper summarizes the activities of the TRACE code validation at the Institute for Neutron Physics and Reactor Technology related to supercritical water conditions. In particular, the providing of the thermo physical properties and its appropriate use in the wall-to-fluid heat transfer models in the frame of the TRACE code is the object of this investigation. In a first step, the thermo physical properties of the original TRACE code were modified in order to account for supercritical conditions. In a second step, existing Nusselt correlations were reviewed and implemented into TRACE and available experiments were simulated to identify the most suitable Nusselt correlation(s).

Heat transfer with geometric shape of micro-fin tubes (I) - Condensing heat transfer

Energy Technology Data Exchange (ETDEWEB)

Kwak, K M; Chang, J S; Bai, C H; Chung, M [Yeungnam University, Kyungsan (Korea)

1999-11-01

To examine the enhancement mechanism of condensing heat transfer through microfin tube, the condensation experiments with refrigerant HCFC 22 are performed using 4 and 6 kinds of microfin tubes with outer diameter of 9.52 mm and 7.0 mm, respectively. Used microfin tubes have different shape and number of fins with each other. The main heat transfer enhancement mechanism is known to be the enlargement of heat transfer area and turbulence promotion. Together with these main factors, we can find other enhancement factors by the experimental data, which are the overflow of the refrigerant over the microfin and microfin arrangement. The overflow of the refrigerant over the microfin can be analyzed by the geometric shape of the microfin. microfin tubes having a shape which can give much overflow over the microfin show large condensing heat transfer coefficients. The effect of microfin arrangement is related to the heat transfer resistance of liquid film of refrigerant. The condensing heat transfer coefficients are high for the microfin tube with even distribution of liquid film. 17 refs., 14 figs., 3 tabs.

Impact of coupled heat and moisture transfer effects on buildings energy consuption

Directory of Open Access Journals (Sweden)

Ferroukhi Mohammed Yacine

2017-01-01

Full Text Available Coupled heat, air, and moisture transfers through building envelope have an important effect on prediction of building energy requirements. Several works were conducted in order to integrate hygrothermal transfers in dynamic buildings simulations codes. However, the incorporation of multidirectional hygrothermal transfer analysis in the envelope into building simulation tools is rarely considered. In this work, coupled heat, air, and moisture (HAM transfer model in multilayer walls was established. Thereafter, the HAM model is coupled dynamically to a building behavior code (BES.The coupling concerns a co-simulation between COMSOL Multiphysics and TRNSYS software. Afterward, the HAM-BES co-simulation accuracy was verified. Then, HAM-BES co-simulation platform was applied to a case study with various types of climates (temperate, hot and humid, cold and humid. Three simulations cases were carried out. The first simulation case consists of the TRNSYS model without HAM transfer model. The second simulation case, 1-D HAM model for the envelope was integrated in TRNSYS code. For the third one, 1-D HAM model for the wall and 2-D HAM model for thermal bridges were coupled to the thermal building model of TRNSYS. Analysis of the results confirms the significant impact of 2-D envelope hygrothermal transfers on the indoor thermal and moisture behavior of building as well as on the energy building assessment. These conclusions are shown for different studied climates.

Measuring of heat transfer coefficient

DEFF Research Database (Denmark)

Henningsen, Poul; Lindegren, Maria

Subtask 3.4 Measuring of heat transfer coefficient Subtask 3.4.1 Design and setting up of tests to measure heat transfer coefficient Objective: Complementary testing methods together with the relevant experimental equipment are to be designed by the two partners involved in order to measure...... the heat transfer coefficient for a wide range of interface conditions in hot and warm forging processes. Subtask 3.4.2 Measurement of heat transfer coefficient The objective of subtask 3.4.2 is to determine heat transfer values for different interface conditions reflecting those typically operating in hot...

HEXEREI: a multi-channel heat conduction convection code for use in transient thermal hydraulic analysis of high-temperature, gas-cooled reactors. Interim report

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

Capillary-Condenser-Pumped Heat-Transfer Loop

Science.gov (United States)

Silverstein, Calvin C.

1989-01-01

Heat being transferred supplies operating power. Capillary-condenser-pumped heat-transfer loop similar to heat pipe and to capillary-evaporator-pumped heat-transfer loop in that heat-transfer fluid pumped by evaporation and condensation of fluid at heat source and sink, respectively. Capillary condenser pump combined with capillary evaporator pump to form heat exchanger circulating heat-transfer fluids in both loops. Transport of heat more nearly isothermal. Thermal stress in loop reduced, and less external surface area needed in condenser section for rejection of heat to heat sink.

The droplet injection system used in the rod bundle heat transfer facility

International Nuclear Information System (INIS)

Frepoli, C.; Andrew, A.J.; Hochreiter, L.E.; Cheung, F.B.

2001-01-01

The full text follows. The US Nuclear Regulatory Commission (NRC) and the Pennsylvania State University are currently funding a research program entitled ''Rod Bundle Heat Transfer'' (RBHT). The main objective of the program is to investigate heat transfer during the core reflood period of a hypothetical Large Break Loss of Coolant Accident in a typical nuclear power plant. The RBHT test facility consists of a full-length 7 x 7 rod bundle. Information gathered by the RBHT test facility will be used for improvement of the reflood heat transfer models in the NRC's thermal hydraulic codes. In particular the RBHT data will be used to improve the understanding of individual heat transfer effects to the total rod heat transfer such that compensating errors present in current Best Estimate codes can be significantly reduced. The strategy in developing the test matrix is to use a ''building block'' approach in which simpler experiments are performed first to quantify a particular heat transfer mechanism alone and then the additional complications of the full two-phase flow, reflood film boiling behavior of the test facility are added in later experiments. One of these ''simpler'' experiments will be the injection of known size and velocity liquid droplets into the main stream of superheated steam. The droplet injection system consists of small diameter tubes inserted across the bundle at a given elevation. A number of equal size holes are drilled perpendicular to the surface in a triangular pitch. Water is forced into opposite ends of the tube and ejected from the holes. The injection system was tested using a digital imaging system known as VisiSizer. This system is capable of determining the diameter and velocity of small water droplets using a laser-illuminated digital camera system (LIDCS). Imaging software analyzes the digital images in real time to determine the distributions of droplet size and velocity. Pre-test analysis using COBRA-TF have been conducted to

Thermal radiation heat transfer in participating media by finite volume discretization using collimated beam incidence

Science.gov (United States)

Harijishnu, R.; Jayakumar, J. S.

2017-09-01

The main objective of this paper is to study the heat transfer rate of thermal radiation in participating media. For that, a generated collimated beam has been passed through a two dimensional slab model of flint glass with a refractive index 2. Both Polar and azimuthal angle have been varied to generate such a beam. The Temperature of the slab and Snells law has been validated by Radiation Transfer Equation (RTE) in OpenFOAM (Open Field Operation and Manipulation), a CFD software which is the major computational tool used in Industry and research applications where the source code is modified in which radiation heat transfer equation is added to the case and different radiation heat transfer models are utilized. This work concentrates on the numerical strategies involving both transparent and participating media. Since Radiation Transfer Equation (RTE) is difficult to solve, the purpose of this paper is to use existing solver buoyantSimlpeFoam to solve radiation model in the participating media by compiling the source code to obtain the heat transfer rate inside the slab by varying the Intensity of radiation. The Finite Volume Method (FVM) is applied to solve the Radiation Transfer Equation (RTE) governing the above said physical phenomena.

Integration of Heat Transfer, Stress, and Particle Trajectory Simulation. Final report

International Nuclear Information System (INIS)

Bui, Thuc; Read, Michael; Ives, Lawrence

2012-01-01

Calabazas Creek Research, Inc. developed and currently markets Beam Optics Analyzer (BOA) in the United States and abroad. BOA is a 3D, charged particle optics code that solves the electric and magnetic fields with and without the presence of particles. It includes automatic and adaptive meshing to resolve spatial scales ranging from a few millimeters to meters. It is fully integrated with CAD packages, such as SolidWorks, allowing seamless geometry updates. The code includes iterative procedures for optimization, including a fully functional, graphical user interface. Recently, time dependent, particle in cell capability was added, pushing particles synchronically under quasistatic electromagnetic fields to obtain particle bunching under RF conditions. A heat transfer solver was added during this Phase I program. Completed tasks include: (1) Added a 3D finite element heat transfer solver with adaptivity; (2) Determined the accuracy of the linear heat transfer field solver to provide the basis for development of higher order solvers in Phase II; (3) Provided more accurate and smoother power density fields; and (4) Defined the geometry using the same CAD model, while maintaining different meshes, and interfacing the power density field between the particle simulator and heat transfer solvers. These objectives were achieved using modern programming techniques and algorithms. All programming was in C++ and parallelization in OpenMP, utilizing state-of-the-art multi-core technology. Both x86 and x64 versions are supported. The GUI design and implementation used Microsoft Foundation Class.

Radiation effects on heat transfer in heat exchangers, (2)

International Nuclear Information System (INIS)

Mori, Yasuo; Watanabe, Kenji; Taira, Tatsuji.

1980-01-01

In a high temperature gas-cooled reactor system, in which the working fluid exchanges heat at high temperature near 1000 deg C, the heat transfer acceleration by positively utilizing the radiation heat transfer between solid surfaces should be considered. This paper reports on the results of experiment and analysis for the effects of radiant heat on the heat transfer performance at elevated temperature by applying the heat transfer-accelerating method using radiators to the heat exchanger with tube bundle composed of two channels of heating and heated sides. As the test heat exchangers, a parallel counter flow exchanger and the cross flow exchanger simulating helical tubes were employed, and the results studied on the characteristics of each heat exchanger are described. The plates placed in parallel to flow in every space of the tube bundle arranged in a matrix were used as the heat transfer accelerator. The effects of acceleration with the plates were the increase of heat transmission from 12 to 24% and 12 to 38% in the parallel flow and cross flow heat exchangers, respectively. Also, it was clarified that the theoretical analysis, in which it was assumed that the region within pitch S and two radiator plates, with a heat-transferring tube placed at the center, is the minimum domain for calculation, and that the heat exchange by radiation occurs only between the domain and the adjacent domains, can estimate the heat transfer-accelerating effect and the temperature distribution in a heat exchanger with sufficient accuracy. (Wakatsuki, Y.)

A three-dimensional analyses of fluid flow and heat transfer for moderator integrity assessment in PHWR

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

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

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

Heat Transfer and Fluid Dynamics Measurements in the Expansion Space of a Stirling Cycle Engine

Science.gov (United States)

Jiang, Nan; Simon, Terrence W.

2006-01-01

The heater (or acceptor) of a Stirling engine, where most of the thermal energy is accepted into the engine by heat transfer, is the hottest part of the engine. Almost as hot is the adjacent expansion space of the engine. In the expansion space, the flow is oscillatory, impinging on a two-dimensional concavely-curved surface. Knowing the heat transfer on the inside surface of the engine head is critical to the engine design for efficiency and reliability. However, the flow in this region is not well understood and support is required to develop the CFD codes needed to design modern Stirling engines of high efficiency and power output. The present project is to experimentally investigate the flow and heat transfer in the heater head region. Flow fields and heat transfer coefficients are measured to characterize the oscillatory flow as well as to supply experimental validation for the CFD Stirling engine design codes. Presented also is a discussion of how these results might be used for heater head and acceptor region design calculations.

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

Multiregional coupled conduction--convection model for heat transfer in an HTGR core

International Nuclear Information System (INIS)

Giles, G.E. Jr.; Childs, K.W.; Sanders, J.P.

1978-01-01

HEXEREI is a three-dimensional, coupled conduction-convection heat transfer and multichannel fluid dynamic analysis computer code with both steady-state and transient capabilities. The program was developed to provide thermal-fluid dynamic analysis of a core following the general design for high-temperature gas-cooled reactors (HTGRs); its purpose was to provide licensing evaluations for the U.S. Nuclear Regulatory Commission. In order to efficiently model the HTGR core, the nodal geometry of HEXEREI was chosen as a regular hexagonal array perpendicular to the axis of and bounded by a right circular cylinder. The cylindrical nodal geometry surrounds the hexagonal center portion of the mesh; these two different types of nodal geometries must be connected by interface nodes to complete the accurate modeling of the HTGR core. HEXEREI will automatically generate a nodal geometry that will accurately model a complex assembly of hexagonal and irregular prisms. The accuracy of the model was proven by a comparison of computed values with analytical results for steady-state and transient heat transfer problems. HEXEREI incorporates convective heat transfer to the coolant in many parallel axial flow channels. Forced and natural convection (which permits different flow directions in parallel channels) is included in the heat transfer and fluid dynamic models. HEXEREI incorporates a variety of steady-state and transient solution techniques that can be matched with a particular problem to minimize the computational time. HEXEREI was compared with a code of similar capabilities that was based on a Cartesian mesh. This code modeled only one specific core design, and the mesh spacing was closer than that generated by HEXEREI. Good agreement was obtained with the detail provided by the representations

Heat transfer from internally heated hemispherical pools

International Nuclear Information System (INIS)

Gabor, J.D.; Ellsion, P.G.; Cassulo, J.C.

1980-01-01

Experiments were conducted on heat transfer from internally heated ZnSO 4 -H 2 O pools to the walls of hemispherical containers. This experimental technique provides data for a heat transfer system that has to date been only theoretically treated. Three different sizes of copper hemispherical containers were used: 240, 280, 320 mm in diameter. The pool container served both as a heat transfer surface and as an electrode. The opposing electrode was a copper disk, 50 mm in diameter located at the top of the pool in the center. The top surface of the pool was open to the atmosphere

Review and proposal for heat transfer predictions at supercritical water conditions using existing correlations and experiments

International Nuclear Information System (INIS)

Jaeger, Wadim; Sanchez Espinoza, Victor Hugo; Hurtado, Antonio

2011-01-01

Highlights: → Implementation of heat transfer correlations for supercritical water into TRACE. → Simulation of several heat transfer experiments with modified TRACE version. → Most correlations are not able to reproduce the experimental results. → Bishop, Sandberg and Tong correlation is most suitable for TRACE applications. - Abstract: This paper summarizes the activities of the TRACE code validation at the Institute for Neutron Physics and Reactor Technology related to supercritical water conditions. In particular, the providing of the thermo physical properties and its appropriate use in the wall-to-fluid heat transfer models in the frame of the TRACE code is the object of this investigation. In a first step, the thermo physical properties of the original TRACE code were modified in order to account for supercritical conditions. In a second step, existing Nusselt correlations were reviewed and implemented into TRACE and available experiments were simulated to identify the most suitable Nusselt correlation(s).

Use of CATHENA to model calandria-tube/moderator heat transfer after pressure-tube/calandria-tube ballooning contact

International Nuclear Information System (INIS)

Fan, H.Z.; Bilanovic, Z.; Nitheanandan, T.

2004-01-01

A study was performed to assess the effect of the calandria-tube/moderator heat transfer after pressure-tube/calandria tube ballooning contact using CATHENA. Results of this study indicated that the analytical tool, CATHENA, can be applied for pool boiling heat transfer on the external surface of a large diameter tube, such as the calandria tube used in CANDU reactors. The methodology in such CANDU-generic study can be used to simulate the tube surface with multiple boiling regimes and to assess the benefits of closely coupling thermalhydraulics modelling and fuel/fuel channel behaviour modelling. CATHENA (Canadian Algorithm for THErmalhydraulic Network Analysis) is a one-dimensional, two-fluid thermalhydraulic simulation code designed by AECL to analyse two-phase flow and heat transfer in piping networks. The detailed heat transfer package in CATHENA allows a connection to be established from the multiple solid surfaces of tubes to the surrounding large amount of moderator water, which acts as a heat sink during a postulated loss of coolant event. The generalized heat transfer package within CATHENA allows the tube walls to be divided into several layers in the radial direction and several sectors in the circumferential direction, to account for heat transfer conditions in these two directions. The CATHENA code with the generalized heat transfer package is capable of capturing key pool-boiling phenomena such as nucleate, transition and film boiling heat transfer as well as an ability to model the rewet phenomenon to some extent. A CATHENA input model was generated and used in simulations of selected contact boiling experiment test cases. The transient wall temperatures have been calculated in different portions of the calandria tube. By using this model an adequate agreement was achieved between CATHENA calculation and experimental measurement The CATHENA code enables one to investigate the transient and local thermal-mechanical behaviour of the calandria tube

Flow regimes and heat transfer modes identification in ANGRA 2 core, during small break in the primary loop with area of 100 cm2, simulated with RELAP5 code

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)

Heat transfer from two-side heated helical channels

International Nuclear Information System (INIS)

Shimonis, V.; Ragaishis, V.; Poshkas, P.

