WorldWideScience

Sample records for reactor heat transfer

  1. Neutron behavior, reactor control, and reactor heat transfer. Volume four

    International Nuclear Information System (INIS)

    Anon.

    1986-01-01

    Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

  2. 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)

  3. Internally Heated Screw Pyrolysis Reactor (IHSPR) heat transfer performance study

    Science.gov (United States)

    Teo, S. H.; Gan, H. L.; Alias, A.; Gan, L. M.

    2018-04-01

    1.5 billion end-of-life tyres (ELT) were discarded globally each year and pyrolysis is considered the best solution to convert the ELT into valuable high energy-density products. Among all pyrolysis technologies, screw reactor is favourable. However, conventional screw reactor risks plugging issue due to its lacklustre heat transfer performance. An internally heated screw pyrolysis reactor (IHSPR) was developed by local renewable energy industry, which serves as the research subject for heat transfer performance study of this particular paper. Zero-load heating test (ZLHT) was first carried out to obtain the operational parameters of the reactor, followed by the one dimensional steady-state heat transfer analysis carried out using SolidWorks Flow Simulation 2016. Experiments with feed rate manipulations and pyrolysis products analyses were conducted last to conclude the study.

  4. Seminar on Heat-transfer fluids for fast neutron reactors

    International Nuclear Information System (INIS)

    Brechet, Yves; Dautray, Robert; Friedel, Jacques; Brezin, Edouard; Martin, Georges; Pineau, Andre; Carre, Francois; Gauche, Francois; Rodriguez, Guillaume; Latge, Christian; Cabet, Celine; Garnier, Jean-Claude; Bamberger, Yves; Sauvage, Jean-Francois; Buisine, Denis; Agostini, Pietro; Ulyanov, Vladimir; Auger, Thierry; Heuer, Daniel; Ghetta, Veronique; Bubelis, Evaldas; Charlaix, Elisabeth; Barrat, Jean-Louis; Boquet, Lyderic; Glickman, Evgueny; Escaravage, Claude

    2014-03-01

    This book reports the content of a two-day meeting held by the Academy of Sciences on the use of heat-transfer fluids in fast neutron reactors. After a first part which proposes an overview of scientific and technical problems related to these heat-transfer fluids (heat transfer process, nuclear properties, chemistry, materials, risks), a contribution proposes a return on experience on the use of heat-transfer fluids in the different design options of reactors of fourth generation: from mercury to NaK in the first fast neutron reactor projects, specific assets and constraints of sodium used as heat-transfer fluid, concepts of fast neutron reactors cooled by something else than sodium, perspectives for projects and research in fast neutron reactors. The next contribution discusses the specifications of future fast-neutron reactors: expectations for fourth-generation reactors, expectations in terms of performance and of safety, specific challenges. The last contribution addresses actions to be undertaken in the field of research and development: actions regarding all reactor types or specific types as sodium-cooled reactors, lead cooled reactors, molten salt reactors, and gas-cooled fast reactors

  5. Heat transfer for ultrahigh flux reactor

    International Nuclear Information System (INIS)

    Wadkins, R.P.; Lake, J.A.; Oh, C.H.

    1987-01-01

    The use of a uniquely designed nuclear reactor to supply neutrons for materials research is the focus of recent reactor design efforts. The biological, materials, and fundamental physics aspects of research require neutron fluxes much higher than present research and testing facilities can produce. The most advanced research using neutrons as probing detectors is being done in the High Flux Reactor at the Institut Laue Langeuin, France. The design of a reactor that can produce neutron fluxes of 1.0 x 10 16 n/cm 2 .s requires a relatively high power (300 MW range) and a small core volume (approximately 30 liters). This combination of power and volume leads to a high power density which places increased demands on thermal hydraulic margins

  6. Prediction of heat and mass transfer in innovative nuclear reactors

    International Nuclear Information System (INIS)

    Ambrosini, W.; Forgione, N.; Manfredini, A.; Oriolo, F.

    2000-01-01

    This paper proposes a short review of the different forms adopted to express the analogy between heat and mass transfer for application in correlating data from condensation and evaporation experiments. In particular, the assumptions at the basis of the various forms presented by classical textbooks as well as recent research work are qualitatively discussed, proposing a unified treatment of the different models. On this background, the results of the application of one of the considered forms of the analogy to a problem having relevance for nuclear reactor safety are then discussed. The work performed in this frame is related to condensation on finned tube heat exchangers, proposed as key components in passive containment cooling systems adopted in some innovative reactor concepts. The application of the model to the experimental dana also allowed to obtain interesting information about the effect of different parameters on the cooling capabilities of this compact heat exchangers. (author)

  7. Improvement of Reactor Fuel Element Heat Transfer by Surface Roughness

    International Nuclear Information System (INIS)

    Kjellstroem, B.; Larsson, A.E.

    1967-04-01

    In heat exchangers with a limited surface temperature such as reactor fuel elements, rough heat transfer surfaces may give lower pumping power than smooth. To obtain data for choice of the most advantageous roughness for the superheater elements in the Marviken reactor, measurements were made of heat transfer and pressure drop in an annular channel with a smooth or rough test rod in a smooth adiabatic shroud. 24 different roughness geometries were tested. The results were transformed to rod cluster geometry by the method of W B Hall, and correlated by the friction and heat transfer similarity laws as suggested by D F Dipprey and R H Sabersky with RMS errors of 12.5 % in the friction factor and 8.1 % in the Stanton number. The relation between the Stanton number and the friction factor could be described by a relation of the type suggested by W Nunner, with a mean error of 3.1 % and an RMS error of 11.6 %. Application of the results to fuel element calculations is discussed, and the great gains in economy which can be obtained with rough surfaces are demonstrated by two examples

  8. Improvement of Reactor Fuel Element Heat Transfer by Surface Roughness

    Energy Technology Data Exchange (ETDEWEB)

    Kjellstroem, B; Larsson, A E

    1967-04-15

    In heat exchangers with a limited surface temperature such as reactor fuel elements, rough heat transfer surfaces may give lower pumping power than smooth. To obtain data for choice of the most advantageous roughness for the superheater elements in the Marviken reactor, measurements were made of heat transfer and pressure drop in an annular channel with a smooth or rough test rod in a smooth adiabatic shroud. 24 different roughness geometries were tested. The results were transformed to rod cluster geometry by the method of W B Hall, and correlated by the friction and heat transfer similarity laws as suggested by D F Dipprey and R H Sabersky with RMS errors of 12.5 % in the friction factor and 8.1 % in the Stanton number. The relation between the Stanton number and the friction factor could be described by a relation of the type suggested by W Nunner, with a mean error of 3.1 % and an RMS error of 11.6 %. Application of the results to fuel element calculations is discussed, and the great gains in economy which can be obtained with rough surfaces are demonstrated by two examples.

  9. 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

  10. 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)

  11. Convective heat transfer analysis in aggregates rotary drum reactor

    International Nuclear Information System (INIS)

    Le Guen, Laurédan; Huchet, Florian; Dumoulin, Jean; Baudru, Yvan; Tamagny, Philippe

    2013-01-01

    Heat transport characterisation inside rotary drum dryer has a considerable importance linked to many industrial applications. The present paper deals with the heat transfer analysis from experimental apparatus installed in a large-scale rotary drum reactor applied to the asphalt materials production. The equipment including in-situ thermal probes and external visualization by mean of infrared thermography gives rise to the longitudinal evaluation of inner and external temperatures. The assessment of the heat transfer coefficients by an inverse methodology is resolved in order to accomplish a fin analysis of the convective mechanism inside baffled (or flights) rotary drum. The results are discussed and compared with major results of the literature. -- Highlights: ► A thermal and flow experimentation is performed on a large-scale rotary drum. ► Four working points is chosen in the frame of asphalt materials production. ► Evaluation of the convective transfer mechanisms is calculated by inverse method. ► The drying stage is performed in the combustion area. ► Wall/aggregates heat exchanges have a major contribution in the heating stage

  12. Conjugate heat transfer simulations of advanced research reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Piro, M.H.A., E-mail: pirom@aecl.ca; Leitch, B.W.

    2014-07-01

    Highlights: • Temperature predictions are enhanced by coupling heat transfer in solid and fluid zones. • Seven different cases are considered to observe trends in predicted temperature and pressure. • The seven cases consider high/medium/low power, flow, burnup, fuel material and geometry. • Simulations provide temperature predictions for performance/safety. Boiling is unlikely. • Simulations demonstrate that a candidate geometry can enhance performance/safety. - Abstract: The current work presents numerical simulations of coupled fluid flow and heat transfer of advanced U–Mo/Al and U–Mo/Mg research reactor fuels in support of performance and safety analyses. The objective of this study is to enhance predictions of the flow regime and fuel temperatures through high fidelity simulations that better capture various heat transfer pathways and with a more realistic geometric representation of the fuel assembly in comparison to previous efforts. Specifically, thermal conduction, convection and radiation mechanisms are conjugated between the solid and fluid regions. Also, a complete fuel element assembly is represented in three dimensional space, permitting fluid flow and heat transfer to be simulated across the entire domain. Seven case studies are examined that vary the coolant inlet conditions, specific power, and burnup to investigate the predicted changes in the pressure drop in the coolant and the fuel, clad and coolant temperatures. In addition, an alternate fuel geometry is considered with helical fins (replacing straight fins in the existing design) to investigate the relative changes in predicted fluid and solid temperatures. Numerical simulations predict that the clad temperature is sensitive to changes in the thermal boundary layer in the coolant, particularly in simultaneously developing flow regions, while the temperature in the fuel is anticipated to be unaffected. Finally, heat transfer between fluid and solid regions is enhanced with

  13. Heat transfer in inertial confinement fusion reactor systems

    International Nuclear Information System (INIS)

    Hovingh, J.

    1979-01-01

    The transfer of energy produced by the interaction of the intense pulses of short-ranged fusion microexplosion products with materials is one of the most difficult problems in inertially-confined fusion (ICF) reactor design. The short time and deposition distance for the energy results in local peak power densities on the order of 10 18 watts/m 3 . High local power densities may cause change of state or spall in the reactor materials. This will limit the structure lifetimes for ICF reactors of economic physical sizes, increasing operating costs including structure replacement and radioactive waste management. Four basic first wall protection methods have evolved: a dry-wall, a wet-wall, a magnetically shielded wall, and a fluid wall. These approaches are distinguished by the way the reactor wall interfaces with fusion debris as well as the way the ambient cavity conditions modify the fusion energy forms and spectra at the first wall. Each of these approaches requires different heat transfer considerations

  14. Heat Transfer Phenomena in Supercritical Water Nuclear Reactors

    International Nuclear Information System (INIS)

    Mark H. Anderson; MichaelL. Corradini; Riccardo Bonazza; Jeremy R. Licht

    2007-01-01

    A supercritical water heat transfer facility has been built at the University of Wisconsin to study heat transfer in a circular and square annular flow channel. A series of integral heat transfer measurements has been carried out over a wide range of heat flux, mass velocity and bulk water temperatures at a pressure of 25 MPa. The circular annular test section geometry is a 1.07 cm diameter heater rod within a 4.29 diameter flow channel

  15. Heat Transfer Phenomena in Supercritical Water Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mark H. Anderson; MichaelL. Corradini; Riccardo Bonazza; Jeremy R. Licht

    2007-10-03

    A supercritical water heat transfer facility has been built at the University of Wisconsin to study heat transfer in ancircular and square annular flow channel. A series of integral heat transfer measurements has been carried out over a wide range of heat flux, mas velocity and bulk water temperatures at a pressure of 25 MPa. The circular annular test section geometry is a 1.07 cm diameter heater rod within a 4.29 diameter flow channel.

  16. 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

  17. E-learning modules for nuclear reactor heat transfer

    Science.gov (United States)

    Jayaram, Praveen Bharadwaj

    E learning in engineering education is becoming popular at several universities as it allows instructors to create content that the students may view and interact with at his/her own convenience. Web-based simulation and what-if analysis are examples of such educational content and has proved to be extremely beneficial for engineering students. Such pedagogical content promote active learning and encourage students to experiment and be more creative. The main objective of this project is to develop web based learning modules, in the form of analytical simulations, for the Reactor Thermal Hydraulics course offered by the College of Engineering at UT Arlington. These modules seek to comprehensively transform the traditional education structure. The simulations are built to supplement the class lectures and are divided into categories like Fundamentals, Heat generation, Heat transfer and Heat removal categories. Each category contains modules which are sub-divided chapter wise and further into section wise. Some of the important sections from the text book are taken and calculations for a particular functionality are implemented. Since it is an interactive tool, it allows user to input certain values, which are then processed with the traditional equations, and output results either in the form of a number or graphs.

  18. Heat transfer characteristics and operation limit of pressurized hybrid heat pipe for small modular reactors

    International Nuclear Information System (INIS)

    Kim, Kyung Mo; Bang, In Cheol

    2017-01-01

    Highlights: • Thermal performances and operation limits of hybrid heat pipe were experimentally studied. • Models for predicting the operation limit of the hybrid heat pipe was developed. • Non-condensable gas affected heat transfer characteristics of the hybrid heat pipe. - Abstract: In this paper, a hybrid heat pipe is proposed for use in advanced nuclear power plants as a passive heat transfer device. The hybrid heat pipe combines the functions of a heat pipe and a control rod to simultaneously remove the decay heat generated from the core and shutdown the reactor under accident conditions. Thus, the hybrid heat pipe contains a neutron absorber in the evaporator section, which corresponds to the core of the reactor pressure vessel. The presence of the neutron absorber material leads to differences in the heated diameter and hydraulic diameter of the heat pipe. The cross-sectional areas of the vapor paths through the evaporator, adiabatic, and condenser sections are also different. The hybrid heat pipe must operate in a high-temperature, high-pressure environment to remove the decay heat. In other words, the operating pressure must be higher than those of the commercially available thermosyphons. Hence, the thermal performances, including operation limit of the hybrid heat pipe, were experimentally studied in the operating pressure range of 0.2–20 bar. The operating pressure of the hybrid heat pipe was controlled by charging the non-condensable gas which is unused method to achieve the high saturation pressure in conventional thermosyphons. The effect of operating pressure on evaporation heat transfer was negligible, while condensation heat transfer was affected by the amount of non-condensable gas in the test section. The operation limit of the hybrid heat pipe increased with the operating pressure. Maximum heat removal capacity of the hybrid heat pipe was up to 6 kW which is meaningful value as a passive decay heat removal device in the nuclear power

  19. SCEPTIC, Pressure Drop, Flow Rate, Heat Transfer, Temperature in Reactor Heat Exchanger

    International Nuclear Information System (INIS)

    Kattchee, N.; Reynolds, W.C.

    1975-01-01

    1 - Nature of physical problem solved: SCEPTIC is a program for calculating pressure drop, flow rates, heat transfer rates, and temperature in heat exchangers such as fuel elements of typical gas or liquid cooled nuclear reactors. The effects of turbulent and heat interchange between flow passages are considered. 2 - Method of solution: The computation procedure amounts to a nodal of lumped parameter type of calculation. The axial mesh size is automatically selected to assure that a prescribed accuracy of results is obtained. 3 - Restrictions on the complexity of the problem: Maximum number of subchannels is 25, maximum number of heated surfaces is 46

  20. Heat exchanger for transfering heat produced in a high temperature reactor to an intermediate circuit gas

    International Nuclear Information System (INIS)

    Barchewitz, E.; Baumgaertner, H.

    1985-01-01

    The invention is concerned with improving the arrangement of a heat exchanger designed to transfer heat from the coolant gas circuit of a high temperature reactor to a gas which is to be used for a process heat plant. In the plant the material stresses are to be kept low at high differential pressures and temperatures. According to the invention the tube bundles designed as boxes are fixed within the heat exchanger closure by means of supply pipes having got loops. For conducting the hot gas the heat exchanger has got a central pipe leading out of the reactor vessel through the pod closure and having got only one point of fixation, lying in this closure. Additional advantageous designs are mentioned. (orig./PW)

  1. Two-phase flow heat transfer in nuclear reactor systems

    International Nuclear Information System (INIS)

    Koncar, Bostjan; Krepper, Eckhard; Bestion, Dominique; Song, Chul-Hwa; Hassan, Yassin A.

    2013-01-01

    Complete text of publication follows: Heat transfer and phase change phenomena in two-phase flows are often encountered in nuclear reactor systems and are therefore of paramount importance for their optimal design and safe operation.The complex phenomena observed especially during transient operation of nuclear reactor systems necessitate extensive theoretical and experimental investigations. This special issue brings seven research articles of high quality. Though small in number, they cover a wide range of topics, presenting high complexity and diversity of heat transfer phenomena in two-phase flow. In the last decades a vast amount of research has been devoted to theoretical work and computational simulations, yet the experimental work remains indispensable for understanding of two-phase flow phenomena and for model validation purposes. This is reflected also in this issue, where only one article is purely experimental, while three of them deal with theoretical modelling and the remaining three with numerical simulations. The experimental investigation of the critical heat flux (CHF) phenomena by means of photographic study is presented in the paper of J. Park et al. They have used a high-speed camera system to observe the transient boiling characteristics on a thin horizontal cylinder submerged in a pool of water or highly wetting liquid. Experiments show that the initial boiling process is strongly affected by the properties and wettability of the liquid. The authors have stressed the importance of the local scale observation leading to better understanding of the transient CHF phenomena. In the article of G. Espinosa-Paredes et al. a theoretical work concerning the derivation of transport equations for two-phase flow is presented. The author proposes a novel approach based on derivation of nonlocal volume averaged equations which contain new terms related to nonlocal transport effects. These non-local terms act as coupling elements between the phenomena

  2. Heat transfer burnout in tube-type fuel elements of nuclear power reactors

    International Nuclear Information System (INIS)

    Subbotin, V.; Alexeev, G.; Peskov, O.; Sapankevic, A.

    1976-01-01

    The conditions are formulated under which the results of the experimental research of the boilino. water heat transfer burnout carried out on models may be applied to fuel elements of nuclear reactors. Experimental material providing data on the heat transfer burnout was expanded by the results of measurements of the uneven (cosine) longitudinal distribution of heat sources. The results of the effects of helical fins or wires on heat transfer burnout are presented. (F.M.)

  3. Heat transfer burnout in tube-type fuel elements of nuclear power reactors

    Energy Technology Data Exchange (ETDEWEB)

    Subbotin, V; Alexeev, G; Peskov, O; Sapankevic, A

    1976-08-01

    The conditions are formulated under which the results of the experimental research of the boiling. water heat transfer burnout carried out on models may be applied to fuel elements of nuclear reactors. Experimental material providing data on the heat transfer burnout was expanded by the results of measurements of the uneven (cosine) longitudinal distribution of heat sources. The results of the effects of helical fins or wires on heat transfer burnout are presented.

  4. Design guide for heat transfer equipment in water-cooled nuclear reactor systems

    International Nuclear Information System (INIS)

    1975-07-01

    Information pertaining to design methods, material selection, fabrication, quality assurance, and performance tests for heat transfer equipment in water-cooled nuclear reactor systems is given in this design guide. This information is intended to assist those concerned with the design, specification, and evaluation of heat transfer equipment for nuclear service and the systems in which this equipment is required. (U.S.)

  5. Effect of reactor heat transfer limitations on CO preferential oxidation

    Science.gov (United States)

    Ouyang, X.; Besser, R. S.

    Our recent studies of CO preferential oxidation (PrOx) identified systematic differences between the characteristic curves of CO conversion for a microchannel reactor with thin-film wall catalyst and conventional mini packed-bed lab reactors (m-PBR's). Strong evidence has suggested that the reverse water-gas-shift (r-WGS) side reaction activated by temperature gradients in m-PBR's is the source of these differences. In the present work, a quasi-3D tubular non-isothermal reactor model based on the finite difference method was constructed to quantitatively study the effect of heat transport resistance on PrOx reaction behavior. First, the kinetic expressions for the three principal reactions involved were formed based on the combination of experimental data and literature reports and their parameters were evaluated with a non-linear regression method. Based on the resulting kinetic model and an energy balance derived for PrOx, the finite difference method was then adopted for the quasi-3D model. This model was then used to simulate both the microreactor and m-PBR's and to gain insights into their different conversion behavior. Simulation showed that the temperature gradients in m-PBR's favor the reverse water-gas-shift (r-WGS) reaction, thus causing a much narrower range of permissible operating temperature compared to the microreactor. Accordingly, the extremely efficient heat removal of the microchannel/thin-film catalyst system eliminates temperature gradients and efficiently prevents the onset of the r-WGS reaction.

  6. Heat and mass transfer in the HYLIFE ICF reactor cavity

    International Nuclear Information System (INIS)

    Glenn, L.A.

    1981-01-01

    A quasi-one dimensional method was developed for calculating transient, compressible, viscous flow across a complex array of tubes or jets. The method also accounts for the diffusion of radiation and for heat and mass exchange between the fluid and the jets. The application was to the impulsive crossflow of a lithium plasma through a close-packed annular arrangement of liquid lithium jets, a problem that arises in the design of inertial confinement fusion reactors. It was found that approximately 2/3 of the energy initially contained in the plasma will diffuse into the liquid jets, not including an additional 7-10% which will go towards jet surface vaporization. Nevertheless, the peak hoop stress in the first wall of the reactor appears to derive from direct impact of the plasma, rather than from the subsequent impact of the jets or fragments thereof. (orig.)

  7. Heat transfer in reactor cavity during core-concrete interaction

    International Nuclear Information System (INIS)

    Adroguer, B.; Cenerino, G.

    1989-08-01

    In the unlikely event of a severe accident in a nuclear power plant, the core may melt through the vessel and slump into the concrete reactor cavity. The hot mixture of the core material called corium interacts thermally with the concrete basemat. The WECHSL code, developed at K.f.K. Karlsruhe in Germany is used at the Protection and Nuclear Safety Institute (I.P.S.N.) of CEA to compute this molten corium concrete interaction (MCCI). Some uncertainties remain in the partition of heat from the corium between the basemat and the upper surrounding structures in the cavity where the thermal conditions are not computer. The CALTHER code, under development to perform a more mechanistic evaluation of the upward heat flux has been linked to WECHSL-MOD2 code. This new version enables the modelling of the feedback effects from the conditions in the cavity to the MCCI and the computation of the fraction of upward flux directly added to the cavity atmosphere. The present status is given in the paper. Preliminary calculations of the reactor case for silicate and limestone common sand (L.C.S.) concretes are presented. Significant effects are found on concrete erosion, gases release and temperature of the upper part of corium, particularly for L.C.S. concrete

  8. Correlations of Nucleate Boiling Heat Transfer and Critical Heat Flux for External Reactor Vessel Cooling

    International Nuclear Information System (INIS)

    J. Yang; F. B. Cheung; J. L. Rempe; K. Y. Suh; S. B. Kim

    2005-01-01

    Four types of steady-state boiling experiments were conducted to investigate the efficacy of two distinctly different heat transfer enhancement methods for external reactor vessel cooling under severe accident conditions. One method involved the use of a thin vessel coating and the other involved the use of an enhanced insulation structure. By comparing the results obtained in the four types of experiments, the separate and integral effect of vessel coating and insulation structure were determined. Correlation equations were obtained for the nucleate boiling heat transfer and the critical heat flux. It was found that both enhancement methods were quite effective. Depending on the angular location, the local critical heat flux could be enhanced by 1.4 to 2.5 times using vessel coating alone whereas it could be enhanced by 1.8 to 3.0 times using an enhanced insulation structure alone. When both vessel coating and insulation structure were used simultaneously, the integral effect on the enhancement was found much less than the product of the two separate effects, indicating possible competing mechanisms (i.e., interference) between the two enhancement methods

  9. 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.

  10. Heat transfer and flow characteristics of a cooling thimble in a molten salt reactor residual heat removal system

    Directory of Open Access Journals (Sweden)

    Zonghao Yang

    2017-12-01

    Full Text Available In the passive residual heat removal system of a molten salt reactor, one of the residual heat removal methods is to use the thimble-type heat transfer elements of the drain salt tank to remove the residual heat of fuel salts. An experimental loop is designed and built with a single heat transfer element to analyze the heat transfer and flow characteristics. In this research, the influence of the size of a three-layer thimble-type heat transfer element on the heat transfer rate is analyzed. Two methods are used to obtain the heat transfer rate, and a difference of results between methods is approximately 5%. The gas gap width between the thimble and the bayonet has a large effect on the heat transfer rate. As the gas gap width increases from 1.0 mm to 11.0 mm, the heat transfer rate decreases from 5.2 kW to 1.6 kW. In addition, a natural circulation startup process is described in this paper. Finally, flashing natural circulation instability has been observed in this thimble-type heat transfer element.

  11. Measuring heat transfer through TR-0 reactor fuel element

    International Nuclear Information System (INIS)

    Nemec, V.; Turzik, Z.; Vitek, M.

    1977-05-01

    The time course of temperatures of the peripheral and the central fuel pins of the TR-O reactor was studied during moderator temperature changes using a model. The formula T=Tsub(e)+(Tsub(o)-Tsub(e)).exp(-t/tsub(e)) applies, where T is the pin temperature, Tsub(o) the initial pin temperature, Tsub(e) is the steady-state bath temperature, tsub(e) the time constant of temperature equilibration and t the time required for a temperature change from value Tsub(o) to T. For the bath level height H=1 m the tsub(e) value for the central pin was determined to be 1.05 hours, for the peripheral pin 0.96 hour; for level height H=2 m the values were 2.1 and 2.12 hours, respectively. The dependence found will allow correcting the experimental results in measurements with heated moderator for fuel temperature changes. (Ha)

  12. 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

  13. Review of heat transfer problems associated with magnetically-confined fusion reactor concepts

    International Nuclear Information System (INIS)

    Hoffman, M.A.; Werner, R.W.; Carlson, G.A.; Cornish, D.N.

    1976-01-01

    Conceptual design studies of possible fusion reactor configurations have revealed a host of interesting and sometimes extremely difficult heat transfer problems. The general requirements imposed on the coolant system for heat removal of the thermonuclear power from the reactor are discussed. In particular, the constraints imposed by the fusion plasma, neutronics, structure and magnetic field environment are described with emphasis on those aspects which are unusual or unique to fusion reactors. Then the particular heat transfer characteristics of various possible coolants including lithium, flibe, boiling alkali metals, and helium are discussed in the context of these general fusion reactor requirements. Some specific areas where further experimental and/or theoretical work is necessary are listed for each coolant along with references to the pertinent research already accomplished. Specialized heat transfer problems of the plasma injection and removal systems are also described. Finally, the challenging heat transfer problems associated with the superconducting magnets are reviewed, and once again some of the key unsolved heat transfer problems are enumerated

  14. Fluid flow and heat transfer investigation of pebble bed reactors using mesh-adaptive LES

    International Nuclear Information System (INIS)

    Pavlidis, Dimitrios; Lathouwers, Danny

    2013-01-01

    The very high temperature reactor is one of the designs currently being considered for nuclear power generation. One its variants is the pebble bed reactor in which the coolant passes through complex geometries (pores) at high Reynolds numbers. A computational fluid dynamics model with anisotropic mesh adaptivity is used to investigate coolant flow and heat transfer in such reactors. A novel method for implicitly incorporating solid boundaries based on multi-fluid flow modelling is adopted. The resulting model is able to resolve and simulate flow and heat transfer in randomly packed beds, regardless of the actual geometry, starting off with arbitrarily coarse meshes. The model is initially evaluated using an orderly stacked square channel of channel-height-to-particle diameter ratio of unity for a range of Reynolds numbers. The model is then applied to the face-centred cubical geometry. coolant flow and heat transfer patterns are investigated

  15. Modeling of condensation heat transfer in a reactor containment

    International Nuclear Information System (INIS)

    Kim, M.H.; Corradini, M.L.

    1990-01-01

    This paper proposes a set of condensation models for forced and natural convection in the presence of a noncondensable gas. A simple model is derived by using the analogy between mass, momentum and energy transfer. The effects of a wavy interface are implemented in this model by using correlations for a rough wall surface. A two-dimensional condensation model using a k-ε model for the turbulent vapor-air flow was also developed to investigate the effect of two-dimensional flow and to provide a sound theoretical basis for the simple model. Each model is compared with the available 'separate effects' experimental data. The forced convection model is then compared to the Carolinas Virginia Tube Reactor (CVTR) integral test by using the vapor-air velocity predicted by a separate two-dimensional fluid dynamics model. The effect of counter-current flow is also considered in this comparison. The natural convection model is also compared to the steady-state integral data of Tagami. The comparison shows good agreement with both sets of experimental data. (orig.)

  16. 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.

  17. ZrH reactor lattice spacing (heat transfer considerations)

    International Nuclear Information System (INIS)

    Felten, L.D.

    1970-01-01

    Temperature calculations for a 295 element ZrH reactor at fuel element spacings from 0.010'' to 0.065'' showed a very small dependence of reactor temperature on element spacing. It was found that one variation in coolant channel area (2 zones) was sufficient to satisfactorily shape the radial flow profile for the core. (U.S.)

  18. 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)

  19. CHAP-2 heat-transfer analysis of the Fort St. Vrain reactor core

    International Nuclear Information System (INIS)

    Kotas, J.F.; Stroh, K.R.

    1983-01-01

    The Los Alamos National Laboratory is developing the Composite High-Temperature Gas-Cooled Reactor Analysis Program (CHAP) to provide advanced best-estimate predictions of postulated accidents in gas-cooled reactor plants. The CHAP-2 reactor-core model uses the finite-element method to initialize a two-dimensional temperature map of the Fort St. Vrain (FSV) core and its top and bottom reflectors. The code generates a finite-element mesh, initializes noding and boundary conditions, and solves the nonlinear Laplace heat equation using temperature-dependent thermal conductivities, variable coolant-channel-convection heat-transfer coefficients, and specified internal fuel and moderator heat-generation rates. This paper discusses this method and analyzes an FSV reactor-core accident that simulates a control-rod withdrawal at full power

  20. Heat transfer from the roughened surface of gas cooled fast breeder reactor fuel element

    International Nuclear Information System (INIS)

    Tang, I.M.

    1979-01-01

    The temperature distributions and the augmentation of heat transfer performance by artificial roughening of a gas cooled fast breeder reactor (GCFR) fuel rod cladding are studied. Numerical solutions are based on the axisymmetric assumption for a two-dimensional model for one rib pitch of axial distance. The local and axial clad temperature distributions are obtained for both the rectangular and ramp rib roughened surface geometries. The transformation of experimentally measured convective heat transfer coefficients, in terms of Stanton number, into GCFR values is studied. In addition, the heat transfer performance of a GCFR fuel rod cladding roughened surface design is evaluated. Approximate analytical solution for correlating an average Stanton number is also obtained and satisfactorily compared with the corresponding numerical result for a GCFR design. The analytical correlation is useful in assessing roughened surface heat transfer performance in scoping studies and conceptual design

  1. 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

  2. Heat transfer and the continuous production of hydroxypropyl starch in a static mixer reactor

    NARCIS (Netherlands)

    Lammers, Gerard; Beenackers, Antonie A. C. M.

    1994-01-01

    A novel continuous reactor for the chemical derivation of aqueous starch solutions based on static mixers is proposed. Both the experimentally observed axial and radial temperature gradients in the static mixer could be accurately described by a pseudohomogeneous two-dimensional heat transfer (PTHT)

  3. Modeling of heat and mass transfer processes during core melt discharge from a reactor pressure vessel

    Energy Technology Data Exchange (ETDEWEB)

    Dinh, T.N.; Bui, V.A.; Nourgaliev, R.R. [Royal Institute of Technology, Stockholm (Sweden)] [and others

    1995-09-01

    The objective of the paper is to study heat and mass transfer processes related to core melt discharge from a reactor vessel is a severe light water reactor accident. The phenomenology of the issue includes (1) melt convection in and heat transfer from the melt pool in contact with the vessel lower head wall; (2) fluid dynamics and heat transfer of the melt flow in the growing discharge hole; and (3) multi-dimensional heat conduction in the ablating lower head wall. A program of model development, validation and application is underway (i) to analyse the dominant physical mechanisms determining characteristics of the lower head ablation process; (ii) to develop and validate efficient analytic/computational methods for estimating heat and mass transfer under phase-change conditions in irregular moving-boundary domains; and (iii) to investigate numerically the melt discharge phenomena in a reactor-scale situation, and, in particular, the sensitivity of the melt discharge transient to structural differences and various in-vessel melt progression scenarios. The paper presents recent results of the analysis and model development work supporting the simulant melt-structure interaction experiments.

  4. Modeling of heat transfer in wall-cooled tubular reactors

    NARCIS (Netherlands)

    Koning, G.W.; Westerterp, K.R.

    1999-01-01

    In a pilot scale wall-cooled tubular reactor, temperature profiles have been measured with and without reaction. As a model reaction oxidation of carbon monoxide in air over a copper chromite catalyst has been used. The kinetics of this reaction have been determined separately in two kinetic

  5. Modelling the behaviour of corrosion products in the primary heat transfer circuits of pressurised water reactors

    International Nuclear Information System (INIS)

    Rodliffe, R.S.; Polley, M.V.; Thornton, E.W.

    1985-05-01

    The redistribution of corrosion products from the primary circuit surfaces of a water reactor can result in increased flow resistance, poorer heat transfer performance, fuel failure and radioactive contamination of circuit surfaces. The environment is generally sufficiently well controlled to ensure that the first three effects are not limiting. The last effect is of particular importance since radioactive corrosion products are major contributors to shutdown fields and since it is necessary to ensure that the radiation exposure of personnel is as low as reasonably achievable. This review focusses attention on the principles which must form the basis for any mechanistic model describing the formation, transport and deposition of radioactive corrosion products. It is relevant to all water reactors in which the primary heat transfer medium is predominantly single-phase water and in which steam is generated in a secondary circuit, i.e. including CANDU pressurised heavy water reactors, Sovient VVERs, etc. (author)

  6. Studies of heat transfer having relevance to nuclear reactor containment cooling by buoyancy-driven air flow

    Energy Technology Data Exchange (ETDEWEB)

    Jackson, J. D.; Li, J.; Wang, J. [The Univ., of Manchester, Manchester (United Kingdom)

    2003-07-01

    Two separate effects experiments concerned with buoyancy-influenced convective heat transfer in vertical passages which have relevance to the problem of nuclear reactor containment cooling by means of buoyancy-driven airflow are described. A feature of each is that local values of heat transfer coefficient are determined on surfaces maintained at uniform temperature. Experimental results are presented which highlight the need for buoyancy-induced impairment of turbulent convective heat transfer to be accounted for in the design of such passive cooling systems. A strategy is presented for predicting the heat removal by combined convective and radiative heat transfer from a full scale nuclear reactor containment shell using such experimental results.

  7. Numerical investigation of heat transfer in high-temperature gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Chen, g.; Anghaie, S. [Univ. of Florida, Gainesville, FL (United States)

    1995-09-01

    This paper proposes a computational model for analysis of flow and heat transfer in high-temperature gas-cooled reactors. The formulation of the problem is based on using the axisymmetric, thin layer Navier-Stokes equations. A hybrid implicit-explicit method based on finite volume approach is used to numerically solve the governing equations. A fast converging scheme is developed to accelerate the Gauss-Siedel iterative method for problems involving the wall heat flux boundary condition. Several cases are simulated and results of temperature and pressure distribution in the core are presented. Results of a parametric analysis for the assessment of the impact of power density on the convective heat transfer rate and wall temperature are discussed. A comparative analysis is conducted to identify the Nusselt number correlation that best fits the physical conditions of the high-temperature gas-cooled reactors.

  8. 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.

  9. 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.