1995-01-01

Experimental results are presented on the heat transfer from two-side heated helical channels to gas (air) flows. The study covered six configurations and wide ranges of geometrical (D/h=5.5 to 84.2) and performance (Re=10 3 to 2*10 5 ) parameters. Under the influence of Re and of the channel curvature, the heat transfer from both the convex and the concave surfaces for two-side heating (q w1 ≅ q w2 ) is augmented by 20-30% over one-side heating. Improved relations to predict the critical values of Reynolds Re cr1 and Re cr2 are suggested. They enable more exact predictions of the heat transfer from convex surface in transient flows for one-side heating. The relation for annular channels is suggested for the turbulent heat transfer from the convex and concave surfaces of two-side heated helical channels. It can be adapted by introducing earlier expresions for one-side heated helical channels. (author). 6 refs., 2 tabs., 3 figs

An assessment of the critical heat flux approaches of thermal-hydraulic system analysis codes using bundle data from the Heat Transfer Research Facility

International Nuclear Information System (INIS)

Min Lee

1994-01-01

Critical heat flux (CHF) bundle data from the Heat Transfer Research Facility of Columbia University are used to check the validity of the CHF approaches used in thermal-hydraulic system analysis codes for light water reactors. The CHF approaches assessed include the Biasi et al. correlation of TRAC, the Groeneveld et al. CHF table lookup approach of RELAP5/MOD3, the CHF table lookup approach of CATHARE, and the CHF approach of RETRAN. Depending on system pressure, RETRAN uses the B and W2, Barnett, and modified Barnett correlations and a linear interpolation scheme to predict CHF. Results show that among these CHF approaches, the Groeneveld et al. approach has the best prediction accuracy and the smallest uncertainty in the estimation of the HTRF bundle data. On the average, the Groeneveld et al. approach overpredicts the uniform axial heat flux distribution by 3.6% and the nonuniform axial heat flux distribution by 0.9%. The performance of the RETRAN approach is comparable with that of the Groenevel et al. Approach for uniform axial heat flux. In general, the accuracy and the uncertainty of all the approaches, except that of CATHARE, are worse under a nonuniform axial heat distribution than under a uniform axial heat distribution. All the CHF approaches assessed have a tendency to overpredict the HTRF bundle data at low pressure, low measured CHF, and high CHF quality. The performance of the Groenevel et al. approach is improved through a CHF table update and modification of the bundle correction factor using the HTRF bundle data

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)

Convection heat transfer

CERN Document Server

Bejan, Adrian

2013-01-01

Written by an internationally recognized authority on heat transfer and thermodynamics, this second edition of Convection Heat Transfer contains new and updated problems and examples reflecting real-world research and applications, including heat exchanger design. Teaching not only structure but also technique, the book begins with the simplest problem solving method (scale analysis), and moves on to progressively more advanced and exact methods (integral method, self similarity, asymptotic behavior). A solutions manual is available for all problems and exercises.

Assessment of the MELCOR 1.8.6 condensation heat transfer model under the presence of noncondensable gases

Energy Technology Data Exchange (ETDEWEB)

Yoo, Ji Min; Lee, Dong Hun; Jeong, Jae Jun [Pusan National University, Busan (Korea, Republic of)

2016-05-15

Condensation heat transfer under the presence of noncondensable gases (NCGs) is an important issue in nuclear safety because the presence of even a small quantity of NC gases in the vapor largely reduces the condensation rate. The extensive assessment of the condensation model of the safety analysis codes has been also performed. When NCGs are present, the condensation phenomenon is largely reduced by accumulated NCGs near the condensing surface. Since the total pressure remains constant, the partial pressure of vapor at the liquid-vapor interface is lower than that in the bulk mixture, providing the driving force for vapor diffusion towards the liquid-vapor interface. The main objective of the present study is the assessment of the condensation heat transfer model of the severe accident code MELCOR 1.8.6 under the presence of NCGs. In this study, the condensation heat transfer model of the MELCOR 1.8.6 is assessed using various experiments which have 4 different types of geometry. Through the comparison of the results, it was shown that the MELCOR code generally under-predicts the condensation heat transfer except the condensation on outer surface of vertical pipes and improvement is needed for other geometries.

Heat transfer characteristics of a direct contact heat exchanger

International Nuclear Information System (INIS)

Kinoshita, I.; Nishi, Y.

1993-01-01

As a first step for development of a direct contact steam generator for FBRs, fundamental heat transfer characteristics of a liquid-liquid contact heat exchanger were evaluated by heat transfer experiment with low melting point alloy and water. Distinctive characteristics of direct contact heat transfer with liquid metal and water was obtained. (author)

Heat transfer 1990. Proceedings of the ninth international heat transfer conference

International Nuclear Information System (INIS)

Hetsroni, G.

1990-01-01

This book contains the proceedings of the Ninth International Heat Transfer Conference. Included in Volume 3 are the following chapters: Refrigerant vapor condensation on a horizontal tube bundle. Local heat transfer in a reflux condensation inside a closed two-phase thermosyphon and surface temperature by means of a pulsed photothermal effects

Selection of Rational Heat Transfer Intensifiers in the Heat Exchanger

Directory of Open Access Journals (Sweden)

S. A. Burtsev

2016-01-01

Full Text Available The paper considers the applicability of different types of heat transfer intensifiers in the heat exchange equipment. A review of the experimental and numerical works devoted to the intensification of the dimpled surface, surfaces with pins and internally ribbed surface were presented and data on the thermal-hydraulic characteristics of these surfaces were given. We obtained variation of thermal-hydraulic efficiency criteria for 4 different objective functions and 15 options for the intensification of heat transfer. This makes it possible to evaluate the advantages of the various heat transfer intensifiers. These equations show influence of thermal and hydraulic characteristics of the heat transfer intensifiers (the values of the relative heat transfer and drag coefficients on the basic parameters of the shell-and-tube heat exchanger: the number and length of the tubes, the volume of the heat exchanger matrix, the coolant velocity in the heat exchanger matrix, coolant flow rate, power to pump coolant (or pressure drop, the amount of heat transferred, as well as the average logarithmic temperature difference. The paper gives an example to compare two promising heat transfer intensifiers in the tubes and shows that choosing the required efficiency criterion to search for optimal heat exchanger geometry is of importance. Analysis is performed to show that a dimpled surface will improve the effectiveness of the heat exchanger despite the relatively small value of the heat transfer intensification, while a significant increase in drag of other heat transfer enhancers negatively affects their thermalhydraulic efficiency. For example, when comparing the target functions of reducing the heat exchanger volume, the data suggest that application of dimpled surfaces in various fields of technology is possible. But there are also certain surfaces that can reduce the parameters of a heat exchanger. It is shown that further work development should be aimed at

Heat transfer in an asymmetrically heated duct, 2

International Nuclear Information System (INIS)

Satoh, Isao; Kurosaki, Yasuo

1986-01-01

The objective of this article is to study theoretically and experimentally the effects of nonuniform heating on turbulent heat transfer characteristics for flow in a horizontal rectangular duct ; a vertical side wall was uniformly heated, and the other wall were insulated. In our theoretical approach, the zero-equation model for turbulent eddy viscosity was employed. The effects of mesh size of finite difference on the calculation results were examined, and some refined compensation for wall temperatures and wall shear stresses by no use of fine mesh were proposed to reduce the calculation time. The heat transfer coefficients in thermally developing region for a nonuniformly heated duct obtained from numerical solutions are larger than the one for uniformly heated case. The buoyancy effects on heat transfer were evaluated. However, it was seen that the secondary flow due to buoyancy force was hardly expected to enhance heat transfer in a turbulent duct flow. Experiments were performed to measure the velocity and temperature profiles in a turbulent duct flow with a nonuniform heated wall. The experimental results were in good agreement with the theoretical ones. (author)

Core heat transfer analysis during a BWR LOCA simulation experiment at ROSA-III

International Nuclear Information System (INIS)

Yonomoto, T.; Koizumi, Y.; Tasaka, K.

1987-01-01

The ROSA-III test facility is a 1/424-th volumetrically scaled BWR/6 simulator with an electrically heated core to study the thermal-hydraulic response during a postulated loss-of-coolant accident (LOCA). Heat transfer analyses for 5, 15, 50 and 200% break tests were conducted to understand the basic heat transfer behavior in the core under BWR LOCA conditions and to obtain a data base of post-critical heat flux (CHF) heat transfer coefficients and quench temperature. The results show that the convective heat transfer coefficient of dried-out rods at the core midplane during a steam cooling period is less than approximately 120 W/m 2 K. It is larger than existing data measured at lower pressures during a spray cooling period. Bottom-up quench temperatures are given by a simple equations: The sum of the saturation temperature and a constant of 262 K. Then the heat transfer model in the RELAP4/MOD6/U4/J3 code was revised using the present results. The rod surface temperature behavior in the 200% break test was calculated better by using the revised model although the model is very simple. (orig.)

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

Heat transfer measurements of internally heated liquids in cylindrical convection cells

International Nuclear Information System (INIS)

Fieg, G.

1978-10-01

In hypothetical reactor accidents, the thermohydraulic behaviour of core melts heated by the after-heat must be analyzed. For this purpose model experiments have been performed to study the stationary, natural convective heat transfer of internally heated fluids in cylindrical convertion cells investigating also the influence of geometry (aspect ratio) as well as of difference thermal wall conditions on to the heat transport characteristics. Axial temperature profiles, local heat flux densities at the vertical walls and their dependence, on the external Rayleigh number ar in detail reported, besides the Nusselt vs Rayleigh correlations for the aspect ratios HID=1 and 0,25. The results of these experiments are compared, as for ar possible, with existing thermohydraulic codes and simpler model asoumptions like the zone-model of Baker et. al. and after experimental verification, be used to study realistic PAHR situations. Velocity measurements by means of Laser-Doppler-Method yield information about the flow characteristics near the vertical walls and within the central part of the convecting fluid. (GL) [de

Effect of longitudinal pitch on the convection heat transfer from the tube banks in crossflow

International Nuclear Information System (INIS)

Kim, Tae-Wan; Hwang, Dae-Hyun; Lee, Chung-Chan; Kim, Keung-Ku

2006-01-01

When the tube banks in the heat exchanger are compactly designed, it is known that the average heat transfer coefficient is reduced compared with that of widely-designed tube banks. Thus, the heat transfer rate calculated by the usual heat transfer correlation will be over-estimated more than the actual one and the heat exchanger with such a design will have insufficient heat transfer capacity. Therefore, it is necessary to evaluate the effect of longitudinal and transverse pitches on the heat transfer, quantitatively. Zukauskas correlated various experimental data for aligned and staggered arrangements of tube banks as a function of Reynolds number and Prandtl number. In addition, Grimison suggested the heat transfer correlation for tube banks whose coefficients are determined by geometrical characteristics. However, Zukauskas correlation does not consider the effect of longitudinal and transverse pitches in the case of the aligned arrangement and Grimison correlation can only be used for specific geometrical arrangement such as 1.25X1.25, 1.50X1.50, and so on. Therefore, additional correlation for a heat transfer coefficient which covers a wide range of a pitch is required to predict the heat transfer rate appropriately. In this study, as a first step, the effect of a longitudinal pitch on the heat transfer is investigated for aligned tube banks by using CFD (Computational Fluid Dynamics) code

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

Energy Technology Data Exchange (ETDEWEB)

Triplett, C.E.

1996-12-01

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

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

International Nuclear Information System (INIS)

Triplett, C.E.

1996-12-01

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

Heat transfer--Orlando (Symposium), 1980

International Nuclear Information System (INIS)

Stein, R.P.

1980-01-01

This conference proceedings contains 36 papers of which 3 appear as abstracts. 23 papers are indexed separately. Topics covered include: thermodynamics of PWR and LMFBR Steam Generators; two-phase flow in parallel channels; geothermal heat transfer; natural circulation in complex geometries; heat transfer in non-Newtonian systems; and process heat transfer

Description of a heat transfer model suitable to calculate transient processes of turbocharged diesel engines with one-dimensional gas-dynamic codes

Energy Technology Data Exchange (ETDEWEB)

Galindo, J.; Lujan, J.M.; Serrano, J.R.; Dolz, V. [CMT-Motores Termicos, Universidad Politecnica de Valencia, Valencia (Spain); Guilain, S. [Renault s.a.s., Lardy (France)

2006-01-15

This paper describes a heat transfer model to be implemented in a global engine 1-D gas-dynamic code to calculate reciprocating internal combustion engine performance in steady and transient operations. A trade off between simplicity and accuracy has been looked for, in order to fit with the stated objective. To validate the model, the temperature of the exhaust manifold wall in a high-speed direct injection (HSDI) turbocharged diesel engine has been measured during a full load transient. In addition, an indirect assessment of the exhaust gas temperature during this transient process has been carried out. The results show good agreement between the measured and modelled data with good accuracy to predict the engine performance. A dual-walled air gap exhaust manifold has been tested in order to quantify the potential of exhaust gas thermal energy saving on engine transient performance. The experimental results together with the heat transfer model have been used to analyse the influence of thermal energy saving on dynamic performance during the load transient of an HSDI turbocharged diesel engine. (author)

Containment condensing heat transfer

International Nuclear Information System (INIS)

Gido, R.G.; Koestel, A.

1983-01-01

This report presents a mechanistic heat-transfer model that is valid for large scale containment heat sinks. The model development is based on the determination that the condensation is controlled by mass diffusion through the vapor-air boundary layer, and the application of the classic Reynolds' analogy to formulate expressions for the transfer of heat and mass based on hydrodynamic measurements of the momentum transfer. As a result, the analysis depends on the quantification of the shear stress (momentum transfer) at the interface between the condensate film and the vapor-air boundary layer. In addition, the currently used Tagami and Uchida test observations and their range of applicability are explained

Heat transfer enhancement for fin-tube heat exchanger using vortex generators

International Nuclear Information System (INIS)

Yoo, Seong Yeon; Park, Dong Seong; Chung, Min Ho; Lee, Sang Yun

2002-01-01

Vortex generators are fabricated on the fin surface of a fin-tube heat exchanger to augment the convective heat transfer. In addition to horseshoe vortices formed naturally around the tube of the fin-tube heat exchanger, longitudinal vortices are artificially created on the fin surface by vortex generators. The purpose of this study is to investigate the local heat transfer phenomena in the fin-tube heat exchangers with and without vortex generators, and to evaluate the effect of vortices on the heat transfer enhancement. Naphthalene sublimation technique is employed to measure local mass transfer coefficients, then analogy equation between heat and mass transfer is used to calculate heat transfer coefficients. Experiments are performed for the model of fin-circular tube heat exchangers with and without vortex generators, and of fin-flat tube heat exchangers with and without vortex generators. Average heat transfer coefficients of fin-flat tube heat exchanger without vortex generator are much lower than those of fin-circular tube heat exchanger. On the other hand, fin-flat tube heat exchanger with vortex generators has much higher heat transfer value than conventional fin-circular tube heat exchanger. At the same time, pressure losses for four types of heat exchanger is measured and compared

Heat Transfer Analysis of Localized Heat-Treatment for Grade 91 Steel

Science.gov (United States)

Walker, Jacob D.

through an inverse method by collecting actual data from different conditions and temperatures. Then the heat transfer coefficients are used to set up a model to determine the appropriate post-weld heat treatment conditions for Grade 91 steel. This will enable one to use the derived coefficients to run a forward analysis with the specific geometry and conditions they will encounter in the heat treatment process for their application. The analysis will provide a theoretical determination of time and temperatures needed to maintain the temperature for the proper time needed to properly heat treat the welded section in the desired areas that have been joined together through a welding process. Finally time and temperature combinations are compared with experimentally measured data. The forward model code applied to the parameters of the heat-treatment can then appropriately assist to determine the proper post-weld heat treatment conditions for the desired toughness and creep properties. This research is very beneficial to the joining of metals industry because it provides a way to ensure the method used to heat treat the welded section is being properly done, and the required heat treatment is achieved. It is applicable to many different geometries so that it can be modified to specific situations.

Flow regimes and heat transfer modes identification in ANGRA 2 core, during small break in the primary loop with area of 100 cm{sup 2}, simulated with RELAP5 code

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)

Radiation and combined heat transfer in channels

International Nuclear Information System (INIS)

Tamonis, M.