  10. A Study on Conjugate Heat Transfer Analysis of Reactor Vessel including Irradiated Structural Heat Source

    Energy Technology Data Exchange (ETDEWEB)

    Yi, Kunwoo; Cho, Hyuksu; Im, Inyoung; Kim, Eunkee [KEPCO EnC, Daejeon (Korea, Republic of)

    2015-10-15

    Though Material reliability programs (MRPs) have a purpose to provide the evaluation or management methodologies for the operating RVI, the similar evaluation methodologies can be applied to the APR1400 fleet in the design stage for the evaluation of neutron irradiation effects. The purposes of this study are: to predict the thermal behavior whether or not irradiated structure heat source; to evaluate effective thermal conductivity (ETC) in relation to isotropic and anisotropic conductivity of porous media for APR1400 Reactor Vessel. The CFD simulations are performed so as to evaluate thermal behavior whether or not irradiated structure heat source and effective thermal conductivity for APR1400 Reactor Vessel. In respective of using irradiated structure heat source, the maximum temperature of fluid and core shroud for isotropic ETC are 325.8 .deg. C, 341.5 .deg. C. The total amount of irradiated structure heat source is about 5.41 MWth and not effect to fluid temperature.

  11. Decay heat removal and heat transfer under normal and accident conditions in gas cooled reactors

    International Nuclear Information System (INIS)

    1994-08-01

    The meeting was convened by the International Atomic Energy Agency on the recommendation of the IAEA's International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Poland, the Russian Federation, Switzerland, the United Kingdom and the United States of America. The meeting was chaired by Prof. Dr. K. Kugeler and Prof. Dr. E. Hicken, Directors of the Institute for Safety Research Technology of the KFA Research Center, and covered the following: Design and licensing requirements for gas cooled reactors; concepts for decay heat removal in modern gas cooled reactors; analytical methods for predictions of thermal response, accuracy of predictions; experimental data for validation of predictive methods - operational experience from gas cooled reactors and experimental data from test facilities. Refs, figs and tabs

  12. 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)

  13. Heat transfer and pressure drop of a reactor fuel element model with polyzonal spiral finning

    Energy Technology Data Exchange (ETDEWEB)

    Oka, S; Becirspahic, S [Institute of Nuclear Sciences Boris Kidric, Heat Transfer Department, Vinca, Beograd (Serbia and Montenegro)

    1964-10-15

    Heat transfer and pressure drop of a reactor fuel element model with polyzonal spiral finning have been investigated. The St-number distribution over length and perimeter of he finning are given. The mean and minimum St{sub k}-number are plotted against the Re-number. The influence of the gap between two fuel elements upon heat transfer and pressure drop, in dependence on the Re-number, and the influence of the length of the fuel element on pressure drop across the gap are shown. The influence of the relative position of the splitters of two neighboring fuel elements on pressure drop and heat transfer is shown. The investigations were performed in the Re-number range 15,000 to 100,000 (author)

  14. A computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hu, Rui, E-mail: rhu@anl.gov; Yu, Yiqi

    2016-11-15

    Highlights: • Developed a computationally efficient method for full-core conjugate heat transfer modeling of sodium fast reactors. • Applied fully-coupled JFNK solution scheme to avoid the operator-splitting errors. • The accuracy and efficiency of the method is confirmed with a 7-assembly test problem. • The effects of different spatial discretization schemes are investigated and compared to the RANS-based CFD simulations. - Abstract: For efficient and accurate temperature predictions of sodium fast reactor structures, a 3-D full-core conjugate heat transfer modeling capability is developed for an advanced system analysis tool, SAM. The hexagon lattice core is modeled with 1-D parallel channels representing the subassembly flow, and 2-D duct walls and inter-assembly gaps. The six sides of the hexagon duct wall and near-wall coolant region are modeled separately to account for different temperatures and heat transfer between coolant flow and each side of the duct wall. The Jacobian Free Newton Krylov (JFNK) solution method is applied to solve the fluid and solid field simultaneously in a fully coupled fashion. The 3-D full-core conjugate heat transfer modeling capability in SAM has been demonstrated by a verification test problem with 7 fuel assemblies in a hexagon lattice layout. Additionally, the SAM simulation results are compared with RANS-based CFD simulations. Very good agreements have been achieved between the results of the two approaches.

  15. USING LIGA BASED MICROFABRICATION TO IMPROVE OVERALL HEAT TRANSFER EFFICIENCY OF PRESSURIZED WATER REACTOR: I. Effects of Different Micro Pattern on Overall Heat Transfer

    International Nuclear Information System (INIS)

    Zhang, M.; Ibekwe, S.; Li, G.; Pang, S.S.; Lian, K.

    2006-01-01

    The Pressurized Water Reactors (PWRs in Figure 1) were originally developed for naval propulsion purposes, and then adapted to land-based applications. It has three parts: the reactor coolant system, the steam generator and the condenser. The Steam generator (a yellow area in Figure 1) is a shell and tube heat exchanger with high-pressure primary water passing through the tube side and lower pressure secondary feed water as well as steam passing through the shell side. Therefore, a key issue in increasing the efficiency of heat exchanger is to improve the design of steam generator, which is directly translated into economic benefits. The past research works show that the presence of a pin-fin array in a channel enhances the heat transfer significantly. Hence, using microfabrication techniques, such as LIGA, micro-molding or electroplating, some special microstructures can be fabricated around the tubes in the heat exchanger to increase the heat-exchanging efficiency and reduce the overall size of the heat-exchanger for the given heat transfer rates. In this paper, micro-pin fins of different densities made of SU-8 photoresist are fabricated and studied to evaluate overall heat transfer efficiency. The results show that there is an optimized micro pin-fin configuration that has the best overall heat transfer effects

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

    International Nuclear Information System (INIS)

    Rice, R.E.

    1976-09-01

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

  17. 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

  18. 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

  19. 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)

  20. Fuel-to-cladding heat transfer coefficient into reactor fuel element

    International Nuclear Information System (INIS)

    Lassmann, K.

    1979-01-01

    Models describing the fuel-to-cladding heat transfer coefficient in a reactor fuel element are reviewed critically. A new model is developed with contributions from solid, fluid and radiation heat transfer components. It provides a consistent description of the transition from an open gap to the contact case. Model parameters are easily available and highly independent of different combinations of material surfaces. There are no restrictions for fast transients. The model parameters are fitted to 388 data points under reactor conditions. For model verification another 274 data points of steel-steel and aluminium-aluminium interfaces, respectively, were used. The fluid component takes into account peak-to-peak surface roughnesses and, approximatively, also the wavelengths of surface roughnesses. For minor surface roughnesses normally prevailing in reactor fuel elements the model asymptotically yields Ross' and Stoute's model for the open gap, which is thus confirmed. Experimental contact data can be interpreted in very different ways. The new model differs greatly from Ross' and Stoute's contact term and results in better correlation coefficients. The numerical algorithm provides an adequate representation for calculating the fuel-to-cladding heat transfer coefficient in large fuel element structural analysis computer systems. (orig.) [de

  1. Heat Transfer Analysis of Methane Hydrate Sediment Dissociation in a Closed Reactor by a Thermal Method

    Directory of Open Access Journals (Sweden)

    Mingjun Yang

    2012-05-01

    Full Text Available The heat transfer analysis of hydrate-bearing sediment involved phase changes is one of the key requirements of gas hydrate exploitation techniques. In this paper, experiments were conducted to examine the heat transfer performance during hydrate formation and dissociation by a thermal method using a 5L volume reactor. This study simulated porous media by using glass beads of uniform size. Sixteen platinum resistance thermometers were placed in different position in the reactor to monitor the temperature differences of the hydrate in porous media. The influence of production temperature on the production time was also investigated. Experimental results show that there is a delay when hydrate decomposed in the radial direction and there are three stages in the dissociation period which is influenced by the rate of hydrate dissociation and the heat flow of the reactor. A significant temperature difference along the radial direction of the reactor was obtained when the hydrate dissociates and this phenomenon could be enhanced by raising the production temperature. In addition, hydrate dissociates homogeneously and the temperature difference is much smaller than the other conditions when the production temperature is around the 10 °C. With the increase of the production temperature, the maximum of ΔToi grows until the temperature reaches 40 °C. The period of ΔToi have a close relation with the total time of hydrate dissociation. Especially, the period of ΔToi with production temperature of 10 °C is twice as much as that at other temperatures. Under these experimental conditions, the heat is mainly transferred by conduction from the dissociated zone to the dissociating zone and the production temperature has little effect on the convection of the water in the porous media.

  2. Saclay Reactor: acquired knowledge by two years experience in heat transfer using compressed gas

    International Nuclear Information System (INIS)

    Yvon, J.

    1955-01-01

    Describes the conception and functioning of a new reactor (EL-2) using compressed gas as primary coolant. The aim of the use of compressed gas as primary coolant is to reduce the quantity of heavy water used in the functioning of the reactor. Description of the reactor vessel (dimensions, materials, reflector and protection). Description of the cells and the circulation of the gas within the cells. A complete explanation of the control and regulating of the reaction by the ionization chamber is given. Heavy water is used as modulator: it describes the heavy water system and its recombination system. The fuel slugs are cooled by compressed gas: its system is described as well as the blower and the heat exchanger system. Water is supplied by a cooling tower which means the reactor power is dependant of the atmospheric conditions. Particular attention has been given to the tightness of the different systems used. The relation between neutron flow and the thermal output is discussed: the thermal output can be calculated by measuring the gas flow and its heating or by measuring the neutron flow within the reactor, both methods gives closed results. Reactivity study: determination of the different factors which induce a variation of reactivity. Heat transfer: discussion on the use of different heat transfer systems, determination of the required chemical and physical properties of the primary coolant as well as the discussion of the nuclear and thermal requirements for the choice of it. A comparison between the use of nitrogen and carbon dioxide gas shows an advantage in using nitrogen with the existing knowledge. Reflexion on the relevance of this work and the future perspectives of the use of compressed gas as primary coolant. (M.P.)

  3. An experimental investigation of heat transfer from a reactor fuel channel to surrounding water

    International Nuclear Information System (INIS)

    Gillespie, G.E.

    An important feature of the CANDU-PHW reactor is that each fuel channel is surrounded by cool heavy-water moderator that can act as a sink for heat generated in the fuel if other means of heat removal were to fail. During postulated loss-of-coolant accidents there are two scenarios in which the primary cooling system may not prevent fuel-channel overheating. These situations arise when: (1) for a particular break size and location, called the critical break, the coolant flow through a portion of the reactor core stagnates before the emergency coolant injection system restores circulation, or, (2) the emergency coolant injection system fails to operate. In either case, the heat generated in the fuel is transferred mainly by radiation to the pressure tube and calandria tube, and then by boiling heat transfer to the moderator. This paper describes a simple one-dimensional model developed to analyse the thermal behaviour of a fuel channel when the internal pressure is high. Also described is a series of experiments in which the pressure-tube segment is pressurized and heated at a constant rate until it contacts a surrounding calandria-tube segment. Predictions of the one-dimensional model are compared with the experimental results

  4. Numerical Simulation of a Coolant Flow and Heat Transfer in a Pebble Bed Reactor

    International Nuclear Information System (INIS)

    In, Wang-Kee; Kim, Min-Hwan; Lee, Won-Jae

    2008-01-01

    Pebble Bed Reactor(PBR) is one of the very high temperature gas cooled reactors(VHTR) which have been reviewed in the Generation IV International Forum as potential sources for future energy needs, particularly for a hydrogen production. The pebble bed modular reactor(PBMR) exhibits inherent safety features due to the low power density and the large amount of graphite present in the core. PBR uses coated fuel particles(TRISO) embedded in spherical graphite fuel pebbles. The fuel pebbles flow down through the PBR core during a reactor operation and the coolant flows around randomly distributed spheres. For the reliable operation and the safety of the PBR, it is important to understand the coolant flow structure and the fuel pebble temperature in the PBR core. There have been few experimental and numerical studies to investigate the fluid and heat transfer phenomena in the PBR core. The objective of this paper is to predict the fluid and heat transfer in the PBR core. The computational fluid dynamics (CFD) code, STAR-CCM+(V2.08) is used to perform the CFD analysis using the design data for the PBMR400

  5. Modeling of Heat Transfer in the Helical-Coil Heat Exchanger for the Reactor Facility "UNITERM"

    Directory of Open Access Journals (Sweden)

    V. I. Solonin

    2014-01-01

    Full Text Available Circuit heat sink plays an important role in the reactor system. Therefore it imposes high requirements for quality of determining thermal-hydraulic parameters. This article is aimed at modeling of heat exchange process of the helical-coil heat exchanger, which is part of the heat sink circuit of the reactor facility "UNITERM."The simulation was performed using hydro-gas-dynamic software package ANSYS CFX. Computational fluid dynamics of this package allows us to perform calculations in a threedimensional setting, giving an idea of the fluid flow nature. The purpose of the simulation was to determine the parameters of the helical-coil heat exchanger (temperature, velocity at the outlet of the pipe and inter-tubular space, pressure drop, and the nature of the fluid flow of primary and intermediate coolants. Geometric parameters of the model were determined using the preliminary calculations performed by the criterion equations. In calculations Turbulence models k-ε RNG, Shear Stress Transport (SST are used. The article describes selected turbulence models, and considers relationship with wall function.The calculation results allow us to give the values obtained for thermal-hydraulic parameters, to compare selected turbulence models, as well as to show distribution patterns of the coolant temperature, pressure, and velocity at the outlet of the intermediate cooler.Calculations have shown that:- maximum values of primary coolant temperature at the outlet of the heat exchanger surface are encountered in the space between the helical-coil tubes;- higher temperatures of intermediate coolant at the outlet of the coils (in space of helicalcoil tubes are observed for the peripheral row;- primary coolant movement in the inter-tubular space of helical-coil surface is formed as a spiral flow, rather than as a in-line tube bank cross flow.

  6. Pulse Star Inertial Confinement Fusion Reactor: Heat transfer loop and balance-of-plant considerations

    International Nuclear Information System (INIS)

    McDowell, M.W.; Blink, J.A.; Curlander, K.A.

    1983-01-01

    A conceptual heat transfer loop and balance-of-plant design for the Pulse Star Inertial Confinement Fusion Reactor has been investigated and the results are presented. The Pulse Star reaction vessel, a perforated steel bell jar about11 m in diameter, is immersed in Li 17 Pb 83 coolant, which flows through the perforations and forms a 1.5-m-thick plenum of droplets around a 8-m-diameter inner chamber. The bell jar and associated pumps, piping, and steam generators are contained within a 17-m-diameter pool of Li 17 Pb 83 coolant to minimize structural requirements and occupied space, resulting in reduced cost. Four parallel heat transfer loops, each with a flow rate of 5.5 m 3 /s, are necessary to transfer 3300 MWt of power. Liquid metal is pumped to the top of the pool, where it flows downward through eight vertical steam generators. Double-walled tubes are used in the steam generators to assure tritium containment without intermediate heat transfer loops. Each pump is a mixed flow type and has a required NPSH of 3.4 m, a speed of 278 rpm, and an impeller diameter of 1.2 m. The steam generator design was optimized by finding the most cost-effective combination of heat exchanger area and pumping power. The design minimizes the total cost (heat exchanger area plus pumping) for the plant lifetime. The power required for the pumps is 36 MWe. Each resulting steam generator is 12 m high and 1.6 m in diameter, with 2360 tubes. The steam generators and pumps fit easily in the pool between the reactor chamber and the pool wall

  7. Pulse*Star Inertial Confinement Fusion Reactor: heat transfer loop and balance of plant considerations

    International Nuclear Information System (INIS)

    McDowell, M.W.; Murray, K.A.

    1984-01-01

    A conceptual heat transfer loop and balance of plant design for the Pulse*Star Inertial Confinement Fusion Reactor has been investigated and results are presented. The Pulse*Star reaction vessel, a perforated steel bell jar approximately 11 m in diameter, is immersed in Li 17 Pb 83 coolant which flows through the perforations and forms a 1.5 m thick plenum of droplets around an 8 m diameter inner chamber. The reactor and associated pumps, piping, and steam generators are contained within a 17 m diameter pool of Li 17 Pb 83 coolant to minimize structural requirements and occupied space, resulting in reduced cost. Four parallel heat transfer loops with flow rates of 5.5 m 3 /s each are necessary to transfer 3300 MWt of power. The steam generator design was optimized by finding the most cost-effective combination of heat exchanger area and pumping power. Power balance calculations based on an improved electrical conversion efficiency revealed a net electrical output of 1260 MWe to the bus bar and a resulting net efficiency of 39%. Suggested balance-of-plant layouts are also presented

  8. Heat transfer tests of ribbed surfaces for gas-cooled reactors

    International Nuclear Information System (INIS)

    Klepper, O.H.

    1975-07-01

    The performance of gas-cooled reactors is often limited by the heat transfer in the reactor core. Means for modifying core heat transfer surfaces to improve their performance were investigated. The 0.3-in.-OD stainless steel clad heater rods were photo-etched to produce external ribs 0.006 in. high and 0.12 in. wide with a pitch of 0.072 in. Helical ribs with a helix angle of 37 0 (to promote interchannel flow mixing in a multirod array) were provided on one surface. For comparison purposes, a transversely ribbed surface and a smooth rod were also studied. The test surfaces were 49 in. long with a 24-in. heated region, concentrically arranged inside a smooth 0.602-in.-ID stainless steel tube. Nitrogen gas at pressures up to 400 psig was used as the coolant; the linear heat rating ranged to 6.8 kW/ft at surface temperatures up to 1400 0 F; T/sub w/T/sub b/ varied from 1.2 to 2.4 at Re values up to 450,000. Annulus results were recalculated for rod geometry using two different transformations. Good agreement was observed with applicable literature values. The effectiveness of the surfaces was assessed as the ratio E of the heat transfer coefficients of the roughened rods to that of a smooth rod at the same pumping power. The effectiveness of the spiral ribs ranged from 1.3 to 1.4, and from 1.2 to 1.4 for the transverse ribs, spanning Re values from 60,000 to 400,000. These data include variations introduced by alternate transformation methods that were used to make annulus test results applicable to rod geometry. The surfaces investigated in these tests were considered for fast gas-cooled reactors; however, the range of parameters studied also applies to heat transfer from ribbed rod-type fuel elements in thermal gas-cooled reactors. (U.S.)

  9. Perspectives of heat transfer enhancement in nuclear reactors toward nanofluids applications

    International Nuclear Information System (INIS)

    Rocha, Marcelo S.; Cabral, Eduardo L.L.; Sabundjian, Gaiane

    2013-01-01

    Nanofluids are colloidal suspensions of nanoparticles in a base fluid with interesting physical properties and large potential for heat transfer enhancement in thermal systems among other applications. There are an increasing number of nanofluids investigations concerning many aspects of synthesis and fabrication technologies, physical properties, and special applications. Results demonstrate that physical properties like high thermal conductivities and high critical heat flux (CHF) of some nanofluids classifies them as potential working fluids for high heat flux transportation in special systems, including thermal management of microelectronic devices (MEMS) and nuclear reactors. Understanding the importance of such investigations for the knowledge development of nuclear engineering a new research is being conducted at the Nuclear Engineering Center (CEN) of the Nuclear and Energy Research Institute (IPEN/CNEN-SP) to analyze the application potentiality of some nanofluids in nuclear systems for heat transfer enhancement under ionizing radiation influence. In this work a revision of theoretical and experimental studies of nanofluids is performed and its potentiality for using in future generations of nuclear reactors is highlighted showing the status of the research at present. (author)

  10. Numerical modelling of heat and mass transfer in adsorption solar reactor of ammonia on active carbon

    Science.gov (United States)

    Aroudam, El. H.

    In this paper, we present a modelling of the performance of a reactor of a solar cooling machine based carbon-ammonia activated bed. Hence, for a solar radiation, measured in the Energetic Laboratory of the Faculty of Sciences in Tetouan (northern Morocco), the proposed model computes the temperature distribution, the pressure and the ammonia concentration within the activated carbon bed. The Dubinin-Radushkevich formula is used to compute the ammonia concentration distribution and the daily cycled mass necessary to produce a cooling effect for an ideal machine. The reactor is heated at a maximum temperature during the day and cool at the night. A numerical simulation is carried out employing the recorded solar radiation data measured locally and the daily ambient temperature for the typical clear days. Initially the reactor is at ambient temperature, evaporating pressure; Pev=Pst(Tev=0 ∘C) and maintained at uniform concentration. It is heated successively until the threshold temperature corresponding to the condensing pressure; Pcond=Pst(Tam) (saturation pressure at ambient temperature; in the condenser) and until a maximum temperature at a constant pressure; Pcond. The cooling of the reactor is characterised by a fall of temperature to the minimal values at night corresponding to the end of a daily cycle. We use the mass balance equations as well as energy equation to describe heat and mass transfer inside the medium of three phases. A numerical solution of the obtained non linear equations system based on the implicit finite difference method allows to know all parameters characteristic of the thermodynamic cycle and consider principally the daily evolution of temperature, ammonia concentration for divers positions inside the reactor. The tube diameter of the reactor shows the dependence of the optimum value on meteorological parameters for 1 m2 of collector surface.

  11. Heat Transfer in Pebble-Bed Nuclear Reactor Cores Cooled by Fluoride Salts

    Science.gov (United States)

    Huddar, Lakshana Ravindranath

    With electricity demand predicted to rise by more than 50% within the next 20 years and a burgeoning world population requiring reliable emissions-free base-load electricity, can we design advanced nuclear reactors to help meet this challenge? At the University of California, Berkeley (UCB) Fluoride-salt-cooled High Temperature Reactors (FHR) are currently being investigated. FHRs are designed with better safety and economic characteristics than conventional light water reactors (LWR) currently in operation. These reactors operate at high temperature and low pressure making them more efficient and safer than LWRs. The pebble-bed FHR (PB-FHR) variant includes an annular nuclear reactor core that is filled with randomly packed pebble fuel. It is crucial to characterize the heat transfer within this unique geometry as this informs the safety limits of the reactor. The work presented in this dissertation focused on furthering the understanding of heat transfer in pebble-bed nuclear reactor cores using fluoride salts as a coolant. This was done through experimental, analytical and computational techniques. A complex nuclear system with a coolant that has never previously been in commercial use requires experimental data that can directly inform aspects of its design. It is important to isolate heat transfer phenomena in order to understand the underlying physics in the context of the PB-FHR, as well as to make decisions about further experimental work that needs to be done in support of developing the PB-FHR. Certain organic oils can simulate the heat transfer behaviour of the fluoride salt if relevant non-dimensional parameters are matched. The advantage of this method is that experiments can be done at a much lower temperature and at a smaller geometric scale compared to FHRs, thereby lowering costs. In this dissertation, experiments were designed and performed to collect data demonstrating similitude. The limitations of these experiments were also elucidated by

  12. Integrated conjugate heat transfer analysis method for in-vessel retention with external reactor vessel cooling - 15477

    International Nuclear Information System (INIS)

    Park, J.W.; Bae, J.H.; Seol, W.C.

    2015-01-01

    An integrated 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. The method calculates steady-state 3-dimensional temperature distribution of a reactor vessel using coupled conjugate heat transfer between in-vessel 3-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 3-layer corium heat transfer model is utilizing lumped-parameter thermal-resistance circuit method and ex-vessel boiling regimes are parametrically considered. The thermal integrity of a reactor vessel is addressed in terms of un-molten thickness profile. The vessel 3-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 inside the oxide pool and the 3-dimensional vessel heat transfer provides a much larger minimum vessel wall thickness. (authors)

  13. Heat transfer in the lithium-cooled blanket of a pulsed fusion reactor

    International Nuclear Information System (INIS)

    Cort, G.E.; Krakowski, R.A.

    1978-01-01

    The transient temperature distribution in the lithium-cooled blanket of a pulsed fusion reactor has been calculated using a finite-element heat-conduction computer program. An auxiliary program was used to predict the coolant transient velocity in a network of parallel and series flow passages with constant driving pressure and varying magnetic field. The coolant velocity was calculated by a Runge-Kutta numerical integration of the conservation equations. The lithium coolant was part of the finite-element heat-conduction mesh with the velocity terms included in the total matrix. The matrix was solved implicitly at each time step for the nodal point temperatures. Slug flow was assumed in the coolant passages and the Boussinesq analogy was used to calculate turbulent heat transfer when the magnetic field was not present

  14. Expandable antivibration bar for heat transfer tubes of a pressurized water reactor steam generator

    International Nuclear Information System (INIS)

    Appleman, R.H.

    1985-01-01

    An expandable antivibration bar for use in stabilizing the U-bend portion of heat transfer tubes in a pressurized water reactor steam generator comprises two adjustable rods connected together by an arcuate connector. The two adjustable rods preferably comprise two mating rod sections having complementary angular sliding surfaces thereon, with means provided to move the rod sections relative to each other along the sliding surfaces so as to expand the rods from a first mated cross-sectional width to a second larger cross-sectional width. The ends of the rod sections have means for aligning the two rod sections and maintaining them in alignment during expansion. (author)

  15. Evaluation of radiation heat transfer in porous medial: Application for a pebble bed modular reactor cooled by CO2 gas

    Directory of Open Access Journals (Sweden)

    Sidi-Ali Kamel

    2013-01-01

    Full Text Available This work analyses the contribution of radiation heat transfer in the cooling of a pebble bed modular reactor. The mathematical model, developed for a porous medium, is based on a set of equations applied to an annular geometry. Previous major works dealing with the subject have considered the forced convection mode and often did not take into account the radiation heat transfer. In this work, only free convection and radiation heat transfer are considered. This can occur during the removal of residual heat after shutdown or during an emergency situation. In order to derive the governing equations of radiation heat transfer, a steady-state in an isotropic and emissive porous medium (CO2 is considered. The obtained system of equations is written in a dimensionless form and then solved. In order to evaluate the effect of radiation heat transfer on the total heat removed, an analytical method for solving the system of equations is used. The results allow quantifying both radiation and free convection heat transfer. For the studied situation, they show that, in a pebble bed modular reactor, more than 70% of heat is removed by radiation heat transfer when CO2 is used as the coolant gas.

  16. A Computational Fluid Dynamic and Heat Transfer Model for Gaseous Core and Gas Cooled Space Power and Propulsion Reactors

    Science.gov (United States)

    Anghaie, S.; Chen, G.

    1996-01-01

    A computational model based on the axisymmetric, thin-layer Navier-Stokes equations is developed to predict the convective, radiation and conductive heat transfer in high temperature space nuclear reactors. An implicit-explicit, finite volume, MacCormack method in conjunction with the Gauss-Seidel line iteration procedure is utilized to solve the thermal and fluid governing equations. Simulation of coolant and propellant flows in these reactors involves the subsonic and supersonic flows of hydrogen, helium and uranium tetrafluoride under variable boundary conditions. An enthalpy-rebalancing scheme is developed and implemented to enhance and accelerate the rate of convergence when a wall heat flux boundary condition is used. The model also incorporated the Baldwin and Lomax two-layer algebraic turbulence scheme for the calculation of the turbulent kinetic energy and eddy diffusivity of energy. The Rosseland diffusion approximation is used to simulate the radiative energy transfer in the optically thick environment of gas core reactors. The computational model is benchmarked with experimental data on flow separation angle and drag force acting on a suspended sphere in a cylindrical tube. The heat transfer is validated by comparing the computed results with the standard heat transfer correlations predictions. The model is used to simulate flow and heat transfer under a variety of design conditions. The effect of internal heat generation on the heat transfer in the gas core reactors is examined for a variety of power densities, 100 W/cc, 500 W/cc and 1000 W/cc. The maximum temperature, corresponding with the heat generation rates, are 2150 K, 2750 K and 3550 K, respectively. This analysis shows that the maximum temperature is strongly dependent on the value of heat generation rate. It also indicates that a heat generation rate higher than 1000 W/cc is necessary to maintain the gas temperature at about 3500 K, which is typical design temperature required to achieve high

  17. Turbulent heat transfer in a coolant channel of a pressurized water reactor (PWR) core

    International Nuclear Information System (INIS)

    Kumar, Sanjeev; Saha, Arun K.; Munshi, Prabhat

    2016-01-01

    Exact predictions in nuclear reactors are more crucial, because of the safety aspects. It necessitates the appropriate modeling of heat transfer phenomena in the reactors core. A two-dimensional thermal-hydraulics model is used to study the detailed analysis of the coolant region of a fuel pin. Governing equations are solved using Marker and Cell (MAC) method. Standard wall functions k-ε turbulence model is incorporated to consider the turbulent behaviour of the flow field. Validation of the code and a few results for a typical PWR running at normal operating conditions reported earlier. There were some discrepancies in the old calculations. These discrepancies have been resolved and updated results are presented in this work. 2D thermal-hydraulics model results have been compared with the 1D thermal-hydraulics model results and conclusions have been drawn. (author)

  18. Study of heat transfer in 3D fuel rods of the EPRI-9R reactor modified

    International Nuclear Information System (INIS)

    Affonso, Renato Raoni Werneck; Lava, Deise Diana; Borges, Diogo da Silva; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes

    2014-01-01

    This paper aims to conduct a case study of the fuel rods that have the highest and the lowest average power of the EPRI-9R 3D reactor modified , for various positions of the control rods banks. For this, will be addressed the verification of computer code, comparing the results obtained with analytical solutions. This check is important so that, subsequently, it is possible use the program to understand the behavior of the fuel rods and the coolant channel of the EPRI-9R 3D reactor modified. Thus, in view of the scope of this paper, first a brief introducing on the heat transfer is done, including the rod equations and the equation of energy in the channel to allow the analysis of the results

  19. Enhancement of nuclear heat transfer in a typical pressurized water reactor by new spacer grids

    International Nuclear Information System (INIS)

    Nazifi, M.; Nematollahi, M.

    2007-01-01

    The fuel element geometry typically used in nuclear reactor is rod bundle whose rod-to-rod clearance is maintained by grid spacer. The heat generated in the rod by nuclear reaction is removed by coolant, usually in turbulent flow. The coolant moves axially through the subchannels. Fuel spacer grid affects the coolant flow distribution in a fuel rod bundle, and so spacer geometry has a strong influence on a bundle's thermal-hydraulic characteristics such as critical heat flux and pressure drop. An understanding of the detailed structure of the turbulent flow and heat transfer in the rod bundle, used especially as nuclear fuel elements, is of major interest to the nuclear power industry for their safe and reliable operation. The flow mixing devices on grid spacer would enhance the mixing rate between sub-channels and promote the turbulence in subchannel. The present study evaluates the effects of mixing vane shape on flow structure and heat transfer downstream of mixing vane in a sub-channel of fuel assembly, by obtaining velocity and pressure fields, turbulent intensity, flow mixing factors, heat transfer coefficient and friction factor using three-dimensional RANS analysis. Six new shapes mixing vane designed by the authors, are simulated numerically to evaluate the performance in enhancing the heat transfer, in comparison with commercialized split vane. Standard K-epsilon model are used as a turbulence closure model and periodic and symmetry condition are set as boundary conditions. The capability of the model to predict the coolant flow distribution inside rod bundles is shown and discussed on the base of comparison with experimental data for a variety of geometrical and Reynolds number conditions. It is conformed that the turbulence in the sub-channel was significantly promoted by spacer and mixing devices but rapidly decreased to a fully developed level approximately 10 time of hydraulic diameter downstream of the top of spacer. Ring type mixer showed a high

  20. Heat transfer. V. 1

    International Nuclear Information System (INIS)

    1992-01-01

    This volume contains the 4 key-note lectures and 83 of the 148 papers presented at the 3rd UK National Conference on Heat Transfer. The papers are grouped under the following broad headings: boiling and condensation; heat exchangers; refrigeration and air-conditioning; natural convection; process safety and nuclear reactors; two-phase flow; post dry-out; combustion, radiation and chemical reaction. Separate abstracts have been prepared for 13 papers of relevance to nuclear reactors. (UK)

  1. Studies of thermal hydraulics and heat transfer in cascade subcritical molten salt reactor

    International Nuclear Information System (INIS)

    Aysen, E.M.; Sedov, A.A.; Subbotin, A.S.

    2005-01-01

    Full text of publication follows: Cascade Subcritical Molten Salt Reactor (CSMSR) consists of three main parts: accelerator-driven proton-bombarded target, central and peripheral zones. External neutrons generated in the result of interaction of protons with the target nuclei are multiplied then in the central zone and leak farther into the peripheral reactor zone, where an efficient burning of Minor Actinides dissolved in a molten salt fluoride composition is produced. The bunch of target and two zones is designed so that preset subcriticality of reactor would not be less than 1% of k eff . A characteristic feature of the reactor is a high density of neutron flux (2.10 15 n/cm 2 s) in the central zone and target and very high volumetric power rate (2000 - 6000 W/cm 3 ) in all the parts of CSMSR. To provide a workability of the core structures under condition of so big level of power rate it is necessary to impose strict limitations on the temperatures and temperature gradients developed in the coolants and constructions. In this reason it has been arranged a calculational-designing study to reveal the problems of heat transfer in the coolant and core structures and to find more appropriate variant of the core and target design, which is a compromise of contradictory requirements: provision of high neutron flux and coolability of the core structures. In this paper the results of studies of thermal hydraulics and heat transfer in the core zones and proton-beam target are presented. Different variants of the target and central zone design as well as application of different kind of coolants in them are discussed and the main problems of heat removal in their structures are analyzed. Multidimensional fields of velocity and temperature got in thermal hydraulics calculations for free flow of fuelled molten salt in cylindrical-cave peripheral CSMSR zone without structures inside are demonstrated. The role of turbulent exchange of momentum and heat for free flow in the

  2. Tests of the heat transfer characteristic of air cooler during cooling by natural convection of the Fast Breeder Reactor plants

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    The purpose of this study is to confirm the heat transfer characteristics of the air cooler (AC) of the Fast Breeder Reactor(FBR) which has a function to remove the residual heat of the reactor by heat exchange between sodium and air in natural convection region if electric power would be lost. In order to confirm the characteristics of the AC installed in the FBR plant, the heat transfer test by using the AC which is installed in the sodium test loop owned by Toshiba Corporation has been planned. In this study, the heat transfer characteristic tests were performed by using the AC in sodium test loop, and the CFD analyses were conducted to evaluate the test results and the heat transfer characteristics of the plant scale AC at the condition of natural convection. In addition, the elemental tests to confirm the influence of the heat transfer tube placement by using the heat transfer tube of the same specification as the AC of Monju were performed. (author)

  3. 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)

  4. Improving the understanding of thermal-hydraulics and heat transfer for super critical water cooled reactors

    International Nuclear Information System (INIS)

    Bilbao y Leon, S.; Aksan, N.

    2010-01-01

    Ensuring the exchange of information and fostering the collaboration among Member States on the development of technology advances for future nuclear power plants are among the key roles of the IAEA. There is high interest internationally in both developing and industrialized countries in the design of innovative super-critical water-cooled reactors (SCWRs). This interest arises from the high thermal efficiencies (44-45%) and improved economic competitiveness promised by for this concept, utilizing and building on the recent developments of highly efficient fossil power plants. The SCWR is one of the six concepts included in the Generation-IV International Forum (GIF). Following the advice of the IAEA Nuclear Energy Dept.'s Technical Working Groups on Advanced Technologies for LWRs and HWRs (the TWG-LWR and TWG-HWR), with the feedback from the Gen-IV SCWR Steering Committee, and in coordination with the OECD-NEA, IAEA is working on a Coordinated Research Project (CRP) in the areas of heat transfer behaviour and testing of thermo-hydraulic computer methods for Supercritical Water-Cooled Reactors. The second Research Coordination Meeting (RCM) of the CRP was held at the IAEA Headquarters, in Vienna (Austria)) in August 2009. This paper summarizes the current status of the CRP, as well as the major achievements to date. (authors)

  5. Expandable antivibration bar for heat transfer tubes of a pressurized water reactor steam generator

    International Nuclear Information System (INIS)

    Appleman, R.H.