1986-01-01

This book presents numerical methods of calculation of radiative and combined heat transfer in channel flows of radiating as well as nonradiating media. Results obtained in calculations for flow conditions of combustion products from organic fuel products are given and methods used in determining the spectral optical properties of molecular gases are analyzed. The book presents applications of heat transfer in solving problems. Topic covered are as follows: optical properties of molecular gases; transfer equations for combined heat transfer; experimental technique; convective heat transfer in heated gas flows; radiative heat transfer in gaseous media; combined heat transfer; and radiative and combined heat transfer in applied problems

Heat transfer between adsorbate and laser-heated hot electrons

International Nuclear Information System (INIS)

Ueba, H; Persson, B N J

2008-01-01

Strong short laser pulses can give rise to a strong increase in the electronic temperature at metal surfaces. Energy transfer from the hot electrons to adsorbed molecules may result in adsorbate reactions, e.g. desorption or diffusion. We point out the limitations of an often used equation to describe the heat transfer process in terms of a friction coupling. We propose a simple theory for the energy transfer between the adsorbate and hot electrons using a newly introduced heat transfer coefficient, which depends on the adsorbate temperature. We calculate the transient adsorbate temperature and the reaction yield for a Morse potential as a function of the laser fluency. The results are compared to those obtained using a conventional heat transfer equation with temperature-independent friction. It is found that our equation of energy (heat) transfer gives a significantly lower adsorbate peak temperature, which results in a large modification of the reaction yield. We also consider the heat transfer between different vibrational modes excited by hot electrons. This mode coupling provides indirect heating of the vibrational temperature in addition to the direct heating by hot electrons. The formula of heat transfer through linear mode-mode coupling of two harmonic oscillators is applied to the recent time-resolved study of carbon monoxide and atomic oxygen hopping on an ultrafast laser-heated Pt(111) surface. It is found that the maximum temperature of the frustrated translation mode can reach high temperatures for hopping, even when direct friction coupling to the hot electrons is not strong enough

Heat transfer characteristics of horizontally oriented multi-layered annular insulation, (1)

International Nuclear Information System (INIS)

Hino, Ryutaro; Simomura, Hiroaki

1985-04-01

A computer code has been developed to analyze the natural convection heat transfer in a horizontal annular insulation layer of a hot gas duct when local gaps and inhomogeneity of filling density of insulation materials exist. This computer code simulates local gaps and inhomogeneity of filling density by a multi-layer model. This report describes an analytical model, a numerical method, an outline of program and some calculation results. (author)

Generalized irreversible heat-engine experiencing a complex heat-transfer law

International Nuclear Information System (INIS)

Chen Lingen; Li Jun; Sun Fengrui

2008-01-01

The fundamental optimal relation between optimal power-output and efficiency of a generalized irreversible Carnot heat-engine is derived based on a generalized heat-transfer law, including a generalized convective heat-transfer law and a generalized radiative heat-transfer law, q ∝ (ΔT n ) m . The generalized irreversible Carnot-engine model incorporates several internal and external irreversibilities, such as heat resistance, bypass heat-leak, friction, turbulence and other undesirable irreversibility factors. The added irreversibilities, besides heat resistance, are characterized by a constant parameter and a constant coefficient. The effects of heat-transfer laws and various loss terms are analyzed. The results obtained corroborate those in the literature

Transient heat transfer analysis up to dryout in 3D fuel rods under unideal conditions through the development of a computer code

International Nuclear Information System (INIS)

Martins, Rodolfo I.; Affonso, Renato R.W.; Moreira, Maria de Lourdes; Sampaio, Paulo A. B. de

2017-01-01

In this paper we analyze a conjugated transient heat transfer problem consisting of a nuclear reactor's fuel rod and its intrinsic coolant channel. Our analysis is made possible through a computer code being developed at the Instituto de Engenharia Nuclear (IEN/CNEN). This code is meant to study the temperature behavior in fuel rods which exhibit deviation from their ideal conditions, that is, rods in which the cladding is deformed or the fuel is dislocated. It is also designed to avoid the use of the computationally expensive Navier-Stokes equations. For these reasons, its physical model has as basis a three-dimensional fuel rod coupled to a one-dimensional coolant channel, which are discretized using the finite element method. Intending to study accidental conditions in which the coolant (light water) transcends its saturation temperature, turning into vapor, a homogeneous mixture is used to represent the two-phase flow, and so the coolant channel's energy equation is described using enthalpy. Owing to the fact that temperature and enthalpy are used in the physical model, it became impractical to generate a fully coupled method for solving the pertinent equations. Thus, the conjugated heat transfer problem is solved in a segregated manner through the implementation of an iterative method. Finally, as study cases for this paper we present analyses concerning the behavior of the hottest fuel rod in a Pressurized Water Reactor during a shutdown wherein the residual heat removal system is lost (loss of the reactor's coolant pumps). These studies contemplate cases in which the fuel rod's geometry is ideal or curved. Analyses are also performed for two circumstances of positioning of the fuel inside the rod: concentric and eccentric. (author)

Phase change heat transfer device for process heat applications

International Nuclear Information System (INIS)

Sabharwall, Piyush; Patterson, Mike; Utgikar, Vivek; Gunnerson, Fred

2010-01-01

The next generation nuclear plant (NGNP) will most likely produce electricity and process heat, with both being considered for hydrogen production. To capture nuclear process heat, and transport it to a distant industrial facility requires a high temperature system of heat exchangers, pumps and/or compressors. The heat transfer system is particularly challenging not only due to the elevated temperatures (up to ∼1300 K) and industrial scale power transport (≥50 MW), but also due to a potentially large separation distance between the nuclear and industrial plants (100+ m) dictated by safety and licensing mandates. The work reported here is the preliminary analysis of two-phase thermosyphon heat transfer performance with alkali metals. A thermosyphon is a thermal device for transporting heat from one point to another with quite extraordinary properties. In contrast to single-phased forced convective heat transfer via 'pumping a fluid', a thermosyphon (also called a wickless heat pipe) transfers heat through the vaporization/condensing process. The condensate is further returned to the hot source by gravity, i.e., without any requirement of pumps or compressors. With this mode of heat transfer, the thermosyphon has the capability to transport heat at high rates over appreciable distances, virtually isothermally and without any requirement for external pumping devices. Two-phase heat transfer by a thermosyphon has the advantage of high enthalpy transport that includes the sensible heat of the liquid, the latent heat of vaporization, and vapor superheat. In contrast, single-phase forced convection transports only the sensible heat of the fluid. Additionally, vapor-phase velocities within a thermosyphon are much greater than single-phase liquid velocities within a forced convective loop. Thermosyphon performance can be limited by the sonic limit (choking) of vapor flow and/or by condensate entrainment. Proper thermosyphon requires analysis of both.

Heat transfer study under supercritical pressure conditions

International Nuclear Information System (INIS)

Yamashita, Tohru; Yoshida, Suguru; Mori, Hideo; Morooka, Shinichi; Komita, Hideo; Nishida, Kouji

2003-01-01

Experiments were performed on heat transfer and pressure drop of a supercritical pressure fluid flowing upward in a uniformly heated vertical tube of a small diameter, using HCFC22 as a test fluid. Following results were obtained. (1) Characteristics of the heat transfer are similar to those for the tubes of large diameter. (2) The effect of tube diameter on the heat transfer was seen for a 'normal heat transfer, but not for a 'deteriorated' heat transfer. (3) The limit heat flux for the occurrence of deterioration in heat transfer becomes larger with smaller diameter tube. (4) The Watts and Chou correlation has the best prediction performance for the present data in the 'normal' heat transfer region. (5) Frictional pressure drop becomes smaller than that for an isothermal flow in the region near the pseudocritical point, and this reduction was more remarkable for the deteriorated' heat transfer. (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.

Heat transfer in Rockwool modelling and method of measurement. Modelling radiative heat transfer in fibrous materials

Energy Technology Data Exchange (ETDEWEB)

Dyrboel, Susanne

1998-05-01

Fibrous materials are some of the most widely used materials for thermal insulation. In this project the focus of interest has been on fibrous materials for building application. Interest in improving the thermal properties of insulation materials is increasing as legislation is being tightened to reduce the overall energy consumption. A knowledge of the individual heat transfer mechanisms - whereby heat is transferred within a particular material is an essential tool to improve continuously the thermal properties of the material. Heat is transferred in fibrous materials by four different transfer mechanisms: conduction through air, conduction through fibres, thermal radiation and convection. In a particular temperature range the conduction through air can be regarded as a constant, and conduction through fibres is an insignificant part of the total heat transfer. Radiation, however, constitutes 25-40% of the total heat transfer in light fibrous materials. In Denmark and a number of other countries convection in fibrous materials is considered as non-existent when calculating heat transmission as well as when designing building structures. Two heat transfer mechanisms have been the focus of the current project: radiation heat transfer and convection. The radiation analysis serves to develop a model that can be used in further work to gain a wider knowledge of the way in which the morphology of the fibrous material, i.e. fibre diameter distribution, fibre orientation distribution etc., influences the radiation heat transfer under different conditions. The convection investigation serves to examine whether considering convection as non-existent is a fair assumption to use in present and future building structures. The assumption applied in practically is that convection makes a notable difference only in very thick insulation, at external temperatures below -20 deg. C, and at very low densities. For lager thickness dimensions the resulting heat transfer through the

Numerical investigation of fluid flow and heat transfer under high heat flux using rectangular micro-channels

KAUST Repository

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.

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.

Investigations on post-dryout heat transfer in bilaterally heated annular channels

International Nuclear Information System (INIS)

Tian, W.X.; Qiu, S.Z.; Jia, D.N.

2006-01-01

Post-dryout heat transfer in bilaterally heated vertical narrow annular channels with 1.0, 1.5 and 2.0 mm gap size has been experimentally investigated with deionized water under the condition of pressure ranging from 1.38 to 5.9 MPa and low mass flow rate from 42.9 to 150.2 kg/m 2 s. The experimental data was compared with well known empirical correlations including Groeneveld, Mattson, etc., and none of them gave an ideal prediction. Theoretical investigations were also carried out on post-dryout heat transfer in annular channels. Based on analysis of heat exchange processes arising among the droplets, the vapor and two tube walls of annular channel, a non-equilibrium mechanistic heat transfer model was developed. Comparison indicated that the present model prediction showed a good agreement with our experimental data. Theoretical calculation result showed that the forced convective heat transfer between the heated wall and vapor dominate the overall heat transfer. The heat transfer caused by the droplets direct contact to the wall and the interfacial convection/evaporation of droplets in superheated vapors also had an indispensable contribution. The radiation heat transfer would be neglected because of its small contribution (less than 0.11%) to the total heat transfer

Influence of taking into account in-pressurizer convective heat- and mass transfer influence effects at the transients in VVER with code RELAP 5/MOD 3.2

International Nuclear Information System (INIS)

Konovalyuk, L.N.; Shevelev, D.V.; Kravchenko, V.G.

2003-01-01

PRZ model is proposed which allows taking into account in pressurizer convective heat- and mass transfer influence effects at the transients in VVER (PWR) Type Reactors case when calculations performed with using 1D thermohydraulic codes. The theoretical backgrounds are given to define the transients with the convective coolant instability in PRZ. The instability threshold is given for real PRZ geometry

Numerical Investigation on Supercritical Heat Transfer of RP3 Kerosene Flowing inside a Cooling Channel of Scramjet

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.

Study on enhancement of heat transfer of reactor vessel auxiliary cooling system of fast breeder reactor

International Nuclear Information System (INIS)

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

1996-01-01

A reactor vessel auxiliary cooling system (RVACS), which is one of the decay heat removal systems of the fast breeder reactor (FBR), has passive safety as well as high reliability. However, the heat removal capability is relatively small, because its heat exchange is dependent on the natural convection of the air. The objectives of this report are to propose a heat transfer medium to enhance the heat transfer and to confirm the heat transfer performance of this system by experimental and analytical studies. From these studies, the following main results were obtained. (1) A porous plate with 5 mm thickness, 5 mm pore diameter, 92% porosity, was found to have the highest enhancement of heat transfer. (2) The heat transfer enhancement was demonstrated by large scale heat transfer experiments. Also, the heat transfer correlations, which can be used in the plant transient analyses, were derived from the experimental results. (3) Analysing the transient conditions of conventional pool-type FBR by means of the system analysis code, the applicable range of this system was assumed from the capability of the RVACS with porous plates. As a result, this type of RVACS was found to be applicable to conventional pool-type FBRs with capacity of about 500 MWe or less. (author)

A literature survey on numerical heat transfer

Science.gov (United States)

Shih, T. M.

1982-12-01

Technical papers in the area of numerical heat transfer published from 1977 through 1981 are reviewed. The journals surveyed include: (1) ASME Journal of Heat Transfer, (2) International Journal of Heat and Mass Transfer, (3) AIAA Journal, (4) Numerical Heat Transfer, (5) Computers and Fluids, (6) International Journal for Numerical Methods in Engineering, (7) SIAM Journal of Numerical Analysis, and (8) Journal of Computational Physics. This survey excludes experimental work in heat transfer and numerical schemes that are not applied to equations governing heat transfer phenomena. The research work is categorized into the following areas: (A) conduction, (B) boundary-layer flows, (C) momentum and heat transfer in cavities, (D) turbulent flows, (E) convection around cylinders and spheres or within annuli, (F) numerical convective instability, (G) radiation, (H) combustion, (I) plumes, jets, and wakes, (J) heat transfer in porous media, (K) boiling, condensation, and two-phase flows, (L) developing and fully developed channel flows, (M) combined heat and mass transfer, (N) applications, (O) comparison and properties of numerical schemes, and (P) body-fitted coordinates and nonuniform grids.

Turbulent heavy liquid metal heat transfer along a heated rod in an annular cavity

International Nuclear Information System (INIS)

Lefhalm, C.-H.; Tak, N.-I.; Piecha, H.; Stieglitz, R.

2004-01-01

Heavy liquid metals (HLM) are considered as coolant and spallation material in accelerator driven systems (ADS), because of their good molecular heat conductivity. This property leads to a separation of the spatial extension of thermal and viscous boundary layers. Commercially available computational fluid dynamic codes (CFD) assume an analogy of momentum and energy transfer, which is problematic for liquid metals flow. Therefore, benchmark experiments are required, in order to validate codes or modify existing models used therein. Within this article an experimental and numerical study of a thermally developing turbulent lead bismuth (PbBi) flow along a uniformly heated rod in a circular tube is presented. Local temperatures and velocity distributions are measured using thermocouples and Pitot tubes. The data are compared to simulation results computed with the CFX code package. The measured velocity profiles coincide nearly perfect with the simulation results. However, discrepancies up to 7% between the measured and computed temperatures appear. A minor part of the deviations can be explained by the imperfect experimental set-up. But, the measured shape of the thermal boundary is different to the calculated one, indicating the inadequateness of the presently used models describing the turbulent heat transport within the thermal boundary layer

Heat transfer bibliography: russian works

Energy Technology Data Exchange (ETDEWEB)

Luikov, A V

1965-02-01

This bibliography of recent Russian publications in heat transfer is divided into the following categories: (1) books; (2) general; (3) experimental methods; (4) analytical calculation methods; (5) thermodynamics; (6) transfer processes involving phase conversions; ((7) transfer processes involving chemical conversions; (8) transfer processes involving very high velocities; (9) drying processes; (10) thermal properties of various materials, heat transfer agents and their determination methods; (11) high temperature physics and magneto- hydrodynamics; and (12) transfer processes in technological apparatuses. (357 refs.)

Advances in heat transfer

CERN Document Server

Hartnett, James P; Cho, Young I; Greene, George A

2001-01-01

Heat transfer is the exchange of heat energy between a system and its surrounding environment, which results from a temperature difference and takes place by means of a process of thermal conduction, mechanical convection, or electromagnetic radiation. Advances in Heat Transfer is designed to fill the information gap between regularly scheduled journals and university-level textbooks by providing in-depth review articles over a broader scope than is allowable in either journals or texts.

Heat transfer, condensation and fog formation in crossflow plastic heat exchangers

NARCIS (Netherlands)

Brouwers, H.J.H.

1996-01-01

In this paper heat transfer of air-water-vapour mixtures in plastic crossflow heat exchangers is studied theoretically and experimentally. First, a model for heat transfer without condensation is derived, resulting in a set of classical differential equations. Subsequently, heat transfer with wall

Boiling Heat Transfer to Halogenated Hydrocarbon Refrigerants

Science.gov (United States)

Yoshida, Suguru; Fujita, Yasunobu

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

Heat exchanger network retrofit optimization involving heat transfer enhancement

International Nuclear Information System (INIS)

Wang Yufei; Smith, Robin; Kim, Jin-Kuk

2012-01-01

Heat exchanger network retrofit plays an important role in energy saving in process industry. Many design methods for the retrofit of heat exchanger networks have been proposed during the last three decades. Conventional retrofit methods rely heavily on topology modifications which often result in a long retrofit duration and high initial costs. Moreover, the addition of extra surface area to the heat exchanger can prove difficult due to topology, safety and downtime constraints. Both of these problems can be avoided through the use of heat transfer enhancement in heat exchanger network retrofit. This paper presents a novel design approach to solve heat exchanger network retrofit problems based on heat transfer enhancement. An optimisation method based on simulated annealing has been developed to find the appropriate heat exchangers to be enhanced and to calculate the level of enhancement required. The physical insight of enhanced exchangers is also analysed. The new methodology allows several possible retrofit strategies using different retrofit methods be determined. Comparison of these retrofit strategies demonstrates that retrofit modification duration and payback time are reduced when heat transfer enhancement is utilised. Heat transfer enhancement can be also used as a substitute for increased heat exchanger network surface area to reduce retrofit investment costs.

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)

Hybrid mesh finite volume CFD code for studying heat transfer in a forward-facing step

Energy Technology Data Exchange (ETDEWEB)

Jayakumar, J S; Kumar, Inder [Bhabha Atomic Research Center, Mumbai (India); Eswaran, V, E-mail: jsjayan@gmail.com, E-mail: inderk@barc.gov.in, E-mail: eswar@iitk.ac.in [Indian Institute of Technology, Kanpur (India)

2010-12-15

Computational fluid dynamics (CFD) methods employ two types of grid: structured and unstructured. Developing the solver and data structures for a finite-volume solver is easier than for unstructured grids. But real-life problems are too complicated to be fitted flexibly by structured grids. Therefore, unstructured grids are widely used for solving real-life problems. However, using only one type of unstructured element consumes a lot of computational time because the number of elements cannot be controlled. Hence, a hybrid grid that contains mixed elements, such as the use of hexahedral elements along with tetrahedral and pyramidal elements, gives the user control over the number of elements in the domain, and thus only the domain that requires a finer grid is meshed finer and not the entire domain. This work aims to develop such a finite-volume hybrid grid solver capable of handling turbulence flows and conjugate heat transfer. It has been extended to solving flow involving separation and subsequent reattachment occurring due to sudden expansion or contraction. A significant effect of mixing high- and low-enthalpy fluid occurs in the reattached regions of these devices. This makes the study of the backward-facing and forward-facing step with heat transfer an important field of research. The problem of the forward-facing step with conjugate heat transfer was taken up and solved for turbulence flow using a two-equation model of k-{omega}. The variation in the flow profile and heat transfer behavior has been studied with the variation in Re and solid to fluid thermal conductivity ratios. The results for the variation in local Nusselt number, interface temperature and skin friction factor are presented.

Hybrid mesh finite volume CFD code for studying heat transfer in a forward-facing step

Science.gov (United States)

Jayakumar, J. S.; Kumar, Inder; Eswaran, V.

2010-12-01

Computational fluid dynamics (CFD) methods employ two types of grid: structured and unstructured. Developing the solver and data structures for a finite-volume solver is easier than for unstructured grids. But real-life problems are too complicated to be fitted flexibly by structured grids. Therefore, unstructured grids are widely used for solving real-life problems. However, using only one type of unstructured element consumes a lot of computational time because the number of elements cannot be controlled. Hence, a hybrid grid that contains mixed elements, such as the use of hexahedral elements along with tetrahedral and pyramidal elements, gives the user control over the number of elements in the domain, and thus only the domain that requires a finer grid is meshed finer and not the entire domain. This work aims to develop such a finite-volume hybrid grid solver capable of handling turbulence flows and conjugate heat transfer. It has been extended to solving flow involving separation and subsequent reattachment occurring due to sudden expansion or contraction. A significant effect of mixing high- and low-enthalpy fluid occurs in the reattached regions of these devices. This makes the study of the backward-facing and forward-facing step with heat transfer an important field of research. The problem of the forward-facing step with conjugate heat transfer was taken up and solved for turbulence flow using a two-equation model of k-ω. The variation in the flow profile and heat transfer behavior has been studied with the variation in Re and solid to fluid thermal conductivity ratios. The results for the variation in local Nusselt number, interface temperature and skin friction factor are presented.

Hybrid mesh finite volume CFD code for studying heat transfer in a forward-facing step

International Nuclear Information System (INIS)

Jayakumar, J S; Kumar, Inder; Eswaran, V

2010-01-01

Computational fluid dynamics (CFD) methods employ two types of grid: structured and unstructured. Developing the solver and data structures for a finite-volume solver is easier than for unstructured grids. But real-life problems are too complicated to be fitted flexibly by structured grids. Therefore, unstructured grids are widely used for solving real-life problems. However, using only one type of unstructured element consumes a lot of computational time because the number of elements cannot be controlled. Hence, a hybrid grid that contains mixed elements, such as the use of hexahedral elements along with tetrahedral and pyramidal elements, gives the user control over the number of elements in the domain, and thus only the domain that requires a finer grid is meshed finer and not the entire domain. This work aims to develop such a finite-volume hybrid grid solver capable of handling turbulence flows and conjugate heat transfer. It has been extended to solving flow involving separation and subsequent reattachment occurring due to sudden expansion or contraction. A significant effect of mixing high- and low-enthalpy fluid occurs in the reattached regions of these devices. This makes the study of the backward-facing and forward-facing step with heat transfer an important field of research. The problem of the forward-facing step with conjugate heat transfer was taken up and solved for turbulence flow using a two-equation model of k-ω. The variation in the flow profile and heat transfer behavior has been studied with the variation in Re and solid to fluid thermal conductivity ratios. The results for the variation in local Nusselt number, interface temperature and skin friction factor are presented.

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

Heat transfer enhancement on nucleate boiling

International Nuclear Information System (INIS)

Zhuang, M.; Guibai, L.

1990-01-01

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

Transient heat transfer analysis up to dryout in 3D fuel rods under unideal conditions through the development of a computer code

Energy Technology Data Exchange (ETDEWEB)

Martins, Rodolfo I.; Affonso, Renato R.W.; Moreira, Maria de Lourdes; Sampaio, Paulo A. B. de, E-mail: rodolfoienny@gmail.com [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

2017-11-01

In this paper we analyze a conjugated transient heat transfer problem consisting of a nuclear reactor's fuel rod and its intrinsic coolant channel. Our analysis is made possible through a computer code being developed at the Instituto de Engenharia Nuclear (IEN/CNEN). This code is meant to study the temperature behavior in fuel rods which exhibit deviation from their ideal conditions, that is, rods in which the cladding is deformed or the fuel is dislocated. It is also designed to avoid the use of the computationally expensive Navier-Stokes equations. For these reasons, its physical model has as basis a three-dimensional fuel rod coupled to a one-dimensional coolant channel, which are discretized using the finite element method. Intending to study accidental conditions in which the coolant (light water) transcends its saturation temperature, turning into vapor, a homogeneous mixture is used to represent the two-phase flow, and so the coolant channel's energy equation is described using enthalpy. Owing to the fact that temperature and enthalpy are used in the physical model, it became impractical to generate a fully coupled method for solving the pertinent equations. Thus, the conjugated heat transfer problem is solved in a segregated manner through the implementation of an iterative method. Finally, as study cases for this paper we present analyses concerning the behavior of the hottest fuel rod in a Pressurized Water Reactor during a shutdown wherein the residual heat removal system is lost (loss of the reactor's coolant pumps). These studies contemplate cases in which the fuel rod's geometry is ideal or curved. Analyses are also performed for two circumstances of positioning of the fuel inside the rod: concentric and eccentric. (author)

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)

Prediction of Heat Transfer Performance on Horizontal U-Shaped Heat Exchanger in Passive Safety System Using MARS

Energy Technology Data Exchange (ETDEWEB)

Jeon, Seong-Su; Hong, Soon-Joon [FNC Tech, Yongin (Korea, Republic of); Cho, Hyoung-Kyu; Park, Goon-Cherl [Seoul National University, Seoul (Korea, Republic of)

2015-10-15

The design and the safety analysis of the passive safety systems are performed mainly using the best-estimate thermal-hydraulic analysis codes such as RELAP5 and MARS. This study developed the heat transfer model package for the horizontal U-shaped HX submerged in a pool by improving the horizontal in-tube condensation model and developing the outside-tube natural convective nucleate boiling model. This paper presents the HX model package and the validation results against the passive safety system-related experimental data of PASCAL and ATLAS-PAFS. This study developed the heat transfer model package of the horizontal U-shaped HX submerged in a pool in order to obtain a reliable prediction of the HX heat removal performance of the passive safety system, especially PAFS, using MARS. From the validation results, the proposed model package provided the improved prediction of HX performance (condensation, natural convective nucleate boiling, and heat removal rate of the HX) compared to the default model in MARS.

Numerical study of turbulent heat transfer along a heated rod in an annular cavity

International Nuclear Information System (INIS)

Batta, A.; Class, A.; Daubner, M.; Gnieser, S.; Stieglitz, R.

2008-01-01

Fundamental knowledge on the turbulent convective heat transfer from a rod into liquid metal is of crucial importance for the design of advanced liquid metal operated nuclear systems since a single rod is the basic element of a fuel rod assembly. Therefore, a numerical investigation of the heated rod experiment at KALLA (KArlsruhe Liquid metal LAboratory) has been performed. This experiment investigates the turbulent heat transfer from a heated rod placed concentrically within in a cylindrical tube in a developing flow of a heavy liquid metal (HLM, here Pb 45 Bi 55 Eutectic) at reactor typical power levels and dimensions. It is set up with thermocouples (TCs), a traversable Pitot tube and three thermocouple rakes consisting of numerous thermocouples (TCs). The concentricity is ensured by means of mechanical spacers placed axially equidistant. This article concentrates on the numerical investigation of the impact of the experimental instrumentation on the developing flow pattern and temperature field. In particular, the influence of spacers which distort the velocity profile as well of a potential contact of the spacer with the heated rod changing the heat conduction regime are considered numerically in this paper using the STAR-CD code. The turbulent flow simulation assumes axis-symmetry and uses the SST turbulence model. The simulation results exhibit a flow pattern that is substantially altered by spacers. Hence, the flow can not be considered to be axis-symmetric. This in turn yields that the convective heat transfer from the heated rod towards the spacer region is reduced leading to a temperature rise in spacer region, which represents the maximum value in this domain. As a consequence the entire three-dimensional test section must be modelled in order to correctly represent the physics and to allow an adequate interpretation of the experimental data. (orig.)

Heat transfer correlations in mantle tanks

DEFF Research Database (Denmark)

Furbo, Simon; Knudsen, Søren

2005-01-01

on calculations with a CFD-model, which has earlier been validated by means of experiments. The CFD-model is used to determine the heat transfer between the solar collector fluid in the mantle and the walls surrounding the mantle in all levels of the mantle as well as the heat transfer between the wall...... transfer correlations are suitable as input for a detailed simulation model for mantle tanks. The heat transfer correlations determined in this study are somewhat different from previous reported heat transfer correlations. The reason is that this study includes more mantle tank designs and operation......Small solar domestic hot water systems are best designed as low flow systems based on vertical mantle tanks. Theoretical investigations of the heat transfer in differently designed vertical mantle tanks during different operation conditions have been carried out. The investigations are based...

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

Study of overall heat transfer coefficient from upper crust to overlaying water during MCCI

International Nuclear Information System (INIS)

Kondo, Masaya; Nishida, Ayumu; Sugimoto, Jun

2015-01-01

A model of the overall heat transfer between the molten core and the overlying coolant above crust during MCCI in severe accident is proposed and confirmed experimentally and analytically. The model assumes that the heat transferred from molten core to the overlaying water is proportional to the amount of water that reaches the molten core surface. The water flow to the molten core surface is assumes to be prevented by the CCFL in the porous crust. Thus, the steam flow and the non-condensable gas flow interact with the water flow. The present model describes the relationship between the overall heat transfer and the water flow, and furthermore, the CCFL effect on the water flow. The non-condensable gas effect on the overall heat transfer predicted by the present model agrees well with experiments. The effects of porosity and hole diameter on the amount of water, which reaches the molten core surface, has also been confirmed using RELAP5 code. (author)

Assessment of reflood heat transfer model of COBRA-TIF against ABB-CE evaluation model

Energy Technology Data Exchange (ETDEWEB)

Lee, S. I.; Lee, S. Y.; Park, C. E.; Choi, H. R.; Choi, C. J. [Korea Power Engineering Company Inc., Taejon (Korea, Republic of)

2000-05-01

According to 10 CFR 50 Appendix K, ECCS performance evaluation model should be based on the experimental data of FLECHT and have the conservatism compared with experimental data. To meet this requirement ABB-CE has the complicate code structure as follows: COMPERC-II calculates three reflood rates, and FLELAPC and HTCOF calculate the reflood heat transfer coefficients, and finally STRIKIN-II calculates the cladding temperature using the reflood heat transfer calculated in previous stage. In this paper, to investigate whether or not COBRA-TF satisfies the requirement of Appendix K, the reflood heat transfer coefficient of COBRA-TF was assessed against ABB-CE MOD-2C model. It was found out that COBRA-TF predicts properly the experimental data and has more conservatism than the results of ABB-CE MOD-2C model. Based on these results, it can be concluded that the reflood heat transfer coefficients calculated by COBRA-TF meet the requirement of Appendix K.

Heat transfer II essentials

CERN Document Server

REA, The Editors of

1988-01-01

REA's Essentials provide quick and easy access to critical information in a variety of different fields, ranging from the most basic to the most advanced. As its name implies, these concise, comprehensive study guides summarize the essentials of the field covered. Essentials are helpful when preparing for exams, doing homework and will remain a lasting reference source for students, teachers, and professionals. Heat Transfer II reviews correlations for forced convection, free convection, heat exchangers, radiation heat transfer, and boiling and condensation.

Transfer of heat to fluidized-solids beds

Energy Technology Data Exchange (ETDEWEB)

1952-10-16

The improvement in the method described and claimed in patent application 14,363/47 (136,186) for supplying heat to a dense turbulent mass of solid fluidized by a gas flowing upwardly therethrough and subjected to a high temperature in a treating zone, by heat transfer through heat-transfer surfaces of heat-transfer elements in contact with the said turbulent mass of finely divided solid and heated by means of a fluid heating medium, including burning fuels comprising contacting the said heat-transfer surfaces with a fuel and a combustion supporting gas under such conditions that the combustion of the fuel is localized in the heat-transfer element near the point of entry of the fuel and combustion-supporting gas and a substantial temperature gradient is maintained along the path of said fuel combustion-supporting gas and combustion products through the said heat-transfer element.

The experimental study of heat transfer around molds inside a model autoclave

Science.gov (United States)

Ghamlouch, Taleb; Roux, Stéphane; Lefèvre, Nicolas; Bailleul, Jean-Luc; Sobotka, Vincent

2018-05-01

The temperature distribution within composite parts manufactured inside autoclaves plays a key role in determining the parts quality at the end of the curing cycle. Indeed, heat transfer between the parts and the surroundings inside an autoclave is strongly coupled with the flow field around the molds and can be modeled through the convective heat transfer coefficient (HTC). The aerodynamically unsuitable geometry of the molds generates complex turbulent non-uniform flows around them accompanied with the presence of dead zones. This heterogeneity can imply non-uniform convective heat transfers leading to temperature gradients inside parts that can be prejudicial. Given this fact, the purpose of this study is to perform experimental measurements in order to describe the flow field and the convective heat transfer behavior around representative industrial molds installed inside a home-made model. A key point of our model autoclave is the ease of use of non-intrusive measuring instruments: the Particle Image Velocimetry (PIV) technique and infrared imaging camera for the study of the flow field and the heat transfer coefficient distribution around the molds respectively. The experimental measurements are then compared to computational fluid dynamics (CFD) calculations performed on the computer code ANSYS Fluent 16.0®. This investigation has revealed, as expected, a non-uniform distribution of the convective heat transfer coefficient around the molds and therefore the presence of thermal gradients which can reduce the composite parts quality during an autoclave process. A good agreement has been achieved between the experimental and the numerical results leading then to the validation of the performed numerical simulations.

Conjugate heat transfer analysis for in-vessel retention with external reactor vessel cooling

International Nuclear Information System (INIS)

Park, Jong-Woon; Bae, Jae-ho; Song, Hyuk-Jin

2016-01-01

Highlights: • A conjugate heat transfer analysis method is applied for in-vessel corium retention. • 3D heat diffusion has a formidable effect in alleviating focusing heat load from metallic layer. • The focusing heat load is decreased by about 2.5 times on the external surface. - Abstract: A conjugate heat transfer analysis method for the thermal integrity of a reactor vessel under external reactor vessel cooling conditions is developed to resolve light metal layer focusing effect issue for in-vessel retention. The method calculates steady-state three-dimensional temperature distribution of a reactor vessel using coupled conjugate heat transfer between in-vessel three-layered stratified corium (metallic pool, oxide pool and heavy metal and polar-angle dependent boiling heat transfer at the outer surface of a reactor vessel). The three-layer corium heat transfer model is utilizing lumped-parameter thermal-resistance circuit method. For the ex-vessel boiling boundary conditions, nucleate, transition and film boiling are considered. The thermal integrity of a reactor vessel is addressed in terms of heat flux at the outer-most nodes of the vessel and remaining thickness profile. The vessel three-dimensional heat conduction is validated against a commercial code. It is found that even though the internal heat flux from the metal layer goes far beyond critical heat flux (CHF) the heat flux from the outermost nodes of the vessel may be maintained below CHF due to massive vessel heat diffusion. The heat diffusion throughout the vessel is more pronounced for relatively low heat generation rate in an oxide pool. Parametric calculations are performed considering thermal conditions such as peak heat flux from a light metal layer, heat generation in an oxide pool and external boiling conditions. The major finding is that the most crucial factor for success of in-vessel retention is not the mass of the molten light metal above the oxide pool but the heat generation rate

Tunable heat transfer with smart nanofluids.

Science.gov (United States)

Bernardin, Michele; Comitani, Federico; Vailati, Alberto

2012-06-01

Strongly thermophilic nanofluids are able to transfer either small or large quantities of heat when subjected to a stable temperature difference. We investigate the bistability diagram of the heat transferred by this class of nanofluids. We show that bistability can be exploited to obtain a controlled switching between a conductive and a convective regime of heat transfer, so as to achieve a controlled modulation of the heat flux.

Postaccident heat removal. II. Heat transfer from an internally heated liquid to a melting solid

International Nuclear Information System (INIS)

Faw, R.E.; Baker, L. Jr.

1976-01-01

Microwave heating has been used in studies of heat transfer from a horizontal layer of internally heated liquid to a melting solid. Experiments were designed to simulate heat transfer and meltthrough processes of importance in the analysis of postaccident heat removal capabilities of nuclear reactors. Glycerin, heated by 2.45-GHz microwave radiation, was used to simulate molten fuel. Paraffin wax was used to simulate a melting barrier confining the fuel. Experimentally measured heat fluxes and melting rates were consistent with a model based on downward heat transfer by conduction through a stagnant liquid layer and upward heat transfer augmented by natural convection. Melting and displacement of the barrier material occurred by upward-moving droplets randomly distributed across the melting surface. Results indicated that the melting and displacement process had no effect on the heat transfer process

Modelling of fluid flow and heat transfer in a reciprocating compressor

Science.gov (United States)

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.

Axial flow heat exchanger devices and methods for heat transfer using axial flow devices

Science.gov (United States)

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.

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.

Boiling heat transfer on horizontal tube bundles

International Nuclear Information System (INIS)

Anon.

1987-01-01

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

Heat transfer and fire spread

Science.gov (United States)

Hal E. Anderson

1969-01-01

Experimental testing of a mathematical model showed that radiant heat transfer accounted for no more than 40% of total heat flux required to maintain rate of spread. A reasonable prediction of spread was possible by assuming a horizontal convective heat transfer coefficient when certain fuel and flame characteristics were known. Fuel particle size had a linear relation...

A Summary of Interfacial Heat Transfer Models and Correlations

Energy Technology Data Exchange (ETDEWEB)

Bae, Sung Won; Cho, Hyung Kyu; Lee, Young Jin; Kim, Hee Chul; Jung, Young Jong; Kim, K. D. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2007-10-15

A long term project has been launched in October 2006 to develop a plant safety analysis code. 5 organizations are joining together for the harmonious coworking to build up the code. In this project, KAERI takes the charge of the building up the physical models and correlations about the transport phenomena. The momentum and energy transfer terms as well as the mass are surveyed from the RELAP5/MOD3, RELAP5-3D, CATHARE, and TRAC-M does. Also the recent papers are surveyed. Among these resources, most of the CATHARE models are based on their own experiment and test results. Thus, the CATHARE models are only used as the comparison purposes. In this paper, a summary of the models and the correlations about the interfacial heat transfer are represented. These surveyed models and correlations will be tested numerically and one correlation is selected finally.