    1987-01-01

    This patent describes a pressurized water reactor steam generator having spaced rows of heat transfer tubes through which primary coolant from the reactor flows, the tubes being of a U-shaped design, with the U-bend portions of the U-shaped tubes stabilized by antivibration bars. The improvement described here comprises expandable antivibration bars for stabilizing the U-bend portions of the U-shaped tubes, the expandable bars having a pair of adjustable rods, formed from a pair of rod sections affixed to a connector, one rod section of each of the pair of rod sections having a plurality of protrusions. Each of the protrusions has slidable surfaces thereon. The other rod section of each of the pair of rod sections has indentations, each of the indentations having slidable surfaces thereon complementary to the sliding surfaces of the protrusions, such that the rods are expandable from a first cross-sectional width less than the spacing between two adjacent rows of the tubes, to a second cross-sectional width greater than the first cross-sectional width. The expanded rods are adapted to contact tubes of the two adjacent rows of the tubes

  6. Oxygen transport membrane system and method for transferring heat to catalytic/process reactors

    Science.gov (United States)

    Kelly, Sean M; Kromer, Brian R; Litwin, Michael M; Rosen, Lee J; Christie, Gervase Maxwell; Wilson, Jamie R; Kosowski, Lawrence W; Robinson, Charles

    2014-01-07

    A method and apparatus for producing heat used in a synthesis gas production is provided. The disclosed method and apparatus include a plurality of tubular oxygen transport membrane elements adapted to separate oxygen from an oxygen containing stream contacting the retentate side of the membrane elements. The permeated oxygen is combusted with a hydrogen containing synthesis gas stream contacting the permeate side of the tubular oxygen transport membrane elements thereby generating a reaction product stream and radiant heat. The present method and apparatus also includes at least one catalytic reactor containing a catalyst to promote the stream reforming reaction wherein the catalytic reactor is surrounded by the plurality of tubular oxygen transport membrane elements. The view factor between the catalytic reactor and the plurality of tubular oxygen transport membrane elements radiating heat to the catalytic reactor is greater than or equal to 0.5.

  7. Oxygen transport membrane system and method for transferring heat to catalytic/process reactors

    Science.gov (United States)

    Kelly, Sean M.; Kromer, Brian R.; Litwin, Michael M.; Rosen, Lee J.; Christie, Gervase Maxwell; Wilson, Jamie R.; Kosowski, Lawrence W.; Robinson, Charles

    2016-01-19

    A method and apparatus for producing heat used in a synthesis gas production process is provided. The disclosed method and apparatus include a plurality of tubular oxygen transport membrane elements adapted to separate oxygen from an oxygen containing stream contacting the retentate side of the membrane elements. The permeated oxygen is combusted with a hydrogen containing synthesis gas stream contacting the permeate side of the tubular oxygen transport membrane elements thereby generating a reaction product stream and radiant heat. The present method and apparatus also includes at least one catalytic reactor containing a catalyst to promote the steam reforming reaction wherein the catalytic reactor is surrounded by the plurality of tubular oxygen transport membrane elements. The view factor between the catalytic reactor and the plurality of tubular oxygen transport membrane elements radiating heat to the catalytic reactor is greater than or equal to 0.5

  8. Transient heat transfer in a directly-irradiated solar chemical reactor for the thermal dissociation of ZnO

    International Nuclear Information System (INIS)

    Mueller, R.; Lipinski, W.; Steinfeld, A.

    2008-01-01

    A numerical and experimental investigation is carried out in a solar thermochemical reactor for the thermal dissociation of ZnO at 2000 K using concentrated solar energy. The reactor consists of a cavity-receiver lined with ZnO particles and directly exposed to high-flux irradiation. A transient heat transfer model is formulated to link the rate of radiation, convection, and conduction heat transfer to the reaction kinetics. The radiosity and Monte Carlo methods are applied to obtain the distribution of net radiative fluxes at the internal surfaces of the reactor cavity and at the surface of the ZnO bed. Validation is accomplished in terms of the calculated and measured transient temperature profiles and chemical reaction rates

  9. Evaluation of an experiment modelling heat transfer from the melt pool for use in VVER 440/213 reactors

    International Nuclear Information System (INIS)

    Skop, J.

    2003-12-01

    The strategy of confining core melt within the reactor vessel is among promising strategies to mitigate severe accidents of VVER 440/213 reactors. This strategy consists in residual heat removal from the melt by external vessel cooling from the outside, using water from the flooded reactor downcomer. This approach can only be successful if the critical heat flux on the external vessel surface is not exceeded. This can be assessed based on the parameters of heat transfer from the core melt pool in the conditions of natural circulation within the pool. Those parameters are the subject of the report. A basic description of the terms and physical basis of the strategy of confining core melt inside the vessel is given in Chapter 2, which also briefly explains similarity theory, based on which the results obtained on experimental facilities, using simulation materials, can be related to the actual situation inside a real reactor. Chapter 3 presents an overview of experimental work addressing the characteristics of heat transfer from the core melt pool in natural circulation conditions and a description of the experimental facilities. An overview of the results emerging from the experiments and their evaluation with respect to their applicability to reactors in Czech nuclear power plants are given in Chapter 4

  10. Development of heat transfer enhancement techniques for external cooling of an advanced reactor vessel

    Science.gov (United States)

    Yang, Jun

    Nucleate boiling is a well-recognized means for passively removing high heat loads (up to ˜106 W/m2) generated by a molten reactor core under severe accident conditions while maintaining relatively low reactor vessel temperature (Critical Heat Flux (CHF), becomes the key to the success of external passive cooling of reactor vessel undergoing core disrupture accidents. In the present study, two boiling heat transfer enhancement methods have been proposed, experimentally investigated and theoretically modelled. The first method involves the use of a suitable surface coating to enhance downward-facing boiling rate and CHF limit so as to substantially increase the possibility of reactor vessel surviving high thermal load attack. The second method involves the use of an enhanced vessel/insulation design to facilitate the process of steam venting through the annular channel formed between the reactor vessel and the insulation structure, which in turn would further enhance both the boiling rate and CHF limit. Among the various available surface coating techniques, metallic micro-porous layer surface coating has been identified as an appropriate coating material for use in External Reactor Vessel Cooling (ERVC) based on the overall consideration of enhanced performance, durability, the ease of manufacturing and application. Since no previous research work had explored the feasibility of applying such a metallic micro-porous layer surface coating on a large, downward facing and curved surface such as the bottom head of a reactor vessel, a series of characterization tests and experiments were performed in the present study to determine a suitable coating material composition and application method. Using the optimized metallic micro-porous surface coatings, quenching and steady-state boiling experiments were conducted in the Sub-scale Boundary Layer Boiling (SBLB) test facility at Penn State to investigate the nucleate boiling and CHF enhancement effects of the surface

  11. Scaling analysis of the coupled heat transfer process in the high-temperature gas-cooled reactor core

    International Nuclear Information System (INIS)

    Conklin, J.C.

    1986-08-01

    The differential equations representing the coupled heat transfer from the solid nuclear core components to the helium in the coolant channels are scaled in terms of representative quantities. This scaling process identifies the relative importance of the various terms of the coupled differential equations. The relative importance of these terms is then used to simplify the numerical solution of the coupled heat transfer for two bounding cases of full-power operation and depressurization from full-system operating pressure for the Fort St. Vrain High-Temperature Gas-Cooled Reactor. This analysis rigorously justifies the simplified system of equations used in the nuclear safety analysis effort at Oak Ridge National Laboratory

  12. Pressure drop and heat transfer in the sodium to air heat exchanger tube banks on advanced sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kang, H.; Eoh, J.; Cha, J.; Kim, S.

    2011-01-01

    A numerical study was performed to investigate the thermal and hydraulic characteristics and build up design model of the AHX (sodium-to-air heat exchanger) unit of a sodium-cooled fast reactor. Helical-coiled tube banks in the AHX were modeled as porous media and simulated heat and momentum transfer. Two-dimensional flow characteristic appeared at the most region of AHX annulus. Pressure drop and heat transfer coefficient for rectangular, parallelogram and staggered tube banks as the main components of the AHX were evaluated and compared with Zhukauskas empirical correlations. (author)

  13. Radiation heat transfer through the gas of a sodium cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Pradel, P.; Frachet, S.; Petit, D.

    1984-04-01

    Analysis based on results from the COCA test campaign and Germinal mockup of Super Phenix upper shuttings, of the heat transfers and radiation attenuation due to sodium aerosols between the free surface of sodium and the upper shuttings

  14. Development of an ISI robot for the fast breeder reactor MONJU primary heat transfer system piping

    International Nuclear Information System (INIS)

    Tagawa, Akihiro; Ueda, Masashi; Yamashita, Takuya; Narisawa, Masataka; Haga, Kouichi

    2011-01-01

    This paper describes the development of a new inspection robot for the In-Service Inspection of the heat transfer system of the Fast Breeder Reactor MONJU. The inspection was carried out using a tire-type ultrasonic sensor for volumetric tests at high temperature (atmosphere, 55degC; piping surface, 80degC) and radiation exposure condition (dose rate, 10 mGy/h; piping surface dose rate, 15 mGy/h). An inspection robot using a new tire type for the ultrasonic testing sensor and a new control method was developed. A signal-to-noise ratio S/N over 2 was obtained during the functional test for a calibration defect with a depth of 50%t (from the tube wall thickness). In the automatic inspection test, an EDM slit with a depth of 9% from the pipe thickness was detectable and with an S/N ratio = 4.0 (12.0 dB). (author)

  15. Experiments and Modelling Techniques for Heat and Mass Transfer in Light Water Reactors

    International Nuclear Information System (INIS)

    Ambrosini, W.; Bucci, M.; Forgione, N.; Manfredini, A.; Oriolo, F.

    2009-01-01

    The paper summarizes the lesson learned from theoretical and experimental activities performed at the University of Pisa, Pisa, Italy, in past decades in order to develop a general methodology of analysis of heat and mass transfer phenomena of interest for nuclear reactor applications. An overview of previously published results is proposed, highlighting the rationale at the basis of the performed work and its relevant conclusions. Experimental data from different sources provided information for model development and assessment. They include condensation experiments performed at SIET (Piacenza, Italy) on the PANTHERS prototypical PCCS module, falling film evaporation tests for simulating AP600-like outer shell spraying conditions, performed at the University of Pisa, experimental data concerning condensation on finned tubes, collected by CISE (Piacenza, Italy) in the frame of the INCON EU Project, and experimental tests performed in the CONAN experimental facility installed at the University of Pisa. The experience gained in these activities is critically reviewed and discussed to highlight the relevant obtained conclusions and the perspectives for future work

  16. Heat and momentum transfer in a gas coolant flow through a circular pipe in a high temperature gas cooled reactor

    International Nuclear Information System (INIS)

    Ogawa, Masuro

    1989-07-01

    In Japan Atomic Energy Research Institute (JAERI), a very high temperature gas cooled reactor (VHTR) has been researched and developed with a purpose of attaining a coolant temperature of around 1000degC at the reactor outlet. In order to design VHTR, comprehensive knowledge is required on thermo-hydraulic characteristics of laminar-turbulent transition, of coolant flow with large thermal property variation due to temperature difference, and of heat transfer deterioration. In the present investigation, experimental and analytical studies are made on a gas flow in a circular tube to elucidate the thermo-hydraulic characteristics. Friction factors and heat transfer coefficients in transitional flows are obtained. Influence of thermal property variation on the friction factor is qualitatively determined. Heat transfer deterioration in the turbulent flow subjected to intense heating is experimentally found to be caused by flow laminarization. The analysis based on a k-kL two-equation model of turbulence predicts well the experimental results on friction factors and heat transfer coefficients in flows with thermal property variation and in laminarizing flows. (author)

  17. Heat transfer coefficient for lead matrixing in disposal containers for used reactor fuel

    International Nuclear Information System (INIS)

    Mathew, P.M.; Taylor, M.; Krueger, P.A.

    1985-02-01

    In the Canadian Nuclear Fuel Waste Management Program, metal matrices with low melting points are being evaluated for their potential to provide support for the shell of disposal containers for used fuel, and to act as an additional barrier to the release of radionuclides. The metal matrix would be incorporated into the container by casting. To study the heat transfer processes during solidification, a steady-state technique was used, involving lead as the cast metal, to determine the overall heat transfer coefficient between the lead and some of the candidate container materials. The existence of an air gap between the cast lead and the container material appeared to control the overall heat transfer coefficient. The experimental observations indicated that the surface topography of the container material influences the heat transfer and that a smoother surface results in a greater heat transfer than a rough surface. The experimental results also showed an increasing heat transfer coefficient with increasing temperature difference across the container base plates; a model developed to base-plate bending can explain the observed results

  18. Experimental study on the heat transfer characteristics of a nuclear reactor containment wall cooled by gravitationally falling water

    Science.gov (United States)

    Pasek, Ari D.; Umar, Efrison; Suwono, Aryadi; Manalu, Reinhard E. E.

    2012-06-01

    Gravitationally falling water cooling is one of mechanism utilized by a modern nuclear Pressurized Water Reactor (PWR) for its Passive Containment Cooling System (PCCS). Since the cooling is closely related to the safety, water film cooling characteristics of the PCCS should be studied. This paper deals with the experimental study of laminar water film cooling on the containment model wall. The influences of water mass flow rate and wall heat rate on the heat transfer characteristic were studied. This research was started with design and assembly of a containment model equipped with the water cooling system, and calibration of all measurement devices. The containment model is a scaled down model of AP 1000 reactor. Below the containment steam is generated using electrical heaters. The steam heated the containment wall, and then the temperatures of the wall in several positions were measure transiently using thermocouples and data acquisition. The containment was then cooled by falling water sprayed from the top of the containment. The experiments were done for various wall heat rate and cooling water flow rate. The objective of the research is to find the temperature profile along the wall before and after the water cooling applied, prediction of the water film characteristic such as means velocity, thickness and their influence to the heat transfer coefficient. The result of the experiments shows that the wall temperatures significantly drop after being sprayed with water. The thickness of water film increases with increasing water flow rate and remained constant with increasing wall heat rate. The heat transfer coefficient decreases as film mass flow rate increase due to the increases of the film thickness which causes the increasing of the thermal resistance. The heat transfer coefficient increases slightly as the wall heat rate increases. The experimental results were then compared with previous theoretical studied.

  19. Fluid flow and heat transfer investigation of pebble bed reactors using mesh adaptive large-eddy simulation

    International Nuclear Information System (INIS)

    Pavlidis, D.; Lathouwers, D.

    2011-01-01

    A computational fluid dynamics model with anisotropic mesh adaptivity is used to investigate coolant flow and heat transfer in pebble bed reactors. A novel method for implicitly incorporating solid boundaries based on multi-fluid flow modelling is adopted. The resulting model is able to resolve and simulate flow and heat transfer in randomly packed beds, regardless of the actual geometry, starting off with arbitrarily coarse meshes. The model is initially evaluated using an orderly stacked square channel of channel-height-to-particle diameter ratio of unity for a range of Reynolds numbers. The model is then applied to the face-centred cubical geometry. Coolant flow and heat transfer patterns are investigated. (author)

  20. Microcomputer based program for predicting heat transfer under reactor accident conditions. Volume I

    International Nuclear Information System (INIS)

    Cheng, S.C.; Groeneveld, D.C.; Leung, L.K.H.; Wong, Y.L.; Nguyen, C.

    1987-07-01

    A microcomputer based program called Heat Transfer Prediction Software (HTPS) has been developed. It calculates the heat transfer for the tube and bundle geometries for steady state and transient conditions. This program is capable of providing the best estimated of the hot pin temperatures during slow transients for 37- and 28-element CANDU type fuel bundles. The program is designed for an IBM-PC AT/XT (or IBM-PC compatible computer) equipped with a Math Co-processor. The following input parameters are required: pressure, mass flux, hydraulic diameter, and quality. For the steady state case, the critical heat flux (CHF), the critical heat flux temperature, the minimum film boiling temperature, and the minimum film boiling heat flux are the primary outputs. With either the surface heat flux or wall temperature specified, the program determines the heat transfer regime and calculates the surface heat flux, wall temperatures and heat transfer coefficient. For the slow transient case, the pressure, mass flux, quality, and volumetric heat generation rate are the time dependent input parameters required to calculate the hot pin sheath temperatures and surface heat fluxes. A simple routine for generating properties has been developed for light water to support the above program. It contains correlations that have been verified for pressures ranging from 0.6kPa to 30 MPa, and temperatures up to 1100 degrees Celcius. The thermodynamic and transport properties that can be generated from this routine are: density, specific volume, enthalpy, specific heat capacity, conductivity, viscosity, surface tension and Prandtl number for saturated liquid, saturated vapour, subcooled liquid for superheated vapour. A software for predicting flow regime has also been developed. It determines the flow pattern at specific flow conditions, and provides a correction factor for calculating the CHF during partially stratified horizontal flow. The technical bases for the program and its

  1. Microcomputer based program for predicting heat transfer under reactor accident conditions. Volume II

    International Nuclear Information System (INIS)

    Cheng, S.C.; Groeneveld, D.C.; Leung, L.K.H.; Wong, Y.L.; Nguyen, C.

    1987-07-01

    A microcomputer based program called Heat Transfer Prediction Software (HTPS) has been developed. It calculates the heat transfer for tube and bundle geometries for steady state and transient conditions. This program is capable of providing the best estimated of the hot pin temperatures during slow transients for 37- and 28-element CANDU type fuel bundles. The program is designed for an IBM-PC AT/XT (or IBM-PC compatible computer) equipped with a Math Co-processor. The following input parameters are required: pressure, mass flux, hydraulic diameter, and quality. For the steady state case, the critical heat flux (CHF), the critical heat flux temperature, the minimum film boiling temperature, and the minimum film boiling heat flux are the primary outputs. With either the surface heat flux or wall temperature specified, the program determines the heat transfer regime and calculates the surface heat flux, wall temperature and heat transfer coefficient. For the slow transient case, the pressure, mass flux, quality, and volumetric heat generation rate are the time dependent input parameters are required to calculate the hot pin sheath temperatures and surface heat fluxes. A simple routine for generating properties has been developed for light water to support the above program. It contains correlations that have been verified for pressures ranging from 0.6kPa to 30 MPa, and temperatures up to 1100 degrees Celcius. The thermodynamic and transport properties that can be generated from this routine are: density, specific volume, enthalpy, specific heat capacity, conductivity, viscosity, surface tension and Prandtle number for saturated liquid, saturated vapour, subcooled liquid of superheated vapour. A software for predicting flow regime has also been developed. It determines the flow pattern at specific flow conditions, and provides a correction factor for calculating the CHF during partially stratified horizontal flow. The technical bases for the program and its structure

  2. Experimental and numerical investigations of high temperature gas heat transfer and flow in a VHTR reactor core

    Science.gov (United States)

    Valentin Rodriguez, Francisco Ivan

    High pressure/high temperature forced and natural convection experiments have been conducted in support of the development of a Very High Temperature Reactor (VHTR) with a prismatic core. VHTRs are designed with the capability to withstand accidents by preventing nuclear fuel meltdown, using passive safety mechanisms; a product of advanced reactor designs including the implementation of inert gases like helium as coolants. The present experiments utilize a high temperature/high pressure gas flow test facility constructed for forced and natural circulation experiments. This work examines fundamental aspects of high temperature gas heat transfer applied to VHTR operational and accident scenarios. Two different types of experiments, forced convection and natural circulation, were conducted under high pressure and high temperature conditions using three different gases: air, nitrogen and helium. The experimental data were analyzed to obtain heat transfer coefficient data in the form of Nusselt numbers as a function of Reynolds, Grashof and Prandtl numbers. This work also examines the flow laminarization phenomenon (turbulent flows displaying much lower heat transfer parameters than expected due to intense heating conditions) in detail for a full range of Reynolds numbers including: laminar, transition and turbulent flows under forced convection and its impact on heat transfer. This phenomenon could give rise to deterioration in convection heat transfer and occurrence of hot spots in the reactor core. Forced and mixed convection data analyzed indicated the occurrence of flow laminarization phenomenon due to the buoyancy and acceleration effects induced by strong heating. Turbulence parameters were also measured using a hot wire anemometer in forced convection experiments to confirm the existence of the flow laminarization phenomenon. In particular, these results demonstrated the influence of pressure on delayed transition between laminar and turbulent flow. The heat

  3. Theoretical and Numerical Study of Heat Transfer Deterioration in High Performance Light Water Reactor

    Directory of Open Access Journals (Sweden)

    David Palko

    2008-01-01

    Full Text Available A numerical investigation of the heat transfer deterioration (HTD phenomena is performed using the low-Re k-ω turbulence model. Steady-state Reynolds-averaged Navier-Stokes equations are solved together with equations for the transport of enthalpy and turbulence. Equations are solved for the supercritical water flow at different pressures, using water properties from the standard IAPWS (International Association for the Properties of Water and Steam tables. All cases are extensively validated against experimental data. The influence of buoyancy on the HTD is demonstrated for different mass flow rates in the heated pipes. Numerical results prove that the RANS low-Re turbulence modeling approach is fully capable of simulating the heat transfer in pipes with the water flow at supercritical pressures. A study of buoyancy influence shows that for the low-mass flow rates of coolant, the influence of buoyancy forces on the heat transfer in heated pipes is significant. For the high flow rates, buoyancy influence could be neglected and there are clearly other mechanisms causing the decrease in heat transfer at high coolant flow rates.

  4. The effect of water vapor in the reactor cavity in a MHTGR [Modular High Temperature Gas Cooled Reactor] on the radiation heat transfer

    International Nuclear Information System (INIS)

    Cappiello, M.W.

    1991-01-01

    Analyses have been completed to determine the effect of the presence of water vapor in the reactor cavity in a modular high temperature gas cooled reactor on the predicted radiation heat transfer from the vessel wall to the reactor cavity cooling system. The analysis involves the radiation heat transfer between two parallel plates with an absorbing and emitting medium present. Because the absorption in the water vapor is spectrally dependent, the solution is difficult even for simple geometries. A computer code was written to solve the problem using the Monte Carlo method. The code was validated against closed form solutions, and shows excellent agreement. In the analysis of the reactor problem, the results show that the reduction in heat transfer, and the consequent increase in the vessel wall temperature, can be significant. This effect can be cast in terms of a reduction in the wall surface emissivities from 0.8 to 0.59. Because of the insulating effect of the water vapor, increasing the gap distance between the vessel wall and the cooling system will cause the vessel wall temperature to increase further. Care should be taken in the design of the facility to minimize the gap distance and keep temperature increase within allowable limits. 3 refs., 6 figs., 4 tabs

  5. Serpentine tube heat transfer characteristic under accident condition in gas cooled reactors

    International Nuclear Information System (INIS)

    Abouhadra, D.S.; Byrne, J.E.

    2004-01-01

    In nuclear reactors of the Magnox or advanced gas Cooled type, serpentine tubing is used in some designs to generate steam in a once through arrangement. The calculation of accident conditions using two phase flow codes requires knowledge of the heat transfer behavior of the boiler steam side. A series of experiments to study the blowdown characteristics of a typical serpentine boiler section was devised in order to validate the MARTHA section of the MACE code used by nuclear Electric. The tests were carried out on the Thermal Hydraulics Experimental Research Assembly (THERA) loop at Manchester University. The Thermal Hydraulic Experimental Research Assembly was designed to operate with pressures up to 180 bar and temperatures of 450degC. The geometry and dimensions of this test section were similar to part of a gas cooled reactor boiler of the Hinkley Point design. Blowdown from a pressure of 60 bar with subcoolings of 5degC, 50degC, 100degC formed the main part of the programme. A set of tests was conducted using discharge orifices of different sizes to produce depressurization times from 30 s to 10 mins, and in a few cases, the duration of blowdown approached 1 hour. These times were defined using the criterion of blowdown end as a final pressure of 10% of the initial pressure. Pressures, wall and fluid temperatures were all measured at average time intervals of 1.1s during the excursion and an inventory of the remaining water content in the serpentine was taken when the blowdown ended. Some tests were also conducted at an initial pressure of 30 bar. The results obtained show interesting stratification effects for the relatively fast discharge, with substantial wall circumferential temperature variations. For these tests, a relatively small water inventory remained after blowdown. The discharge characteristics of the serpentine in terms of orifice size have been mapped, and tests at 30 bar show the equivalence in terms of orifice size have been mapped

  6. Condensation heat transfer with noncondensable gas for passive containment cooling of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Leonardi, Tauna [Schlumberger, 14910 Airline Rd., Rosharon, TX 77583 (United States)]. E-mail: Tleonardi@slb.com; Ishii, Mamoru [School of Nuclear Engineering, Purdue University, West Lafayette, IN 47907 (United States)]. E-mail: Ishii@ecn.purdue.edu

    2006-09-15

    Noncondensable gases that come from the containment and the interaction of cladding and steam during a severe accident deteriorate a passive containment cooling system's performance by degrading the heat transfer capabilities of the condensers in passive containment cooling systems. This work contributes to the area of modeling condensation heat transfer with noncondensable gases in integral facilities. Previously existing correlations and models are for the through-flow of the mixture of steam and the noncondensable gases and this may not be applicable to passive containment cooling systems where there is no clear passage for the steam to escape. This work presents a condensation heat transfer model for the downward cocurrent flow of a steam/air mixture through a condenser tube, taking into account the atypical characteristics of the passive containment cooling system. An empirical model is developed that depends on the inlet conditions, including the mixture Reynolds number and noncondensable gas concentration.

  7. 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

  8. 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

  9. Radiation heat transfer in a pressurized water reactor lower head filled with molten corium

    International Nuclear Information System (INIS)

    Šadek, Siniša; Grgić, Davor; Debrecin, Nenad

    2013-01-01

    Highlights: ► We develop radiation heat exchange model for a reactor pressure vessel lower head. ► Model is used during a late in-vessel phase of severe accidents. ► View factors are calculated automatically for a time-dependent enclosure. ► Model is included in the RELAP5/SCDAPSIM computer code. ► Inclusion of heat radiation causes faster heat-up rate of RPV lower head structures. - Abstract: Following a core melt, molten material may slump to the lower head of a reactor pressure vessel (RPV). In that case, some structures like lower parts of fuel elements and a core support plate would remain intact. Since the melt is at high temperature and there are no obstacles between the melt and the supporting plate, the plate is exposed to an intense radiation heating. The radiation heat exchange model of the lower head was developed and applied to a finite element code COUPLE which is a part of the detailed mechanistic code RELAP5/SCDAPSIM. The radiation enclosure consisted of three surfaces: the upper surface of the relocated material, the inner surface of the RPV wall above the relocated material and the lower surface of the core support plate. View factors were calculated for the enclosure geometry that is changing in time because of intermittent accumulation of molten material. The enclosure surfaces were covered by mesh of polygonal areas and view factors were calculated, for each pair of the element areas, by solving the definite integrals using the algorithms for adaptive integrations by means of Gaussian quadrature. Algebraic equations for radiosity and irradiation vectors were solved by LU decomposition and the radiation model was explicitly coupled with the heat conduction model. The results show that there is a possibility of the core support plate failure after being heated up due to radiation heat exchange with the melt.

  10. 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.

  11. Characterization and modelling of the heat transfers in a pilot-scale reactor during composting under forced aeration

    International Nuclear Information System (INIS)

    Guardia, A. de; Petiot, C.; Benoist, J.C.; Druilhe, C.

    2012-01-01

    The paper focused on the modelling of the heat transfers during composting in a pilot-scale reactor under forced aeration. The model took into account the heat production and the transfers by evaporation, convection between material and gas crossing the material, conduction and surface convection between gas and material in bottom and upper parts of the reactor. The model was adjusted thanks to the measurements practised during fifteen composting experiments in which five organic wastes were, each, composted under three constant aeration rates. Heat production was considered proportional to oxygen consumption rate and the enthalpy per mole oxygen consumed was assumed constant. The convective heat transfer coefficients were determined on basis of the continuous measurements of the temperatures of both the lid and the bottom part of the reactor. The model allowed a satisfying prediction of the temperature of the composting material. In most cases, the mean absolute discard between the experimental and the simulated temperatures was inferior to 2.5 °C and the peaks of temperature occurred with less than 8 h delay. For the half of the experiments the temperature discard between the simulated peak and the experimental one was inferior to 5 °C. On basis of the calculation of a stoichiometric production of water through oxidation of the biodegradable organic matter, the simulation of water going out from material as vapour also allowed a rather satisfying prediction of the mass of water in final mixture. The influence of the aeration rate on every type of heat loss was characterized. Finally, the model was used to evaluate the impacts on material temperature caused by the change of the insulation thickness, the ambient temperature, take the lid away, the increase or the decrease of the mass of waste to compost.

  12. Numerical study on pressure drop and heat transfer for designing sodium-to-air heat exchanger tube banks on advanced sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Kang, Hie-Chan; Eoh, Jae-Hyuk; Cha, Jae-Eun; Kim, Seong-O.

    2013-01-01

    Highlights: ► Numerical simulation for the heat flow characteristic of the sodium-to-air heat exchanger (AHX) and tube banks. ► Parallelogram tube banks showed almost similar thermal and hydraulic characteristics to the rectangular tube banks. ► Pressure drop and heat transfer of the staggered and rectangular tube banks compared with Zhukauskas’ correlation. ► AHX was modeled as porous media and suggested design guide to enhance the performance. - Abstract: A numerical study is performed to investigate the thermal and hydraulic characteristics and build up design model of the AHX (sodium-to-air heat exchanger) unit of a sodium-cooled fast reactor. Helical-coiled tube banks in the AHX are modeled as porous media and simulated heat and momentum transfer by a commercial program. Two-dimensional flow characteristic appears differently at the inlet region of the AHX annulus, and the required length of the inlet region is shorter for an inlet having a 45 degree chamber or a round shape than for one with a perpendicular corner. Pressure drop and heat transfer coefficient for rectangular, parallelogram and staggered tube banks as the main components of the AHX are evaluated and discussed. Pressure drop and heat transfer shows similar trends and underestimated values, respectively, when compared with Zhukauskas empirical correlations. The parallelogram tube bank shows similar results to the rectangular arrangement.

  13. Specialists' meeting on heat and mass transfer in the reactor cover gas, Harwell, England, 8-10 October 1985

    International Nuclear Information System (INIS)

    1986-07-01

    The specialists' meeting on ''Heat and Mass Transfer in the Reactor Cover Gas'' was held at Harwell, the United Kingdom, on 8-10 October 1985. It was attended by 24 participants from all IWGFR member-countries: France, the Federal Republic of Germany, India, Italy, Japan, the Union of Soviet Socialist Republics, the United Kingdom and the United States. The meeting was presided over by Dr K. Eickhoff of the United Kingdom. The following topical areas were reviewed and discussed during the meeting: 1. National review presentations on the status of activities on heat and mass transfer in the reactor cover gas - 2 papers; 2. Aerosol dynamics - 4 papers; 3. Aerosol trapping - 2 papers; 4. Heat and mass transfer through cover gas in annuli - 3 papers; 5. Radiative properties - 4 papers; 6. Modelling of cover gas - 4 papers. A separate abstract was prepared for each of these papers. On the basis of papers presented and discussed by participants, session summaries and conclusions were drafted on the above topical areas. These summaries, as well as general conclusions and recommendations of the meeting were reviewed and agreed upon by consensus at the end of the meeting

  14. Numerical prediction on turbulent heat transfer of a spacer ribbed fuel rod for high temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Takase, Kazuyuki

    1994-11-01

    The turbulent heat transfer of a fuel rod with three-dimensional trapezoidal spacer ribs for high temperature gas-cooled reactors was analyzed numerically using the k-ε turbulence model, and investigated experimentally using a simulated fuel rod under the helium gas condition of a maximum outlet temperature of 1000degC and pressure of 4MPa. From the experimental results, it found that the turbulent heat transfer coefficients of the fuel rod were 18 to 80% higher than those of a concentric smooth annulus at a region of Reynolds number exceeding 2000. On the other hand, the predicted average Nusselt number of the fuel rod agreed well with the heat transfer correlation obtained from the experimental data within a relative error of 10% with Reynolds number of more than 5000. It was verified that the numerical analysis results had sufficient accuracy. Furthermore, the numerical prediction could clarify quantitatively the effects of the heat transfer augmentation by the spacer rib and the axial velocity increase due to a reduction in the annular channel cross-section. (author)

  15. Heat-transfer analysis of the existing HEU and proposed LEU cores of Pakistan research reactor

    International Nuclear Information System (INIS)

    Khan, L.A.; Nabbi, R.

    1987-02-01

    In connection with conversion of Pakistan Research Reactor (PARR) from the use of Highly Enriched Uranium (HEU) fuel to the use of Low Enriched Uranium (LEU) fuel, steady-state thermal hydraulic analysis of both existing HEU and proposed LEU cores has been carried out. Keeping in mind the possibility of power upgrading, the performance of proposed LEU core, under 10 MW operating conditions, has also been evaluated. Computer code HEATHYD has been used for this purpose. In order to verify the reliability of the code, IAEA benchmark 2 MW reactor was analyzed. The cooling parameters evaluated include: coolant velocity, critical velocity, pressure drop, temperature distribution in the core, heat fluxes at onset of nucleate boiling, flow instability and burnout and corresponding safety margins. From the results of the study it can be concluded that the conversion of the core to LEU fuel will result in higher safety margins, as compared to existing HEU core, mainly because the increased number of fuel plates in the proposed design will reduce the average heat flux significantly. Anyhow upgrading of the reactor power to 10 MW will need the flow rate to be adjusted between 850 to 900 m 3 /hr, to achieve reasonable safety margins, at least, comparable with the existing HEU core. (orig.)

  16. Freon Rig design for performing to heat transfer experiments for nuclear reactors fuel bundles

    International Nuclear Information System (INIS)

    Flores, L.F.V.

    1981-01-01

    The main features of a Freon Rig design for performing to heat transfer experiments for PWR and BWR fuel bundles, are presented. The project is based on a Freon Rig pressurized at 30 bar with a flow rate up to 80 m 3 /h. The maximum power fed to test sections is of about 420 KW D.C. The rig was designed to use scaling techniques wich would enable a fluid of low latente heat to be used in place of water, thereby reducing the cost of testes. (Author) [pt

  17. New systematic methodology for incorporating dynamic heat transfer modelling in multi-phase biochemical reactors.

    Science.gov (United States)

    Fernández-Arévalo, T; Lizarralde, I; Grau, P; Ayesa, E

    2014-09-01

    This paper presents a new modelling methodology for dynamically predicting the heat produced or consumed in the transformations of any biological reactor using Hess's law. Starting from a complete description of model components stoichiometry and formation enthalpies, the proposed modelling methodology has integrated successfully the simultaneous calculation of both the conventional mass balances and the enthalpy change of reaction in an expandable multi-phase matrix structure, which facilitates a detailed prediction of the main heat fluxes in the biochemical reactors. The methodology has been implemented in a plant-wide modelling methodology in order to facilitate the dynamic description of mass and heat throughout the plant. After validation with literature data, as illustrative examples of the capability of the methodology, two case studies have been described. In the first one, a predenitrification-nitrification dynamic process has been analysed, with the aim of demonstrating the easy integration of the methodology in any system. In the second case study, the simulation of a thermal model for an ATAD has shown the potential of the proposed methodology for analysing the effect of ventilation and influent characterization. Copyright © 2014 Elsevier Ltd. All rights reserved.

  18. The secure heating reactor

    International Nuclear Information System (INIS)

    Pind, C.

    1987-01-01

    The SECURE heating reactor was designed by ASEA-ATOM as a realistic alternative for district heating in urban areas and for supplying heat to process industries. SECURE has unique safety characteristics, that are based on fundamental laws of physics. The safety does not depend on active components or operator intervention for shutdown and cooling of the reactor. The inherent safety characteristics of the plant cannot be affected by operator errors. Due to its very low environment impact, it can be sited close to heat consumers. The SECURE heating reactor has been shown to be competitive in comparison with other alternatives for heating Helsinki and Seoul. The SECURE heating reactor forms a basis for the power-producing SECURE-P reactor known as PIUS (Process Inherent Ultimate Safety), which is based on the same inherent safety principles. The thermohydraulic function and transient response have been demonstrated in a large electrically heated loop at the ASEA-ATOM laboratories

  19. 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

  20. 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)

  1. Simulation of thermal-hydraulic process in reactor of HTR-PM based on flow and heat transfer network

    International Nuclear Information System (INIS)

    Zhou Kefeng; Zhou Yangping; Sui Zhe; Ma Yuanle

    2012-01-01

    The development of HTR-PM full scale simulator (FSS) is an important part in the project. The simulation of thermal-hydraulic process in reactor is one of the key technologies in the development of FSS. The simulation of thermal-hydraulic process in reactor was studied. According to the geometry structures and the characteristics of thermal-hydraulic process in reactor, the model was setup in components construction way. Based on the established simulation method of flow and heat transfer network, a Fortran code was developed and the simulation of thermal-hydraulic process was achieved. The simulation results of 50% FP steady state, 100% FP steady state and control rod mistakenly ascension accidents were given. The verification of simulation results was carried out by comparing with the design and analysis code THERMIX. The results show that the method and model based on flow and heat transfer network can meet the requirements of FSS and reflect the features of thermal-hydraulic process in HTR-PM. (authors)

  2. 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.