Heat transfer in heterogeneous propellant combustion systems

International Nuclear Information System (INIS)

Brewster, M.Q.

1992-01-01

This paper reports that heat transfer plays an important role in several critical areas of heterogeneous, solid-propellant combustion systems. These areas include heat feedback to the propellant surface, heat transfer between burning aluminum droplets and their surroundings, heat transfer to internal insulation systems, and heat transfer to aft-end equipment. Gas conduction dominates heat feedback to the propellant surface in conventional ammonium perchlorate (AP) composite propellants, although particle radiative feedback also plays a significant role in combustion of metalized propellants. Particle radiation plays a dominant role in heat transfer to internal insulation, compared with that of convection. However, conduction by impingement of burning aluminum particles, which has not been extensively studied, may also be significant. Radiative heat loss plays an important role in determining the burning rate of molten aluminum particles due to a highly luminous, oxide particle-laden, detached flame envelope. Radiation by aluminum oxide smoke particles also plays a dominant role in heat transfer from the exhaust plume to aft-end equipment. Uncertainties in aluminum oxide particle-size distribution and optical properties still make it difficult to predict radiative plume heat transfer accurately from first principles

Condensation heat transfer in plate heat exchangers

International Nuclear Information System (INIS)

Panchal, C.B.

1985-01-01

An Alfa-Laval plate heat exchanger, previously tested as an evaporator, was retested as a condenser. Two series of tests with different chevron-angle plates were carried out using ammonia as a working fluid. The overall heat-transfer coefficient and pressure drop were measured, and the effects of operating parameters were determined. The experimental data were compared with theoretical predictions. In the analysis, a gravity-controlled condensation process was modeled theoretically, and the overall performance was calculated. The analysis shows that the overall heat-transfer coefficient can be predicted with an average uncertainty of about 10%. It is, however, important to consider the interfacial shear stress, because the effective friction factor is high for flow in plate heat exchangers

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)

A moving subgrid model for simulation of reflood heat transfer

International Nuclear Information System (INIS)

Frepoli, Cesare; Mahaffy, John H.; Hochreiter, Lawrence E.

2003-01-01

In the quench front and froth region the thermal-hydraulic parameters experience a sharp axial variation. The heat transfer regime changes from single-phase liquid, to nucleate boiling, to transition boiling and finally to film boiling in a small axial distance. One of the major limitations of all the current best-estimate codes is that a relatively coarse mesh is used to solve the complex fluid flow and heat transfer problem in proximity of the quench front during reflood. The use of a fine axial mesh for the entire core becomes prohibitive because of the large computational costs involved. Moreover, as the mesh size decreases, the standard numerical methods based on a semi-implicit scheme, tend to become unstable. A subgrid model was developed to resolve the complex thermal-hydraulic problem at the quench front and froth region. This model is a Fine Hydraulic Moving Grid (FHMG) that overlies a coarse Eulerian mesh in the proximity of the quench front and froth region. The fine mesh moves in the core and follows the quench front as it advances in the core while the rods cool and quench. The FHMG software package was developed and implemented into the COBRA-TF computer code. This paper presents the model and discusses preliminary results obtained with the COBRA-TF/FHMG computer code

Component Cooling Heat Exchanger Heat Transfer Capability Operability Monitoring

International Nuclear Information System (INIS)

Mihalina, M.; Djetelic, N.

2010-01-01

The ultimate heat sink (UHS) is of highest importance for nuclear power plant safe and reliable operation. The most important component in line from safety-related heat sources to the ultimate heat sink water body is a component cooling heat exchanger (CC Heat Exchanger). The Component Cooling Heat Exchanger has a safety-related function to transfer the heat from the Component Cooling (CC) water system to the Service Water (SW) system. SW systems throughout the world have been the root of many plant problems because the water source, usually river, lake, sea or cooling pond, are conductive to corrosion, erosion, biofouling, debris intrusion, silt, sediment deposits, etc. At Krsko NPP, these problems usually cumulate in the summer period from July to August, with higher Sava River (service water system) temperatures. Therefore it was necessary to continuously evaluate the CC Heat Exchanger operation and confirm that the system would perform its intended function in accordance with the plant's design basis, given as a minimum heat transfer rate in the heat exchanger design specification sheet. The Essential Service Water system at Krsko NPP is an open cycle cooling system which transfers heat from safety and non-safety-related systems and components to the ultimate heat sink the Sava River. The system is continuously in operation in all modes of plant operation, including plant shutdown and refueling. However, due to the Sava River impurities and our limited abilities of the water treatment, the system is subject to fouling, sedimentation buildup, corrosion and scale formation, which could negatively impact its performance being unable to satisfy its safety related post accident heat removal function. Low temperature difference and high fluid flows make it difficult to evaluate the CC Heat Exchanger due to its specific design. The important effects noted are measurement uncertainties, nonspecific construction, high heat transfer capacity, and operational specifics (e

Calculation and validation of heat transfer coefficient for warm forming operations

Science.gov (United States)

Omer, Kaab; Butcher, Clifford; Worswick, Michael

2017-10-01

In an effort to reduce the weight of their products, the automotive industry is exploring various hot forming and warm forming technologies. One critical aspect in these technologies is understanding and quantifying the heat transfer between the blank and the tooling. The purpose of the current study is twofold. First, an experimental procedure to obtain the heat transfer coefficient (HTC) as a function of pressure for the purposes of a metal forming simulation is devised. The experimental approach was used in conjunction with finite element models to obtain HTC values as a function of die pressure. The materials that were characterized were AA5182-O and AA7075-T6. Both the heating operation and warm forming deep draw were modelled using the LS-DYNA commercial finite element code. Temperature-time measurements were obtained from both applications. The results of the finite element model showed that the experimentally derived HTC values were able to predict the temperature-time history to within a 2% of the measured response. It is intended that the HTC values presented herein can be used in warm forming models in order to accurately capture the heat transfer characteristics of the operation.

Estimation of heat transfer and heat source in a molten pool

Energy Technology Data Exchange (ETDEWEB)

Yun, J.I.; Suh, K.Y.; Kang, C.S. [Seoul National Univ., Dept. of Nuclear Engineering (Korea, Republic of)

2001-07-01

Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the pool. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool was estimated by product of the mass concentration and energy conversion factor of each fission product. For the calculation of heat generation rate in the pool, twenty-nine (29) elements were chosen and classified by their chemical properties. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis was performed for the TMI-2 accident. The pool is assumed to be a partially filled hemispherical geometry, 1.45 m in radius and 32,700 kg in mass. The change of pool geometry during the numerical calculation was neglected. The peak temperature sizably decreased by about 60 K as the fission products were released from the pool. (author)

Estimation of heat transfer and heat source in a molten pool

International Nuclear Information System (INIS)

Yun, J.I.; Suh, K.Y.; Kang, C.S.

2001-01-01

Heat transfer and fluid flow in a molten pool are influenced by internal volumetric heat generated from the radioactive decay of fission product species retained in the pool. The pool superheat is determined based on the overall energy balance that equates the heat production rate to the heat loss rate. Decay heat of fission products in the pool was estimated by product of the mass concentration and energy conversion factor of each fission product. For the calculation of heat generation rate in the pool, twenty-nine (29) elements were chosen and classified by their chemical properties. The mass concentration of a fission product is obtained from released fraction and the tabular output of the ORIGEN 2 code. The initial core and pool inventories at each time can also be estimated using ORIGEN 2. The released fraction of each fission product is calculated based on the bubble dynamics and mass transport. Numerical analysis was performed for the TMI-2 accident. The pool is assumed to be a partially filled hemispherical geometry, 1.45 m in radius and 32,700 kg in mass. The change of pool geometry during the numerical calculation was neglected. The peak temperature sizably decreased by about 60 K as the fission products were released from the pool. (author)

Computer aided heat transfer analysis in a laboratory scaled heat exchanger unit

International Nuclear Information System (INIS)

Gunes, M.

1998-01-01

In this study. an explanation of a laboratory scaled heat exchanger unit and a software which is developed to analyze heat transfer. especially to use it in heat transfer courses, are represented. Analyses carried out in the software through sample values measured in the heat exchanger are: (l) Determination of heat transfer rate, logarithmic mean temperature difference and overall heat transfer coefficient; (2)Determination of convection heat transfer coefficient inside and outside the tube and the effect of fluid velocity on these; (3)Investigation of the relationship between Nusselt Number. Reynolds Number and Prandtl Number by using multiple non-linear regression analysis. Results are displayed on the screen graphically

Heat transfer coefficient as parameter describing ability of insulating liquid to heat transfer

Science.gov (United States)

Nadolny, Zbigniew; Gościński, Przemysław; Bródka, Bolesław

2017-10-01

The paper presents the results of the measurements of heat transfer coefficient of insulating liquids used in transformers. The coefficient describes an ability of the liquid to heat transport. On the basis of the coefficient, effectiveness of cooling system of electric power devices can be estimated. Following liquids were used for the measurements: mineral oil, synthetic ester and natural ester. It was assumed that surface heat load is about 2500 W·m-2, which is equal the load of transformer windings. A height of heat element was 1.6 m, because it makes possible steady distribution of temperature on its surface. The measurements of heat transfer coefficient was made as a function of various position of heat element (vertical, horizontal). In frame of horizontal position of heat element, three suppositions were analysed: top, bottom, and side.

Heat transfer coefficient as parameter describing ability of insulating liquid to heat transfer

Directory of Open Access Journals (Sweden)

Nadolny Zbigniew

2017-01-01

Full Text Available The paper presents the results of the measurements of heat transfer coefficient of insulating liquids used in transformers. The coefficient describes an ability of the liquid to heat transport. On the basis of the coefficient, effectiveness of cooling system of electric power devices can be estimated. Following liquids were used for the measurements: mineral oil, synthetic ester and natural ester. It was assumed that surface heat load is about 2500 W·m-2, which is equal the load of transformer windings. A height of heat element was 1.6 m, because it makes possible steady distribution of temperature on its surface. The measurements of heat transfer coefficient was made as a function of various position of heat element (vertical, horizontal. In frame of horizontal position of heat element, three suppositions were analysed: top, bottom, and side.

A meshless method for modeling convective heat transfer

Energy Technology Data Exchange (ETDEWEB)

Carrington, David B [Los Alamos National Laboratory

2010-01-01

A meshless method is used in a projection-based approach to solve the primitive equations for fluid flow with heat transfer. The method is easy to implement in a MATLAB format. Radial basis functions are used to solve two benchmark test cases: natural convection in a square enclosure and flow with forced convection over a backward facing step. The results are compared with two popular and widely used commercial codes: COMSOL, a finite element model, and FLUENT, a finite volume-based model.

Heat transfer pipe shielding device for heat exchanger

International Nuclear Information System (INIS)

Hanawa, Jun.

1991-01-01

The front face and the rear face of a frame that surrounds the circumference of the water chamber body of a multi-tube heat exchanger are covered by a rotational shielding plate. A slit is radially formed to the shielding plate for the insertion of a probe or cleaner to the heat transfer pipe and a deflector is disposed on the side opposite to the slit. The end of the heat transfer pipe to be inspected is exposed to the outer side by way of the slit by the rotation of the shielding plate, and the probe or cleaner is inserted in the heat transfer pipe to conduct an eddy current injury monitoring test or cleaning. The inside of the water chamber and the heat transfer pipe is exhausted by a ventilation nozzle disposed to the frame. Accordingly, a shielding effect upon inspection and cleaning can be obtained and, in addition, inspection and exhaustion at the cleaning position can be conducted easily. Since the operation for attachment and detachment is easy, the effect of reducing radiation dose per unit can be obtained by the shortening of the operation time. (N.H.)

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

Heat transfer from internally-heated molten UO2 pools

International Nuclear Information System (INIS)

Stein, R.P.; Baker, L. Jr.; Gunther, W.H.; Cook, C.

1978-01-01

Experimental measurements of heat transfer from internally heated pools of molten UO 2 have been obtained for two cell sizes: 10 cm x 10 cm and 20 cm x 20 cm. The experiments with the large cell have supported a previous conclusion from early small data that the measured downward heat fluxes are higher than would be expected on the basis of considerations of thermal convection. A convective model underpredicts the downward heat fluxes by a factor of 2.5 to 4.5 for all but one early experiment. Arbitrary assumptions of increased thermal conductivity do not account for the discrepancy. A single model based on internal thermal radiation heat transfer is able to account for the high values. The model uses the optically thick Rosseland approximation. Because of this, it is tentatively concluded that thermal radiation plays a dominant role in controlling the heat transfer from internally heated molted fuel

Theory of Periodic Conjugate Heat Transfer

CERN Document Server

Zudin, Yuri B

2012-01-01

This book presents the theory of periodic conjugate heat transfer in a detailed way. The effects of thermophysical properties and geometry of a solid body on the commonly used and experimentally determined heat transfer coefficient are analytically presented from a general point of view. The main objective of the book is a simplified description of the interaction between a solid body and a fluid as a boundary value problem of the heat conduction equation for the solid body. At the body surface, the true heat transfer coefficient is composed of two parts: the true mean value resulting from the solution of the steady state heat transfer problem and a periodically variable part, the periodic time and length to describe the oscillatory hydrodynamic effects. The second edition is extended by (i) the analysis of stability boundaries in helium flow at supercritical conditions in a heated channel with respect to the interaction between a solid body and a fluid; (ii) a periodic model and a method of heat transfer sim...

Modeling of heat transfer into a heat pipe for a localized heat input zone

International Nuclear Information System (INIS)

Rosenfeld, J.H.

1987-01-01

A general model is presented for heat transfer into a heat pipe using a localized heat input. Conduction in the wall of the heat pipe and boiling in the interior structure are treated simultaneously. The model is derived from circumferential heat transfer in a cylindrical heat pipe evaporator and for radial heat transfer in a circular disk with boiling from the interior surface. A comparison is made with data for a localized heat input zone. Agreement between the theory and the model is good. This model can be used for design purposes if a boiling correlation is available. The model can be extended to provide improved predictions of heat pipe performance

Heat exchanger device and method for heat removal or transfer

Science.gov (United States)

Koplow, Jeffrey P

2013-12-10

Systems and methods for a forced-convection heat exchanger are provided. In one embodiment, heat is transferred to or from a thermal load in thermal contact with a heat conducting structure, across a narrow air gap, to a rotating heat transfer structure immersed in a surrounding medium such as air.

Visualisation of heat transfer in laminar flows

NARCIS (Netherlands)

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

A heat transfer textbook

CERN Document Server

Lienhard, John H

2011-01-01

This introduction to heat transfer offers advanced undergraduate and graduate engineering students a solid foundation in the subjects of conduction, convection, radiation, and phase-change, in addition to the related topic of mass transfer. A staple of engineering courses around the world for more than three decades, it has been revised and updated regularly by the authors, a pair of recognized experts in the field. The text addresses the implications, limitations, and meanings of many aspects of heat transfer, connecting the subject to its real-world applications and developing students' ins

Elementary heat transfer analysis

CERN Document Server

Whitaker, Stephen; Hartnett, James P

1976-01-01

Elementary Heat Transfer Analysis provides information pertinent to the fundamental aspects of the nature of transient heat conduction. This book presents a thorough understanding of the thermal energy equation and its application to boundary layer flows and confined and unconfined turbulent flows. Organized into nine chapters, this book begins with an overview of the use of heat transfer coefficients in formulating the flux condition at phase interface. This text then explains the specification as well as application of flux boundary conditions. Other chapters consider a derivation of the tra

Two Dimensional CFD Analyses on the Heat Transfer for a Supercritical Pressure CO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Cho, Bong Hyun; Kim, Young In; Bae, Yoon Yeong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2005-07-01

The Supercritical Water Cooled Reactor(SCWR) operates in a pressure around 25MPa and temperature of 293{approx}510 .deg. C. In order to study the heat transfer behaviors and good comparisons between the various fluids, a heat transfer test loop(SPHINX) using CO{sub 2} has been constructed in KAERI as a part of international research program, I-NERI. At a supercritical pressure, the heat transfer coefficient is much larger than that estimated from the Dittus-Boelter correlation for a relatively large flow rate with moderate wall heat flux conditions. This phenomenon was explained by the rapid variations of the physical properties near the wall with the temperature. On the contrary, the heat transfer becomes worse when the bulk fluid enthalpy is below the pseudo-critical enthalpy under a low flow rate with large heat flux conditions. This phenomenon is called 'deteriorated heat transfer', and which is explained as the modification of the shear stress distribution across the tube to a buoyancy and/or acceleration in a low density layer near the wall, with the consequence of a turbulence. The upward vertical flow of CO{sub 2} through a uniformly heated tube of 4.4 mm in diameter and 3m long(heated length is 2.1m) was investigated numerically using the CFD code, FLUENT. Through the numerical simulations, we have attempted to obtain a physically meaningful insight into the heat transfer mechanisms at a supercritical pressure.

Research on heat and mass transfer model for passive containment cooling system

International Nuclear Information System (INIS)

Jiang Xiaowei; Yu Hongxing; Sun Yufa; Huang Daishun

2013-01-01

Different with the traditional dry style containment design without external cooling, the PCCS design increased the temperature difference between the wall and the containment atmosphere significantly, and also the absolute temperature of the containment surfaces will be lower, affecting properties relevant in the condensation process. A research on the heat and mass transfer model has been done in this paper, especially the improvement on the condensation and evaporation model in the presence of noncondensable gases. Firstly, the Peterson's diffusion layer model was proved to equivalent to the stagnant film model adopted by CONTAIN code using the Clausius-Clapeyron equation, then a factor which can be used to stagnant film model was derived from the comparison between the Y.Liao's generalized diffusion layer model and the Peterson's diffusion layer model. Finally, the model in CONTAIN code used to compute the condensation and evaporation mass flux was modified using the factor, and the Wisconsin condensation tests and Westinghouse film evaporation on heated plate tests were simulated which had proved the improved model can predict more closer value of the heat and mass transfer coefficient to experimental value than original model. (authors)

Modeling of the heat transfer in bypass transitional boundary-layer flows

Science.gov (United States)

Simon, Frederick F.; Stephens, Craig A.

1991-01-01

A low Reynolds number k-epsilon turbulence model and conditioned momentum, energy and turbulence equations were used to predict bypass transition heat transfer on a flat plate in a high-disturbance environment with zero pressure gradient. The use of conditioned equations was demonstrated to be an improvement over the use of the global-time-averaged equations for the calculation of velocity profiles and turbulence intensity profiles in the transition region of a boundary layer. The approach of conditioned equations is extended to include heat transfer and a modeling of transition events is used to predict transition onset and the extent of transition on a flat plate. The events, which describe the boundary layer at the leading edge, result in boundary-layer regions consisting of: (1) the laminar, (2) pseudolaminar, (3) transitional, and (4) turbulent boundary layers. The modeled transition events were incorporated into the TEXSTAN 2-D boundary-layer code which is used to numerically predict the heat transfer. The numerical predictions in general compared well with the experimental data and revealed areas where additional experimental information is needed.

Heat and mass transfer

CERN Document Server

Karwa, Rajendra

2017-01-01

This textbook presents the classical treatment of the problems of heat transfer in an exhaustive manner with due emphasis on understanding of the physics of the problems. This emphasis is especially visible in the chapters on convective heat transfer. Emphasis is laid on the solution of steady and unsteady two-dimensional heat conduction problems. Another special feature of the book is a chapter on introduction to design of heat exchangers and their illustrative design problems. A simple and understandable treatment of gaseous radiation has been presented. A special chapter on flat plate solar air heater has been incorporated that covers thermo-hydraulic modeling and simulation. The chapter on mass transfer has been written looking specifically at the needs of the students of mechanical engineering. The book includes a large number and variety of solved problems with supporting line diagrams. The author has avoided duplicating similar problems, while incorporating more application-based examples. All the end-...

Heat Transfer Basics and Practice

CERN Document Server

Böckh, Peter

2012-01-01

The book provides an easy way to understand the fundamentals of heat transfer. The reader will acquire the ability to design and analyze heat exchangers. Without extensive derivation of the fundamentals, the latest correlations for heat transfer coefficients and their application are discussed. The following topics are presented - Steady state and transient heat conduction - Free and forced convection - Finned surfaces - Condensation and boiling - Radiation - Heat exchanger design - Problem-solving After introducing the basic terminology, the reader is made familiar with the different mechanisms of heat transfer. Their practical application is demonstrated in examples, which are available in the Internet as MathCad files for further use. Tables of material properties and formulas for their use in programs are included in the appendix. This book will serve as a valuable resource for both students and engineers in the industry. The author’s experience indicates that students, after 40 lectures and exercises ...

Characterizing convective heat transfer using infrared thermography and the heated-thin-foil technique

International Nuclear Information System (INIS)

Stafford, Jason; Walsh, Ed; Egan, Vanessa

2009-01-01

Convective heat transfer, due to axial flow fans impinging air onto a heated flat plate, is investigated with infrared thermography to assess the heated-thin-foil technique commonly used to quantify two-dimensional heat transfer performance. Flow conditions generating complex thermal profiles have been considered in the analysis to account for dominant sources of error in the technique. Uncertainties were obtained in the measured variables and the influences on the resultant heat transfer data are outlined. Correction methods to accurately account for secondary heat transfer mechanisms were developed and results show that as convective heat transfer coefficients and length scales decrease, the importance of accounting for errors increases. Combined with flow patterns that produce large temperature gradients, the influence of heat flow within the foil on the resultant heat transfer becomes significant. Substantial errors in the heat transfer coefficient are apparent by neglecting corrections to the measured data for the cases examined. Methods to account for these errors are presented here, and demonstrated to result in an accurate measurement of the local heat transfer map on the surface

Heat transfer characteristics of a helical heat exchanger

International Nuclear Information System (INIS)

San, Jung-Yang; Hsu, Chih-Hsiang; Chen, Shih-Hao

2012-01-01

Heat transfer performance of a helical heat exchanger was investigated. The heat exchanger is composed of a helical tube with rectangular cross section and two cover plates. The ε–Ntu relation of the heat exchanger was obtained using a numerical method. In the analysis, the flow in the tube (helical flow) was considered to be mixed and the flow outside the tube (radial flow) was unmixed. In the experiment, the Darcy friction factor (f) and convective heat transfer coefficient (h) of the radial flow were measured. The radial flow was air and the helical flow was water. Four different channel spacing (0.5, 0.8, 1.2 and 1.6 mm) were individually considered. The Reynolds numbers were in the range 307–2547. Two correlations, one for the Darcy friction factor and the other for the Nusselt number, were proposed. - Highlights: ► We analyze the heat transfer characteristics of a helical heat exchanger and examine the effectiveness–Ntu relation. ► Increasing number of turns of the heat exchanger would slightly increase the effectiveness. ► There is an optimum Ntu value corresponding to a maximum effectiveness. ► We measure the Darcy friction factor and Nusselt number of the radial flow and examine the correlations.

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Heat Transfer in a Thermoacoustic Process

Science.gov (United States)

Beke, Tamas

2012-01-01

Thermoacoustic instability is defined as the excitation of acoustic modes in chambers with heat sources due to the coupling between acoustic perturbations and unsteady heat addition. The major objective of this paper is to achieve accurate theoretical results in a thermoacoustic heat transfer process. We carry out a detailed heat transfer analysis…

Analysis of panthers full-scale heat transfer tests with RELAP5

International Nuclear Information System (INIS)

Parlatan, Y.; Boyer, B.D.; Jo, J.; Rohatgi, S.

1996-01-01

The RELAP5 code is being assessed on the full-scale Passive Containment Cooling System (PCCS) in the Performance ANalysis and Testing of HEat Removal Systems (PANTHERS) facility at Societa Informazioni Termoidrauliche (SIET) in Italy. PANTHERS is a test facility with fall-size prototype beat exchangers for the PCCS in support of the General Electric's (GE) Simplified Boiling Water Reactor (SBWR) program. PANTHERS tests with a low noncondensable gas concentration and with a high noncondensable gas concentration were analyzed with RELAP5. The results showed that beat transfer rate decreases significantly along the PCCS tubes. In the test case with a higher inlet noncondensable gas fraction, the PCCS removed 35% less heat than in the test case with the lower noncondensable gas fraction. The dominant resistance to the overall heat transfer is the condensation beat transfer resistance inside the tubes. This resistance increased by about 5-fold between the inlet and exit of the tube due to the build up of noncondensable gases along the tube. The RELAP5 calculations also predicted that 4% to 5% of the heat removed to the PCCS pool occurs in the inlet steam piping and PCCS upper and lower headers. These piping needs to be modeled for other tests systems. The full-scale PANTHERS predictions are also compared against 1/400 scale GIRAFFE tests. GIRAFFE has 33% larger heat surface area, but its efficiency is only 15% and 23% higher than PANTHERS for the two cases analyzed This was explained by the high heat transfer resistance inside the tubes near the exit

"Nanotechnology Enabled Advanced Industrial Heat Transfer Fluids"

Energy Technology Data Exchange (ETDEWEB)

Dr. Ganesh Skandan; Dr. Amit Singhal; Mr. Kenneth Eberts; Mr. Damian Sobrevilla; Prof. Jerry Shan; Stephen Tse; Toby Rossmann

2008-06-12

ABSTRACT Nanotechnology Enabled Advanced industrial Heat Transfer Fluids” Improving the efficiency of Industrial Heat Exchangers offers a great opportunity to improve overall process efficiencies in diverse industries such as pharmaceutical, materials manufacturing and food processing. The higher efficiencies can come in part from improved heat transfer during both cooling and heating of the material being processed. Additionally, there is great interest in enhancing the performance and reducing the weight of heat exchangers used in automotives in order to increase fuel efficiency. The goal of the Phase I program was to develop nanoparticle containing heat transfer fluids (e.g., antifreeze, water, silicone and hydrocarbon-based oils) that are used in transportation and in the chemical industry for heating, cooling and recovering waste heat. Much work has been done to date at investigating the potential use of nanoparticle-enhanced thermal fluids to improve heat transfer in heat exchangers. In most cases the effect in a commercial heat transfer fluid has been marginal at best. In the Phase I work, we demonstrated that the thermal conductivity, and hence heat transfer, of a fluid containing nanoparticles can be dramatically increased when subjected to an external influence. The increase in thermal conductivity was significantly larger than what is predicted by commonly used thermal models for two-phase materials. Additionally, the surface of the nanoparticles was engineered so as to have a minimal influence on the viscosity of the fluid. As a result, a nanoparticle-laden fluid was successfully developed that can lead to enhanced heat transfer in both industrial and automotive heat exchangers

Experimental and computational investigations of heat and mass transfer of intensifier grids

International Nuclear Information System (INIS)

Kobzar, Leonid; Oleksyuk, Dmitry; Semchenkov, Yuriy

2015-01-01

The paper discusses experimental and numerical investigations on intensification of thermal and mass exchange which were performed by National Research Centre ''Kurchatov Institute'' over the past years. Recently, many designs of heat mass transfer intensifier grids have been proposed. NRC ''Kurchatov Institute'' has accomplished a large scope of experimental investigations to study efficiency of intensifier grids of various types. The outcomes of experimental investigations can be used in verification of computational models and codes. On the basis of experimental data, we derived correlations to calculate coolant mixing and critical heat flux mixing in rod bundles equipped with intensifier grids. The acquired correlations were integrated in subchannel code SC-INT.

A porous media calculation for the isolation condenser heat transfer and circulation

International Nuclear Information System (INIS)

Jaakko, Miettinen; Ismo, Karppinen

2003-01-01

In the development of advanced light water reactors, thermohydraulic phenomena are versatile in comparison with the present concepts. The new features include, for example, passive safety systems, where energy transport takes place by natural circulation instead of forced flow. In the isolation condenser, the steam generated in the reactor vessel is conduced into the heat transfer tubes. The tube bundle has been submerged into a large water pool, where the heat flux through the tube wall initially is heating the subcooled water, but rather soon boiling take place. The temperature differences and void fraction in the pool create large two-phase circulation. For modeling of the entire condenser a combined application of two types of simulation models has been selected. For the whole geometry, a porous media solution has been developed, where the existence of the heat transfer tubes in the water pool and their heat generation is described by the porous media approach. The 3-dimensional solution of two-phase equations is based on the drift-flux formalism. The condensation and liquid film generation inside the heat transfer tube is modelled using a 1-dimensional model considering the steam core, liquid film and heat transfer tube. The heat flux through the tube wall defines the boundary conditions for the water pool. Parallel to the porous media development for the entire process facility, the phase change models have been improved for Fluent 6 mixture model, and the code is used for analysing in detail the heat transfer around the tubing. The purpose in the analyses is to obtain more detailed information of the flow field and vapour distribution around the tube bundle. By combining the porous media model for the entire facility, with the CFD models for the two-phase flow details around the heat transfer tubes and experimental studies the most important mechanisms around the condensation pool can be gathered. In this context the porous media model is considered. (author)

Magnetohydrodynamic flow and heat transfer around a heated cylinder of arbitrary conductivity

Science.gov (United States)

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.

Heat transfer entropy resistance for the analyses of two-stream heat exchangers and two-stream heat exchanger networks

International Nuclear Information System (INIS)

Cheng, XueTao; Liang, XinGang

2013-01-01

The entropy generation minimization method is often used to analyze heat transfer processes from the thermodynamic viewpoint. In this paper, we analyze common heat transfer processes with the concept of entropy generation, and propose the concept of heat transfer entropy resistance. It is found that smaller heat transfer entropy resistance leads to smaller equivalent thermodynamic force difference with prescribed heat transfer rate and larger heat transfer rate with prescribed equivalent thermodynamic force difference. With the concept of heat transfer entropy resistance, the performance of two-stream heat exchangers (THEs) and two-stream heat exchanger networks (THENs) is analyzed. For the cases discussed in this paper, it is found that smaller heat transfer entropy resistance always leads to better heat transfer performance for THEs and THENs, while smaller values of the entropy generation, entropy generation numbers and revised entropy generation number do not always. -- Highlights: • The concept of entropy resistance is defined. • The minimum entropy resistance principle is developed. • Smaller entropy resistance leads to better heat transfer

SCDAP/RELAP5 modeling of fluid heat transfer and flow losses through porous debris in a light water reactor

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

Heat transfer coeffcient for boiling carbon dioxide

DEFF Research Database (Denmark)

Knudsen, Hans Jørgen Høgaard; Jensen, Per Henrik

1997-01-01

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

Creep collapse of thick-walled heat transfer tube subjected to external pressure at high temperature

International Nuclear Information System (INIS)

Ioka, Ikuo; Kaji, Yoshiyuki; Terunuma, Isao; Nekoya, Shin-ichi; Miyamoto, Yoshiaki

1994-09-01

A series of creep collapse tests of thick-walled heat transfer tube were examined experimentally and analytically to confirm an analytical method for creep deformation behavior of a heat transfer tube of an intermediate heat exchanger (IHX) at a depressurization accident of secondary cooling system of HTTR (High Temperature Engineering Test Reactor). The tests were carried out using thick-walled heat transfer tubes made of Hastelloy XR at 950degC in helium gas environment. The predictions of creep collapse time obtained by a general purpose FEM-code ABAQUS were in good agreement with the experimental results. A lot of cracks were observed on the outer surface of the test tubes after the creep collapse. However, the cracks did not pass through the tube wall and, therefore, the leak tightness was maintained regardless of a collapse deformation for all tubes tested. (author)

Supervised Transfer Sparse Coding

KAUST Repository

Al-Shedivat, Maruan

2014-07-27

A combination of the sparse coding and transfer learn- ing techniques was shown to be accurate and robust in classification tasks where training and testing objects have a shared feature space but are sampled from differ- ent underlying distributions, i.e., belong to different do- mains. The key assumption in such case is that in spite of the domain disparity, samples from different domains share some common hidden factors. Previous methods often assumed that all the objects in the target domain are unlabeled, and thus the training set solely comprised objects from the source domain. However, in real world applications, the target domain often has some labeled objects, or one can always manually label a small num- ber of them. In this paper, we explore such possibil- ity and show how a small number of labeled data in the target domain can significantly leverage classifica- tion accuracy of the state-of-the-art transfer sparse cod- ing methods. We further propose a unified framework named supervised transfer sparse coding (STSC) which simultaneously optimizes sparse representation, domain transfer and classification. Experimental results on three applications demonstrate that a little manual labeling and then learning the model in a supervised fashion can significantly improve classification accuracy.

Estimation of Surface Temperature and Heat Flux by Inverse Heat Transfer Methods Using Internal Temperatures Measured While Radiantly Heating a Carbon/Carbon Specimen up to 1920 F

Science.gov (United States)

Pizzo, Michelle; Daryabeigi, Kamran; Glass, David

2015-01-01

The ability to solve the heat conduction equation is needed when designing materials to be used on vehicles exposed to extremely high temperatures; e.g. vehicles used for atmospheric entry or hypersonic flight. When using test and flight data, computational methods such as finite difference schemes may be used to solve for both the direct heat conduction problem, i.e., solving between internal temperature measurements, and the inverse heat conduction problem, i.e., using the direct solution to march forward in space to the surface of the material to estimate both surface temperature and heat flux. The completed research first discusses the methods used in developing a computational code to solve both the direct and inverse heat transfer problems using one dimensional, centered, implicit finite volume schemes and one dimensional, centered, explicit space marching techniques. The developed code assumed the boundary conditions to be specified time varying temperatures and also considered temperature dependent thermal properties. The completed research then discusses the results of analyzing temperature data measured while radiantly heating a carbon/carbon specimen up to 1920 F. The temperature was measured using thermocouple (TC) plugs (small carbon/carbon material specimens) with four embedded TC plugs inserted into the larger carbon/carbon specimen. The purpose of analyzing the test data was to estimate the surface heat flux and temperature values from the internal temperature measurements using direct and inverse heat transfer methods, thus aiding in the thermal and structural design and analysis of high temperature vehicles.

Conjugated heat transfer and temperature distributions in a gas turbine combustion liner under base-load operation

International Nuclear Information System (INIS)

Kim, Kyung Min; Yun, Nam Geon; Jeon, Yun Heung; Lee, Dong Hyun; Cho, Yung Hee

2010-01-01

Prediction of temperature distributions on hot components is important in development of a gas turbine combustion liner. The present study investigated conjugated heat transfer to obtain temperature distributions in a combustion liner with six combustion nozzles. 3D numerical simulations using FVM commercial codes, Fluent and CFX were performed to calculate combustion and heat transfer distributions. The temperature distributions in the combustor liner were calculated by conjugation of conduction and convection (heat transfer coefficients) obtained by combustion and cooling flow analysis. The wall temperature was the highest on the attachment points of the combustion gas from combustion nozzles, but the temperature gradient was high at the after shell section with low wall temperature

Heat transfer and critical heat flux in a asymmetrically heated tube helicoidal flow

International Nuclear Information System (INIS)

Boscary, J.