  3. The Radiative Heat Transfer Properties of Molten Salts and Their Relevance to the Design of Advanced Reactors

    Science.gov (United States)

    Chaleff, Ethan Solomon

    Molten salts, such as the fluoride salt eutectic LiF-NaF-KF (FLiNaK) or the transition metal fluoride salt KF-ZrF4, have been proposed as coolants for numerous advanced reactor concepts. These reactors are designed to operate at high temperatures where radiative heat transfer may play a significant role. If this is the case, the radiative heat transfer properties of the salt coolants are required to be known for heat transfer calculations to be performed accurately. Chapter 1 describes the existing literature and experimental efforts pertaining to radiative heat transfer in molten salts. The physics governing photon absorption by halide salts is discussed first, followed by a more specific description of experimental results pertaining to salts of interest. The phonon absorption edge in LiF-based salts such as FLiNaK is estimated and the technique described for potential use in other salts. A description is given of various spectral measurement techniques which might plausibly be employed in the present effort, as well as an argument for the use of integral techniques. Chapter 2 discusses the mathematical treatments required to approximate and solve for the radiative flux in participating materials. The differential approximation and the exact solutions to the radiative flux are examined, and methods are given to solve radiative and energy equations simultaneously. A coupled solution is used to examine radiative heat transfer to molten salt coolants. A map is generated of pipe diameters, wall temperatures, and average absorption coefficients where radiative heat transfer will increase expected heat transfer by more than 10% compared to convective methods alone. Chapter 3 presents the design and analysis of the Integral Radiative Absorption Chamber (IRAC). The IRAC employs an integral technique for the measurement of the entire electromagnetic spectrum, negating some of the challenges associated with the methods discussed in Chapter 1 at the loss of spectral

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  5. Influence of coolant pH on corrosion of 6061 aluminum under reactor heat transfer conditions

    International Nuclear Information System (INIS)

    Pawel, S.J.; Felde, D.K.; Pawel, R.E.

    1995-10-01

    To support the design of the Advanced Neutron Source (ANS), an experimental program was conducted wherein aluminum alloy specimens were exposed at high heat fluxes to high-velocity aqueous coolants in a corrosion test loop. The aluminum alloys selected for exposure were candidate fuel cladding materials, and the loop system was constructed to emulate the primary coolant system for the proposed ANS reactor. One major result of this program has been the generation of an experimental database defining oxide film growth on 6061 aluminum alloy cladding. Additionally, a data correlation was developed from the database to permit the prediction of film growth for any reasonable thermal-hydraulic excursion. This capability was utilized effectively during the conceptual design stages of the reactor. During the course of this research, it became clear that the kinetics of film growth on the aluminum alloy specimens were sensitively dependent on the chemistry of the aqueous coolant and that relatively small deviations from the intended pH 5 operational level resulted in unexpectedly large changes in the corrosion behavior. Examination of the kinetic influences and the details of the film morphology suggested that a mechanism involving mass transport from other parts of the test loop was involved. Such a mechanism would also be expected to be active in the operating reactor. This report emphasizes the results of experiments that best illustrate the influence of the nonthermal-hydraulic parameters on film growth and presents data to show that comparatively small variations in pH near 5.0 invoke a sensitive response. Simply, for operation in the temperature and heat flux range appropriate for the ANS studies, coolant pH levels from 4.5 to 4.9 produced significantly less film growth than those from pH 5.1 to 6. A mechanism for this behavior based on the concept of treating the entire loop as an active corrosion system is presented

  6. Scrap tyre recycling process with molten zinc as direct heat transfer and solids separation fluid: A new reactor concept.

    Science.gov (United States)

    Riedewald, Frank; Goode, Kieran; Sexton, Aidan; Sousa-Gallagher, Maria J

    2016-01-01

    Every year about 1.5 billion tyres are discarded worldwide representing a large amount of solid waste, but also a largely untapped source of raw materials. The objective of the method was to prove the concept of a novel scrap tyre recycling process which uses molten zinc as the direct heat transfer fluid and, simultaneously, uses this media to separate the solids products (i.e. steel and rCB) in a sink-float separation at an operating temperature of 450-470 °C. This methodology involved: •construction of the laboratory scale batch reactor,•separation of floating rCB from the zinc,•recovery of the steel from the bottom of the reactor following pyrolysis.

  7. 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.

  8. Nuclear reactor plant for production process heat

    International Nuclear Information System (INIS)

    Weber, M.

    1979-01-01

    The high temperature reactor is suitable as a heat source for carrying out endothermal chemical processes. A heat exchanger is required for separating the reactor coolant gases and the process medium. The heat of the reactor is transferred at a temperature lower than the process temperature to a secondary gas and is compressed to give the required temperature. The compression energy is obtained from the same reactor. (RW) [de

  9. Predictions for heat transfer characteristics in a natural draft reactor cooling system using a second moment closure turbulence model

    International Nuclear Information System (INIS)

    Nishimura, M.; Maekawa, I.

    2004-01-01

    A numerical study is performed on the natural draft reactor cavity cooling system (RCCS). In the cooling system, buoyancy driven heated upward flow could be in the mixed convection regime that is accompanied by heat transfer impairment. Also, the heating wall condition is asymmetric with regard to the channel cross section. These flow regime and thermal boundary conditions may invalidate the use of design correlation. To precisely simulate the flow and thermal fields within the RCCS, the second moment closure turbulence model is applied. Two types of the RCCS channel geometry are selected to make a comparison: an annular duct with fins on the outer surface of the inner circular wall, and a multi-rectangular duct. The prediction shows that the local heat transfer coefficient on the RCCS with finned annular duct is less than 1/6 of that estimated with Dittus-Boelter correlation. Much portion of the natural draft airflow does not contribute cooling at all because mainstream escapes from the narrow gaps between the fins. This result and thus the finned annulus design are unacceptable from the viewpoint for structural integrity of the RCCS wall boundary. The performance of the multi-rectangular duct design is acceptable that the RCCS maximum temperature is less than 400 degree centigrade even when the flow rate is halved from the designed condition. (author)

  10. Effects of transient and non-uniform distribution of heat flux on intensity of heat transfer and burnout conditions in the channels of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Vitaly Osmachkin [Russian Research Center ' Kurchatov Institute' 1, Kurchatov sq, Moscow 123182 (Russian Federation)

    2005-07-01

    Full text of publication follows: The influence of power transient, changes of flow rate, inlet temperatures or pressure in cores of nuclear reactors on heat transfer and burnout conditions in channels depend on rate of such violations. Non-uniform distribution of the heat flux is also important factor for heat transfer and development of crisis phenomenon. Such effects may be significant for NPPs safety. But they have not yet generally accepted interpretation. Steady state approach is often recommended for use in calculations. In the paper a review of experimental observed so-called non-equilibrium effects is presented. The effects of space and time factors are displaying due delay in reformation turbulence intensity, velocity, temperatures or void fraction profiles, water film flow on the surface of heated channels. For estimation of such effect different methods are used. Modern computer codes based on two or three fluids approaches are considered as most effective. But simple and clear correlations may light up the mechanics of effects on heat transfer and improve general understanding of scale and significance of the transient events. In the paper the simplified methods for assessment the influence of lags in the development of distributions of parameters of flow, the relaxation of temporal or space violations are considered. They are compared with more sophisticated approaches. Velocities of disturbance fronts moving along the channels are discussed also. (author)

  11. Analyses of fluid flow and heat transfer inside calandria vessel of CANDU-6 reactor using CFD

    International Nuclear Information System (INIS)

    Yu, Seon Oh; Kim, Man Woong; Kim, Hho Jung

    2005-01-01

    In a CANDU (CANada Deuterium Uranium) reactor, fuel channel integrity depends on the coolability of the moderator as an ultimate heat sink under transient conditions such as a Loss Of Coolant Accident (LOCA) with coincident Loss Of Emergency Core Cooling (LOECC). as well as normal operating conditions. This study presents assessments of moderator thermal-hydraulic characteristics in the normal operating conditions and one transient condition for CANDU-6 reactors, using a general purpose three-dimensional computational fluid dynamics code. First, an optimized calculation scheme is obtained by many-sided comparisons of the predicted results with the related experimental data, and by evaluating the fluid flow and temperature distributions. Then, using the optimized scheme, analyses of real CANDU-6 in normal operating conditions and the transition condition have been performed. The present model successfully predicted the experimental results and also reasonably assessed the thermal-hydraulic characteristics of a real CANDU-6 with 380 fuel channels. A flow regime map with major parameters representing the flow pattern inside a calandria vessel has also proposed to be used as operational and/or regulatory guidelines

  12. Heat transfer from rough surfaces

    International Nuclear Information System (INIS)

    Dalle Donne, M.

    1977-01-01

    Artificial roughness is often used in nuclear reactors to improve the thermal performance of the fuel elements. Although these are made up of clusters of rods, the experiments to measure the heat transfer and friction coefficients of roughness are performed with single rods contained in smooth tubes. This work illustrated a new transformation method to obtain data applicable to reactor fuel elements from these annulus experiments. New experimental friction data are presented for ten rods, each with a different artificial roughness made up of two-dimensional rectangular ribs. For each rod four tests have been performed, each in a different outer smooth tube. For two of these rods, each for two different outer tubes, heat transfer data are also given. The friction and heat transfer data, transformed with the present method, are correlated by simple equations. In the paper, these equations are applied to a case typical for a Gas Cooled Fast Reactor fuel element. (orig.) [de

  13. 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....

  14. 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.)

  15. Heat Transfer Analysis of the European Pressurized Water Reactor (EPR) Core Catcher Test Facility Volley

    Energy Technology Data Exchange (ETDEWEB)

    Pikkarainen, Mika; Laine, Jani; Purhonen, Heikki; Kyrki-Rajamaeki, Riitta [Lappeenranta University of Technology, P.O. 20 53851 Lappeenranta (Finland); Sairanen, Risto [Radiation and Nuclear Safety Authority, P.O. 14 00881 Helsinki (Finland)

    2008-07-01

    The EPR is designed to cope with severe accidents, involving core meltdown. A specific melt spreading area has been designed within the containment. This core catcher will be flooded by water, which transfers the decay heat to the containment heat removal system. To improve cooling, horizontal flow channels made of cast iron are located also below the core catcher. STUK, the radiation and nuclear safety authority in Finland, wanted an independent study of the functionality of the core catcher design. Effect of the presence of insulation material and boric acid in the cooling water was to be studied, as well as the general behavior of the system in different phases of the flooding of the core melt spreading area. To verify the function of the core catcher design, a scaled down test facility was built at Lappeenranta University of Technology. Since there are some physical restrictions of a test facility computational tools were applied especially for the tests where steady state conditions could not be reached without endangering the integrity of the test facility. This paper introduces the Volley test facility, computational simulations and compares them with the test results. Simulated temperatures of those Volley tests, which could be run until steady state conditions, are very close to the measured temperatures. It can be concluded also, that the temperatures are evidently below the cast iron melting point with heat fluxes used in the tests, if there is a small flow inside the cooling channels or even in case when only a few adjacent cooling channels are totally dry. (authors)

  16. Heat Transfer Analysis of the European Pressurized Water Reactor (EPR) Core Catcher Test Facility Volley

    International Nuclear Information System (INIS)

    Pikkarainen, Mika; Laine, Jani; Purhonen, Heikki; Kyrki-Rajamaeki, Riitta; Sairanen, Risto

    2008-01-01

    The EPR is designed to cope with severe accidents, involving core meltdown. A specific melt spreading area has been designed within the containment. This core catcher will be flooded by water, which transfers the decay heat to the containment heat removal system. To improve cooling, horizontal flow channels made of cast iron are located also below the core catcher. STUK, the radiation and nuclear safety authority in Finland, wanted an independent study of the functionality of the core catcher design. Effect of the presence of insulation material and boric acid in the cooling water was to be studied, as well as the general behavior of the system in different phases of the flooding of the core melt spreading area. To verify the function of the core catcher design, a scaled down test facility was built at Lappeenranta University of Technology. Since there are some physical restrictions of a test facility computational tools were applied especially for the tests where steady state conditions could not be reached without endangering the integrity of the test facility. This paper introduces the Volley test facility, computational simulations and compares them with the test results. Simulated temperatures of those Volley tests, which could be run until steady state conditions, are very close to the measured temperatures. It can be concluded also, that the temperatures are evidently below the cast iron melting point with heat fluxes used in the tests, if there is a small flow inside the cooling channels or even in case when only a few adjacent cooling channels are totally dry. (authors)

  17. Application of automatic inspection system to nondestructive test of heat transfer tubes of primary pressurized water cooler in the high temperature engineering test reactor. Joint research

    International Nuclear Information System (INIS)

    Takeda, Takeshi; Furusawa, Takayuki

    2001-07-01

    Heat transfer tubes of a primary pressurized water cooled (PPWC) in the high temperature engineering test reactor (HTTR) form the reactor pressure boundary of the primary coolant, therefore are important from the viewpoint of safety. To establish inspection techniques for the heat transfer tubes of the PPWC, an automatic inspection system was developed. The system employs a bobbin coil probe, a rotating probe for eddy current testing (ECT) and a rotating probe for ultrasonic testing (UT). Nondestructive test of a half of the heat transfer tubes of the PPWC was carried out by the automatic inspection system during reactor shutdown period of the HTTR (about 55% in the maximum reactor power in this paper). The nondestructive test results showed that the maximum signal-to-noise ratio was 1.8 in ECT. Pattern and phase of Lissajous wave, which were obtained for the heat transfer tube of the PPWC, were different from those obtained for the artificially defected tube. In UT echo amplitude of the PPWC tubes inspected was lower than 20% of distance-amplitude calibration curve. Thus, it was confirmed that there was no defect in depth, which was more than the detecting standard of the probes, on the outer surface of the heat transfer tubes of the PPWC inspected. (author)

  18. 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

  19. Influence of single-phase heat transfer correlations on safety analysis of research reactors with narrow rectangular fuel channels

    International Nuclear Information System (INIS)

    Rawashdeh, A.; Altamimi, R.; Lee, B.; Chung, Y. J.; Park, S.

    2013-01-01

    The influence of different single-phase heat transfer correlations on the fuel temperature and minimum critical heat flux ratio (MCHFR) during a typical accident of a 5 MW research reactor is investigated. A reactor uses plate type fuel, of which the cooling channels have a narrow rectangular shape. RELAP5/MOD3.3 tends to over-predict the Nusselt number (Nu) at a low Reynolds number (Re) region, and therefore the correlation set is modified to properly describe the thermal behavior at that region. To demonstrate the effect of Nu at a low-Re region on an accident analysis, a two-pump failure accident was chosen as a sample problem. In the accident, the downward core flow decreases by a pump coast-down, and then reverses upward by natural convection. During the pump coast-down and flow reversal, the flow undergoes a laminar flow regime which has a different Nu with respect to the correlation sets. Compared to the results by the original RELAP5/MOD3.3, the modified correlation set predicts the fuel temperature to be a little higher than the original value, and the MCHFR to be a little lower than the original value. Although the modified correlation set predicts the fuel temperature and the MCHFR to be less conservative than those calculated from the original correlation of RELAP5/MOD3.3, the maximum fuel temperature and the MCHFR still satisfy the safety acceptance criteria

  20. Large Eddy Simulation of Fluid flow and Heat Transfer in the Upper Plenum of Fast Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Seokki; Lee, Taeho; Kim, Dongeun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ko, Sungho [Chungnam National Univ., Daejeon (Korea, Republic of)

    2014-05-15

    The important parameters in the thermal striping are the frequency and the amplitude of the temperature fluctuation. Since the sodium used as coolant in the PGSFR has a high thermal conductivity, the temperature fluctuation can be easily transferred to the solid walls of the components in the upper plenum. To remedy these problems, numerical studies are performed in the present study to analyze the thermal striping for possible improvement of the design and safety of the reactor. For the numerical works, Chacko et al. performed LES for the experiment by Nam and Kim, and found that the LES can produce the oscillation of temperature fluctuation properly, while the realizable k - ε model predicts the amplitude and frequency of the temperature fluctuation very poorly indicating that the LES method is an appropriate calculation method for the thermal striping. In this paper, the simulation of thermal striping in the upper plenum of PGSFR is performed using the LES method. The WALE eddy viscosity model by Nicoud and Ducros built in CFX-13 commercial code is employed for the LES eddy viscosity model. The numerical investigation of the thermal striping is performed with the LES method using the CFX-13 commercial code, where the solution domain is the upper plenum of the PGSFR. As the first step, dozens of monitoring points are set to locations that are anticipated to cause thermal striping. Then, the temperature fluctuations were calculated along with the time-averaged variables such as the velocity and temperature. From these results we have obtained the following conclusions. At the side wall of IHX, a slight fluctuation is observed, but it seems that there is no risk of thermal striping. The flows from the reactor core are not mixed when reaching the UIS. So both the first and second plates need to be considered. Among the first grid plate regions, the shape region is the weakest region for thermal striping. The second weakest region for thermal striping is the shape

  1. Investigation of heat transfer in liquid-metal flows under fusion-reactor conditions

    Energy Technology Data Exchange (ETDEWEB)

    Poddubnyi, I. I., E-mail: poddubnyyii@nikiet.ru [Joint Stock Company Dollezhal Research and Development Institute of Power Engineering (JSC NIKIET) (Russian Federation); Pyatnitskaya, N. Yu.; Razuvanov, N. G.; Sviridov, V. G.; Sviridov, E. V. [Russian Academy of Science, Joint Institute of High Temperatures (Russian Federation); Leshukov, A. Yu. [Joint Stock Company Dollezhal Research and Development Institute of Power Engineering (JSC NIKIET) (Russian Federation); Aleskovskiy, K. V. [National Research University Moscow Power Engineering Institute (MPEI) (Russian Federation); Obukhov, D. M. [Joint Stock Company Efremov Institute of Electrophysical Apparatus (Russian Federation)

    2016-12-15

    The effect discovered in studying a downward liquid-metal flow in vertical pipe and in a channel of rectangular cross section in, respectively, a transverse and a coplanar magnetic field is analyzed. In test blanket modules (TBM), which are prototypes of a blanket for a demonstration fusion reactor (DEMO) and which are intended for experimental investigations at the International Thermonuclear Experimental Reactor (ITER), liquid metals are assumed to fulfil simultaneously the functions of (i) a tritium breeder, (ii) a coolant, and (iii) neutron moderator and multiplier. This approach to testing experimentally design solutions is motivated by plans to employ, in the majority of the currently developed DEMO blanket projects, liquid metals pumped through pipes and/or rectangular channels in a transvers magnetic field. At the present time, experiments that would directly simulate liquid-metal flows under conditions of ITER TBM and/or DEMO blanket operation (irradiation with thermonuclear neutrons, a cyclic temperature regime, and a magnetic-field strength of about 4 to 10 T) are not implementable for want of equipment that could reproduce simultaneously the aforementioned effects exerted by thermonuclear plasmas. This is the reason why use is made of an iterative approach to experimentally estimating the performance of design solutions for liquid-metal channels via simulating one or simultaneously two of the aforementioned factors. Therefore, the investigations reported in the present article are of considerable topical interest. The respective experiments were performed on the basis of the mercury magneto hydrodynamic (MHD) loop that is included in the structure of the MPEI—JIHT MHD experimental facility. Temperature fields were measured under conditions of two- and one-sided heating, and data on averaged-temperature fields, distributions of the wall temperature, and statistical fluctuation features were obtained. A substantial effect of counter thermo gravitational

  2. Heat transfer tests conducted on full-scale model, to investigate cooling conditions of EL.3 experimental reactor

    International Nuclear Information System (INIS)

    Raievski, R.; Bousquet, M.; Braudeau, M.; Milliat, M.

    1958-01-01

    For such high heat flux density as is released in the channels of EL3 reactor (2.10 6 kcal/m 2 h on the hottest point) cooling conditions have proved to be satisfactory, that is free from nucleate boiling. The arrangements provided for these tests and the technique used for measurements (of temperature particularly) are specified. Two fields have been investigated: in the former (forced convection without nucleate boiling) a good agreement is found with Colburn's formula. The influence of the ratio L/D is pointed out. The latter field is of forced convection with beginning of nucleate boiling; there the observed raise of the transfer coefficient has been shown occurring with some delay. (author) [fr

  3. Numerical investigation of a heat transfer within the prismatic fuel assembly of a very high temperature reactor

    International Nuclear Information System (INIS)

    Tak, Nam-il; Kim, Min-Hwan; Lee, Won Jae

    2008-01-01

    The complex geometry of the hexagonal fuel blocks of the prismatic fuel assembly in a very high temperature reactor (VHTR) hinders accurate evaluations of the temperature profile within the fuel assembly without elaborate numerical calculations. Therefore, simplified models such as a unit cell model have been widely applied for the analyses and designs of prismatic VHTRs since they have been considered as effective approaches reducing the computational efforts. In a prismatic VHTR, however, the simplified models cannot consider a heat transfer within a fuel assembly as well as a coolant flow through a bypass gap between the fuel assemblies, which may significantly affect the maximum fuel temperature. In this paper, a three-dimensional computational fluid dynamics (CFD) analysis has been carried out on a typical fuel assembly of a prismatic VHTR. Thermal behaviours and heat transfer within the fuel assembly are intensively investigated using the CFD solutions. In addition, the accuracy of the unit cell approach is assessed against the CFD solutions. Two example situations are illustrated to demonstrate the deficiency of the unit cell model caused by neglecting the effects of the bypass gap flow and the radial power distribution within the fuel assembly

  4. Verification of Heat and Mass Transfer Closures in Industrial Scale Packed Bed Reactor Simulations

    Directory of Open Access Journals (Sweden)

    Arpit Singhal

    2018-03-01

    Full Text Available Particle-resolved direct numerical simulation (PR-DNS is known to provide an accurate detailed insight into the local flow phenomena in static particle arrays. Most PR-DNS studies in literature do not account for reactions taking place inside the porous particles. In this study, PR-DNS is performed for catalytic reactions inside the particles using the multifluid approach where all heat and mass transfer phenomena are directly resolved both inside and outside the particles. These simulation results are then used to verify existing 1D model closures from literature over a number of different reaction parameters including different reaction orders, multiple reactions and reactants, interacting reactions, and reactions involving gas volume generation/consumption inside the particle. Results clearly showed that several modifications to existing 1D model closures are required to reproduce PR-DNS results. The resulting enhanced 1D model was then used to accurately simulate steam methane reforming, which includes all of the aforementioned reaction complexities. The effect of multiple reactants was found to be the most influential in this case.

  5. Heat transfer study of a submerged reactor channel under boil-off condition

    Energy Technology Data Exchange (ETDEWEB)

    Mukhopadhyay, Deb [Bhabha Atomic Research Centre, Mumbai (India). Reactor Safety Div.; Sahoo, P.K. [Indian Institute of Technology, Roorkee (India). Dept. of Mechanical and Industrial Engineering; Ghosh, A.K. [Bhabha Atomic Research Centre, Mumbai (India). Health, Safety and Environment Group

    2012-12-15

    Experiments have been carried out to study the heatup behavior of a single segmented reactor channel for Pressurized Heavy Water Reactor under submerged, partially submerged and exposed conditions. This situation may arise from a severe accident scenario of Pressurised Heavy Water Reactors where full or segmented reactor channels are likely to be disassembled and form a submerged debris bed. An assembly of electrical heater rod, simulating fuel bundle and channel components like Pressure Tube and Calandria Tube constitutes the segmented reactor channel. Heatup of this assembly is observed with respect to different water levels ranging from full submergence to totally exposed and power levels of 6-8 kW, typical to decay power level. It has been observed from the set of experiment that fuel bundle local dry out followed by heatup does not happen till the bundle is partially submerged. Temperature excursion of the bundle is evident when the bundle is exposed to steam-air environment. (orig.)

  6. Liquid metal reactor head designs in the USA - heat and mass transfer considerations

    International Nuclear Information System (INIS)

    Burke, T.M.

    1986-01-01

    Development of liquid metal reactor plants in the United States over the past 30 years has resulted in an evolution of reactor head designs as reflected in the SRE, Hallam, EBR-II and FFTF plants. This evolution has probably been affected to some extent by the fact that, in contrast to most other countries, there is no single organization in the United States which has been responsible for the design of liquid metal reactor plants. The current U.S. LMR design efforts involve two innovative design consortiums (guided by the US Department of Energy) and a joint industry venture on the Large Scale Prototype Breeder. It is therefore somewhat difficult to provide a statement on the philosophy of the reactor head design in the U.S. This paper however briefly describes the existing and proposed U.S. liquid metal reactor head designs and in the process, attempt to provide some insight on the basis for those designs

  7. IAEA coordinated research programme on heat transfer behavior and thermo-hydraulics code testing for super critical water cooled reactors

    International Nuclear Information System (INIS)

    Bilbao y Leon, Sama; Aksan, Nusret

    2009-01-01

    One of the key roles of the IAEA is to foster the collaboration among Member States on the development of advances in technology for advanced nuclear power plants. There is high international interest, both in developing and industrialized countries, in innovative supercritical water-cooled reactors (SCWRs), primarily because such concepts will achieve high thermal efficiencies (44-45%) and promise improved economic competitiveness utilizing and building upon the recent developments for highly efficient fossil power plants. The SCWR has been selected as one of the promising concepts for development by the Generation-IV International Forum. Following the advice of the IAEA Nuclear Energy Department's Technical Working Groups on Advanced Technologies for LWRs and HWRs (the TWG-LWR and TWG-HWR), with the feedback from the Gen-IV SCWR Steering Committee, and in coordination with the OECD-NEA, IAEA has recently started a Coordinated Research Programme (CRP) in the areas of heat transfer behaviour and testing of thermo-hydraulic computer methods for Supercritical Water-Cooled Reactors. The first Research Coordination Meeting (RCM) of the CRP was held at the IAEA Headquarters, in Vienna, Austria in July 2008. This paper summarizes the current status of the CRP, including the Integrated Research Plan and the general schedule for the CRP. (author)

  8. Supercritical heat transfer phenomena in nuclear system

    International Nuclear Information System (INIS)

    Seo, Kyoung Woo; Kim, Moo Hwan; Anderson, Mark H.; Corradini, Michael L.

    2005-01-01

    A supercritical water (SCW) power cycle has been considered as one of the viable candidates for advanced fission reactor designs. However, the dramatic variation of thermo-physical properties with a modest change of temperature near the pseudo-critical point make existing heat transfer correlations such as the Dittus-Boelter correlation not suitably accurate to calculate the heat transfer in supercritical fluid. Several other correlations have also been suggested but none of them are able to predict the heat transfer over a parameter range, needed for reactor thermal-hydraulics simulation and design. This has prompted additional research to understand the characteristic of supercritical fluid heat transfer

  9. Conjugate heat transfer analysis of multiple enclosures in prototype fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Velusamy, K.; Balaubramanian, V.; Vaidyanathan, G.; Chetal, S.C. [Indira Gandhi Centre for Atomic Research, Kalpakkam (India)

    1995-09-01

    Prototype Fast Breeder Reactor (PFBR) is a 500 MWe sodium cooled reactor under design. The main vessel of the reactor serves as the primary boundary. It is surrounded by a safety vessel which in turn is surrounded by biological shield. The gaps between them are filled with nitrogen. Knowledge of temperature distribution prevailing under various operating conditions is essential for the assessment of structural integrity. Due to the presence of cover gas over sodium free level within the main vessel, there are sharp gradients in temperatures. Also cover gas height reduces during station blackout conditions due to sodium level rise in main vessel caused by temperature rise. This paper describes the model used to analyse the natural convection in nitrogen, conduction in structures and radiation interaction among them. Results obtained from parametric studies for PFBR are also presented.

  10. Transient core characteristics of small molten salt reactor coupling problem between heat transfer/flow and nuclear fission reaction

    International Nuclear Information System (INIS)

    Yamamoto, Takahisa; Mitachi, Koshi

    2004-01-01

    This paper performed the transient core analysis of a small Molten Salt Reactor (MSR). The emphasis is that the numerical model employed in this paper takes into account the interaction among fuel salt flow, nuclear reaction and heat transfer. The model consists of two group diffusion equations for fast and thermal neutron fluexs, balance equations for six-group delayed neutron precursors and energy conservation equations for fuel salt and graphite moderator. The results of transient analysis are that (1) fission reaction (heat generation) rate significantly increases soon after step reactivity insertion, e.g., the peak of fission reaction rate achieves about 2.7 times larger than the rated power 350 MW when the reactivity of 0.15% Δk/k 0 is inserted to the rated state, and (2) the self-control performance of the small MSR effectively works under the step reactivity insertion of 0.56% Δk/k 0 , putting the fission reaction rate back on the rated state. (author)

  11. MHTGR inherent heat transfer capability

    International Nuclear Information System (INIS)

    Berkoe, J.M.

    1992-01-01

    This paper reports on the Commercial Modular High Temperature Gas-Cooled Reactor (MHTGR) which achieves improved reactor safety performance and reliability by utilizing a completely passive natural convection cooling system called the RCCS to remove decay heat in the event that all active cooling systems fail to operate. For the highly improbable condition that the RCCS were to become non-functional following a reactor depressurization event, the plant would be forced to rely upon its inherent thermo-physical characteristics to reject decay heat to the surrounding earth and ambient environment. A computational heat transfer model was created to simulate such a scenario. Plant component temperature histories were computed over a period of 20 days into the event. The results clearly demonstrate the capability of the MHTGR to maintain core integrity and provide substantial lead time for taking corrective measures

  12. Depressurization study of CAREM 25 reactor considering the structures heat transfer

    International Nuclear Information System (INIS)

    Doval, A.

    1990-01-01

    This work presents the CAREM 25 reactor depressurization analysis results as an alternative form of accidents mitigation. Such results will help to determine design pressure valves for the emergency injection system as well as the depressurization valve diameter. Calculations were made with BLOW.MOD2 program. (Author) [es

  13. Italian position paper on heat and mass transfer in the reactor cover gas

    International Nuclear Information System (INIS)

    Caponetti, R.; Olivieri, P.; Petrazzuolo, F.

    1986-01-01

    The major effort being made in Italy with regard to the development of fast nuclear reactors is concentrated, as is known, in the construction of the PEC reactor, whose mechanical completion is expected early in 1988. The 116MWt PEC (Prova Elementi di Combustibile; i.e. Fuel Element Testing) reactor is sodium cooled. It is being built to study the behavior of fuel elements under thermal and neutronic conditions similar to those of fast nuclear power stations. Particular attention is being dedicated to safety aspects. This document furnishes a number of construction solutions with regard to that reactor and preparatory approaches to its operation, namely: a brief description of the construction solutions as far as concerns the Closure Head Assembly and the cover gas circuit together with its main components; the description of some test facilities arranged for abatement and measurement of sodium aerosol concentration; a number of preliminary evaluation results obtained thus far with regard to the formation, transport and depositing of sodium aerosols

  14. 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

  15. Numerical investigation of flow and heat transfer in a novel configuration multi-tubular fixed bed reactor for propylene to acrolein process

    Science.gov (United States)

    Jiang, Bin; Hao, Li; Zhang, Luhong; Sun, Yongli; Xiao, Xiaoming

    2015-01-01

    In the present contribution, a numerical study of fluid flow and heat transfer performance in a pilot-scale multi-tubular fixed bed reactor for propylene to acrolein oxidation reaction is presented using computational fluid dynamics (CFD) method. Firstly, a two-dimensional CFD model is developed to simulate flow behaviors, catalytic oxidation reaction, heat and mass transfer adopting porous medium model on tube side to achieve the temperature distribution and investigate the effect of operation parameters on hot spot temperature. Secondly, based on the conclusions of tube-side, a novel configuration multi-tubular fixed-bed reactor comprising 790 tubes design with disk-and-doughnut baffles is proposed by comparing with segmental baffles reactor and their performance of fluid flow and heat transfer is analyzed to ensure the uniformity condition using molten salt as heat carrier medium on shell-side by three-dimensional CFD method. The results reveal that comprehensive performance of the reactor with disk-and-doughnut baffles is better than that of with segmental baffles. Finally, the effects of operating conditions to control the hot spots are investigated. The results show that the flow velocity range about 0.65 m/s is applicable and the co-current cooling system flow direction is better than counter-current flow to control the hottest temperature.

  16. 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.

  17. 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.

  18. 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.

  19. Heat Transfer Salts for Nuclear Reactor Systems - Chemistry Control, Corrosion Mitigation, and Modeling

    International Nuclear Information System (INIS)

    Anderson, Mark; Sridharan, Kumar; Morgan, Dane; Peterson, Per; Calderoni, Pattrick; Scheele, Randall; Casekka, Andrew; McNamara, Bruce

    2015-01-01

    The concept of a molten salt reactor has existed for nearly sixty years. Previously all work was done during a large collaborative effort at Oak Ridge National Laboratory, culminating in a research reactor which operated for 15,000 hours without major error. This technical success has garnished interest in modern, high temperature, reactor schemes. Research using molten fluoride salts for nuclear applications requires a steady supply of high grade molten salts. There is no bulk supplier of research grade fluoride salts in the world, so a facility which could provide all the salt needed for testing at the University of Wisconsin had to be produced. Two salt purification devices were made for this purpose, a large scale purifier, and a small scale purifier, each designed to clean the salts from impurities and reduce their corrosion potential. As of now, the small scale has performed with flibe salt, hydrogen, and hydrogen fluoride, yielding clean salt. This salt is currently being used in corrosion testing facilities at the Massachusetts Institute of Technology and the University of Wisconsin. Working with the beryllium based salts requires extensive safety measures and health monitoring to prevent the development of acute or chronic beryllium disease, two pulmonary diseases created by an allergic reaction to beryllium in the lungs. Extensive health monitoring, engineering controls, and environment monitoring had to be set up with the University of Wisconsin department of Environment, Health and Safety. The hydrogen fluoride required for purification was also an extreme health hazard requiring thoughtful planning and execution. These dangers have made research a slow and tedious process. Simple processes, such as chemical handling and clean-up, can take large amounts of ingenuity and time. Other work has complemented the experimental research at Wisconsin to advance high temperature reactor goals. Modeling work has been performed in house to re

  20. Heat Transfer Salts for Nuclear Reactor Systems - Chemistry Control, Corrosion Mitigation, and Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Mark [Univ. of Wisconsin, Madison, WI (United States); Sridharan, Kumar [Univ. of Wisconsin, Madison, WI (United States); Morgan, Dane [Univ. of Wisconsin, Madison, WI (United States); Peterson, Per [Univ. of Wisconsin, Madison, WI (United States); Calderoni, Pattrick [Univ. of Wisconsin, Madison, WI (United States); Scheele, Randall [Univ. of Wisconsin, Madison, WI (United States); Casekka, Andrew [Univ. of Wisconsin, Madison, WI (United States); McNamara, Bruce [Univ. of Wisconsin, Madison, WI (United States)

    2015-01-22

    The concept of a molten salt reactor has existed for nearly sixty years. Previously all work was done during a large collaborative effort at Oak Ridge National Laboratory, culminating in a research reactor which operated for 15,000 hours without major error. This technical success has garnished interest in modern, high temperature, reactor schemes. Research using molten fluoride salts for nuclear applications requires a steady supply of high grade molten salts. There is no bulk supplier of research grade fluoride salts in the world, so a facility which could provide all the salt needed for testing at the University of Wisconsin had to be produced. Two salt purification devices were made for this purpose, a large scale purifier, and a small scale purifier, each designed to clean the salts from impurities and reduce their corrosion potential. As of now, the small scale has performed with flibe salt, hydrogen, and hydrogen fluoride, yielding clean salt. This salt is currently being used in corrosion testing facilities at the Massachusetts Institute of Technology and the University of Wisconsin. Working with the beryllium based salts requires extensive safety measures and health monitoring to prevent the development of acute or chronic beryllium disease, two pulmonary diseases created by an allergic reaction to beryllium in the lungs. Extensive health monitoring, engineering controls, and environment monitoring had to be set up with the University of Wisconsin department of Environment, Health and Safety. The hydrogen fluoride required for purification was also an extreme health hazard requiring thoughtful planning and execution. These dangers have made research a slow and tedious process. Simple processes, such as chemical handling and clean-up, can take large amounts of ingenuity and time. Other work has complemented the experimental research at Wisconsin to advance high temperature reactor goals. Modeling work has been performed in house to re

  1. Modelling of the heat transfer during oxygen atoms recombination on metallic surfaces in a plasma reactor

    NARCIS (Netherlands)

    Cavadias, S; Cauquot, P; Amouroux, J

    1997-01-01

    Space shuttle overheating during the re-entry phase, due to catalytic oxygen recombination on the thermal protection system, is a problem of practical and theoretical interest. The energy transfer is characterised by the product of the accommodation and the recombination coefficients. Previous

  2. The experimental study on the mass transfer model of boron injection for natural circular heating reactor

    International Nuclear Information System (INIS)

    Zha Meisheng; Nie Mengchen; Zhou Huizhong; Wang Liqun; Guo Weiping; Liu ZHiyong

    1989-09-01

    A pulse injection stimulus-response technique to study the boron mixing and transport performance after boron-loaded liquid was injected into the reactor core is described. The experiment was carried out in a simulation device. The simulation medium was used. The experimental results show that the lower plenum where the injection point located can be simplified to one scale inertial unit and the movement of boron mixture was only transported after it had entered into the fuel elements. The definition of boron initiative mixing fraction η is also given. By using relating data a dimensionless equation is obtained

  3. A Heat Transfer Model for a Stratified Corium-Metal Pool in the Lower Plenum of a Nuclear Reactor

    International Nuclear Information System (INIS)

    Sohal, M.S.; Siefken, L.J.

    1999-01-01

    This preliminary design report describes a model for heat transfer in a corium-metal stratified pool. It was decided to make use of the existing COUPLE model. Currently available correlations for natural convection heat transfer in a pool with and without internal heat generation were obtained. The appropriate correlations will be incorporated in the existing COUPLE model. Heat conduction and solidification modeling will be done with existing algorithms in the COUPLE. Assessment of the new model will be done by simple energy conservation problems

  4. 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;...

  5. Primary heat transfer loop design for the Cascade inertial confinement fusion reactor

    International Nuclear Information System (INIS)

    Murray, K.A.; McDowell, M.W.

    1984-05-01

    This study investigates a heat exchanger and balance of plant design to accompany the Cascade inertial confinement fusion reaction chamber concept. The concept uses solid Li 2 O or other lithium-ceramic granules, held to the wall of a rotating reaction chamber by centrifugal action, as a tritium breeding blanket and first wall protection. The Li 2 O granules enter the chamber at 800 K and exit at 1200 K after absorbing the thermal energy produced by the fusion process

  6. Analysis of reactor material experiments investigating oxide fuel crust stability and heat transfer in jet impingement flow

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Spencer, B.W.

    1985-01-01

    An analysis is presented of the crust stability and heat transfer behavior in the CSTI-1, CSTI-3, and CWTI-11 reactor material experiments in which a jet of molten oxide fuel at approx. 160 0 K above its freezing temperature was impinged normally upon stainless steel plates initially at 300 and 385 K. The major issue is the existence of nonexistence of a stable solidified layer of fuel, or crust, interstitial to the flowing hot fuel and the steel substrate, tending to insulate the steel from the hot molten fuel. A computer model was developed to predict the heatup of thermocouples imbedded immediately beneath the surface of the plate for both of the cases in which a stable crust is assumed to be either present or absent during the impingement phase. Comparison of the model calculations with the measured thermocouple temperatures indicates that a protective crust was present over nearly all of the plate surface area throughout the impingement process precluding major melting of the plate steel. However, the experiments also show evidence for very localized and isolated steel melting as revealed by localized and isolated pitting of the steel surface and the response of thermocouples located within the pitted region

  7. 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.)

  8. CORTAP: a coupled neutron kinetics-heat transfer digital computer program for the dynamic simulation of the high temperature gas cooled reactor core

    International Nuclear Information System (INIS)

    Cleveland, J.C.

    1977-01-01

    CORTAP (Core Transient Analysis Program) was developed to predict the dynamic behavior of the High Temperature Gas Cooled Reactor (HTGR) core under normal operational transients and postulated accident conditions. CORTAP is used both as a stand-alone component simulation and as part of the HTGR nuclear steam supply (NSS) system simulation code ORTAP. The core thermal neutronic response is determined by solving the heat transfer equations for the fuel, moderator and coolant in an average powered region of the reactor core. The space independent neutron kinetics equations are coupled to the heat transfer equations through a rapidly converging iterative technique. The code has the capability to determine conservative fuel, moderator, and coolant temperatures in the ''hot'' fuel region. For transients involving a reactor trip, the core heat generation rate is determined from an expression for decay heat following a scram. Nonlinear effects introduced by temperature dependent fuel, moderator, and coolant properties are included in the model. CORTAP predictions will be compared with dynamic test results obtained from the Fort St. Vrain reactor owned by Public Service of Colorado, and, based on these comparisons, appropriate improvements will be made in CORTAP

  9. Analytical heat transfer

    CERN Document Server

    Han, Je-Chin

    2012-01-01

    … it will complete my library … [and] complement the existing literature on heat transfer. It will be of value for both graduate students and faculty members.-Bengt Sunden, Lund University, Sweden

  10. Development of local heat transfer and pressure drop models for pebble bed high temperature gas-cooled reactor cores - HTR2008-58296

    International Nuclear Information System (INIS)

    McLaughlin, B.; Worsley, M.; Stainsby, R.; Grief, A.; Dennier, A.; Macintosh, S.; Van Heerden, E.

    2008-01-01

    This paper describes pressure drop and heat transfer coefficient predictions for a typical coolant flow within the core of a pebble bed reactor (PBR) by examining a representative group of pebbles remote from the reflector region. The three- dimensional steady state flow and heat transfer predictions utilized in this work are obtained from a computational fluid dynamics (CFD) model created in the commercial software ANSYS FLUENT TM . This work utilizes three RANS turbulence models and the Chilton-Colburn analogy for heat transfer. A methodology is included in this paper for creating a quality unstructured mesh with prismatic surface layers on a random arrangement of touching pebbles. The results of the model are validated by comparing them with the correlations of the German KTA rules for a PBR. (authors)

  11. Introduction to heat transfer

    International Nuclear Information System (INIS)

    Weisman, J.

    1983-01-01

    Heat may be defined as that form of energy which spontaneously flows between two bodies, or two regions of a body, by virtue of a temperature difference. The second law of thermodynamics tells us that we cannot have heat flow from a low temperature to high temperature without doing work. Heat flows spontaneously from a high temperature to a low temperature region. Thermodynamics, which is concerned with equilibrium states, cannot tell us anything about the rate of heat flow in the presence of a finite temperature difference. It is to the discipline of heat transfer to which we must turn for this answer

  12. 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

  13. 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

  14. Thermal-hydraulic analysis of heat transfer in subchannels of the European high performance supercritical Water-Cooled Reactor for different CFD turbulence models

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Landy Y.; Rojas, Leorlen Y.; Gamez, Abel; Rosales, Jesus; Gonzalez, Daniel; Garcia, Carlos, E-mail: lcastro@instec.cu, E-mail: leored1984@gmail.com, E-mail: agamezgmf@gmail.com, E-mail: jrosales@instec.cu, E-mail: danielgonro@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Oliveira, Carlos Brayner de, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Dominguez, Dany S., E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Pos-Graduacao em Modelagem Computacional

    2015-07-01

    Chosen as one of six Generation‒IV nuclear-reactor concepts, Supercritical Water-cooled Reactors (SCWRs) are expected to have high thermal efficiencies within the range of 45 - 50% owing to the reactor's high pressures and outlet temperatures. In this reactor, the primary water enters the core under supercritical-pressure condition (25 MPa) at a temperature of 280 deg C and leaves it at a temperature of up to 510 deg C. Due to the significant changes in the physical properties of water at supercritical-pressure, the system is susceptible to local temperature, density and power oscillations. The behavior of supercritical water into the core of the SCWR, need to be sufficiently studied. Most of the methods available to predict the effects of the heat transfer phenomena within the pseudocritical region are based on empirical one-directional correlations, which do not capture the multidimensional effects and do not provide accurate results in regions such as the deteriorated heat transfer regime. In this paper, computational fluid dynamics (CFD) analysis was carried out to study the thermal-hydraulic behavior of supercritical water flows in sub-channels of a typical European High Performance Light Water Reactor (HPLWR) fuel assembly using commercial CFD code CFX-14. It was determined the steady-state equilibrium parameters and calculated the temperature and density distributions. A comparative study for different turbulence models were carried out and the obtained results are discussed. (author)

  15. Thermal-hydraulic analysis of heat transfer in subchannels of the European high performance supercritical Water-Cooled Reactor for different CFD turbulence models

    International Nuclear Information System (INIS)

    Castro, Landy Y.; Rojas, Leorlen Y.; Gamez, Abel; Rosales, Jesus; Gonzalez, Daniel; Garcia, Carlos; Oliveira, Carlos Brayner de; Dominguez, Dany S.

    2015-01-01

    Chosen as one of six Generation‒IV nuclear-reactor concepts, Supercritical Water-cooled Reactors (SCWRs) are expected to have high thermal efficiencies within the range of 45 - 50% owing to the reactor's high pressures and outlet temperatures. In this reactor, the primary water enters the core under supercritical-pressure condition (25 MPa) at a temperature of 280 deg C and leaves it at a temperature of up to 510 deg C. Due to the significant changes in the physical properties of water at supercritical-pressure, the system is susceptible to local temperature, density and power oscillations. The behavior of supercritical water into the core of the SCWR, need to be sufficiently studied. Most of the methods available to predict the effects of the heat transfer phenomena within the pseudocritical region are based on empirical one-directional correlations, which do not capture the multidimensional effects and do not provide accurate results in regions such as the deteriorated heat transfer regime. In this paper, computational fluid dynamics (CFD) analysis was carried out to study the thermal-hydraulic behavior of supercritical water flows in sub-channels of a typical European High Performance Light Water Reactor (HPLWR) fuel assembly using commercial CFD code CFX-14. It was determined the steady-state equilibrium parameters and calculated the temperature and density distributions. A comparative study for different turbulence models were carried out and the obtained results are discussed. (author)

  16. Heat pipe nuclear reactor for space power

    Science.gov (United States)

    Koening, D. R.

    1976-01-01

    A heat-pipe-cooled nuclear reactor has been designed to provide 3.2 MWth to an out-of-core thermionic conversion system. The reactor is a fast reactor designed to operate at a nominal heat-pipe temperature of 1675 K. Each reactor fuel element consists of a hexagonal molybdenum block which is bonded along its axis to one end of a molybdenum/lithium-vapor heat pipe. The block is perforated with an array of longitudinal holes which are loaded with UO2 pellets. The heat pipe transfers heat directly to a string of six thermionic converters which are bonded along the other end of the heat pipe. An assembly of 90 such fuel elements forms a hexagonal core. The core is surrounded by a thermal radiation shield, a thin thermal neutron absorber, and a BeO reflector containing boron-loaded control drums.

  17. 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-...

  18. Friction pressure drop and heat transfer coefficient of two-phase flow in helically coiled tube once-through steam generator for integrated type marine water reactor

    International Nuclear Information System (INIS)

    Nariai, Hideki; Kobayashi, Michiyuki; Matsuoka, Takeshi.

    1982-01-01

    Two-phase friction pressure drop and heat transfer coefficients in a once-through steam generator with helically coiled tubes were investigated with the model test rig of an integrated type marine water reactor. As the dimensions of the heat transfer tubes and the thermal-fluid conditions are almost the same as those of real reactors, the data applicable directly to the real reactor design were obtained. As to the friction pressure drop, modified Kozeki's prediction which is based on the experimental data by Kozeki for coiled tubes, agreed the best with the experimental data. Modified Martinelli-Nelson's prediction which is based on Martinelli-Nelson's multiplier using Ito's equation for single-phase flow in coiled tube, agreed within 30%. The effect of coiled tube on the average heat transfer coefficients at boiling region were small, and the predictions for straight tube could also be applied to coiled tube. Schrock-Grossman's correlation agreed well with the experimental data at the pressures of lower than 3.5 MPa. It was suggested that dryout should be occurred at the quality of greater than 90% within the conditions of this report. (author)

  19. Design of a natural draft air-cooled condenser and its heat transfer characteristics in the passive residual heat removal system for 10 MW molten salt reactor experiment

    International Nuclear Information System (INIS)

    Zhao, Hangbin; Yan, Changqi; Sun, Licheng; Zhao, Kaibin; Fa, Dan

    2015-01-01

    As one of the Generation IV reactors, Molten Salt Reactor (MSR) has its superiorities in satisfying the requirements on safety. In order to improve its inherent safety, a concept of passive residual heat removal system (PRHRS) for the 10 MW Molten Salt Reactor Experiment (MSRE) was put forward, which mainly consisted of a fuel drain tank, a feed water tank and a natural draft air-cooled condenser (NDACC). Besides, several valves and pipes are also included in the PRHRS. A NDACC for the PRHRS was preliminarily designed in this paper, which contained a finned tube bundle and a chimney. The tube bundle was installed at the bottom of the chimney for increasing the velocity of the air across the bundle. The heat transfer characteristics of the NDACC were investigated by developing a model of the PRHRS using C++ code. The effects of the environmental temperature, finned tube number and chimney height on heat removal capacity of the NDACC were analyzed. The results show that it has sufficient heat removal capacity to meet the requirements of the residual heat removal for MSRE. The effects of these three factors are obvious. With the decay heat reducing, the heat dissipation power declines after a short-time rise in the beginning. The operation of the NDACC is completely automatic without the need of any external power, resulting in a high safety and reliability of the reactor, especially once the accident of power lost occurs to the power plant. - Highlights: • A model to study the heat transfer characteristics of the NDACC was developed. • The NDACC had sufficient heat removal capacity to remove the decay heat of MSRE. • NDACC heat dissipation power depends on outside temperature and condenser geometry. • As time grown, the effects of outside temperature and condenser geometry diminish. • The NDACC could automatically adjust its heat removal capacity

  20. Experimental Measurements of Heat Transfer through a Lunar Regolith Simulant in a Vibro-Fluidized Reactor Oven

    Science.gov (United States)

    Nayagam, Vedha; Berger, Gordon M.; Sacksteder, Kurt R.; Paz, Aaron

    2012-01-01

    Extraction of mission consumable resources such as water and oxygen from the planetary environment provides valuable reduction in launch-mass and potentially extends the mission duration. Processing of lunar regolith for resource extraction necessarily involves heating and chemical reaction of solid material with processing gases. Vibrofluidization is known to produce effective mixing and control of flow within granular media. In this study we present experimental results for vibrofluidized heat transfer in lunar regolith simulants (JSC-1 and JSC-1A) heated up to 900 C. The results show that the simulant bed height has a significant influence on the vibration induced flow field and heat transfer rates. A taller bed height leads to a two-cell circulation pattern whereas a single-cell circulation was observed for a shorter height. Lessons learned from these test results should provide insight into efficient design of future robotic missions involving In-Situ Resource Utilization.

  1. Natural convection heat transfer in SIGMA experiment

    International Nuclear Information System (INIS)

    Lee, Seung Dong; Lee, Gang Hee; Suh, Kune Yull

    2004-01-01

    A loss-of-coolant accident (LOCA) results in core melt formation and relocation at various locations within the reactor core over a considerable period of time. If there is no effective cooling mechanism, the core debris may heat up and commence natural circulation. The high temperature pool of molten core material will threaten the structural integrity of the reactor vessel. The extent and urgency of this threat depend primarily upon the intensity of the internal heat sources and upon the consequent distribution of the heat fluxes on the vessel walls in contact with the molten core material pools. In such a steady molten pool convection state, the thermal loads against the vessel would be determined by the in-vessel heat transfer distribution involving convective and conductive heat transfer from the decay-heated core material pool to the lower head wall in contact with the core material. In this study, upward and downward heat transfer fraction ratio is focused on

  2. 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...

  3. A two-cavity reactor for solar chemical processes: heat transfer model and application to carbothermic reduction of ZnO

    International Nuclear Information System (INIS)

    Wieckert, Christian; Palumbo, Robert; Frommherz, Ulrich

    2004-01-01

    A 5 kW two-cavity beam down reactor for the solar thermal decomposition of ZnO with solid carbon has been developed and tested in a solar furnace. Initial exploratory experiments show that it operates with a solar to chemical energy conversion efficiency of about 15% when the solar flux entering the reactor is 1300 kW/m 2 , resulting in a reaction chamber temperature of about 1500 K. The solid products have a purity of nearly 100% Zn. Furthermore, the reactor has been described by a numerical model that combines radiant and conduction heat transfer with the decomposition kinetics of the ZnO-carbon reaction. The model is based on the radiosity exchange method. For a given solar input, the model estimates cavity temperatures, Zn production rates, and the solar to chemical energy conversion efficiency. The model currently makes use of two parameters which are determined from the experimental results: conduction heat transfer through the reactor walls enters the model as a lumped term that reflects the conduction loss during the experiments, and the rate of the chemical reaction includes an experimentally determined term that reflects the effective amount of ZnO and CO participating in the reactor. The model output matches well the experimentally determined cavity temperatures. It suggests that reactors built with this two-cavity concept already on this small scale can reach efficiencies exceeding 25%, if operated with a higher solar flux or if one can reduce conduction heat losses through better insulation and if one can maintain or improve the effective amount of ZnO and CO that participates in the reaction

  4. 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...

  5. 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.

  6. 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.)

  7. 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)

  8. Enhanced Condensation Heat Transfer

    Science.gov (United States)

    Rose, John Winston

    The paper gives some personal observations on various aspects of enhanced condensation heat transfer. The topics discussed are external condensation (horizontal low-finned tubes and wire-wrapped tubes), internal condensation (microfin tubes and microchannels) and Marangoni condensation of binary mixtures.

  9. Experiments on the Heat Transfer and Natural Circulation Characteristics of the Passive Residual Heat Removal System for the Advanced Integral Type Reactor

    International Nuclear Information System (INIS)

    Park, Hyun-Sik; Choi, Ki-Yong; Cho, Seok; Park, Choon-Kyung; Lee, Sung-Jae; Song, Chul-Hwa; Chung, Moon-Ki; Lee, Un-Chul

    2004-01-01

    Experiments on the heat transfer characteristics and natural circulation performance of the passive residual heat removal system (PRHRS) for the SMART-P have been performed using the high temperature/high pressure thermal-hydraulic test facility (VISTA). The VISTA facility consists of the primary loop, the secondary loop, the PRHRS loop, and auxiliary systems to simulate the SMART-P, a pilot plant of the SMART. The primary loop is composed of the steam generator (SG) primary side, a simulated core, a main coolant pump, and loop piping, and the PRHRS loop consists of the SG secondary side, a PRHRS heat exchanger, and loop piping. The natural circulation performance of the PRHRS, the heat transfer characteristics of the PRHRS heat exchangers and the emergency cooldown tank (ECT), and the thermal-hydraulic behavior of the primary loop are intensively investigated. The experimental results show that the coolant flows steadily in the PRHRS loop and the heat transfers through the PRHRS heat exchanger and the emergency cooldown tank are sufficient enough to enable the natural circulation of the coolant. The results also show that the core decay heat can be sufficiently removed from the primary loop with the operation of the PRHRS. (authors)

  10. Liquid metal heat transfer issues

    International Nuclear Information System (INIS)

    Hoffman, H.W.; Yoder, G.L.

    1984-01-01

    An alkali liquid metal cooled nuclear reactor coupled with an alkali metal Rankine cycle provides a practicable option for space systems/missions requiring power in the 1 to 100 MW(e) range. Thermal issues relative to the use of alkali liquid metals for this purpose are identified as these result from the nature of the alkali metal fluid itself, from uncertainties in the available heat transfer correlations, and from design and performance requirements for system components operating in the earth orbital microgravity environment. It is noted that, while these issues require further attention to achieve optimum system performance, none are of such magnitude as to invalidate this particular space power concept

  11. Effect of fluid-to-structure heat transfer on the structural damage potential to a liquid-metal fast breeder reactor

    International Nuclear Information System (INIS)

    Hakim, S.J.; Abramson, P.B.

    1979-01-01

    Deterministic calculations simulating a hypothetical accident in a liquid-metal fast breeder reactor that leads to a hydrodynamic disassembly of the core have been carried out to estimate the system's damage potential due to the vapor-pressure-driven expansion of molten core material and its dependency on the heat transfer to the remaining structure. These calculations ignored the effect on the work potential of sodium left in the core during the disassembly. Results indicate that steel cladding in the upper axial blankets and fission gas plenum acts as a thermodynamic energy sink that could reduce the total thermodynamic work energy by between one and two orders of magnitude, provided little or no sodium is present in the core at the time of interaction. These results have been found to be insensitive to the rate of heat transferred from the molten fuel to the molten steel that comprises the molten core material

  12. HEAT TRANSFER METHOD

    Science.gov (United States)

    Gambill, W.R.; Greene, N.D.

    1960-08-30

    A method is given for increasing burn-out heat fluxes under nucleate boiling conditions in heat exchanger tubes without incurring an increase in pumping power requirements. This increase is achieved by utilizing a spinning flow having a rotational velocity sufficient to produce a centrifugal acceleration of at least 10,000 g at the tube wall. At this acceleration the heat-transfer rate at burn out is nearly twice the rate which can be achieved in a similar tube utilizing axial flow at the same pumping power. At higher accelerations the improvement over axial flow is greater, and heat fluxes in excess of 50 x 10/sup 6/ Btu/hr/sq ft can be achieved.

  13. Heat Transfer Phenomena of Supercritical Fluids

    Energy Technology Data Exchange (ETDEWEB)

    Krau, Carmen Isabella; Kuhn, Dietmar; Schulenberg, Thomas [Forschungszentrum Karlsruhe, Institute for Nuclear and Energy Technologies, 76021 Karlsruhe (Germany)

    2008-07-01

    In concepts for supercritical water cooled reactors, the reactor core is cooled and moderated by water at supercritical pressures. The significant temperature dependence of the fluid properties of water requires an exact knowledge of the heat transfer mechanism to avoid fuel pin damages. Near the pseudo-critical point a deterioration of heat transfer might happen. Processes, that take place in this case, are not fully understood and are due to be examined systematically. In this paper a general overview on the properties of supercritical water is given, experimental observations of different authors will be reviewed in order to identify heat transfer phenomena and onset of occurrence. The conceptional design of a test rig to investigate heat transfer in the boundary layer will be discussed. Both, water and carbon dioxide, may serve as operating fluids. The loop, including instrumentation and safety devices, is shown and suitable measuring methods are described. (authors)

  14. Predictions of hydrodynamic vibrations for some cylindrical structures of nuclear reactors in contact with the heat-transferring fluid

    International Nuclear Information System (INIS)

    Milan, D.; Berthollon, G.

    1979-01-01

    This paper deals with the study of the vibratory behaviour of thin cylindrical structures under the turbulence of the heat transferring fluid. These simple structures constitute an idealization, often adequate, of several structures used in nuclear hydraulics. It is shown that an empiric representation of the excitation spatial field makes it possible, in the experimented cases, to predict with a fair accuracy the vibratory response of the first modes. The cases of application of such an approach are discussed. (author)

  15. Feasibility study of the university of Utah TRIGA reactor power upgrade - part II: Thermohydraulics and heat transfer study in respect to cooling system requirements and design

    Directory of Open Access Journals (Sweden)

    Babitz Philip

    2013-01-01

    Full Text Available The thermodynamic conditions of the University of Utah's TRIGA Reactor were simulated using SolidWorks Flow Simulation, Ansys, Fluent and PARET-ANL. The models are developed for the reactor's currently maximum operating power of 90 kW, and a few higher power levels to analyze thermohydraulics and heat transfer aspects in determining a design basis for higher power including the cost estimate. It was found that the natural convection current becomes much more pronounced at higher power levels with vortex shedding also occurring. A departure from nucleate boiling analysis showed that while nucleate boiling begins near 210 kW it remains in this state and does not approach the critical heat flux at powers up to 500 kW. Based on these studies, two upgrades are proposed for extended operation and possibly higher reactor power level. Together with the findings from Part I studies, we conclude that increase of the reactor power is highly feasible yet dependable on its purpose and associated investments.

  16. Heat Transfer Analogies

    Energy Technology Data Exchange (ETDEWEB)

    Bhattacharyya, A

    1965-11-15

    This report contains descriptions of various analogues utilised to study different steady-state and unsteady-state heat transfer problems. The analogues covered are as follows: 1 . Hydraulic: a) water flow b) air flow 2. Membrane 3. Geometric Electrical: a) Electrolytic-tank b) Conducting sheet 4. Network; a) Resistance b) R-C A comparison of the different analogues is presented in the form of a table.

  17. Heat Transfer Analogies

    International Nuclear Information System (INIS)

    Bhattacharyya, A.

    1965-11-01

    This report contains descriptions of various analogues utilised to study different steady-state and unsteady-state heat transfer problems. The analogues covered are as follows: 1 . Hydraulic: a) water flow b) air flow 2. Membrane 3. Geometric Electrical: a) Electrolytic-tank b) Conducting sheet 4. Network; a) Resistance b) R-C A comparison of the different analogues is presented in the form of a table

  18. Nucleate boiling heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Saiz Jabardo, J.M. [Universidade da Coruna (Spain). Escola Politecnica Superior], e-mail: mjabardo@cdf.udc.es

    2009-07-01

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

  19. Nucleate boiling heat transfer

    International Nuclear Information System (INIS)

    Saiz Jabardo, J.M.

    2009-01-01

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

  20. Heat extraction from HTGR reactor

    International Nuclear Information System (INIS)

    Balajka, J.; Princova, H.

    1986-01-01

    The analysis of an HTGR reactor energy balance showed that steam reforming of natural gas or methane is the most suitable process of utilizing the high-temperature heat. Basic mathematical relations are derived allowing to perform a general energy balance of the link between steam reforming and reactor heat output. The results of the calculation show that the efficiency of the entire reactor system increases with increasing proportion of heat output for steam reforming as against heat output for the steam generator. This proportion, however, is limited with the output helium temperature from steam reforming. It is thus always necessary to use part of the reactor heat output for the steam cycle involving electric power generation or low-potential heat generation. (Z.M.)

  1. 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

  2. Heat transfer and pressure drop of the reactor fuel element with polyzonal spiral finning; Prelaz toplote i pad pritiska reaktorskog gorivnog elementa sa polizonalno-spiralnim orebrenjem

    Energy Technology Data Exchange (ETDEWEB)

    Oka, S; Becirspahic, S [Institute of Nuclear Sciences Vinca, Beograd (Serbia and Montenegro)

    1964-06-15

    Heat transfer and pressure drop of the reactor fuel element with polyzonal spiral finning were investigated. Longitudinal and circumferential distributions of Sr-number of finnings in the fuel element are given. Dependences of St{sub kmin} and St{sub ksr} on the Re number are derived. The influence of gap between two fuel elements on the heat transfer, pressure drop is presented dependent on the Re number. The influence of mutual position of flow separators of two neighbouring fuel elements on the pressure drop and heat transfer is shown as well. Investigations were performed in the range of Re numbers from 15000 to 100000. Ispitivan je prelaz toplote i pad pritiska modela reaktorskog gorivnog elementa sa polizonalno-spiralnim orebrenjem. Dat je uzduznu i obimni raspored Sr-broja na orebrenju gorivnog elementa. Izvedene su zavisnosti St{sub kmin} i St{sub ksr} u funkciji od Re-broja. Pokazan je uticaj prekida izmedju dva gorivna elementa na prelaz toplote i pad pritiska u zavisnosti od Re-broja. Pokazan je uticaj medjusobnog polozaja razdeljivaca struje dva susedna gorivna elementa na pad pritiska i prelaz toplote. Ispitivanja su vrsena u oblasti Re-brojeva od 15000 do 100000 (author)

  3. 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.)

  4. A susceptor heating structure in MOVPE reactor by induction heating

    International Nuclear Information System (INIS)

    Li, Zhiming; Li, Hailing; Zhang, Jincheng; Li, Jinping; Jiang, Haiying; Fu, Xiaoqian; Han, Yanbin; Xia, Yingjie; Huang, Yimei; Yin, Jianqin; Zhang, Lejuan; Hu, Shigang

    2014-01-01

    A novel susceptor with a revolutionary V-shaped slot of solid of revolution form is proposed in the metalorganic vapor phase epitaxy (MOVPE) reactor by induction heating. This slot changes the heat transfer rate as the generated heat is transferred from the high temperature region of the susceptor to the substrate, which improves the uniformity of the substrate temperature distribution. By using finite element method (FEM), the susceptor with this structure for heating the substrate of six inches in diameter is optimized. It is observed that this optimized susceptor with the V-shaped slot makes the uniformity of the substrate temperature distribution improve more than 80%, which can be beneficial to the film growth. - Highlights: •A novel susceptor with V-shaped slot in MOVPE reactor is proposed. •Temperature in the substrate is optimized. •Great temperature uniformity of the substrate is obtained

  5. The Effective Convectivity Model for Simulation and Analysis of Melt Pool Heat Transfer in a Light Water Reactor Pressure Vessel Lower Head

    International Nuclear Information System (INIS)

    Tran, Chi Thanh

    2009-09-01

    Severe accidents in a Light Water Reactor (LWR) have been a subject of intense research for the last three decades. The research in this area aims to reach understanding of the inherent physical phenomena and reduce the uncertainties in their quantification, with the ultimate goal of developing models that can be applied to safety analysis of nuclear reactors, and to evaluation of the proposed accident management schemes for mitigating the consequences of severe accidents. In a hypothetical severe accident there is likelihood that the core materials will be relocated to the lower plenum and form a decay-heated debris bed (debris cake) or a melt pool. Interactions of core debris or melt with the reactor structures depend to a large extent on the debris bed or melt pool thermal hydraulics. In case of inadequate cooling, the excessive heat would drive the structures' overheating and ablation, and hence govern the vessel failure mode and timing. In turn, threats to containment integrity associated with potential ex-vessel steam explosions and ex-vessel debris uncoolability depend on the composition, superheat, and amount of molten corium available for discharge upon the vessel failure. That is why predictions of transient melt pool heat transfer in the reactor lower head, subsequent vessel failure modes and melt characteristics upon the discharge are of paramount importance for plant safety assessment. The main purpose of the present study is to develop a method for reliable prediction of melt pool thermal hydraulics, namely to establish a computational platform for cost-effective, sufficiently-accurate numerical simulations and analyses of core Melt-Structure-Water Interactions in the LWR lower head during a postulated severe core-melting accident. To achieve the goal, an approach to efficient use of Computational Fluid Dynamics (CFD) has been proposed to guide and support the development of models suitable for accident analysis. The CFD method, on the one hand, is

  6. 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...

  7. Features of an emergency heat-conducting path in reactors about lead-bismuth and lead heat-carriers

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Bokova, T.A.; Molodtsov, A.A.

    2006-01-01

    The reactor emergency heat removal systems should transfer heat from the surface of reactor core fuel element claddings to the primary circuit followed by heat transfer to the environment. One suggests three design approaches for emergency heat removal systems in lead-bismuth and lead cooled reactor circuits that take account of the peculiar nature of their features. Application of the discussed systems for emergency heat removal improves safety of lead-bismuth and lead cooled reactor plants [ru

  8. 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.

  9. System-Level Heat Transfer Analysis, Thermal- Mechanical Cyclic Stress Analysis, and Environmental Fatigue Modeling of a Two-Loop Pressurized Water Reactor. A Preliminary Study

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish [Argonne National Lab. (ANL), Argonne, IL (United States); Soppet, William [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, Saurin [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, Ken [Argonne National Lab. (ANL), Argonne, IL (United States)

    2015-01-03

    This report provides an update on an assessment of environmentally assisted fatigue for light water reactor components under extended service conditions. This report is a deliverable in April 2015 under the work package for environmentally assisted fatigue under DOE's Light Water Reactor Sustainability program. In this report, updates are discussed related to a system level preliminary finite element model of a two-loop pressurized water reactor (PWR). Based on this model, system-level heat transfer analysis and subsequent thermal-mechanical stress analysis were performed for typical design-basis thermal-mechanical fatigue cycles. The in-air fatigue lives of components, such as the hot and cold legs, were estimated on the basis of stress analysis results, ASME in-air fatigue life estimation criteria, and fatigue design curves. Furthermore, environmental correction factors and associated PWR environment fatigue lives for the hot and cold legs were estimated by using estimated stress and strain histories and the approach described in NUREG-6909. The discussed models and results are very preliminary. Further advancement of the discussed model is required for more accurate life prediction of reactor components. This report only presents the work related to finite element modelling activities. However, in between multiple tensile and fatigue tests were conducted. The related experimental results will be presented in the year-end report.

  10. Study on enhancement of heat transfer of RVACS

    International Nuclear Information System (INIS)

    Nishi, Yoshihisa; Kinoshita, Izumi

    1989-01-01

    As for the enhancement of heat transfer on Reactor Vessel Auxiliary Cooling System (RVACS), utilization of high porosity porous bodies have been proposed by the last report. This report describe the experimental results to evaluate heat transfer performance of the porous bodies and to estimate the extrapolation to long heat transfer surface such as reactor scale. Following are typical results. (1) Usually the Heat Transfer coefficient at the lower reaches is smoller than that of the upper reaches. But Using with the high porosity porous bodies, the Heat Transfer coefficient at the lower reaches remains a constant value against distance from entrance point or a increase slightly compared to that of the upper reaches because of the effect of thermal radiation. (2) From the results of Heat Transfer coefficients against distance from the entrance point, the increasing ratio of enhancement of heat removal in the case of reactor scale is about 1.3. (author)

  11. 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.

  12. Measurement of cooling coil film heat transfer coefficient with polymer reaction proceeding in a stirred batch reactor; Jugo sonai ni okeru hanno shinko ni tomonau reikyaku coil no kyomaku netsudentatsu keisu no keiji henka

    Energy Technology Data Exchange (ETDEWEB)

    Yamamoto, K [Soken Chemical and Engineering Co. Ltd., Saitama (Japan); Nishi, K; Kaminoyama, M; Kamiwano, M [Yokohama National University, Yokohama (Japan). Faculty of Engineering

    1996-09-10

    In radical additional solution polymerization, the viscosity increases with reaction progress. It is important to evaluate beforehand the cooling capacity of the reactor, which worsens with the process. In this study, a stirred batch reactor with both a paddle and a helical screw impeller were studied, and measurements were made for the dynamic changes of the film heat transfer coefficient of the cooling coil with progress of the polymer reaction. We found the change could be evaluated by the calculating heat balance of the generated heat, the viscous dissipation energy and the sensible heat change under conditions of monomer conversion and changing viscosity. 11 refs., 7 figs.

  13. 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

  14. Radiative heat transfer in the Na mist dispersion over the hot surface of liquid Na in the cooling system of nuclear reactor

    International Nuclear Information System (INIS)

    Kunitomo, T.; Shafey, H.M.

    1980-01-01

    The analysis has been carried out for the radiative heat transfer in the Na mist dispersion enclosed between the hot surface of liquid Na at temperature Tsub(n) and the cold surface of Na deposit at Tsub(c). The model selected for the present study represents the Na mist formed in a sodium cooled fast breeder reactor in which the condensed liquid particles are dispersed in the mixture of the Ar cover gas and the Na vapor. The analysis is based on replacing the inhomogeneous dispersing medium by three discrete homogeneous layers, and formulating the transfer equation for the monochromatic radiation in each layer according to the Chandrasekhar theory. The numerical calculations of the radiative qsub(r) and convective qsub(c) heat transfers have been performed for the wave length range lambda=1.6-30 μm and are compared. The qsub(r) has the same order of magnitude as the qsub(c) for all conditions of the mist dispersions. Both qsub(r) and qsub(c) increase by nearly equal rates with the increase of Tsub(H) and decrease by different rates with increasing Tsub(c). Variations of the particle diameter of the Na mist do not change substantially the qsub(r). Both qsub(r) and qsub(c) decrease slightly with the increase in the total thickness of the Na mist dispersion

  15. 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...