1995-10-01

The design of plasma facing components is crucial for plasma performance in next fusion reactors. These elements will be submitted to very high heat flux. They will be actively water-cooled by swirl tubes in the subcooled boiling regime. High heat flux experiments were conducted in order to analyse the heat transfer and to evaluate the critical heat flux. Water-cooled mock-ups were one-side heated by an electron beam gun for different thermal-hydraulic conditions. The critical heat flux was detected by an original method based on the isotherm modification on the heated surface. The wall heat transfer law including forced convection and subcooled boiling regimes was established. Numerical calculations of the material heat transfer conduction allowed the non-homogeneous distribution of the wall temperature and of the wall heat flux to be evaluated. The critical heat flux value was defined as the wall maximum heat flux. A critical heat flux model based on the liquid sublayer dryout under a vapor blanket was established. A good agreement with test results was found. (author). 198 refs., 126 figs., 21 tabs

Supercritical heat transfer in an annular channel with external heating

International Nuclear Information System (INIS)

Remizov, O.V.; Gal'chenko, Eh.F.; Shurkin, N.G.; Sergeev, V.V.

1980-01-01

Results are presented of experimental studies of the burnout heat transfer in a 32x28x3000 mm annular channel with a uniform distribution of a heat flow at pressures of 6.9-19.6 MPa and mass rates of 350-1000 kg/m 2 xs. The heating is electrical, external, one-sided. It is shown that dependencies of the heat-transfer coefficient on rated parameters in the annular channel and tube are similar. An empirical equation has been obtained for the calculation of the burnout heat transfer in the annual channels with external heating in the following range: pressure, 6.9 -13.7 MPa; mass rate 350-700 kg/m 2 xs, and steam content ranging from Xsub(crit) to 1

CFD investigation of flow and heat transfer of nanofluids in isoflux spirally fluted tubes

KAUST Repository

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

Heat and mass transfer in particulate suspensions

CERN Document Server

Michaelides, Efstathios E (Stathis)

2013-01-01

Heat and Mass Transfer in Particulate Suspensions is a critical review of the subject of heat and mass transfer related to particulate Suspensions, which include both fluid-particles and fluid-droplet Suspensions. Fundamentals, recent advances and industrial applications are examined. The subject of particulate heat and mass transfer is currently driven by two significant applications: energy transformations –primarily combustion – and heat transfer equipment. The first includes particle and droplet combustion processes in engineering Suspensions as diverse as the Fluidized Bed Reactors (FBR’s) and Internal Combustion Engines (ICE’s). On the heat transfer side, cooling with nanofluids, which include nanoparticles, has attracted a great deal of attention in the last decade both from the fundamental and the applied side and has produced several scientific publications. A monograph that combines the fundamentals of heat transfer with particulates as well as the modern applications of the subject would be...

Prediction of transpiration effects on heat and mass transfer by different turbulence models

International Nuclear Information System (INIS)

Bucci, M.; Sharabi, M.; Ambrosini, W.; Forgione, N.; Oriolo, F.; He, S.

2008-01-01

The paper reports the results of a study related to transpirating flows, stimulated by the interest that these phenomena, occurring in the presence of simultaneous heat and mass transfer, have for nuclear reactor applications. The work includes a summary and the follow-up of previous experimental and numerical investigations on filmwise condensation and falling film evaporation and of a recent review of different forms of the heat and mass transfer analogy. The particular objective here pursued is to compare transpiration effects as predicted by different turbulence models with classical suction and blowing multipliers based on stagnant layer theories, in the attempt to clarify their quantitative implications on the predicted mass transfer rates. A commercial and an in-house CFD code have been adopted for evaluating the heat and mass transfer rates occurring over a flat plate exposed to an air-vapour stream, with uniform bulk steam mass fraction and temperature boundary conditions at the wall. This simple configuration was purposely selected since it is a simplified representation of the test section of an experimental facility presently in operation at the University of Pisa. This allows a direct comparison between the heat and mass transfer coefficients predicted by CFD models and classical correlations for Nusselt and Sherwood numbers

Thermal radiation heat transfer

CERN Document Server

Howell, John R; Mengüç, M Pinar

2011-01-01

Providing a comprehensive overview of the radiative behavior and properties of materials, the fifth edition of this classic textbook describes the physics of radiative heat transfer, development of relevant analysis methods, and associated mathematical and numerical techniques. Retaining the salient features and fundamental coverage that have made it popular, Thermal Radiation Heat Transfer, Fifth Edition has been carefully streamlined to omit superfluous material, yet enhanced to update information with extensive references. Includes four new chapters on Inverse Methods, Electromagnetic Theory, Scattering and Absorption by Particles, and Near-Field Radiative Transfer Keeping pace with significant developments, this book begins by addressing the radiative properties of blackbody and opaque materials, and how they are predicted using electromagnetic theory and obtained through measurements. It discusses radiative exchange in enclosures without any radiating medium between the surfaces-and where heat conduction...

Modification and validation of the natural heat convection and subcooled void formation models in the code PARET

International Nuclear Information System (INIS)

Hainoun, A.; Alhabit, F.; Ghazi, N.

2008-01-01

Two new modifications have been included in the current PARET code that is widely applied in the dynamic and safety analysis of research reactors. A new model was implemented for the simulation of void formation in the subcooled boiling regime, the other modification dealt with the implementation of a new approach to improve the prediction of heat transfer coefficient under natural circulation condition. The modified code was successfully validated using adequate single effect tests covering the physical phenomena of interest for both natural circulation and subcooled void formation at low pressure and low heat flux. The validation results indicate significant improvement of the code compared to the default version. Additionally, to simplify the code application an interactive user interface was developed enabling pre and post-processing of the code predictions. (author)

GOTHIC-IST 6.1b code validation exercises relating to heat removal by dousing and air coolers in CANDU containment

International Nuclear Information System (INIS)

Ramachandran, S.; Krause, M.; Nguyen, T.

2003-01-01

This paper presents validation results relating to the use of the GOTHIC containment analysis code for CANDU safety analysis. The validation results indicate that GOTHIC predicts heat removal by dousing and air cooler heat transfer with reasonable accuracy. (author)

Effect of surface etching on condensing heat transfer

Energy Technology Data Exchange (ETDEWEB)

Seok, Sung Chul; Park, Jae Won; Jung, Jiyeon; Choi, Chonggun; Choi, Gyu Hong; Hwang, Seung Sik; Chung, Tae Yong; Shin, Donghoon [Kookmin University, Seoul (Korea, Republic of); Kim, Jin Jun [Hoseo University, Asan (Korea, Republic of)

2016-02-15

This study conducted experiments on humid air condensation during heat transfer in an air preheating exchanger attached to a home condensing boiler to improve thermal efficiency. An etchant composed of sulfuric acid and sodium nitrate was used to create roughness on the heat exchanger surface made from STS430J1L. A counter flow heat exchanger was fabricated to test the performance of heat transfer. Results showed that the overall heat transfer coefficients of all specimens treated with etchant improved with respect to the original specimens (not treated with etchant), and the overall heat transfer coefficient of the 60 s etching specimen increased by up to 15%. However, the increasing rate of the heat transfer coefficient was disproportional to the etching time. When the etching time specifically increased above 60 s, the heat transfer coefficient decreased. This effect was assumed to be caused by surface characteristics such as contact angle. Furthermore, a smaller contact angle or higher hydrophilicity leads to higher heat transfer coefficient.

Suppression of the sonic heat transfer limit in high-temperature heat pipes

Science.gov (United States)

Dobran, Flavio

1989-08-01

The design of high-performance heat pipes requires optimization of heat transfer surfaces and liquid and vapor flow channels to suppress the heat transfer operating limits. In the paper an analytical model of the vapor flow in high-temperature heat pipes is presented, showing that the axial heat transport capacity limited by the sonic heat transfer limit depends on the working fluid, vapor flow area, manner of liquid evaporation into the vapor core of the evaporator, and lengths of the evaporator and adiabatic regions. Limited comparisons of the model predictions with data of the sonic heat transfer limits are shown to be very reasonable, giving credibility to the proposed analytical approach to determine the effect of various parameters on the axial heat transport capacity. Large axial heat transfer rates can be achieved with large vapor flow cross-sectional areas, small lengths of evaporator and adiabatic regions or a vapor flow area increase in these regions, and liquid evaporation in the evaporator normal to the main flow.

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

Three dimensional conjugated heat transfer analysis in sodium fast reactor wire-wrapped fuel assembly

International Nuclear Information System (INIS)

Peniguel, C.; Rupp, I.; Juhel, JP.; Rolfo, S.; Guillaud, M.; Gervais, N.

2009-01-01

Fast reactors with liquid metal coolant have recently received a renewed interest owing to a more efficient usage of the primary uranium resources, and they are one of the proposal for the next Generation IV. In order to evaluate nuclear power plant design and safety, 3D analysis of the flow and heat transfer in a wire spacer fuel assembly are ongoing at EDF. The introduction of the wire wrapped spacers, helically wound along the pin axis, enhances the mixing of the coolant between sub-channels and prevents contact between the fuel pins. The mesh generation step constitutes a challenging task if a reasonable amount of cells in conjunction with a suitable spatial discretization is wanted. Several approaches have been investigated and will be presented. Quite complex global flow patterns are found using either k-ε or preferably Reynolds Stress turbulent models. Preliminary conjugated heat transfer calculations using a coupling between the finite element thermal code SYRTHES and the finite volume CFD code Code Saturne are also shown. (author)

Calculation code of mass and heat transfer in a pulsed column for Purex process

International Nuclear Information System (INIS)

Tsukada, Takeshi; Takahashi, Keiki

1993-01-01

A calculation code for extraction behavior analysis in a pulsed column employed at an extraction process of a reprocessing plant was developed. This code was also combined with our previously developed calculation code for axial temperature profiles in a pulsed column. The one-dimensional dispersion model was employed for both of the extraction behavior analysis and the axial temperature profile analysis. The reported values of the fluid characteristics coefficient, the transfer coefficient and the diffusivities in the pulsed column were used. The calculated concentration profiles of HNO 3 , U and Pu for the steady state have a good agreement with the reported experimental results. The concentration and temperature profiles were calculated under the operation conditions which induce the abnormal U extraction behavior, i.e. U extraction zone is moved to the bottom of the column. Thought there is slight difference between calculated and experimental value, it is appeared that our developed code can be applied to the simulation under the normal operation condition and the relatively slowly transient condition. Pu accumulation phenomena was analyzed with this code and the accumulation tendency is similar to the reported analysis results. (author)

Heat transfer enhancement using 2MHz ultrasound.

Science.gov (United States)

Bulliard-Sauret, Odin; Ferrouillat, Sebastien; Vignal, Laure; Memponteil, Alain; Gondrexon, Nicolas

2017-11-01

The present work focuses on possible heat transfer enhancement from a heating plate towards tap water in forced convection by means of 2MHz ultrasound. The thermal approach allows to observe the increase of local convective heat transfer coefficients in the presence of ultrasound and to deduce a correlation between ultrasound power and Nusselt number. Heat transfer coefficient under ultrasound remains constant while heat transfer coefficient under silent conditions increases with Reynolds number from 900 up to 5000. Therefore, heat transfer enhancement factor ranges from 25% up to 90% for the same energy conditions (supplied ultrasonic power=110W and supplied thermal power=450W). In the same time cavitational activity due to 2MHz ultrasound emission was characterized from mechanical and chemical viewpoints without significant results. At least, Particle Image Velocimetry (PIV) measurements have been performed in order to investigate hydrodynamic modifications due to the presence of 2MHz ultrasound. It was therefore possible to propose a better understanding of heat transfer enhancement mechanism with high frequency ultrasound. Copyright © 2017 Elsevier B.V. All rights reserved.

Numerical investigation of heat transfer enhancement in ribbed channel for the first wall of DFLL-TBM in ITER

International Nuclear Information System (INIS)

Jin Qiang; Liu Songlin; Li Min; Wang Weihua

2012-01-01

As an important component of Dual Functional Lithium Lead-Test Blanket Module (DFLL-TBM), the first wall (FW) must withstand and remove the heat flux from the plasma (q″ = 0.3 MW/m 2 ) and high nuclear power deposited in the structure at normal plasma operation scenario of ITER. In this paper the transverse ribs arranged along the plasma facing inner wall surface were used to enhance the heat transfer capability. After the validation compared with empirical correlations the Standard k–ω model was employed to do the numerical simulation using FLUENT code to investigate the heat transfer efficiency and flow performance of coolant in the ribbed channel preliminarily. The perforation on the bottom of rib was proposed near the lower heat transfer area (LHTA) to improve the heat transfer performance according to results of analyses.

Engineering heat transfer

International Nuclear Information System (INIS)

Welty, J.R.

1974-01-01

The basic concepts of heat transfer are covered with special emphasis on up-to-date techniques for formulating and solving problems in the field. The discussion progresses logically from phenomenology to problem solving, and treats numerical, integral, and graphical methods as well as traditional analytical ones. The book is unique in its thorough coverage of the fundamentals of numerical analysis appropriate to solving heat transfer problems. This coverage includes several complete and readable examples of numerical solutions, with discussions and interpretations of results. The book also contains an appendix that provides students with physical data for often-encountered materials. An index is included. (U.S.)

Laboratory simulation of heat exchange for liquids with Pr > 1: Heat transfer

Science.gov (United States)

Belyaev, I. A.; Zakharova, O. D.; Krasnoshchekova, T. E.; Sviridov, V. G.; Sukomel, L. A.

2016-02-01

Liquid metals are promising heat transfer agents in new-generation nuclear power plants, such as fast-neutron reactors and hybrid tokamaks—fusion neutron sources (FNSs). We have been investigating hydrodynamics and heat exchange of liquid metals for many years, trying to reproduce the conditions close to those in fast reactors and fusion neutron sources. In the latter case, the liquid metal flow takes place in a strong magnetic field and strong thermal loads resulting in development of thermogravitational convection in the flow. In this case, quite dangerous regimes of magnetohydrodynamic (MHD) heat exchange not known earlier may occur that, in combination with other long-known regimes, for example, the growth of hydraulic drag in a strong magnetic field, make the possibility of creating a reliable FNS cooling system with a liquid metal heat carrier problematic. There exists a reasonable alternative to liquid metals in FNS, molten salts, namely, the melt of lithium and beryllium fluorides (Flibe) and the melt of fluorides of alkali metals (Flinak). Molten salts, however, are poorly studied media, and their application requires detailed scientific substantiation. We analyze the modern state of the art of studies in this field. Our contribution is to answer the following question: whether above-mentioned extremely dangerous regimes of MHD heat exchange detected in liquid metals can exist in molten salts. Experiments and numerical simulation were performed in order to answer this question. The experimental test facility represents a water circuit, since water (or water with additions for increasing its electrical conduction) is a convenient medium for laboratory simulation of salt heat exchange in FNS conditions. Local heat transfer coefficients along the heated tube, three-dimensional (along the length and in the cross section, including the viscous sublayer) fields of averaged temperature and temperature pulsations are studied. The probe method for measurements in

Calculation of Post-Closure Natural Convection Heat and Mass Transfer in Yucca Mountain Drifts

International Nuclear Information System (INIS)

Webb, S.; Itamura, M.

2004-01-01

Natural convection heat and mass transfer under post-closure conditions has been calculated for Yucca Mountain drifts using the computational fluid dynamics (CFD) code FLUENT. Calculations have been performed for 300, 1000, 3000, and 10,000 years after repository closure. Effective dispersion coefficients that can be used to calculate mass transfer in the drift have been evaluated as a function of time and boundary temperature tilt

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

Research progress on microgravity boiling heat transfer

International Nuclear Information System (INIS)

Xiao Zejun; Chen Bingde

2003-01-01

Microgravity boiling heat transfer is one of the most basic research topics in aerospace technology, which is important for both scientific research and engineering application. Research progress on microgravity boiling heat transfer is presented, including terrestrial simulation technique, terrestrial simulation experiment, microgravity experiment, and flow boiling heat transfer

Experimental investigation of heat transfer performance for a novel microchannel heat sink

International Nuclear Information System (INIS)

Wang, Y; Ding, G-F

2008-01-01

We demonstrated a novel microchannel heat sink with a high local heat transfer efficiency contributed by a complicated microchannel system, which comprises parallel longitudinal microchannels etched in a silicon substrate and transverse microchannels electroplated on a copper heat spreader. The thermal boundary layer develops in transverse microchannels. Meanwhile, the heat transfer area is increased compared with the conventional microchannel heat sink only having parallel longitudinal microchannels. Both benefits yield high local heat transfer efficiency and enhance the overall heat transfer, which is attractive for the cooling of high heat flux electronic devices. Infrared tests show the temperature distribution in the test objects. The effects of flow rate and heat flux levels on heat transfer characteristics are presented. A uniform temperature distribution is obtained through the heating area. The reference temperatures decrease with the increasing flow rate from 0.64 ml min −1 to 6.79 ml min −1 for a constant heat flux of 10.4 W cm −2 . A heat flux of 18.9 W cm −2 is attained at a flow rate of 6.79 ml min −1 for assuring the maximum temperature of the microchannel heat sink less than the maximum working temperature of electronic devices

Heat transfer capability analysis of heat pipe for space reactor

International Nuclear Information System (INIS)

Li Huaqi; Jiang Xinbiao; Chen Lixin; Yang Ning; Hu Pan; Ma Tengyue; Zhang Liang

2015-01-01

To insure the safety of space reactor power system with no single point failures, the reactor heat pipes must work below its heat transfer limits, thus when some pipes fail, the reactor could still be adequately cooled by neighbor heat pipes. Methods to analyze the reactor heat pipe's heat transfer limits were presented, and that for the prevailing capillary limit analysis was improved. The calculation was made on the lithium heat pipe in core of heat pipes segmented thermoelectric module converter (HP-STMC) space reactor power system (SRPS), potassium heat pipe as radiator of HP-STMC SRPS, and sodium heat pipe in core of scalable AMTEC integrated reactor space power system (SAIRS). It is shown that the prevailing capillary limits of the reactor lithium heat pipe and sodium heat pipe is 25.21 kW and 14.69 kW, providing a design margin >19.4% and >23.6%, respectively. The sonic limit of the reactor radiator potassium heat pipe is 7.88 kW, providing a design margin >43.2%. As the result of calculation, it is concluded that the main heat transfer limit of HP-STMC SRPS lithium heat pipe and SARIS sodium heat pipe is prevailing capillary limit, but the sonic limit for HP-STMC SRPS radiator potassium heat pipe. (authors)

Homogenization of some radiative heat transfer models: application to gas-cooled reactor cores

International Nuclear Information System (INIS)

El Ganaoui, K.