  16. Heat transfer physics

    CERN Document Server

    Kaviany, Massoud

    2014-01-01

    This graduate textbook describes atomic-level kinetics (mechanisms and rates) of thermal energy storage, transport (conduction, convection, and radiation), and transformation (various energy conversions) by principal energy carriers. The approach combines the fundamentals of molecular orbitals-potentials, statistical thermodynamics, computational molecular dynamics, quantum energy states, transport theories, solid-state and fluid-state physics, and quantum optics. The textbook presents a unified theory, over fine-structure/molecular-dynamics/Boltzmann/macroscopic length and time scales, of heat transfer kinetics in terms of transition rates and relaxation times, and its modern applications, including nano- and microscale size effects. Numerous examples, illustrations, and homework problems with answers that enhance learning are included. This new edition includes applications in energy conversion (including chemical bond, nuclear, and solar), expanded examples of size effects, inclusion of junction quantum tr...

  17. Application of heat pipes in nuclear reactors for passive heat removal

    Energy Technology Data Exchange (ETDEWEB)

    Haque, Z.; Yetisir, M., E-mail: haquez@aecl.ca [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2013-07-01

    This paper introduces a number of potential heat pipe applications in passive (i.e., not requiring external power) nuclear reactor heat removal. Heat pipes are particularly suitable for small reactors as the demand for heat removal is significantly less than commercial nuclear power plants, and passive and reliable heat removal is required. The use of heat pipes has been proposed in many small reactor designs for passive heat removal from the reactor core. This paper presents the application of heat pipes in AECL's Nuclear Battery design, a small reactor concept developed by AECL. Other potential applications of heat pipes include transferring excess heat from containment to the atmosphere by integrating low-temperature heat pipes into the containment building (to ensure long-term cooling following a station blackout), and passively cooling spent fuel bays. (author)

  18. 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)

  19. A Secondary Flow Effect on the Heat and Mass Transfer Processes in the Finned Rod Bundles of Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    A. A. Dunaitsev

    2017-01-01

    Full Text Available In nuclear power engineering a need to justify an operability of products and their components is of great importance. In high-temperature gas reactors, the critical element affecting the facility reliability is the fuel rod cladding, which in turn leads to the need to gain knowledge in the field of gas dynamics and heat transfer in the reactor core and to increase the detail of the calculation results. For the time being, calculations of reactor core are performed using the proven techniques of per-channel calculations, which show good representativeness and count rate. However, these techniques require additional experimental studies to describe correctly the inter-channel exchange, which, being taken into account, largely affects the pattern of the temperature fields in the region under consideration. Increasingly more relevant and demandable are numerical simulation methods of fluid and gas dynamics, as well as of heat exchange, which consist in the direct solution of the system of differential equations of mass balance, kinetic moment, and energy. Calculation of reactor cores or rod bundles according these techniques does not require additional experimental studies and allows us to obtain the local distributions of flow characteristics in the bundle and the flow characteristics that are hard to measure in the physical experiment.The article shows the calculation results and their analysis for an infinite rod lattice of the reactor core. The results were obtained by the technique of modelling one rod of a regular lattice using the periodic boundary conditions, followed by translating the results to the neighbouring rods. In channels of complex shape, there are secondary flows caused by changes in the channel geometry along the flow and directed across the main front of the flow. These secondary flows in the reactor cores with rods spaced by the winding wire lead to a redistribution of the coolant along the channel section, which in turn

  20. CFD simulation of a coolant flow and a heat transfer in a pebble bed reactor - HTR2008-58334

    International Nuclear Information System (INIS)

    In, W. K.; Lee, W. J.; Hassan, Y. A.

    2008-01-01

    This CFD study is to simulate a coolant(gas) flow and heat transfer in a PBR core during a normal operation. This study used a pebble array with direct area contacts among the pebbles which is one of the pebbles arrangements for a detailed simulation of PBR core CFD studies. A CFD model is developed to more adequately represent the pebbles randomly stacked in the PBR core. The CFD predictions showed a large variation of the temperature on the pebble surface as well as in the pebble core. The temperature drop in the outer graphite layer is smaller than that in the pebble-core region. This is because the thermal conductivity of graphite is higher than the fuel (UO, mixture) conductivity in the pebble core. Higher pebble surface temperature is predicted downstream of the pebble contact due to a reverse flow. Multiple vortices are predicted to occur downstream of the spherical pebbles due to a flow separation. The coolant flow structure and fuel temperature in the PBR core appears to largely depend on the in-core distribution of the pebbles. (authors)

  1. Effect of lattice deformation on temperature fields and heat transfer in the fuel elements of characteristic zones for a model of fast reactor fuel assembly

    International Nuclear Information System (INIS)

    Zhukov, A.V.; Matyukhin, N.M.; Sviridenko, E.Ya.

    1980-01-01

    Given are the experimental results for temperature fields in the model assembly in nonribbed simulators of the BN-600-type reactor fuel elements in the course of deformation of the lattice caused by shifting of the central and peripheral (lateral, angular) fuel elements by the value of the gap between the fuel elements (the limiting case when the fuel elements touch each other along the whole length). An assembly consisting of 37 electroheated pipes arranged in a triangular lattice with a relative step of S/d=1.185 is used as a model. The experiments were carried out on the sodium stand at constant energy release along the length of the fuel element simulators and at the Pe number changing in the 14-700 range. The data obtained show considerable increase of nonuniformities of the fuel element temperatures for characteristic zones of the fuel cassette assembly models of the fast reactor at deviations of the lattice geometric sizes from the nominal ones. For the central nonribbed element the temperature nonuniformity increases approximately 7.5 times and for the lateral element approximately 6 times when the elements touch each other along the whole length. The shift the central nonribbed element by the value of the gap between the fu.el elements leads to the decrease of heat transfer in comparison with heat transfer at the nominal geometry approximately 3-7 times in the 10-450 range for the Pe numbers. It is shown that the coolant temperature distribution along the assembly radius has a complex character (with a peak between the centre and the perifery) caused by redistribution of coolant consumptions due to fuel element lattice deformation

  2. Development of monitoring system using acoustic emission for detection of helium gas leakage for primary cooling system and flow-induced vibration for heat transfer tube of heat exchangers for the High Temperature Engineering Test Reactor (HTTR)

    International Nuclear Information System (INIS)

    Tachibana, Yukio; Kunitomi, Kazuhiko; Furusawa, Takayuki; Shinozaki, Masayuki; Satoh, Yoshiyuki; Yanagibashi, Minoru

    1998-10-01

    The High Temperature Engineering Test Reactor (HTTR) uses helium gas for its primary coolant, whose leakage inside reactor containment vessel is considered in design of the HTTR. It is necessary to detect leakage of helium gas at an early stage so that total amount of the leakage should be as small as possible. On the other hand, heat transfer tubes of heat exchangers of the HTTR are designed not to vibrate at normal operation, but the flow-induced vibration is to be monitored to provide against an emergency. Thus monitoring system of acoustic emission for detection of primary coolant leakage and vibration of heat transfer tubes was developed and applied to the HTTR. Before the application to the HTTR, leakage detection test was performed using 1/4 scaled model of outer tube of primary concentric hot gas duct. Result of the test covers detectable minimum leakage rate and effect of difference in gas, pressure, shape of leakage path and distance from the leaking point. Detectable minimum leakage rate was about 5 Ncc/sec. The monitoring system is promising in leakage detection, though countermeasure to noise is to be needed after the HTTR starts operating. (author)

  3. Heat dissipating nuclear reactor

    Science.gov (United States)

    Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    Disclosed is a nuclear reactor containment adapted to retain and cool core debris in the unlikely event of a core meltdown and subsequent breach in the reactor vessel. The reactor vessel is seated in a cavity which has a thick metal sidewall that is integral with a thick metal basemat at the bottom of the cavity. The basemat extends beyond the perimeter of the cavity sidewall. Underneath the basemat is a porous bed with water pipes and steam pipes running into it. Water is introduced into the bed and converted into steam which is vented to the atmosphere. A plurality of metal pilings in the form of H-beams extend from the metal base plate downwardly and outwardly into the earth.

  4. Heat transfer phenomena revelant to severe accidents

    International Nuclear Information System (INIS)

    Dallman, R.J.; Duffey, R.B.

    1990-01-01

    A number of aspects of severe accidents have been reviewed, particularly in relation to the heat transfer characteristics and the important phenomena. It is shown that natural circulation, forced convection, and entrainment phenomena are important for both the reactor system and ex-vessel events. It is also shown that the phenomena related to two component enhanced heat transfer is important in the pool of molten core debris, in relation to the potential for attack of the liner structure and the concrete. These mechanisms are discussed within the general context of severe accident progression

  5. Heat transfer phenomena relevant to severe accidents

    International Nuclear Information System (INIS)

    Dallman, R.J.; Duffey, R.B.

    1990-01-01

    A number of aspects of severe accidents have been reviewed, particularly in relation to the heat transfer characteristics and the important phenomena. It is shown that natural circulation, forced convection, and entrainment phenomena are important for both the reactor system and ex-vessel events. It is also shown that the phenomena related to two component enhanced heat transfer is important in the pool of molten core debris, in relation to the potential for attack of the liner structure and the concrete. These mechanisms are discussed within the general context of severe accident progression. 26 refs

  6. Handbook of heat and mass transfer. Volume 2

    International Nuclear Information System (INIS)

    Cheremisinoff, N.P.

    1986-01-01

    This two-volume series, the work of more than 100 contributors, presents advanced topics in industrial heat and mass transfer operations and reactor design technology. Volume 2 emphasizes mass transfer and reactor design. Some of the contents discussed are: MASS TRANSFER PRINCIPLES - Effect of turbulence promoters on mass transfer. Mass transfer principles with homogeneous and heterogeneous reactions. Convective diffusion with reactions in a tube. Transient mass transfer onto small particles and drops. Modeling heat and mass transport in falling liquid films. Heat and mass transfer in film absorption. Multicomponent mass transfer: theory and applications. Diffusion limitation for reaction in porous catalysts. Kinetics and mechanisms of catalytic deactivation. DISTILLATION AND EXTRACTION - Generalized equations of state for process design. Mixture boiling. Estimating vapor pressure from normal boiling points of hydrocarbons. Estimating liquid and vapor molar fractions in distillation columns. Principles of multicomponent distillation. Generalized design methods for multicomponent distillation. Interfacial films in inorganic substances extraction. Liquid-liquid extraction in suspended slugs. MULTIPHASE REACTOR SYSTEMS - Reaction and mass transport in two-phase reactors. Mass transfer and kinetics in three-phase reactors. Estimating liquid film mass transfer coefficients in randomly packed columns. Designing packed tower wet scrubbers - emphasis on nitrogen oxides. Gas absorption in aerated mixers. Axial dispersion and heat transfer in gas-liquid bubble columns. Operation and design of trickle-bed reactors

  7. Heat exchangers in heavy water reactor systems

    International Nuclear Information System (INIS)

    Mehta, S.K.

    1988-01-01

    Important features of some major heat exchange components of pressurized heavy water reactors and DHRUVA research reactor are presented. Design considerations and nuclear service classifications are discussed

  8. 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 ...

  9. Heat transfer in high-level waste management

    International Nuclear Information System (INIS)

    Dickey, B.R.; Hogg, G.W.

    1979-01-01

    Heat transfer in the storage of high-level liquid wastes, calcining of radioactive wastes, and storage of solidified wastes are discussed. Processing and storage experience at the Idaho Chemical Processing Plant are summarized for defense high-level wastes; heat transfer in power reactor high-level waste processing and storage is also discussed

  10. The Thermos process heat reactor

    International Nuclear Information System (INIS)

    Lerouge, Bernard

    1979-01-01

    The THERMOS process heat reactor was born from the following idea: the hot water energy vector is widely used for heating purposes in cities, so why not save on traditional fossil fuels by simply substituting a nuclear boiler of comparable power for the classical boiler installed in the same place. The French Atomic Energy Commission has techniques for heating in the big French cities which provide better guarantees for national independence and for the environment. This THERMOS technique would result in a saving of 40,000 to 80,000 tons of oil per year [fr

  11. Proceedings of the ANS/ASME/NRC international topical meeting on nuclear reactor thermal-hydraulics: fundamental aspects of two-phase flow and boiling heat transfer

    International Nuclear Information System (INIS)

    1980-08-01

    Separate abstracts are included for each of the papers presented concerning critical flow of two-phase mixtures; two-phase flow instrumentation; critical heat flux and effects of local disturbances; heat transfer and rewetting during reflood; hydrodynamic mechanisms in boiling heat transfer; and entrainment and droplet deposition in two-phase flow. Five papers have been previously abstracted and input to the data base

  12. 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.

  13. 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

  14. Heat transfer to accelerating gas flows

    International Nuclear Information System (INIS)

    Kennedy, T.D.A.

    1978-01-01

    The development of fuels for gas-cooled reactors has resulted in a number of 'gas loop' experiments in materials-testing research reactors. In these experiments, efforts are made to reproduce the conditions expected in gas-cooled power reactors. Constant surface temperatures are sought over a short (300 mm) fuelled length, and because of entrance effects, an accelerating flow is required to increase the heat transfer down-stream from the entrance. Strong acceleration of a gas stream will laminarise the flow even at Reynolds Numbers up to 50000, far above values normally associated with laminar flow. A method of predicting heat transfer in this situation is presented here. An integral method is used to find the velocity profile; this profile is then used in an explicit finite-difference solution of the energy equation to give a temperature profile and resultant heat-transfer coefficient values. The Kline criterion, which compares viscous and disruptive forces, is used to predict whether the flow will be laminar. Experimental results are compared with predictions, and good agreement is found to exist. (author)

  15. Heat-pipe thermionic reactor concept

    DEFF Research Database (Denmark)

    Storm Pedersen, E.

    1967-01-01

    Main components are reactor core, heat pipe, thermionic converter, secondary cooling system, and waste heat radiator; thermal power generated in reactor core is transported by heat pipes to thermionic converters located outside reactor core behind radiation shield; thermionic emitters are in direct...

  16. 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

  17. Advances in heat transfer enhancement

    CERN Document Server

    Saha, Sujoy Kumar; Sundén, Bengt; Wu, Zan

    2016-01-01

    This Brief addresses the phenomena of heat transfer enhancement. A companion edition in the SpringerBrief Subseries on Thermal Engineering and Applied Science to three other monographs including “Critical Heat Flux in Flow Boiling in Microchannels,” this volume is idea for professionals, researchers, and graduate students concerned with electronic cooling.

  18. Heat removal by natural convection in a RPR reactor

    International Nuclear Information System (INIS)

    Sampaio, P.A.B. de

    1987-01-01

    In this paper natural convection in RPR reactor is analysed. The effect of natural convection valves size on cladding temperature is studied. The reactor channel heat transfer problem is solved using finite elements in a two-dimensional analysis. Results show that two valves with Φ = 0.16 m are suited to keep coolant and cladding temperatures below 73 0 C. (author) [pt

  19. Fundamental principles of heat transfer

    CERN Document Server

    Whitaker, Stephen

    1977-01-01

    Fundamental Principles of Heat Transfer introduces the fundamental concepts of heat transfer: conduction, convection, and radiation. It presents theoretical developments and example and design problems and illustrates the practical applications of fundamental principles. The chapters in this book cover various topics such as one-dimensional and transient heat conduction, energy and turbulent transport, forced convection, thermal radiation, and radiant energy exchange. There are example problems and solutions at the end of every chapter dealing with design problems. This book is a valuable int

  20. Heat transfer enhancement with nanofluids

    CERN Document Server

    Bianco, Vincenzo; Nardini, Sergio; Vafai, Kambiz

    2015-01-01

    Properties of NanofluidSamuel Paolucci and Gianluca PolitiExact Solutions and Their Implications in Anomalous Heat TransferWenhao Li, Chen Yang and Akira NakayamaMechanisms and Models of Thermal Conductivity in NanofluidsSeung-Hyun Lee and Seok Pil JangExperimental Methods for the Characterization of Thermophysical Properties of NanofluidsSergio Bobbo and Laura FedeleNanofluid Forced ConvectionGilles RoyExperimental Study of Convective Heat Transfer in NanofluidsEhsan B. Haghighi, Adi T. Utomo, Andrzej W. Pacek and Björn E. PalmPerformance of Heat Exchangers Using NanofluidsBengt Sundén and Za

  1. Heat transfer for plasma facing components

    International Nuclear Information System (INIS)

    Boyd, R.D.; Meng, X.; Maughan, H.

    1995-01-01

    Although the high heat flux requirements for plasma-facing components have been reduced drastically from 40.0 MW/m 2 to near 10.0 MW/m 2 , there are still some refinements needed. This paper highlights: (1) recent accomplishments and pinpoints new thermal solutions and problem areas of immediate concern to the development of plasma-facing components, and (2) next generation thermal hydraulic problems which must be addressed to insure safety and reliability in component operation. More specifically the near-term thermal hydraulic problems entail: (1) generating an appropriate data base to insure the development of single-side heat flux correlations; and (2) adapting the existing vast uniform heat flux literature to the case of non-uniform heat flux distributions found in plasma facing components in fusion reactors. Results are presented for the latter task which includes: (a) an accurate subcooled flow boiling curve correlation for the partial nucleate boiling regime which can be adapted using previously proposed correlations relating single-side boundary heat flux to heat transfer, in uniformly heated channels, (b) the evaluation of the possibility of using the existing literature directly with redefined parameters, and (c) an estimation of circumferential variations in the heat transfer coefficient

  2. Reliable prediction of heat transfer coefficient in three-phase bubble column reactor via adaptive neuro-fuzzy inference system and regularization network

    Science.gov (United States)

    Garmroodi Asil, A.; Nakhaei Pour, A.; Mirzaei, Sh.

    2018-04-01

    In the present article, generalization performances of regularization network (RN) and optimize adaptive neuro-fuzzy inference system (ANFIS) are compared with a conventional software for prediction of heat transfer coefficient (HTC) as a function of superficial gas velocity (5-25 cm/s) and solid fraction (0-40 wt%) at different axial and radial locations. The networks were trained by resorting several sets of experimental data collected from a specific system of air/hydrocarbon liquid phase/silica particle in a slurry bubble column reactor (SBCR). A special convection HTC measurement probe was manufactured and positioned in an axial distance of 40 and 130 cm above the sparger at center and near the wall of SBCR. The simulation results show that both in-house RN and optimized ANFIS due to powerful noise filtering capabilities provide superior performances compared to the conventional software of MATLAB ANFIS and ANN toolbox. For the case of 40 and 130 cm axial distance from center of sparger, at constant superficial gas velocity of 25 cm/s, adding 40 wt% silica particles to liquid phase leads to about 66% and 69% increasing in HTC respectively. The HTC in the column center for all the cases studied are about 9-14% larger than those near the wall region.

  3. Measurements of Critical Heat Flux using Mass Transfer System

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Seung Hyun; Chung Bum Jin [Kyunghee University, Yongin (Korea, Republic of)

    2016-05-15

    In a severe accident, the reactor vessel is heated by the decay heat from core melts and the outer surface of reactor vessel is cooled by the natural convection of water pool. When the heat flux increases, boiling will start. Further increase of the heat flux may result in the CHF, which is generated by the bubble combinations. The CHF means that the reactor vessel was separated with coolant and wall temperature is raised rapidly. It may damage the reactor vessel. Also the CHF indicates the maximum cooling capability of the system. Therefore, the CHF has been used as a criterion for the regulatory and licensing. Mechanism of hydrogen vapor bubbles generated and combined can be simulated water bubbles mechanism. And also the both heat and mass transfer mechanism of CHF can be identified in the same methods. Therefore, the CHF phenomena can be simulated enough by mass transfer.

  4. Heat transfer and mechanical interactions in fusion nuclear systems

    International Nuclear Information System (INIS)

    Nygren, R.E.

    1984-01-01

    This general review of design issues in heat transfer and mechanical interactions of the first wall, blanket and shield systems of tokamak and mirror fusion reactors begins with a brief introduction to fusion nuclear systems. The design issues are summarized in tables and the following examples are described to illustrate these concerns: the surface heating of limiters, heat transfer from solid breeders, MHD effects in liquid metal blankets, mechanical loads from electromagnetic transients and remote maintenance

  5. Homogenization of some radiative heat transfer models: application to gas-cooled reactor cores; Homogeneisation de modeles de transferts thermiques et radiatifs: application au coeur des reacteurs a caloporteur gaz

    Energy Technology Data Exchange (ETDEWEB)

    El Ganaoui, K

    2006-09-15

    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)

  6. Gasification with nuclear reactor heat

    International Nuclear Information System (INIS)

    Weisbrodt, I.A.

    1977-01-01

    The energy-political ultimate aims for the introduction of nuclear coal gasification and the present state of technology concerning the HTR reactor, concerning gasification and heat exchanging components are outlined. Presented on the plans a) for hydro-gasification of lignite and for steam gasification of pit coal for the production of synthetic natural gas, and b) for the introduction of a nuclear heat system. The safety and environmental problems to be expected are portrayed. The main points of development, the planned prototype plant and the schedule of the project Pototype plant Nuclear Process heat (PNP) are specified. In a market and economic viability study of nuclear coal gasification, the application potential of SNG, the possible construction programme for the FRG, as well as costs and rentability of SNG production are estimated. (GG) [de

  7. 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.

  8. Engineering heat transfer

    CERN Document Server

    Annaratone, Donatello

    2010-01-01

    This book is a generalist textbook; it is designed for anybody interested in heat transmission, including scholars, designers and students. Two criteria constitute the foundation of Annaratone's books, including the present one. The first one consists of indispensable scientific rigor without theoretical exasperation. The inclusion in the book of some theoretical studies, even if admirable for their scientific rigor, would have strengthened the scientific foundation of this publication, yet without providing the reader with further applicable know-how. The second criterion is to deliver practi

  9. Essentials of radiation heat transfer

    CERN Document Server

    Balaji

    2014-01-01

    Essentials of Radiation Heat Transfer is a textbook presenting the essential, fundamental information required to gain an understanding of radiation heat transfer and equips the reader with enough knowledge to be able to tackle more challenging problems. All concepts are reinforced by carefully chosen and fully worked examples, and exercise problems are provided at the end of every chapter. In a significant departure from other books on this subject, this book completely dispenses with the network method to solve problems of radiation heat transfer in surfaces. It instead presents the powerful radiosity-irradiation method and shows how this technique can be used to solve problems of radiation in enclosures made of one to any number of surfaces. The network method is not easily scalable. Secondly, the book introduces atmospheric radiation, which is now being considered as a potentially important area, in which engineers can contribute to the technology of remote sensing and atmospheric sciences in general, b...

  10. Heat transfer and critical heat flux in a spiral flow in an asymmetrical heated tube

    International Nuclear Information System (INIS)

    Boscary, J.; Association Euratom-CEA, Centre d'Etudes Nucleaires de Cadarache, 13 - Saint-Paul-lez-Durance

    1997-03-01

    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)

  11. 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

  12. OMR type process heat reactor

    International Nuclear Information System (INIS)

    Franzetti, Franco.

    1974-01-01

    A description is given of an OMR type reactor for heat generation. It includes a vessel the upper part of which is shut by a plug. The lower part of the vessel includes a core of fuel elements and is filled with an organic liquid. Over this there is a middle area filled with an inert gas. The plug includes an upper part forming a closure and resting around its edge on the vessel, and a lower part fixed under the closure and composed of a hollow cylindrical tank fitted with a bottom and filled with another organic liquid. The height of the cylindrical tank is such that, increased by the height of the first organic liquid in the lower area and above the core, it provides biological protection. The cooling system includes a heat exchanger and a pump to move the liquid from the lower part of the core and to inject some as spray into that part of the vessel filled with the inert gas. When loading and unloading, after the reactor is shut down, the clear organic liquid contained in the plug is discharged into the reactor vessel in such a way that it does not mix with the opaque organic liquid already contained in the vessel, and in that the opaque organic liquid is emptied out [fr

  13. Basic heat and mass transfer

    CERN Document Server

    Mills, A F

    1999-01-01

    The Second Edition offers complete coverage of heat transfer with broad up-to-date coverage that includes an emphasis on engineering relevance and on problem solving. Integrates software to assist the reader in efficiently calculations. Carefully orders material to make book more reader-friendly and accessible. Offers an extensive introduction to heat exchange design to enhance the engineering and design content of course to satisfy ABET requirements. For professionals in engineering fields.

  14. Nuclear reactor auxiliary heat removal system

    International Nuclear Information System (INIS)

    Thompson, R.E.; Pierce, B.L.

    1977-01-01

    An auxiliary heat removal system to remove residual heat from gas-cooled nuclear reactors is described. The reactor coolant is expanded through a turbine, cooled in a heat exchanger and compressed by a compressor before reentering the reactor coolant. The turbine powers both the compressor and the pump which pumps a second fluid through the heat exchanger to cool the reactor coolant. A pneumatic starter is utilized to start the turbine, thereby making the auxiliary heat removal system independent of external power sources

  15. Heat transfer between immiscible liquids enhanced by gas bubbling

    International Nuclear Information System (INIS)

    Greene, G.A.; Schwarz, C.E.; Klages, J.; Klein, J.

    1982-08-01

    The phenomena of core-concrete interactions impact upon containment integrity of light water reactors (LWR) following postulated complete meltdown of the core by containment pressurization, production of combustible gases, and basemat penetration. Experiments have been performed with non-reactor materials to investigate one aspect of this problem, heat transfer between overlying immiscible liquids whose interface is disturbed by a transverse non-condensable gas flux emanating from below. Hydrodynamic studies have been performed to test a criterion for onset of entrainment due to bubbling through the interface and subsequent heat transfer studies were performed to assess the effect of bubbling on interfacial heat transfer rates, both with and without bubble induced entrainment. Non-entraining interfacial heat transfer data with mercury-water/oil fluid pairs were observed to be bounded from below within a factor of two to three by the Szekeley surface renewal heat transfer model. However heat transfer data for fluid pairs which are found to entrain (water-oil), believed to be characteristic of molten reactor core-concrete conditions, were measured to be up to two orders of magnitude greater than surface renewal predictions and are calculated by a simple entrainment heat transfer model

  16. Five MW Nuclear Heating Reactor

    International Nuclear Information System (INIS)

    Zhang Dafang; Dong Duo; Su Qingshan

    1997-01-01

    The 5 MW Nuclear Heating Reactor (NHR-5) developed and designed by the Institute of Nuclear Energy Technology (INET) has been operated for four winter seasons since 1989. During the time of commissioning and operation a number of experiments including self-stability, self-regulation, and simulation of ATWS etc. were carried out. Some operating experiences such as water chemistry, radiation protection and environmental impacts and so on were also obtained at the same time. All of these results demonstrate the design of the NHR-5 is successful. (author). 9 refs, 11 figs, 5 tabs

  17. Five MW Nuclear Heating Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dafang, Zhang; Duo, Dong; Qingshan, Su [Institute of Nuclear Energy and Technology, Tsingua Univ., Beijing (China)

    1997-09-01

    The 5 MW Nuclear Heating Reactor (NHR-5) developed and designed by the Institute of Nuclear Energy Technology (INET) has been operated for four winter seasons since 1989. During the time of commissioning and operation a number of experiments including self-stability, self-regulation, and simulation of ATWS etc. were carried out. Some operating experiences such as water chemistry, radiation protection and environmental impacts and so on were also obtained at the same time. All of these results demonstrate the design of the NHR-5 is successful. (author). 9 refs, 11 figs, 5 tabs.

  18. Supercritical water gasification with decoupled pressure and heat transfer modules

    KAUST Repository

    Dibble, Robert

    2017-09-14

    The present invention discloses a system and method for supercritical water gasification (SCWG) of biomass materials wherein the system includes a SCWG reactor and a plurality of heat exchangers located within a shared pressurized vessel, which decouples the function of containing high pressure from the high temperature function. The present invention allows the heat transfer function to be conducted independently from the pressure transfer function such that the system equipment can be designed and fabricated in manner that would support commercial scaled-up SCWG operations. By using heat exchangers coupled to the reactor in a series configuration, significant efficiencies are achieved by the present invention SCWG system over prior known SCWG systems.

  19. Heat Transfer Characteristics of SiC-coated Heat Pipe for Passive Decay Heat Removal

    International Nuclear Information System (INIS)

    Kim, Kyung Mo; Kim, In Guk; Jeong, Yeong Shin; Bang, In Cheol

    2014-01-01

    The main concern with the Fukushima accident was the failure of active and passive core cooling systems. The main function of existing passive decay heat removal systems is feeding additional coolant to the reactor core. Thus, an established emergency core cooling system (ECCS) cannot operate properly because of impossible depressurization under the station blackout (SBO) condition. Therefore, a new concept for passive decay heat removal system is required. In this study, an innovative hybrid control rod concept is considered for passive in-core decay heat removal that differs from the existing direct vessel injection core cooling system and passive auxiliary feedwater system (PAFS). The heat transfer between the evaporator and condenser sections occurs by phase change of the working fluid and capillary action induced by wick structures installed on the inner wall of the heat pipe. In this study, a hybrid control rod is developed to take the roles of both neutron absorption and heat removal by combining the functions of a heat pipe and control rod. Previous studies on enhancing the heat removal capacity of heat pipes used nanofluids, self-rewetting fluids, various wick structures and condensers. Many studies have examined the thermal performances of heat pipes using various nanofluids. They concluded that the enhanced thermal performance of the heat pipe using nanofluids is due to nanoparticle deposition on the wick structures. Thus, the wick structure of heat pipes has been modified by nanoparticle deposition to enhance the heat removal capacity. However, previous studies used relatively small heat pipes and narrow ranges of heat loads. The environment of a nuclear reactor is very specific, and the decay heat produced by fission products after shutdown is relatively large. Thus, this study tested a large-scale heat pipe over a wide range of power. The concept of a hybrid heat pipe for an advanced in-core decay heat removal system was introduced for complete

  20. Heat Transfer Characteristics of SiC-coated Heat Pipe for Passive Decay Heat Removal

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kyung Mo; Kim, In Guk; Jeong, Yeong Shin; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

    2014-10-15

    The main concern with the Fukushima accident was the failure of active and passive core cooling systems. The main function of existing passive decay heat removal systems is feeding additional coolant to the reactor core. Thus, an established emergency core cooling system (ECCS) cannot operate properly because of impossible depressurization under the station blackout (SBO) condition. Therefore, a new concept for passive decay heat removal system is required. In this study, an innovative hybrid control rod concept is considered for passive in-core decay heat removal that differs from the existing direct vessel injection core cooling system and passive auxiliary feedwater system (PAFS). The heat transfer between the evaporator and condenser sections occurs by phase change of the working fluid and capillary action induced by wick structures installed on the inner wall of the heat pipe. In this study, a hybrid control rod is developed to take the roles of both neutron absorption and heat removal by combining the functions of a heat pipe and control rod. Previous studies on enhancing the heat removal capacity of heat pipes used nanofluids, self-rewetting fluids, various wick structures and condensers. Many studies have examined the thermal performances of heat pipes using various nanofluids. They concluded that the enhanced thermal performance of the heat pipe using nanofluids is due to nanoparticle deposition on the wick structures. Thus, the wick structure of heat pipes has been modified by nanoparticle deposition to enhance the heat removal capacity. However, previous studies used relatively small heat pipes and narrow ranges of heat loads. The environment of a nuclear reactor is very specific, and the decay heat produced by fission products after shutdown is relatively large. Thus, this study tested a large-scale heat pipe over a wide range of power. The concept of a hybrid heat pipe for an advanced in-core decay heat removal system was introduced for complete

  1. The Development of an INL Capability for High Temperature Flow, Heat Transfer, and Thermal Energy Storage with Applications in Advanced Small Modular Reactors, High Temperature Heat Exchangers, Hybrid Energy Systems, and Dynamic Grid Energy Storage C

    International Nuclear Information System (INIS)

    Sun, Xiaodong; Zhang, Xiaoqin; Kim, Inhun; O'Brien, James; Sabharwall, Piyush

    2014-01-01

    The overall goal of this project is to support Idaho National Laboratory in developing a new advanced high temperature multi fluid multi loop test facility that is aimed at investigating fluid flow and heat transfer, material corrosion, heat exchanger characteristics and instrumentation performance, among others, for nuclear applications. Specifically, preliminary research has been performed at The Ohio State University in the following areas: 1. A review of fluoride molten salts' characteristics in thermal, corrosive, and compatibility performances. A recommendation for a salt selection is provided. Material candidates for both molten salt and helium flow loop have been identified. 2. A conceptual facility design that satisfies the multi loop (two coolant loops [i.e., fluoride molten salts and helium]) multi purpose (two operation modes [i.e., forced and natural circulation]) requirements. Schematic models are presented. The thermal hydraulic performances in a preliminary printed circuit heat exchanger (PCHE) design have been estimated. 3. An introduction of computational methods and models for pipe heat loss analysis and cases studies. Recommendations on insulation material selection have been provided. 4. An analysis of pipe pressure rating and sizing. Preliminary recommendations on pipe size selection have been provided. 5. A review of molten fluoride salt preparation and chemistry control. An introduction to the experience from the Molten Salt Reactor Experiment at Oak Ridge National Laboratory has been provided. 6. A review of some instruments and components to be used in the facility. Flowmeters and Grayloc connectors have been included. This report primarily presents the conclusions drawn from the extensive review of literatures in material selections and the facility design progress at the current stage. It provides some useful guidelines in insulation material and pipe size selection, as well as an introductory review of facility process and

  2. The Development of an INL Capability for High Temperature Flow, Heat Transfer, and Thermal Energy Storage with Applications in Advanced Small Modular Reactors, High Temperature Heat Exchangers, Hybrid Energy Systems, and Dynamic Grid Energy Storage C

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Xiaodong [The Ohio State Univ., Columbus, OH (United States); Zhang, Xiaoqin [The Ohio State Univ., Columbus, OH (United States); Kim, Inhun [The Ohio State Univ., Columbus, OH (United States); O' Brien, James [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

    The overall goal of this project is to support Idaho National Laboratory in developing a new advanced high temperature multi fluid multi loop test facility that is aimed at investigating fluid flow and heat transfer, material corrosion, heat exchanger characteristics and instrumentation performance, among others, for nuclear applications. Specifically, preliminary research has been performed at The Ohio State University in the following areas: 1. A review of fluoride molten salts’ characteristics in thermal, corrosive, and compatibility performances. A recommendation for a salt selection is provided. Material candidates for both molten salt and helium flow loop have been identified. 2. A conceptual facility design that satisfies the multi loop (two coolant loops [i.e., fluoride molten salts and helium]) multi purpose (two operation modes [i.e., forced and natural circulation]) requirements. Schematic models are presented. The thermal hydraulic performances in a preliminary printed circuit heat exchanger (PCHE) design have been estimated. 3. An introduction of computational methods and models for pipe heat loss analysis and cases studies. Recommendations on insulation material selection have been provided. 4. An analysis of pipe pressure rating and sizing. Preliminary recommendations on pipe size selection have been provided. 5. A review of molten fluoride salt preparation and chemistry control. An introduction to the experience from the Molten Salt Reactor Experiment at Oak Ridge National Laboratory has been provided. 6. A review of some instruments and components to be used in the facility. Flowmeters and Grayloc connectors have been included. This report primarily presents the conclusions drawn from the extensive review of literatures in material selections and the facility design progress at the current stage. It provides some useful guidelines in insulation material and pipe size selection, as well as an introductory review of facility process and components.

  3. 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)

  4. 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)

  5. 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

  6. Saclay Reactor: acquired knowledge by two years experience in heat transfer using compressed gas; La pile de Saclay experience acquise en deux ans sur le transfert de chaleur par gaz comprime

    Energy Technology Data Exchange (ETDEWEB)

    Yvon, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1955-07-01

    Describes the conception and functioning of a new reactor (EL-2) using compressed gas as primary coolant. The aim of the use of compressed gas as primary coolant is to reduce the quantity of heavy water used in the functioning of the reactor. Description of the reactor vessel (dimensions, materials, reflector and protection). Description of the cells and the circulation of the gas within the cells. A complete explanation of the control and regulating of the reaction by the ionization chamber is given. Heavy water is used as modulator: it describes the heavy water system and its recombination system. The fuel slugs are cooled by compressed gas: its system is described as well as the blower and the heat exchanger system. Water is supplied by a cooling tower which means the reactor power is dependant of the atmospheric conditions. Particular attention has been given to the tightness of the different systems used. The relation between neutron flow and the thermal output is discussed: the thermal output can be calculated by measuring the gas flow and its heating or by measuring the neutron flow within the reactor, both methods gives closed results. Reactivity study: determination of the different factors which induce a variation of reactivity. Heat transfer: discussion on the use of different heat transfer systems, determination of the required chemical and physical properties of the primary coolant as well as the discussion of the nuclear and thermal requirements for the choice of it. A comparison between the use of nitrogen and carbon dioxide gas shows an advantage in using nitrogen with the existing knowledge. Reflexion on the relevance of this work and the future perspectives of the use of compressed gas as primary coolant. (M.P.)

  7. Heat transfer in intermediate heat exchanger under low flow rate conditions

    International Nuclear Information System (INIS)

    Mochizuki, H.

    2008-01-01

    The present paper describes the heat transfer in intermediate heat exchangers (IHXs) of liquid metal cooled fast reactors when flow rate is low such as a natural circulation condition. Although empirical correlations of heat transfer coefficients for IHX were derived using test data at the fast reactor 'Monju' and 'Joyo' and also at the 50 MW steam generator facility, the heat transfer coefficient was very low compared to the well known correlation for liquid metals proposed by Seban-Shimazaki. The heat conduction in IHX was discussed as a possible cause of the low Nusselt number. As a result, the heat conduction is not significant under the natural circulation condition, and the heat conduction term in the energy equation can be neglected in the one-dimensional plant dynamics calculation. (authors)

  8. Validation of heat transfer models for gap cooling

    International Nuclear Information System (INIS)

    Okano, Yukimitsu; Nagae, Takashi; Murase, Michio

    2004-01-01

    For severe accident assessment of a light water reactor, models of heat transfer in a narrow annular gap between overheated core debris and a reactor pressure vessel are important for evaluating vessel integrity and accident management. The authors developed and improved the models of heat transfer. However, validation was not sufficient for applicability of the gap heat flux correlation to the debris cooling in the vessel lower head and applicability of the local boiling heat flux correlations to the high-pressure conditions. Therefore, in this paper, we evaluated the validity of the heat transfer models and correlations by analyses for ALPHA and LAVA experiments where molten aluminum oxide (Al 2 O 3 ) at about 2700 K was poured into the high pressure water pool in a small-scale simulated vessel lower head. In the heating process of the vessel wall, the calculated heating rate and peak temperature agreed well with the measured values, and the validity of the heat transfer models and gap heat flux correlation was confirmed. In the cooling process of the vessel wall, the calculated cooling rate was compared with the measured value, and the validity of the nucleate boiling heat flux correlation was confirmed. The peak temperatures of the vessel wall in ALPHA and LAVA experiments were lower than the temperature at the minimum heat flux point between film boiling and transition boiling, so the minimum heat flux correlation could not be validated. (author)

  9. Study on boiling heat transfer of high temperature liquid sodium

    International Nuclear Information System (INIS)

    Sakurai, Akira

    1978-01-01

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

  10. Experimental investigation on Heat Transfer Performance of Annular Flow Path Heat Pipe

    International Nuclear Information System (INIS)

    Kim, In Guk; Kim, Kyung Mo; Jeong, Yeong Shin; Bang, In Cheol

    2015-01-01

    Mochizuki et al. was suggested the passive cooling system to spent nuclear fuel pool. Detail analysis of various heat pipe design cases was studied to determine the heat pipes cooling performance. Wang et al. suggested the concept PRHRS of MSR using sodium heat pipes, and the transient performance of high temperature sodium heat pipe was numerically simulated in the case of MSR accident. The meltdown at the Fukushima Daiichi nuclear power plants alarmed to the dangers of station blackout (SBO) accident. After the SBO accident, passive decay heat removal systems have been investigated to prevent the severe accidents. Mochizuki et al. suggested the heat pipes cooling system using loop heat pipes for decay heat removal cooling and analysis of heat pipe thermal resistance for boiling water reactor (BWR). The decay heat removal systems for pressurized water reactor (PWR) were suggested using natural convection mechanisms and modification of PWR design. Our group suggested the concept of a hybrid heat pipe with control rod as Passive IN-core Cooling System (PINCs) for decay heat removal for advanced nuclear power plant. Hybrid heat pipe is the combination of the heat pipe and control rod. In the present research, the main objective is to investigate the effect of the inner structure to the heat transfer performance of heat pipe containing neutron absorber material, B 4 C. The main objective is to investigate the effect of the inner structure in heat pipe to the heat transfer performance with annular flow path. ABS pellet was used instead of B 4 C pellet as cylindrical structures. The thermal performances of each heat pipes were measured experimentally. Among them, concentric heat pipe showed the best performance compared with others. 1. Annular evaporation section heat pipe and annular flow path heat pipe showed heat transfer degradation. 2. AHP also had annular vapor space and contact cooling surface per unit volume of vapor was increased. Heat transfer coefficient of

  11. Small heating reactors for local heating of communities

    International Nuclear Information System (INIS)

    Seifritz, W.

    1985-08-01

    The incentives to introduce relatively small heating reactors for local heating of communities are presented and the reasons why this vertically integrated energy system will meet the requirement of an emission - free substitution system are outlined. (author)

  12. Heat transfer calculations for the High Flux Isotope Reactor (HFIR). Technical specifications: bases for safety limits and limiting safety system settings

    International Nuclear Information System (INIS)

    Sims, T.M.; Swanks, J.H.

    1977-09-01

    Heat transfer analyses, in support of the preparation of the HFIR technical specifications, were made to establish the bases for the safety limits and limiting safety system settings applicable to the HFIR. The results of these analyses, along with the detailed bases, are presented

  13. Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two phase. 11. meeting of the International Association for Hydraulic Research (IAHR) Working Group. Working material

    International Nuclear Information System (INIS)

    2005-01-01

    This Working Material includes the papers presented at the International Meeting 'Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two-phase', which was held 5-9 July 2004 at the State Scientific Center of Russian Federation - Institute for Physics and Power Engineering named after A.I. Leypunsky, in Obninsk near Moscow. The objectives of the meeting were to discuss new results obtained in the field of liquid metal coolant and to recommend the lines of further general physics and applied investigations, with the purpose of validating existing and codes under development for liquid metal cooled advanced and new generation nuclear reactors. Most of the contributions present results of experimental and numerical investigations into velocity, temperature and heat transfer in fuel subassemblies of fast reactors cooled by sodium or lead. In the frame of the meeting a benchmark problem devoted to heat transfer in the model subassembly of the fast reactor BREST-OD-300 was proposed. Experts from 5 countries (Japan, Netherlands, Spain, Republic of Korea, and Russia) took part in this benchmark exercise. The results of the benchmark calculations are summarized in the Working Material. The results of hydrodynamic studies of pressure head chambers and collector systems of liquid metal cooled reactors are presented in a number of papers. Also attention was given to the generalization of experimental data on hydraulic losses in the pipelines in case of mutual influence of local pressure drops, and to the modeling of natural convection in the fuel subassemblies and circuits with liquid metal cooling. Special emphasis at the meeting was placed on thermal hydraulics issues related to the development and design of target systems, such as heat removal in the target unit of the cascade subcritical reactor cooled by liquid salt; the target complex MK-1 for accelerator driven systems cooled by eutectic lead-bismuth alloy; and the test

  14. Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two phase. 11. meeting of the International Association for Hydraulic Research (IAHR) Working Group. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    This Working Material includes the papers presented at the International Meeting 'Hydrodynamics and heat transfer in reactor components cooled by liquid metal coolants in single/two-phase', which was held 5-9 July 2004 at the State Scientific Center of Russian Federation - Institute for Physics and Power Engineering named after A.I. Leypunsky, in Obninsk near Moscow. The objectives of the meeting were to discuss new results obtained in the field of liquid metal coolant and to recommend the lines of further general physics and applied investigations, with the purpose of validating existing and codes under development for liquid metal cooled advanced and new generation nuclear reactors. Most of the contributions present results of experimental and numerical investigations into velocity, temperature and heat transfer in fuel subassemblies of fast reactors cooled by sodium or lead. In the frame of the meeting a benchmark problem devoted to heat transfer in the model subassembly of the fast reactor BREST-OD-300 was proposed. Experts from 5 countries (Japan, Netherlands, Spain, Republic of Korea, and Russia) took part in this benchmark exercise. The results of the benchmark calculations are summarized in the Working Material. The results of hydrodynamic studies of pressure head chambers and collector systems of liquid metal cooled reactors are presented in a number of papers. Also attention was given to the generalization of experimental data on hydraulic losses in the pipelines in case of mutual influence of local pressure drops, and to the modeling of natural convection in the fuel subassemblies and circuits with liquid metal cooling. Special emphasis at the meeting was placed on thermal hydraulics issues related to the development and design of target systems, such as heat removal in the target unit of the cascade subcritical reactor cooled by liquid salt; the target complex MK-1 for accelerator driven systems cooled by eutectic lead-bismuth alloy; and the test

  15. Direct contact heat transfer characteristics between melting alloy and water

    International Nuclear Information System (INIS)

    Kinoshita, Izumi; Nishi, Yoshihisa; Furuya, Masahiro

    1995-01-01

    As a candidate for an innovative steam generator for fast breeder reactors, a heat exchanger with direct contact heat transfer between melting alloy and water was proposed. The evaluation of heat transfer characteristics of this heat exchanger is one of the research subjects for the design and development of the steam generator. In this study, the effect of the pressure on heat transfer characteristics and the required degree of superheating of melting alloy above water saturation temperature are evaluated during the direct contact heat transfer experiment by injecting water into Wood's alloy. In the experiment, the pressure, the temperature of the Wood's alloy, the flow rate of feed water, and the depth of the feed water injection point are varied as parameters. As a result of the experiment, the product of the degree of Wood's alloy superheating above water saturation temperature and the depth of the feed water injection point is constant for each pressure. This constant increases as the pressure rises. (author)

  16. Experimental heat transfer in tube bundle

    International Nuclear Information System (INIS)

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

    1983-01-01

    Previous work has looked for the problem of heat transfer with flow parallel to rod bundle either by treating each rod individually as a separate channel or by treating the bundle as one unit. The present work will consider the existence of both the central and corner rods simultaneously inside the cluster itself under the same working conditions. The test section is geometrically similar to the fuel assembly of the Egyptian Research Reactor-1. The hydro-thermal performance of bundle having 16 - stainless steel tubes arranged in square array of 1.5 pitch to diameter ratio is investigated. Surface temperature and pressure distributions are determined. Average heat transfer coefficient for both central and corner tubes are correlated. Also, pressure drop and friction factor correlations are predicted. The maximum experimental range of the measured parameters are determined in the nonboiling region at 1400 Reynolds number and 3.64 W/cm 2 . It is found that the average heat transfer coefficient of the central tube is higher than that of the corner tube by 27%. Comparison with the previous work shows satisfactory agreement particularly with the circular tubes correlation - Dittus et al. - at 104 Reynolds number

  17. PWR type process heat reactor

    International Nuclear Information System (INIS)

    Aubert, Gilles; Petit, Guy.

    1974-01-01

    The nuclear reactor described is of the pressurized water type. It includes a prestressed concrete vessel, the upper part of which is shut by a closure, and a core surrounded by a core ring. The core fuel assemblies are supported by an initial set of vertical tubes integral with the bottom of the vessel, which serve to guide the rods of the control system. Over the core there is a second set of vertical tubes, able to receive the absorbing part of a control rod when this is raised above the core. An annular pressurizer around the core ring keeps the water in a liquid state. A pump is located above the second set of tubes and is integral with the closure. It circulates the water between the core and the intake of at least one primary heat exchanger, the exchanger (s) being placed between the wall of the vessel and the core ring [fr

  18. Facility with a nuclear district heating reactor

    International Nuclear Information System (INIS)

    Straub, H.

    1988-01-01

    The district heating reactor has a pressure vessel which contains the reactor core and at least one coolant conducting primary heat carrier surrounded by a heat sink. The pressure vessel has two walls with a space between them. This space is connected with a container which contains air as heat isolating medium and water as heat conducting medium. During the normal reactor operation the space is filled by air from the container with the aid of a blower, whereas in the case of a break-down of the cooling system it is filled by water which flows out of the container by gravity after the blower has been switched off. The after-heat, generated in the reactor core during cooling break-down, is removed into the heat sink surrounding the pressure vessel in a safe and simple way. 6 figs

  19. Size control synthesis and characterization of ZnO nanoparticles and its application as ZnO-water based nanofluid in heat transfer enhancement in light water nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sharma, Deepak; Pandey, Krishna Murari [National Institute of Technology Silchar, Assam (India). Dept. of Mechanical Engineering

    2017-03-15

    A novel and facile approach for size-tunable synthesis of ZnO nanoparticle (NPs) is reported. Size-tuning was attained by using PEG (polyethylene glycol) of molecular weights 400 and 4000. ZnO NPs was synthesized using homogeneous chemical precipitation followed by hydrothermal. Here triethylamine (TEA) was used as a hydroxylating agent. As-synthesized ZnO NPs were characterized by powder X-ray diffraction (XRD), transmission electron microscopy (TEM) and Energy Dispersive Spectroscopy (EDS) analysis. Synthesized ZnO nanoparticle was used for the preparation of ZnO-water based nanofluid and its application in heat transfer enhancement in light water nuclear reactor. In this work, ZnO-water based nanofluid of different volume concentration (1%, 2% and 3%) and particle size of 10 nm and 20 nm is used for enhancement in heat transfer in annular channel by using two phase approach. The particle size of 10 nm gives better result for enhancing the heat transfer rate in comparison to 20 nm particle size in nuclear reactor.

  20. 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

  1. Post CHF heat transfer and quenching

    International Nuclear Information System (INIS)

    Nelson, R.A.; Condie, K.G.

    1980-01-01

    This paper describes quantitatively new mechanisms in the post-CHF regime which provide understanding and predictive capability for several current two-phase forced convective heat transfer problems. These mechanisms are important in predicting rod temperature turnaround and quenching during the reflood phase of either a hypothetical loss-of-coolant accident (LOCA) or the FLECHT and Semiscale experiments. The mechanisms are also important to the blowdown phase of a LOCA or the recent Loss-of-Fluid Test (LOFT) experiments L2-2 and L2-3, which were 200% cold leg break transients. These LOFT experiments experienced total core quenching in the early part of the blowdown phase at high (1000 psia) pressures. The mechanisms are also important to certain pressurized water reactor (PWR) operational transients where the reactor may operate in the post-CHF regime for short periods of time. Accurate prediction of the post-CHF heat transfer including core quench during these transients is of prime importance to limit maximum cladding temperatures and prevent cladding deformation

  2. 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.

  3. Dimensional analysis of boiling heat transfer burnout conditions

    International Nuclear Information System (INIS)

    El-Mitwally, E.S.; Raafat, N.M.; Darwish, M.A.

    1979-01-01

    The first criteria in boiling water systems design, such as boiling water reactors, is that no burnout in the core is allowed to exist under any conditions of the reactor operation either during steady state operation or during any of the several postulated accidental transients, such as sudden interruption of coolant flow in the reactor core (due to pump failure or blockage of fuel channel). The aim of the present work is to obtain a correlation for the critical heat flux based on a theoretical study where the mechanism of burn out and the related hydrodynamic and heat transfer equations are considered. 8 refs

  4. 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

  5. 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

  6. 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.

  7. Transient heat transfer characteristics of liquid helium

    International Nuclear Information System (INIS)

    Tsukamoto, Osami

    1976-01-01

    The transient heat transfer characteristics of liquid helium are investigated. The critical burnout heat fluxes for pulsive heating are measured, and empirical relations between the critical burnout heat flux and the length of the heat pulse are given. The burnout is detected by observing the super-to-normal transition of the temperature sensor which is a thin lead film prepared on the heated surface by vacuum evaporation. The mechanism of boiling heat transfer for pulsive heating is discussed, and theoretical relations between the critical burnout heat flux and the length of the heat pulse are derived. The empirical data satisfy the theoretical relations fairly well. (auth.)

  8. High-temperature gas-cooled reactors and process heat

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1980-01-01

    High-Temperature Gas-Cooled Reactors (HTGRs) are fueled with ceramic-coated microspheres of uranium and thorium oxides/carbides embedded in graphite blocks which are cooled with helium. Promising areas of HTGR application are in cogeneration, energy transport using Heat Transfer Salt, recovery of oils from oil shale, steam reforming of methane for chemical production, coal gasification, and in energy transfer using chemical heat jpipes in the long term. Further, HTGRs could be used as the energy source for hydrogen production through thermochemical water splitting in the long term. The potential market for Process Heat HTGRs is 100-200 large units by about the year 2020

  9. 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…

  10. HTGR type reactors for the heat market

    International Nuclear Information System (INIS)

    Oesterwind, D.

    1981-01-01

    Information about the standard of development of the HTGR type reactor are followed by an assessment of its utilization on the heat market. The utilization of HTGR type reactors is considered suitable for the production of synthesis gas, district heat, and industrial process heat. A comparison with a pit coal power station shows the economy of the HTGR. Finally, some aspects of introducing new technologies into the market, i.e. small plants in particular are investigated. (UA) [de

  11. Verification and Validation of Heat Transfer Model of AGREE Code

    Energy Technology Data Exchange (ETDEWEB)

    Tak, N. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Seker, V.; Drzewiecki, T. J.; Downar, T. J. [Department of Nuclear Engineering and Radiological Sciences, Univ. of Michigan, Michigan (United States); Kelly, J. M. [US Nuclear Regulatory Commission, Washington (United States)

    2013-05-15

    The AGREE code was originally developed as a multi physics simulation code to perform design and safety analysis of Pebble Bed Reactors (PBR). Currently, additional capability for the analysis of Prismatic Modular Reactor (PMR) core is in progress. Newly implemented fluid model for a PMR core is based on a subchannel approach which has been widely used in the analyses of light water reactor (LWR) cores. A hexagonal fuel (or graphite block) is discretized into triangular prism nodes having effective conductivities. Then, a meso-scale heat transfer model is applied to the unit cell geometry of a prismatic fuel block. Both unit cell geometries of multi-hole and pin-in-hole types of prismatic fuel blocks are considered in AGREE. The main objective of this work is to verify and validate the heat transfer model newly implemented for a PMR core in the AGREE code. The measured data in the HENDEL experiment were used for the validation of the heat transfer model for a pin-in-hole fuel block. However, the HENDEL tests were limited to only steady-state conditions of pin-in-hole fuel blocks. There exist no available experimental data regarding a heat transfer in multi-hole fuel blocks. Therefore, numerical benchmarks using conceptual problems are considered to verify the heat transfer model of AGREE for multi-hole fuel blocks as well as transient conditions. The CORONA and GAMMA+ codes were used to compare the numerical results. In this work, the verification and validation study were performed for the heat transfer model of the AGREE code using the HENDEL experiment and the numerical benchmarks of selected conceptual problems. The results of the present work show that the heat transfer model of AGREE is accurate and reliable for prismatic fuel blocks. Further validation of AGREE is in progress for a whole reactor problem using the HTTR safety test data such as control rod withdrawal tests and loss-of-forced convection tests.

  12. High-temperature process heat reactor with solid coolant and radiant heat exchange

    International Nuclear Information System (INIS)

    Alekseev, A.M.; Bulkin, Yu.M.; Vasil'ev, S.I.

    1984-01-01

    The high temperature graphite reactor with the solid coolant in which heat transfer is realized by radiant heat exchange is described. Neutron-physical and thermal-technological features of the reactor are considered. The reactor vessel is made of sheet carbon steel in the form of a sealed rectangular annular box. The moderator is a set of graphite blocks mounted as rows of arched laying Between the moderator rows the solid coolant annular layings made of graphite blocks with high temperature nuclear fuel in the form of coated microparticles are placed. The coolant layings are mounted onto ring movable platforms, the continuous rotation of which is realizod by special electric drives. Each part of the graphite coolant laying consecutively passes through the reactor core neutron cut-off zones and technological zone. In the core the graphite is heated up to the temperature of 1350 deg C sufficient for effective radiant heat transfer. In the neutron cut-off zone the chain reaction and further graphite heating are stopped. In the technological zone the graphite transfers the accumulated heat to the walls of technological channels in which the working medium moves. The described reactor is supposed to be used in nuclear-chemical complex for ammonia production by the method of methane steam catalytic conversion

  13. Facilities of fuel transfer for nuclear reactors

    International Nuclear Information System (INIS)

    Wade, E.E.

    1977-01-01

    This invention relates to sodium cooled fast breeder reactors. It particularly concerns facilities for the transfer of fuel assemblies between the reactor core and a fuel transfer area. The installation is simple in construction and enables a relatively small vessel to be used. In greater detail, the invention includes a vessel with a head, fuel assemblies housed in this vessel, and an inlet and outlet for the coolant covering these fuel assemblies. The reactor has a fuel transfer area in communication with this vessel and gear inside the vessel for the transfer of these fuel assemblies. These facilities are borne by the vessel head and serve to transfer the fuel assemblies from the vessel to the transfer area; whilst leaving the fuel assemblies completely immersed in a continuous mass of coolant. A passageway is provided between the vessel and this transfer area for the fuel assemblies. Facilities are provided for closing off this passageway so that the inside of the reactor vessel may be isolated as desired from this fuel transfer area whilst the reactor is operating [fr

  14. Turbulence model for melt pool natural convection heat transfer

    International Nuclear Information System (INIS)

    Kelkar, K.M.; Patankar, S.V.

    1994-01-01

    Under severe reactor accident scenarios, pools of molten core material may form in the reactor core or in the hemispherically shaped lower plenum of the reactor vessel. Such molten pools are internally heated due to the radioactive decay heat that gives rise to buoyant flows in the molten pool. The flow in such pools is strongly influenced by the turbulent mixing because the expected Rayleigh numbers under accidents scenarios are very high. The variation of the local heat flux over the boundaries of the molten pools are important in determining the subsequent melt progression behavior. This study reports results of an ongoing effort towards providing a well validated mathematical model for the prediction of buoyant flow and heat transfer in internally heated pool under conditions expected in severe accident scenarios

  15. Comparison of correlations for heat transfer in sphere-pac beds

    International Nuclear Information System (INIS)

    Fundamenski, W.R.; Gierszewski, P.J.

    1991-08-01

    The design of a tritium breeding blanket for a fusion reactor requires the knowledge of heat transfer within the blanket. In this paper three models for effective bed heat transfer are compared against the experimental database in order to choose a reference correlation to be used in blanket design. Two parameters are used to describe heat transfer in a packed bed: effective thermal conductivity of the bed, and a heat transfer coefficient at the bed-solid interface

  16. Heat removing device for reactor container

    International Nuclear Information System (INIS)

    Hisamochi, Kohei; Matsumoto, Tomoyuki; Matsumoto, Masayoshi; Sato, Ken-ichi.

    1996-01-01

    A recycling loop for reactor water is disposed in a reactor pressure vessel of a BWR type reactor. Extracted reactor water from the recycling loop passes through a extracted reactor water pipeline and flows into a reactor coolant cleanup system. A pipeline for connecting the extracted reactor water pipeline and a suppression pool is disposed, and a discharged water pressurizing pump is disposed to the pipeline. Upon occurrence of emergency, discharged water from the suppression pool is pressurized by a discharged water pressurizing pump and sent to a reactor coolant cleanup system. The discharged water is cooled while passing through a sucking water cooling portion of a regenerative heat exchanger and a non-regenerative heat exchanger. Then, it is sent to a feed water pipeline passing a bypass line of a filtering desalter and a bypass line of the sucked water cooling portion of the regenerative heat exchanger, injected to the inside of the pressure vessel to cool the reactor core and remove after-heat. Then, it cools the inside of the reactor container together with coolants flown out of the pressure vessel and then returns to the suppression pool. (I.N.)

  17. Single-phase convection heat transfer characteristics of pebble-bed channels with internal heat generation

    International Nuclear Information System (INIS)

    Meng Xianke; Sun Zhongning; Xu Guangzhan

    2012-01-01

    Graphical abstract: The core of the water-cooled pebble bed reactor is the porous channels which stacked with spherical fuel elements. The gaps between the adjacent fuel elements are complex because they are stochastic and often shift. We adopt electromagnetic induction heating method to overall heat the pebble bed. By comparing and analyzing the experimental data, we get the rule of power distribution and the rule of heat transfer coefficient with particle diameter, heat flux density, inlet temperature and working fluid's Re number. Highlights: ► We adopt electromagnetic induction heating method to overall heat the pebble bed to be the internal heat source. ► The ball diameter is smaller, the effect of the heat transfer is better. ► With Re number increasing, heat transfer coefficient is also increasing and eventually tends to stabilize. ► The changing of heat power makes little effect on the heat transfer coefficient of pebble bed channels. - Abstract: The reactor core of a water-cooled pebble bed reactor includes porous channels that are formed by spherical fuel elements. This structure has notably improved heat transfer. Due to the variability and randomness of the interstices in pebble bed channels, heat transfer is complex, and there are few studies regarding this topic. To study the heat transfer characters of pebble bed channels with internal heat sources, oxidized stainless steel spheres with diameters of 3 and 8 mm and carbon steel spheres with 8 mm diameters are used in a stacked pebble bed. Distilled water is used as a refrigerant for the experiments, and the electromagnetic induction heating method is used to heat the pebble bed. By comparing and analyzing the experimental results, we obtain the governing rules for the power distribution and the heat transfer coefficient with respect to particle diameter, heat flux density, inlet temperature and working fluid Re number. From fitting of the experimental data, we obtain the dimensionless average

  18. 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)

  19. Blowdown heat transfer experiment, (1)

    International Nuclear Information System (INIS)

    Soda, Kunihisa; Yamamoto, Nobuo; Osaki, Hideki; Shiba, Masayoshi

    1976-09-01

    Blowdown heat transfer experiment has been carried out with a transparent test section to observe phenomena in coolant behavior during blowdown process. Experimental parameters are discharge position, initial system pressure, initial coolant temperature, power supply to heater rods and number of heater rods. At initial pressure 7-12 ata and initial power 6-50 kw per one heater rod, the flow condition in the test section is a major factor in determining time of DNB occurrence and physical process to DNB during blowdown. (auth.)

  20. 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

  1. 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.

  2. 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.

  3. A swivelling transfer device for nuclear reactors

    International Nuclear Information System (INIS)

    Allain, Albert; Mulot, Pierre; Filloleau, Etienne

    1974-01-01

    The invention relates to a swivelling transfer device for fuel-assemblies. According to the invention, the device comprises, within a protective enclosure, a swivelling system comprising two sets of rails rotatable about an axis and so arranged that the lower and thereof penetrates into the extensions of the extremities of ramps dipped into the reactor and into a storage enclosure. This can apply to the transfer of nuclear reactor fuel assemblies, in particular for reactors of the molten sodium fast neutron type [fr

  4. Experimental study on single-phase convection heat transfer characteristics of pebble bed channels with internal heat generation

    International Nuclear Information System (INIS)

    Meng Xianke; Sun Zhongning; Zhou Ping; Xu Guangzhan

    2012-01-01

    The water-cooled pebble bed reactor core is the porous channels stacked with spherical fuel elements, having evident effect on enhancing heat transfer. Owing to the variability and randomness characteristics of it's interstice, pebble bed channels have a very complex heat transfer situation and have little correlative research. In order to research the heat transfer characters of pebble bed channels with internal heat source, electromagnetic induction heating method was adopted for overall heating the pebble bed which was composed of 8 mm diameter steel balls, and the internal heat transfer characteristics were researched. By comparing and analyzing the experimental data, the rule of power distribution and heat transfer coefficient with heat flux density, inlet temperature and working fluid's Re were got. According to the experimental data fitting, the dimensionless average heat transfer coefficient correlation criteria was got. The fitting results are good agreement with the experimental results within 12% difference. (authors)

  5. Reassessment of forced convection heat transfer correlations for refrigerant-12

    International Nuclear Information System (INIS)

    Celata, G.P.; Cuomo, M.; D'Annibale, F.; Farello, G.E.; Setaro, T.

    1986-01-01

    In the frame of a Refrigerant-12 experiment on postulated accidental transients in Pressurized Water Reactors under way at Heat Transfer Laboratory (ENEA Casaccia Research Center), an assessment of the main correlation available in scientific literature, for the different heat transfer regions encountered when a liquid is boiled in a confined heated channel, has been performed. Considering a vertical tube uniformly heated over its length with CHF at the exit, the following heat transfer regimes may be individuated: convective heat transfer to liquid, subcooled boiling, saturated nucleate boiling, forced convective heat transfer through liquid film (annular flow regime) and thermal crisis. From the comparison of computed values with an original ENEA dataset, the best correlations in predicting Refrigerant-12 data have been individuated. In a few cases, though preserving the original structure of the correlations, mainly developed for water, it was necessary to adjust some coefficients by means of best-fit procedures through our experimental data. The work has been performed in the frame of the ENEA Thermal Reactor Department Safety Research Project

  6. Decay heat removal and heat transfer under normal and accident conditions in gas cooled reactors. Proceedings of a specialists meeting held in Juelich, Germany, 6-8 July 1992

    Energy Technology Data Exchange (ETDEWEB)

    1994-08-15

    The meeting was convened by the International Atomic Energy Agency on the recommendation of the IAEA`s International Working Group on Gas Cooled Reactors. It was attended by participants from China, France, Germany, Japan, Poland, the Russian Federation, Switzerland, the United Kingdom and the United States of America. The meeting was chaired by Prof. Dr. K. Kugeler and Prof. Dr. E. Hicken, Directors of the Institute for Safety Research Technology of the KFA Research Center, and covered the following: Design and licensing requirements for gas cooled reactors; concepts for decay heat removal in modern gas cooled reactors; analytical methods for predictions of thermal response, accuracy of predictions; experimental data for validation of predictive methods - operational experience from gas cooled reactors and experimental data from test facilities. Refs, figs and tabs.

  7. 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)

  8. Production of molten UO2 pools by internal heating: apparatus and preliminary experimental heat transfer results

    International Nuclear Information System (INIS)

    Chasanov, M.G.; Gunther, W.H.; Baker, L. Jr.

    1977-01-01

    The capability for removal of heat from a pool of molten fuel under postaccident conditions is an important consideration in liquid-metal fast breeder reactor safety analysis. No experimental data for pool heat transfer from molten UO 2 under conditions simulating internal heat generation by fission product decay have been reported previously in the literature. An apparatus to provide such data was developed and used to investigate heat transfer from pools containing up to 7.5 kg of UO 2 ; the internal heat generation rates and pool depths attained cover most of the ranges of interest for postaccident heat removal analysis. It was also observed in these studies that the presence of simulated fission products corresponding to approximately 150,000 kW-day/kg burnup had no significant effect on the observed heat transfer

  9. Heat Transfer in Complex Fluids

    Energy Technology Data Exchange (ETDEWEB)

    Mehrdad Massoudi

    2012-01-01

    fluids show Newtonian (linear) behavior for a given range of parameters or geometries; there are many empirical or semi-empirical constitutive equations suggested for these fluids. There have also been many non-linear constitutive relations which have been derived based on the techniques of continuum mechanics. The non-linearities oftentimes appear due to higher gradient terms or time derivatives. When thermal and or chemical effects are also important, the (coupled) momentum and energy equations can give rise to a variety of interesting problems, such as instability, for example the phenomenon of double-diffusive convection in a fluid layer. In Conclusion, we have studied the flow of a compressible (density gradient type) non-linear fluid down an inclined plane, subject to radiation boundary condition. The heat transfer is also considered where a source term, similar to the Arrhenius type reaction, is included. The non-dimensional forms of the equations are solved numerically and the competing effects of conduction, dissipation, heat generation and radiation are discussed. It is observed that the velocity increases rapidly in the region near the inclined surface and is slower in the region near the free surface. Since R{sub 7} is a measure of the heat generation due to chemical reaction, when the reaction is frozen (R{sub 7}=0.0) the temperature distributions would depend only on R{sub 1}, and R{sub 2}, representing the effects of the pressure force developed in the material due to the distribution, R{sub 3} and R{sub 4} viscous dissipation, R{sub 5} the normal stress coefficient, R{sub 6} the measure of the emissivity of the particles to the thermal conductivity, etc. When the flow is not frozen (RP{sub 7} > 0) the temperature inside the flow domain is much higher than those at the inclined and free surfaces. As a result, heat is transferred away from the flow toward both the inclined surface and the free surface with a rate that increases as R{sub 7} increases. For a

  10. Study of heat transfer in 3D fuel rods of the EPRI-9R reactor modified; Estudo da transferencia de calor em varetas combustiveis 3D do reator EPRI-9R 3D modificado

    Energy Technology Data Exchange (ETDEWEB)

    Affonso, Renato Raoni Werneck; Lava, Deise Diana; Borges, Diogo da Silva; Sampaio, Paulo Augusto Berquo de; Moreira, Maria de Lourdes, E-mail: raoniwa@yahoo.com.br, E-mail: deisedy@gmail.com, E-mail: diogosb@outlook.com, E-mail: sampaio@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    This paper aims to conduct a case study of the fuel rods that have the highest and the lowest average power of the EPRI-9R 3D reactor modified , for various positions of the control rods banks. For this, will be addressed the verification of computer code, comparing the results obtained with analytical solutions. This check is important so that, subsequently, it is possible use the program to understand the behavior of the fuel rods and the coolant channel of the EPRI-9R 3D reactor modified. Thus, in view of the scope of this paper, first a brief introducing on the heat transfer is done, including the rod equations and the equation of energy in the channel to allow the analysis of the results.

  11. HTCC - a heat transfer model for gas-steam mixtures

    International Nuclear Information System (INIS)

    Papadimitriou, P.

    1983-01-01

    The mathematical model HTCC (Heat Transfer Coefficient in Containment) has been developed for RALOC after a loss-of-coolant accident in order to determine the local heat transfer coefficients for transfer between the containment atmosphere and the walls of the reactor building. The model considers the current values of room and wall temperature, the concentration of steam and non-condensible gases, geometry data and those of fluid dynamics together with thermodynamic parameters and from these determines the heat transfer mechanisms due to convection, radiation and condensation. The HTCC is implemented in the RALOC program. Comparative analyses of computed temperature profiles, for HEDL Standard problems A and B on hydrogen distribution, and of computed temperature profiles determined during the heat-up phase in the CSE-A5 experiment show a good agreement with experimental data. (orig.) [de

  12. 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...

  13. 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

  14. 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.

  15. Overview PWR-Blowdown Heat Transfer Separate-Effects Program

    International Nuclear Information System (INIS)

    White, J.D.

    1978-01-01

    The ORNL Pressurized Water Reactor Blowdown Heat Transfer Program (PWR-BDHT) is a separate-effects experimental study of thermal-hydraulic phenomena occurring during the first 20 sec of a hypothetical LOCA. Specific objectives include the determination, for a wide range of parameters, of time to CHF and the following variables for both pre- and post-CHF: heat fluxes, ΔT (temperature difference between pin surface and fluid), heat transfer coefficients, and local fluid properties. A summary of the most interesting results from the program obtained during the past year is presented. These results are in the area of: (1) RELAP verification, (2) electric pin calibration, (3) time to critical heat flux (CHF), (4) heat transfer coefficient comparisons, and (5) nuclear fuel pin simulation

  16. Various methods to improve heat transfer in exchangers

    Directory of Open Access Journals (Sweden)

    Pavel Zitek

    2015-01-01

    Full Text Available The University of West Bohemia in Pilsen (Department of Power System Engineering is working on the selection of effective heat exchangers. Conventional shell and tube heat exchangers use simple segmental baffles. It can be replaced by helical baffles, which increase the heat transfer efficiency and reduce pressure losses. Their usage is demonstrated in the primary circuit of IV. generation MSR (Molten Salt Reactors. For high-temperature reactors we consider the use of compact desk heat exchangers, which are small, which allows the integral configuration of reactor. We design them from graphite composites, which allow up to 1000°C and are usable as exchangers: salt-salt or salt-acid (e.g. for the hydrogen production. In the paper there are shown thermo-physical properties of salts, material properties and principles of calculations.