2006-09-01

In the context of homogenization theory we treat some heat transfer problems involving unusual (according to the homogenization) boundary conditions. These problems are defined in a solid periodic perforated domain where two scales (macroscopic and microscopic) are to be taken into account and describe heat transfer by conduction in the solid and by radiation on the wall of each hole. Two kinds of radiation are considered: radiation in an infinite medium (non-linear problem) and radiation in cavity with grey-diffuse walls (non-linear and non-local problem). The derived homogenized models are conduction problems with an effective conductivity which depend on the considered radiation. Thus we introduce a framework (homogenization and validation) based on mathematical justification using the two-scale convergence method and numerical validation by simulations using the computer code CAST3M. This study, performed for gas cooled reactors cores, can be extended to other perforated domains involving the considered heat transfer phenomena. (author)

Transient heat transfer in liquid helium

International Nuclear Information System (INIS)

Shiotsu, Masahiro

1991-01-01

Detailed knowledge on the steady-state and transient heat transfer from solid surfaces in He I and He II is important as a database for the analysis of the influence of local thermal disturbances on the stability of He I or He II cooled large superconducting magnets. In this paper, an overview of the transient heat transfer characteristics on solid surfaces in He I and He II caused by various large stepwise heat inputs, such as the quasi-steady nucleate boiling with a certain lifetime in He I and the quasi-steady Kapitza conductance heat flux with a certain lifetime in He II, are presented in comparison with their steady-state heat transfer characteristics. (author)

Calculation study of nonequilibrium post-CHF heat transfer in rod bundle test using modified RELAP5/MOD2

International Nuclear Information System (INIS)

Hassan, Y.A.

1987-01-01

To date there is only very limited data for non-equilibrium convective film boiling in rod bundle geometries. A recent nine (3 x 3) rod bundle post-critical-flux (CHF) test from the Lehigh University test facility was simulated using RELAP5/MOD2, to assess its capabilities in predicting the overall convective mechanisms in post-CHF heat transfer in rod bundle geometries. The code calculations were compared with experimental data. The code predicted low vapor superheats and void fraction oscillations. A new interfacial heat transfer between the droplet/steam resulted in a reasonable prediction of vapor superheats. A revised dispersed flow film boiling correlation which accounts for the enhancement of steam convective cooling by droplet-induced turbulence was incorporated in the code. Comparison with the data showed a fair agreement

Conjugate heat and mass transfer in heat mass exchanger ducts

CERN Document Server

Zhang, Li-Zhi

2013-01-01

Conjugate Heat and Mass Transfer in Heat Mass Exchanger Ducts bridges the gap between fundamentals and recent discoveries, making it a valuable tool for anyone looking to expand their knowledge of heat exchangers. The first book on the market to cover conjugate heat and mass transfer in heat exchangers, author Li-Zhi Zhang goes beyond the basics to cover recent advancements in equipment for energy use and environmental control (such as heat and moisture recovery ventilators, hollow fiber membrane modules for humidification/dehumidification, membrane modules for air purification, desi

Experimental and analytical study of natural-convection heat transfer of internally heated liquids

International Nuclear Information System (INIS)

Green, G.A.

1982-08-01

Boundary heat transfer from a liquid pool with a uniform internal heat source to a vertical or inclined boundary was investigated. The experiments were performed in an open rectangular liquid pool in which the internal heat source was generated by electrical heating. The local heat flux was measured to a boron nitride test wall which was able to be continuously inclined from vertical. Gold plated microthermocouples of 0.01 inch outside diameter were developed to measure the local surface temperature, both front and back, of the boron nitride. The local heat flux and, thus, the local heat transfer coefficient was measured at nineteen locations along the vertical axis of the test plate. A theoretical analysis of the coupled nonlinear boundary layer equations was performed. The parametric effect of the Prandtl number and the dimensionless wall temperature on the boundary heat transfer were investigated When the analytical model was used to calculate the boundary heat transfer data, agreement was achieved with the experimental data within 3% for the local heat transfer and within 2% for the average heat transfer

Analysis of the heat transfer in double and triple concentric tube heat exchangers

Science.gov (United States)

Rădulescu, S.; Negoiţă, L. I.; Onuţu, I.

2016-08-01

The tubular heat exchangers (shell and tube heat exchangers and concentric tube heat exchangers) represent an important category of equipment in the petroleum refineries and are used for heating, pre-heating, cooling, condensation and evaporation purposes. The paper presents results of analysis of the heat transfer to cool a petroleum product in two types of concentric tube heat exchangers: double and triple concentric tube heat exchangers. The cooling agent is water. The triple concentric tube heat exchanger is a modified constructive version of double concentric tube heat exchanger by adding an intermediate tube. This intermediate tube improves the heat transfer by increasing the heat area per unit length. The analysis of the heat transfer is made using experimental data obtained during the tests in a double and triple concentric tube heat exchanger. The flow rates of fluids, inlet and outlet temperatures of water and petroleum product are used in determining the performance of both heat exchangers. Principally, for both apparatus are calculated the overall heat transfer coefficients and the heat exchange surfaces. The presented results shows that triple concentric tube heat exchangers provide better heat transfer efficiencies compared to the double concentric tube heat exchangers.

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

Numerical modeling of heat transfer and pasteurizing value during thermal processing of intact egg.

Science.gov (United States)

Abbasnezhad, Behzad; Hamdami, Nasser; Monteau, Jean-Yves; Vatankhah, Hamed

2016-01-01

Thermal Pasteurization of Eggs, as a widely used nutritive food, has been simulated. A three-dimensional numerical model, computational fluid dynamics codes of heat transfer equations using heat natural convection, and conduction mechanisms, based on finite element method, was developed to study the effect of air cell size and eggshell thickness. The model, confirmed by comparing experimental and numerical results, was able to predict the temperature profiles, the slowest heating zone, and the required heating time during pasteurization of intact eggs. The results showed that the air cell acted as a heat insulator. Increasing the air cell volume resulted in decreasing of the heat transfer rate, and the increasing the required time of pasteurization (up to 14%). The findings show that the effect on thermal pasteurization of the eggshell thickness was not considerable in comparison to the air cell volume.

Heat Transfer in Metal Foam Heat Exchangers at High Temperature

Science.gov (United States)

Hafeez, Pakeeza

Heat transfer though open-cell metal foam is experimentally studied for heat exchanger and heat shield applications at high temperatures (˜750°C). Nickel foam sheets with pore densities of 10 and 40 pores per linear inch (PPI), have been used to make the heat exchangers and heat shields by using thermal spray coating to deposit an Inconel skin on a foam core. Heat transfer measurements were performed on a test rig capable of generating hot gas up to 1000°C. The heat exchangers were tested by exposing their outer surface to combustion gases at a temperature of 550°C and 750°C while being cooled by air flowing through them at room temperature at velocities up to 5 m/s. The temperature rise of the air, the surface temperature of the heat exchangers and the air temperature inside the heat exchanger were measured. The volumetric heat transfer coefficient and Nusselt number were calculated for different velocities. The heat transfer performance of the 40PPI sample brazed with the foil is found to be the most efficient. Pressure drop measurements were also performed for 10 and 40PPI metal foam. Thermographic measurements were done on 40PPI foam heat exchangers using a high temperature infrared camera. A high power electric heater was used to produce hot air at 300°C that passed over the foam heat exchanger while the cooling air was blown through it. Heat shields were made by depositing porous skins on metal foam and it was observed that a small amount of coolant leaking through the pores notably reduces the heat transfer from the hot gases. An analytical model was developed based assuming local thermal non-equilibrium that accounts for the temperature difference between solid and fluid phase. The experimental results are found to be in good agreement with the predicted values of the model.

Heat Transfer by Thermo-Capillary Convection. Sounding Rocket COMPERE Experiment SOURCE

Science.gov (United States)

Fuhrmann, Eckart; Dreyer, Michael

2009-08-01

This paper describes the results of a sounding rocket experiment which was partly dedicated to study the heat transfer from a hot wall to a cold liquid with a free surface. Natural or buoyancy-driven convection does not occur in the compensated gravity environment of a ballistic phase. Thermo-capillary convection driven by a temperature gradient along the free surface always occurs if a non-condensable gas is present. This convection increases the heat transfer compared to a pure conductive case. Heat transfer correlations are needed to predict temperature distributions in the tanks of cryogenic upper stages. Future upper stages of the European Ariane V rocket have mission scenarios with multiple ballistic phases. The aims of this paper and of the COMPERE group (French-German research group on propellant behavior in rocket tanks) in general are to provide basic knowledge, correlations and computer models to predict the thermo-fluid behavior of cryogenic propellants for future mission scenarios. Temperature and surface location data from the flight have been compared with numerical calculations to get the heat flux from the wall to the liquid. Since the heat flux measurements along the walls of the transparent test cell were not possible, the analysis of the heat transfer coefficient relies therefore on the numerical modeling which was validated with the flight data. The coincidence between experiment and simulation is fairly good and allows presenting the data in form of a Nusselt number which depends on a characteristic Reynolds number and the Prandtl number. The results are useful for further benchmarking of Computational Fluid Dynamics (CFD) codes such as FLOW-3D and FLUENT, and for the design of future upper stage propellant tanks.

Primary Issues of Mixed Convection Heat Transfer Phenomena

Energy Technology Data Exchange (ETDEWEB)

Chae, Myeong-Seon; Chung, Bum-Jin [Kyung Hee University, Yongin (Korea, Republic of)

2015-10-15

The computer code analyzing the system operating and transient behavior must distinguish flow conditions involved with convective heat transfer flow regimes. And the proper correlations must be supplied to those flow regimes. However the existing safety analysis codes are focused on the Light Water Reactor and they are skeptical to be applied to the GCRs (Gas Cooled Reactors). One of the technical issues raise by the development of the VHTR is the mixed convection, which occur when the driving forces of both forced and natural convection are of comparable magnitudes. It can be encountered as in channel of the stacked with fuel elements and a decay heat removal system and in VHTR. The mixed convection is not intermediate phenomena with natural convection and forced convection but independent complicated phenomena. Therefore, many researchers have been studied and some primary issues were propounded for phenomena mixed convection. This paper is to discuss some problems identified through reviewing the papers for mixed convection phenomena. And primary issues of mixed convection heat transfer were proposed respect to thermal hydraulic problems for VHTR. The VHTR thermal hydraulic study requires an indepth study of the mixed convection phenomena. In this study we reviewed the classical flow regime map of Metais and Eckert and derived further issues to be considered. The following issues were raised: (1) Buoyancy aided an opposed flows were not differentiated and plotted in a map. (2) Experimental results for UWT and UHF condition were also plotted in the same map without differentiation. (3) The buoyancy coefficient was not generalized for correlating with buoyancy coefficient. (4) The phenomenon analysis for laminarization and returbulization as buoyancy effects in turbulent mixed convection was not established. (5) The defining to transition in mixed convection regime was difficult.

On the heat transfer correlation for membrane distillation

International Nuclear Information System (INIS)

Wang, Chi-Chuan

2011-01-01

Research highlights: → Heat transfer coefficients applicable for membrane distillation. → Data reduction for heat transfer coefficient for membrane distillation method. → Uncertainty of permeate side due to large magnitude of membrane resistance. → Increase accuracy of heat transfer coefficient by modified Wilson plot technique. -- Abstract: The present study examines the heat transfer coefficients applicable for membrane distillation. In the available literatures, researchers often adopt some existing correlations and claim the suitability of these correlations to their test data or models. Unfortunately this approach is quite limited and questionable. This is subject to the influences of boundary conditions, geometrical configurations, entry flow conditions, as well as some influences from spacer or support. The simple way is to obtain the heat transfer coefficients from experimentation. However there is no direct experimental data for heat transfer coefficients being reported directly from the measurements. The main reasons are from the uncertainty of permeate side and of the comparatively large magnitude of membrane resistance. Additional minor influence is the effect of mass transfer on the heat transfer performance. In practice, the mass transfer effect is negligible provided the feed side temperature is low. To increase the accuracy of the measured feed side heat transfer coefficient, it is proposed in this study to exploit a modified Wilson plot technique. Through this approach, one can eliminate the uncertainty from permeate side and reduce the uncertainty in membrane to obtain a more reliable heat transfer coefficients at feed side from the experimentation.

Finite element simulation of heat transfer

CERN Document Server

Bergheau, Jean-Michel

2010-01-01

This book introduces the finite element method applied to the resolution of industrial heat transfer problems. Starting from steady conduction, the method is gradually extended to transient regimes, to traditional non-linearities, and to convective phenomena. Coupled problems involving heat transfer are then presented. Three types of couplings are discussed: coupling through boundary conditions (such as radiative heat transfer in cavities), addition of state variables (such as metallurgical phase change), and coupling through partial differential equations (such as electrical phenomena).? A re

Free surface deformation and heat transfer by thermocapillary convection

Science.gov (United States)

Fuhrmann, Eckart; Dreyer, Michael; Basting, Steffen; Bänsch, Eberhard

2016-04-01

Knowing the location of the free liquid/gas surface and the heat transfer from the wall towards the fluid is of paramount importance in the design and the optimization of cryogenic upper stage tanks for launchers with ballistic phases, where residual accelerations are smaller by up to four orders of magnitude compared to the gravity acceleration on earth. This changes the driving forces drastically: free surfaces become capillary dominated and natural or free convection is replaced by thermocapillary convection if a non-condensable gas is present. In this paper we report on a sounding rocket experiment that provided data of a liquid free surface with a nonisothermal boundary condition, i.e. a preheated test cell was filled with a cold but storable liquid in low gravity. The corresponding thermocapillary convection (driven by the temperature dependence of the surface tension) created a velocity field directed away from the hot wall towards the colder liquid and then in turn back at the bottom towards the wall. A deformation of the free surface resulting in an apparent contact angle rather different from the microscopic one could be observed. The thermocapillary flow convected the heat from the wall to the liquid and increased the heat transfer compared to pure conduction significantly. The paper presents results of the apparent contact angle as a function of the dimensionless numbers (Weber-Marangoni and Reynolds-Marangoni number) as well as heat transfer data in the form of a Nusselt number. Experimental results are complemented by corresponding numerical simulations with the commercial software Flow3D and the inhouse code Navier.

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

Evaluation of piping heat transfer, piping flow regimes, and steam generator heat transfer for the Semiscale Mod-1 isothermal tests

International Nuclear Information System (INIS)

French, R.T.

1975-08-01

Selected experimental data pertinent to piping heat transfer, transient fluid flow regimes, and steam generator heat transfer obtained during the Semiscale Mod-1 isothermal blowdown test series (Test Series 1) are analyzed. The tests in this first test series were designed to provide counterparts to the LOFT nonnuclear experiments. The data from the Semiscale Mod-1 intact and broken loop piping are evaluated to determine the surface heat flux and average heat transfer coefficients effective during the blowdown transient and compared with well known heat transfer correlations used in the RELAP4 computer program. Flow regimes in horizontal pipe sections are calculated and compared with data obtained from horizontal and vertical densitometers and with an existing steady state flow map. Effects of steam generator heat transfer are evaluated quantitatively and qualitatively. The Semiscale Mod-1 data and the analysis presented in this report are valuable for evaluating the adequacy and improving the predictive capability of analytical models developed to predict system response to piping heat transfer, piping flow regimes, and steam generator heat transfer during a postulated loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR). 16 references. (auth)

Subcooled boiling heat transfer on a finned surface

International Nuclear Information System (INIS)

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

1992-01-01

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

Engineering calculations in radiative heat transfer

CERN Document Server

Gray, W A; Hopkins, D W

1974-01-01

Engineering Calculations in Radiative Heat Transfer is a six-chapter book that first explains the basic principles of thermal radiation and direct radiative transfer. Total exchange of radiation within an enclosure containing an absorbing or non-absorbing medium is then described. Subsequent chapters detail the radiative heat transfer applications and measurement of radiation and temperature.

Industrial furnace with improved heat transfer

Energy Technology Data Exchange (ETDEWEB)

Hoetzl, M.; Lingle, T.M.

1992-07-07

This patent describes an industrial furnace for heating work which emits volatiles during heating. It comprises a generally cylindrical, closed end furnace section defining a sealable heat transfer chamber for heating work disposed therein; fan means for directing furnace atmosphere as a swirling wind mass about the interior of the furnace section over a portion thereof; heat means for heating the wind mass within the fan chamber; and an incineration track formed as a circumferentially extending groove about the exterior of the furnace section and in heat transfer relationship with and situated at least to extend about a portion of the fan chamber.

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