  17. Designing heat exchangers for process heat reactors

    International Nuclear Information System (INIS)

    Quade, R.N.

    1980-01-01

    A brief account is given of the IAEA specialist meeting on process heat applications technology held in Julich, November 1979. The main emphasis was on high temperature heat exchange. Papers were presented covering design requirements, design construction and prefabrication testing, and selected problems. Primary discussion centered around mechanical design, materials requirements, and structural analysis methods and limits. It appears that high temperature heat exchanges design to nuclear standards, is under extensive development but will require a lengthy concerted effort before becoming a commercial reality. (author)

  18. Dynamic Response Testing in an Electrically Heated Reactor Test Facility

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Morton, T. J.

    2006-01-01

    Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

  19. Heat Transfer in Gas Turbines

    Science.gov (United States)

    Garg, Vijay K.

    2001-01-01

    The turbine gas path is a very complex flow field. This is due to a variety of flow and heat transfer phenomena encountered in turbine passages. This manuscript provides an overview of the current work in this field at the NASA Glenn Research Center. Also, based on the author's preference, more emphasis is on the computational work. There is much more experimental work in progress at GRC than that reported here. While much has been achieved, more needs to be done in terms of validating the predictions against experimental data. More experimental data, especially on film cooled and rough turbine blades, are required for code validation. Also, the combined film cooling and internal cooling flow computation for a real blade is yet to be performed. While most computational work to date has assumed steady state conditions, the flow is clearly unsteady due to the presence of wakes. All this points to a long road ahead. However, we are well on course.

  20. Heat transfer in porous media

    Directory of Open Access Journals (Sweden)

    N. Amanifard

    2007-06-01

    Full Text Available In this work, the effects of electrical double layer (EDL near the solid/ liquid interface, on three dimensional heat transfer characteristic and pressure drop of water flow through a rectangular microchannel numerically are investigated. An additional body force originating from the existence of EDL is considered to modify the conventional Navier-stokes and energy equations. These modified equations are solved numerically for steady laminar flow on the basis of control volume approaches. Fluid velocity distribution and temperature with presence and absence of EDL effects are presented for various geometric cases and different boundary conditions. The results illustrate that, the liquid flow in rectangular microchannels is influenced significantly by the EDL, particularly in the high electric potentials, and hence deviates from flow characteristics described by classical fluid mechanics.

  1. 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.

  2. 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

  3. Study of the moderating effect of salts on the sodium-water reaction on the cleaning of irradiated fuel assemblies from fast neutron reactors, using fluid sodium heat transfer

    International Nuclear Information System (INIS)

    Lacroix, Marie

    2014-01-01

    Within the framework of the development of generation IV reactors one of the research tracks is related to the development of fast neutron reactors using fluid sodium heat transfer. The CEA (French Alternative Energies and Atomic Energy Commission) plans to build a prototype of reactor of this type called 'ASTRID'. To address development requirements for this prototype, research is in progress on the reactor's availability and in particular on the reduction of the washing duration for residual sodium fuel assemblies during their discharge. In fact, because sodium is very reactive with water (presently the only available process), the washing is done, for example, by very gradual addition. A solution currently being studied at the CEA and which is the subject of this thesis report consists of the addition of an aqueous salts solutions to the washing water in order to slow down the kinetic reaction. This doctoral dissertation describes the various salts, which have been evaluated and aims to explain their action mode. (author) [fr

  4. Pre-design stage of the intermediate heat exchanger for experimental fast reactor

    International Nuclear Information System (INIS)

    Luz, M.; Borges, E.M.; Braz Filho, F.A.; Hirdes, V.R.

    1986-09-01

    This report presents the outlines of a thermal-hydraulic calculation procedure for the pre-design stage of the Intermediate Heat Exchanger for a 5 MW Experimental Fast Reactor (EFR), which can be used in other similar projects, at the same stage of evolution. Heat transfer and heat loss computations for the preliminary design of the heat exchanger are presented. (author) [pt

  5. 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

  6. Local heat transfer coefficient in a fluidized bed

    International Nuclear Information System (INIS)

    Al-Busoul, A. M.

    1999-01-01

    This paper presents an experimental study for the local heat transfer coefficient. The experiments was conducted inside a reactor with inner diameter (I D = 142mm) at atmospheric conditions (temperature mean value = 29 deg.) The bed was heated by means of a parochial electric heater with a diameter of (d h = 29 mm) and a constant power of 5W. The following factors varied: particles type and diameter, fluid velocity, bed height and heater position inside the reactor. The results were presented and discussed. (author). 15 refs., 7 figs

  7. Heat transfer and fluid flow in nuclear systems

    CERN Document Server

    Fenech, Henri

    1982-01-01

    Heat Transfer and Fluid in Flow Nuclear Systems discusses topics that bridge the gap between the fundamental principles and the designed practices. The book is comprised of six chapters that cover analysis of the predicting thermal-hydraulics performance of large nuclear reactors and associated heat-exchangers or steam generators of various nuclear systems. Chapter 1 tackles the general considerations on thermal design and performance requirements of nuclear reactor cores. The second chapter deals with pressurized subcooled light water systems, and the third chapter covers boiling water reacto

  8. 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.

  9. Heat transfer coefficient for boiling carbon dioxide

    DEFF Research Database (Denmark)

    Knudsen, Hans Jørgen Høgaard; Jensen, Per Henrik

    1998-01-01

    Heat transfer coefficient and pressure drop for boiling carbon dioxide (R744) flowing in a horizontal pipe has been measured. The calculated heat transfer coeeficient has been compared with the Chart correlation of Shah. The Chart Correlation predits too low heat transfer coefficient but the ratio...... 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 not been compared with correlation's....

  10. Heat dissipating nuclear reactor with metal liner

    Science.gov (United States)

    Gluekler, E.L.; Hunsbedt, A.; Lazarus, J.D.

    1985-11-21

    A nuclear reactor containment including a reactor vessel disposed within a cavity with capability for complete inherent decay heat removal in the earth and surrounded by a cast steel containment member which surrounds the vessel is described in this disclosure. The member has a thick basemat in contact with metal pilings. The basemat rests on a bed of porous particulate material, into which water is fed to produce steam which is vented to the atmosphere. There is a gap between the reactor vessel and the steel containment member. The containment member holds any sodium or core debris escaping from the reactor vessel if the core melts and breaches the vessel.

  11. Single-phase convective heat transfer in rod bundles

    International Nuclear Information System (INIS)

    Holloway, Mary V.; Beasley, Donald E.; Conner, Michael E.

    2008-01-01

    The convective heat transfer for turbulent flow through rod bundles representative of nuclear fuel rods used in pressurized water reactors is examined. The rod bundles consist of a square array of parallel rods that are held on a constant pitch by support grids spaced axially along the rod bundle. Split-vane pair support grids, which create swirling flow in the rod bundle, as well as disc and standard support grids are investigated. Single-phase convective heat transfer coefficients are measured for flow downstream of support grids in a rod bundle. The rods are heated using direct resistance heating, and a bulk axial flow of air is used to cool the rods in the rod bundle. Air is used as the working fluid instead of water to reduce the power required to heat the rod bundle. Results indicate heat transfer enhancement for up to 10 hydraulic diameters downstream of the support grids. A general correlation is developed to predict the heat transfer development downstream of support grids. In addition, circumferential variations in heat transfer coefficients result in hot streaks that develop on the rods downstream of split-vane pair support grids

  12. Single-phase convective heat transfer in rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Holloway, Mary V. [Mechanical Engineering Department, United States Naval Academy, 590 Holloway Rd., Annapolis, MD 21402 (United States)], E-mail: holloway@usna.edu; Beasley, Donald E. [Mechanical Engineering Department, Clemson University, Clemson, SC 29634 (United States); Conner, Michael E. [Westinghouse Nuclear Fuel, 5801 Bluff Road, Columbia, SC 29250 (United States)

    2008-04-15

    The convective heat transfer for turbulent flow through rod bundles representative of nuclear fuel rods used in pressurized water reactors is examined. The rod bundles consist of a square array of parallel rods that are held on a constant pitch by support grids spaced axially along the rod bundle. Split-vane pair support grids, which create swirling flow in the rod bundle, as well as disc and standard support grids are investigated. Single-phase convective heat transfer coefficients are measured for flow downstream of support grids in a rod bundle. The rods are heated using direct resistance heating, and a bulk axial flow of air is used to cool the rods in the rod bundle. Air is used as the working fluid instead of water to reduce the power required to heat the rod bundle. Results indicate heat transfer enhancement for up to 10 hydraulic diameters downstream of the support grids. A general correlation is developed to predict the heat transfer development downstream of support grids. In addition, circumferential variations in heat transfer coefficients result in hot streaks that develop on the rods downstream of split-vane pair support grids.

  13. Secondary heat exchanger design and comparison for advanced high temperature reactor

    International Nuclear Information System (INIS)

    Sabharwall, P.; Kim, E. S.; Siahpush, A.; McKellar, M.; Patterson, M.

    2012-01-01

    Next generation nuclear reactors such as the advanced high temperature reactor (AHTR) are designed to increase energy efficiency in the production of electricity and provide high temperature heat for industrial processes. The efficient transfer of energy for industrial applications depends on the ability to incorporate effective heat exchangers between the nuclear heat transport system and the industrial process heat transport system. This study considers two different types of heat exchangers - helical coiled heat exchanger and printed circuit heat exchanger - as possible options for the AHTR secondary heat exchangers with distributed load analysis and comparison. Comparison is provided for all different cases along with challenges and recommendations. (authors)

  14. 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...

  15. 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

  16. 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

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

    International Nuclear Information System (INIS)

    Consolini, L.; Thome, J.R.

    2009-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Consolini, L.; Thome, J.R. [Ecole Polytechnique Federale de Lausanne (Switzerland). Lab. de Transfert de Chaleur et de Masse], e-mail: lorenzo.consolini@epfl.ch, e-mail: john.thome@epfl.ch

    2009-07-01

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

  19. Heat transfer from humans wearing clothing

    NARCIS (Netherlands)

    Lotens, W.A.

    1993-01-01

    In this monograph the effects of clothing on human heat transfer are described. The description is based on the physics of heat and mass transfer, depending on the design of the clothing, the climate, and the activity of the wearer. The resulting model has been stepwise implemented in computer

  20. Heat Transfer Experiment with Supercritical CO{sub 2} Flowing Upward in a Circular Tube

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyung Rae; Kim, Hwan Yeol; Song, Jin Ho; Kim, Hee Dong; Bae, Yoon Yeong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    SCWR (SuperCritical Water-cooled Reactor) is one of the six reactor candidates selected in the Gen-IV project, which aims at the development of new reactors with enhanced economy and safety. Heat transfer experiments under supercritical conditions are required in relevant geometries for the proper prediction of thermo-hydraulic phenomena in a reactor core. A heat transfer test loop, named as SPHINX (Supercritical Pressure Heat Transfer Investigation for NeXt generation), has been constructed in KAERI. The loop uses carbon dioxide as a surrogate fluid for water since the critical pressure and temperature of CO{sub 2} are much lower those of water. As a first stage of heat transfer experiments, a single tube test is being performed in the test loop. Controlled parameters for the tests are operating pressure, mass flux, and heat flux. Wall temperatures are measured along the tube. Experimental data are compared with existing correlations.

  1. Heat Transfer Experiment with Supercritical CO2 Flowing Upward in a Circular Tube

    International Nuclear Information System (INIS)

    Kim, Hyung Rae; Kim, Hwan Yeol; Song, Jin Ho; Kim, Hee Dong; Bae, Yoon Yeong

    2005-01-01

    SCWR (SuperCritical Water-cooled Reactor) is one of the six reactor candidates selected in the Gen-IV project, which aims at the development of new reactors with enhanced economy and safety. Heat transfer experiments under supercritical conditions are required in relevant geometries for the proper prediction of thermo-hydraulic phenomena in a reactor core. A heat transfer test loop, named as SPHINX (Supercritical Pressure Heat Transfer Investigation for NeXt generation), has been constructed in KAERI. The loop uses carbon dioxide as a surrogate fluid for water since the critical pressure and temperature of CO 2 are much lower those of water. As a first stage of heat transfer experiments, a single tube test is being performed in the test loop. Controlled parameters for the tests are operating pressure, mass flux, and heat flux. Wall temperatures are measured along the tube. Experimental data are compared with existing correlations

  2. A nuclear reactor for district heating

    International Nuclear Information System (INIS)

    Bancroft, A.R.; Fenton, N.

    1989-07-01

    Global energy requirements are expected to double over the next 40 years. In the northern hemisphere, many countries consume in excess of 25 percent of their primary energy supply for building heating. Satisfying this need, within the constraints now being acknowledged for sustainable global development, provides an important opportunity for district heating. Fuel-use flexibility, energy and resource conservation, and reduced atmospheric pollution from acid gases and greenhouse gases, are important features offered by district heating systems. Among the major fuel options, only hydro-electricity and nuclear heat completely avoid emissions of combustion gases. To fill the need for an economical nuclear heat source, Atomic Energy of Canada Limited has designed a 10 MW plant that is suitable as a heat source within a network or as the main supply to large individual users. Producing hot water at temperatures below 100 degrees C, it incorporates a small pool-type reactor based on AECL's successful SLOWPOKE Research Reactor. A 2 MW prototype for the commercial unit is now being tested at the Whiteshell Nuclear Research Establishment in Manitoba. With capital costs of $7 million (Canadian), unit energy costs are projected to be $0.02/kWh for a 10 MW unit operating in a heating grid over a 30-year period. By keeping the reactor power low and the water temperature below 100 degrees C, much of the complexity of the large nuclear power plants can be avoided, thus allowing these small, safe nuclear heating systems to be economically viable

  3. On the heat transfer in packed beds

    International Nuclear Information System (INIS)

    Sordon, G.

    1988-09-01

    The design of a fusion reactor blanket concept based on a bed of lithium containing ceramic pebbles or a mixture of ceramic and beryllium pebbles demands the knowledge of the effective thermal conductivity of pebble beds, including beds formed by a binary mixture of high conducting metallic pebbles and poorly conducting pebbles. In this work, binary mixtures of spheres of same diameter and different conductivities as well as beds formed by one type of spheres were investigated. The experimental apparatus consists of a stainless steel cylinder with a heating rod along the symmetry axis. Experiments with stagnant and flowing gas were performed. The pebbles were of Al 2 O 3 (diameter = 1, 2, 4 mm), of Li 4 SO 4 (diameter = 0.5 mm) of Al (diameter = 2 mm) and of steel (diameter = 2, 4 mm). Experimental values of the thermal conductivity and of the wall heat transfer coefficient are compared with the predicted ones. Modifications of already existing models were suggested. (orig.) [de

  4. Boiling heat transfer modern developments and advances

    CERN Document Server

    Lahey, Jr, RT

    2013-01-01

    This volume covers the modern developments in boiling heat transfer and two-phase flow, and is intended to provide industrial, government and academic researchers with state-of-the-art research findings in the area of multiphase flow and heat transfer technology. Special attention is given to technology transfer, indicating how recent significant results may be used for practical applications. The chapters give detailed technical material that will be useful to engineers and scientists who work in the field of multiphase flow and heat transfer. The authors of all chapters are members of the

  5. Heat transfer coefficient of cryotop during freezing.

    Science.gov (United States)

    Li, W J; Zhou, X L; Wang, H S; Liu, B L; Dai, J J

    2013-01-01

    Cryotop is an efficient vitrification method for cryopreservation of oocytes. It has been widely used owing to its simple operation and high freezing rate. Recently, the heat transfer performance of cryotop was studied by numerical simulation in several studies. However, the range of heat transfer coefficient in the simulation is uncertain. In this study, the heat transfer coefficient for cryotop during freezing process was analyzed. The cooling rates of 40 percent ethylene glycol (EG) droplet in cryotop during freezing were measured by ultra-fast measurement system and calculated by numerical simulation at different value of heat transfer coefficient. Compared with the results obtained by two methods, the range of the heat transfer coefficient necessary for the numerical simulation of cryotop was determined, which is between 9000 W/(m(2)·K) and 10000 W/(m (2)·K).

  6. "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

  7. 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.

  8. 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

  9. Effect of radiant heat transfer on the performance of high temperature heat exchanger

    International Nuclear Information System (INIS)

    Mori, Yasuo; Hijikata, Kunio; Yamada, Yukio

    1975-01-01

    The development of high temperature gas-cooled reactors is motivated by the consideration of the application of nuclear heat for industrial uses or direct steelmaking and chemical processes. For these purposes, reliable and efficient heat exchangers should be developed. This report analyzes the effect of radiant heat transfer on the performance of high temperature heat exchangers. The heat transfer model is as follows: the channel composed with two parallel adiabatic walls is divided with one parallel plate between the walls. Non-radiative fluid flows in the two separated channels in opposite direction. Heat transfer equations for this system were obtained, and these equations were solved by some approximate method and numerical analysis. The effect of radiation on heat transfer became larger as the radiant heat transfer between two walls was larger. In the heat exchangers of counter flow type, the thermal efficiency is controlled with three parameters, namely radiation-convection parameter, Stanton number and temperature difference. The thermal efficiency was larger with the increase of these parameters. (Iwase, T.)

  10. Droplet heat transfer and chemical reactions during direct containment heating

    International Nuclear Information System (INIS)

    Baker, L. Jr.

    1986-01-01

    A simplified model of heat transfer and chemical reaction has been adapted to evaluate the expected behavior of droplets containing unreacted Zircaloy and stainless steel moving through the containment atmosphere during postulated accidents involving direct containment heating. The model includes internal and external diffusive resistances to reaction. The results indicate that reactions will be incomplete for many conditions characteristic of direct containment heating sequences

  11. 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.

  12. The magnetic fluid for heat transfer applications

    International Nuclear Information System (INIS)

    Nakatsuka, K.; Jeyadevan, B.; Neveu, S.; Koganezawa, H.

    2002-01-01

    Real-time visual observation of boiling water-based and ionic magnetic fluids (MFs) and heat transfer characteristics in heat pipe using ionic MF stabilized by citrate ions (JC-1) as working liquid are reported. Irrespective of the presence or absence of magnetic field water-based MF degraded during boiling. However, the degradation of JC-1 was avoided by heating the fluid in magnetic field. Furthermore, the heat transfer capacity of JC-1 heat pipe under applied magnetic field was enhanced over the no field case

  13. 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

  14. Reactor waste heat utilization and district heating reactors. Nuclear district heating in Sweden - Regional reject heat utilization schemes and small heat-only reactors

    International Nuclear Information System (INIS)

    Hannerz, K.; Larsson, Y.; Margen, P.

    1977-01-01

    A brief review is given of the current status of district heating in Sweden. In future, district heating schemes will become increasingly interesting as a means of utilizing heat from nuclear reactors. Present recommendations in Sweden are that large reactors should not be located closer than about 20 km from large population centres. Reject heat from such reactors is cheap at source. To minimize the cost of long distance hot water transmission large heat rates must be transmitted. Only areas with large populations can meet this requirement. The three areas of main interest are Malmoe/Lund/Helsingborg housing close to 0.5 million; Greater Stockholm housing 1 to 1.5 million and Greater Gothenburg housing about 0.5 million people. There is an active proposal that the Malmoe/Lund/Helsingborg region would be served by a third nuclear unit at Barsebaeck, located about 20 km from Malmoe/Lund and supplying 950 MW of base load heat. Preliminary proposals for Stockholm involve a 2000 MW heat supply; proposals for Gothenburg are more tentative. The paper describes progress on these proposals and their technology. It also outlines technology under development to increase the economic range of large scale heat transport and to make distribution economic even for low heat-density family housing estates. Regions apart from the few major urban areas mentioned above require the adoption of a different approach. To this end the development of a small, simple low-temperature reactor for heat-only production suitable for urban location has been started in Sweden in close contact with Finland. Some results of the work in progress are presented, with emphasis on the safety requirements. An outline is given in the paper as to how problems of regional heat planning and institutional and legislative issues are being approached

  15. E-beam heated linear solenoid reactors

    International Nuclear Information System (INIS)

    Benford, J.; Bailey, V.; Oliver, D.

    1976-01-01

    A conceptual design and system analysis shows that electron beam heated linear solenoidal reactors are attractive for near term applications which can use low gain fusion sources. Complete plant designs have been generated for fusion based breeders of fissile fuel over a wide range of component parameters (e.g., magnetic fields, reactor lengths, plasma densities) and design options (e.g., various radial and axial loss mechanisms). It appears possible that a reactor of 100 to 300 meters length operating at power levels of 1000 MWt can economically produce 2000 to 8000 kg/yr of 233 U to supply light water reactor fuel needs beyond 2000 A.D. Pure fusion reactors of 300 to 500 meter lengths are possible. Physics and operational features of reactors are described. Beam heating by classical and anomalous energy deposition is reviewed. The technology of the required beams has been developed to MJ/pulse levels, within a factor of 20 of that needed for full scale production reactors. The required repetitive pulsing appears practical

  16. 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

  17. 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.

  18. Cyro Power and Heat Transfer

    National Research Council Canada - National Science Library

    Chow, L

    1998-01-01

    .... The heat generated from a 9x9-heater array was removed by liquid nitrogen pool boiling. The orientation and space limitation of the array were varied to explore their effects on the critical heat flux (CHF) value...

  19. Heat transfer to liquid sodium in the thermal entrance region

    International Nuclear Information System (INIS)

    Qiu, R.

    1981-01-01

    It is well known that the convective heat transfer in the regions of duct systems where the thermal boundary layers are not yet established can be far superior to heat transfer in the fully developed regions. A quantitative understanding of heat transfer in the thermal entrance region is essential in designing high heat-flux nuclear reactors. More specifically, if the thermal boundary layers have not been fully established in the system, the forced-convection relations for the fully developed regions cannot be used to predict the heat transfer characteristics. The present work is characterized by the following: 1. The behaviours in the thermal entrance region have been examined more completely. 2. To obtain a higher accuracy of analyses, in present study the method of SPARROW et al. for pipe was improved for annulus by utilizing a finite difference technique. Furthermore, an asymptotic solution was developed. 3. This is, in our knowledge, the first experimental investigation about the thermal development effect on turbulent heat transfer from rod element to liquid sodium in annulus with fully developed flow. (MDC)

  20. 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.

  1. Moderator heat recovery of CANDU reactors

    International Nuclear Information System (INIS)

    Fath, H.E.S.; Ahmed, S.T.

    1986-01-01

    A moderator heat recovery scheme is proposed for CANDU reactors. The proposed circuit utilizes all the moderator heat to the first stages of the plant feedwater heating system. CANDU-600 reactors are considered with moderator heat load varying from 120 to 160 MWsub(th), and moderator outlet temperature (from calandria) varying from 80 to 100 0 C. The steam saved from the turbine extraction system was found to produce an additional electric power ranging from 5 to 11 MW. This additional power represents a 0.7-1.7% increase in the plant electric output power and a 0.2-0.7% increase in the plant thermal efficiency. The outstanding features and advantages of the proposed scheme are presented. (author)

  2. Preliminary feasibility study of the heat - pipe ENHS reactor

    International Nuclear Information System (INIS)

    Fratoni, M.; Kim, L.; Mattafirri, S.; Petroski, R.; Greenspan, E.

    2007-01-01

    This preliminary study assesses the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor [1] to have a solid core from which heat is removed by liquid-metal heat pipes (HP). Like the SAFE space nuclear reactor core [2], the HP-ENHS core is comprised of fuel rods and HPs embedded in a solid structure arranged in a hexagonal lattice in a 3:1 ratio. The HPs extend beyond the core length and transfer heat to a secondary coolant that flows by natural circulation. The HP-ENHS reactor is designed to preserve many features of the ENHS reactor including 20-year operation without refueling, very small excess reactivity throughout life, natural circulation cooling, walk-away passive safety, and robust proliferation resistance. The target power level and specific power of the HP-ENHS reactor are those of the reference ENHS reactor [1]. Compared to previous ENHS reactor designs utilizing a lead or lead-bismuth alloy natural circulation cooling system, the HP-ENHS reactor offers a number of possible advantageous features including: (1) significantly enhanced decay heat removal capability; (2) no positive void reactivity coefficients; (3) no direct contact between the fuel clad and coolant, hence, relatively lower wet corrosion of the clad; (4) a core that is more robust for transportation; (5) higher temperature potentially offering higher efficiency and hydrogen production capability. The study focuses on four areas: material compatibility analysis, HP performance analysis, neutronic analysis and thermal-hydraulic analysis. Of four high-temperature structural materials evaluated, Mo TZM alloy is the preferred choice; its upper estimated feasible operating temperature is 1350 K. HP performance is evaluated as a function of working fluid type, operating temperature, wick design and HP diameter and length. Sodium is the preferred working fluid and the HP working temperature is 1300 K. The neutronic analysis found that it is possible to achieve criticality

  3. 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

  4. Thermal striping heat transfer measurements in sodium AKB experiments

    International Nuclear Information System (INIS)

    Sheriff, N.; Sephton, K.P.; Gleave, C.

    1988-01-01

    Temperature fluctuations are produced in the sodium flow of fast reactors where hot and cold flow streams mix. A sodium experiment mounted on the Interatom facility AKB has been used to measure the heat transfer conditions in a flow stream with typical temperature fluctuations. The measurements were made at locations near to the leading edge of a plate, where in practice the most severe conditions are expected. With tests carried out over a wide range of flows good correlations of the heat transfer data with flow have been obtained. A simple theoretical model is proposed to explain the magnitude of the measured heat transfer coefficients, and the use of reasonable assumptions in the model produce good agreement with the experimental measurements

  5. Heat insulation device for reactor pressure vessel in water

    International Nuclear Information System (INIS)

    Nakamura, Heiichiro; Tanaka, Yoshimi.

    1993-01-01

    Outer walls of a reactor pressure vessel are covered with water-tight walls made of metals. A heat insulation metal material is disposed between them. The water tight walls are joined by welding and flanges. A supply pipeline for filling gases and a discharge pipeline are in communication with the inside of the water tight walls. Further, a water detector is disposed in the midway of the gas discharge pipeline. With such a constitution, the following advantages can be attained. (1) Heat transfer from the reactor pressure vessel to water of a reactor container can be suppressed by filled gases and heat insulation metal material. (2) Since the pressure at the inside of the water tight walls can be equalized with the pressure of the inside of the reactor container, the thickness of the water-tight walls can be reduced. (3) Since intrusion of water to the inside of the walls due to rupture of the water tight walls is detected by the water detector, reactor scram can be conducted rapidly. (4) The sealing property of the flange joint portion is sufficient and detaching operation thereof is easy. (I.S.)

  6. Fuel transfer system for a nuclear reactor

    International Nuclear Information System (INIS)

    Katz, L.R.; Marshall, J.R.; Desmarchais, W.E.

    1977-01-01

    Disclosed is a fuel transfer system for moving nuclear reactor fuel assemblies from a new fuel storage pit to a containment area containing the nuclear reactor, and for transferring spent fuel assemblies under water from the reactor to a spent fuel storage area. The system includes an underwater track which extends through a wall dividing the fuel building from the reactor containment and a car on the track serves as the vehicle for moving fuel assemblies between these two areas. The car is driven by a motor and linkage extending from an operating deck to a chain belt drive on the car. A housing pivotally mounted at its center on the car is hydraulically actuated to vertically receive a fuel assembly which then is rotated to a horizontal position to permit movement through the wall between the containment and fuel building areas. Return to the vertical position provides for fuel assembly removal and the reverse process is repeated when transferring an assembly in the opposite direction. Limit switches used in controlling operation of the system are designed to be replaced from the operating deck when necessary by tools designed for this purpose. 5 claims, 8 figures

  7. Endwall convective heat transfer for bluff bodies

    DEFF Research Database (Denmark)

    Wang, Lei; Salewski, Mirko; Sundén, Bengt

    2012-01-01

    The endwall heat transfer characteristics of forced flow past bluff bodies have been investigated using liquid crystal thermography (LCT). The bluff body is placed in a rectangular channel with both its ends attached to the endwalls. The Reynolds number varies from 50,000 to 100,000. In this study......, a single bluff body and two bluff bodies arranged in tandem are considered. Due to the formation of horseshoe vortices, the heat transfer is enhanced appreciably for both cases. However, for the case of two bluff bodies in tandem, it is found that the presence of the second bluff body decreases the heat...... transfer as compared to the case of a single bluff body. In addition, the results show that the heat transfer exhibits Reynolds number similarity. For a single bluff body, the Nusselt number profiles collapse well when the data are scaled by Re0.55; for two bluff bodies arranged in tandem, the heat...

  8. 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

  9. 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.

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

    International Nuclear Information System (INIS)

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

    1976-01-01

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

  11. Study on Heat Transfer Characteristics of One Side Heated Vertical Channel Applied as Vessel Cooling System

    International Nuclear Information System (INIS)

    Kuriyama, Shinji; Takeda, Tetsuaki; Funatani, Shumpei

    2014-01-01

    The inherent properties of the Very-High-Temperature Reactor facilitate the design of the VHTR with high degree of passive safe performances, compared to other type of reactors. However; it is still not clear if the VHTR can maintain a passive safe function during the severe accident, or what would be a design criterion to guarantee the VHTR with the high degree of passive safe performances during the accidents. In the Very High Temperature Reactor (VHTR) which is a next generation nuclear reactor system, ceramics and graphite are used as a fuel coating material and a core structural material, respectively. Even if the depressurization accident occurs and the reactor power goes up instantly, the temperature of the core will change slowly. This is because the thermal capacity of the core is so large. Therefore, the VHTR system can passively remove the decay heat of the core by natural convection and radiation from the surface of the reactor pressure vessel (RPV). This study is to develop the passive cooling system for the VHTR using the vertical channel inserting porous materials. The objective of this study is to investigate heat transfer characteristics of natural convection of a one-side heated vertical channel inserting the porous materials with high porosity. In order to obtain the heat transfer and fluid flow characteristics of a vertical channel inserting porous material, we have also carried out a numerical analysis using the commercial CFD code. From the analytical results obtained in the natural convection cooling, an amount of removed heat enhanced inserting the copper wire. It was found that an amount of removed heat inserting the copper wire (porosity = 0.9972) was about 10% higher than that without the copper wire. This paper describes a thermal performance of the one-side heated vertical channel inserting copper wire with high porosity. (author)

  12. Pumped two-phase heat transfer loop

    Science.gov (United States)

    Edelstein, Fred

    1988-01-01

    A pumped loop two-phase heat transfer system, operating at a nearly constant temperature throughout, includes several independently operating grooved capillary heat exchanger plates supplied with working fluid through independent flow modulation valves connected to a liquid supply line, a vapor line for collecting vapor from the heat exchangers, a condenser between the vapor and the liquid lines, and a fluid circulating pump between the condenser and the heat exchangers.

  13. Forced convection heat transfer in He II

    International Nuclear Information System (INIS)

    Kashani, A.

    1986-01-01

    An investigation of forced convection heat transfer in He II is conducted. The study includes both experimental and theoretical treatments of the problem. The experiment consists of a hydraulic pump and a copper flow tube, 3 mm in ID and 2m long. The system allows measurements of one-dimensional heat and mass transfer in He II. The heat transfer experiments are performed by applying heat at the midpoint along the length of the flow tube. Two modes of heat input are employed, i.e., step function heat input and square pulse heat input. The heat transfer results are discussed in terms of temperature distribution in the tube. The experimental temperature profiles are compared with numerical solutions of an analytical model developed from the He II energy equation. The bath temperature is set at three different values of 1.65, 1.80, and 1.95 K. The He II flow velocity is varied up to 90 cm/s. Pressure is monitored at each end of the flow tube, and the He II pressure drop is obtained for different flow velocities. Results indicate that He II heat transfer by forced convention is considerably higher than that by internal convection. The theoretical model is in close agreement with the experiment. He II pressure drop and friction factor are very similar to those of an ordinary fluid

  14. 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

  15. Design of the prestressed concrete reactor vessel for gas-cooled heating reactors

    International Nuclear Information System (INIS)

    Becker, G.; Notheisen, C.; Steffen, G.

    1987-01-01

    The GHR pebble bed reactor offers a simple, safe and economic possibility of heat generation. An essential component of this concept is the prestressed concrete reactor vessel. A system of cooling pipes welded to the outer surface of the liner is used to transfer the heat from the reactor to the intermediate circuit. The high safety of this vessel concept results from the clear separation of the functions of the individual components and from the design principle of the prestressed conncrete. The prestressed concrete structure is so designed that failure can be reliably ruled out under all operating and accident conditions. Even in the extremely improbable event of failure of all decay heat removal systems when decay heat and accumulated heat are transferred passively by natural convection only, the integrity of the vessel remains intact. For reasons of plant availability the liner and the liner cooling system shall be designed so as to ensure safe elimination of failure over the total operating life. The calculations which were peformed partly on the basis of extremely adverse assumption, also resulted in very low loads. The prestressed concrete vessel is prefabricated to the greatest possible extent. Thus a high quality and optimized fabrication technology can be achieved especially for the liner and the liner cooling system. (orig./HP)

  16. Interfacial stability with mass and heat transfer

    International Nuclear Information System (INIS)

    Hsieh, D.Y.

    1977-07-01

    A simplified formulation is presented to deal with interfacial stability problems with mass and heat transfer. For Rayleigh-Taylor stability problems of a liquid-vapor system, it was found that the effect of mass and heat transfer tends to enhance the stability of the system when the vapor is hotter than the liquid, although the classical stability criterion is still valid. For Kelvin-Holmholtz stability problems, however, the classical stability criterion was found to be modified substantially due to the effect of mass and heat transfer

  17. Present status of heat transfer in narrow gap rectangular channel

    International Nuclear Information System (INIS)

    Sudo, Yukio; Kaminaga, Masanori

    1990-01-01

    In the safety evaluation for research nuclear reactors, at the time of abnormal transient change and accidents, after the tripping of a primary coolant pump, such event that the flow direction of coolant in a core reverses from steady downward flow to rising flow is supposed. In this case, the coexisting convection field, in which free convection and forced convection coexist, arises in place of forced convection, and especially in the research reactors using plate type fuel like JRR-3, it is important to grasp the heat transfer characteristics in the coexisting convection field in a narrow channel. Jackson et al. proposed the heat transfer correlation equation which can be applied to wide conditions including the coexisting convection zone, but its applicability to a narrow channel has not been confirmed. Based on the experimental results, in this study, the effect that the decrease of gap exerts to the convection heat transfer characteristics reported so far was investigated. The experiment and the results are reported. In this experiment on the coexisting convection zone in a narrow gap, the effect of main flow acceleration arose sufficiently large as compared with the effect of buoyancy, and heat transfer was promoted. (K.I.)

  18. 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

  19. 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.

  20. THERMOS, district central heating nuclear reactors

    International Nuclear Information System (INIS)

    Patarin, L.

    1981-02-01

    In order to expand the penetration of uranium in the national energy balance sheet, the C.E.A. has been studying nuclear reactors for several years now, that are capable of providing heat at favourable economic conditions. In this paper the THERMOS model is introduced. After showing the attraction of direct town heating by nuclear energy, the author describes the THERMOS project, defines the potential market, notably in France, and applies the lay-out study to the Grenoble Nuclear Study Centre site with district communal heating in mind. The economic aspects of the scheme are briefly mentioned [fr

  1. Alfven wave heating in a tokamak reactor

    International Nuclear Information System (INIS)

    Borg, G.G.; Appert, K.; Knight, A.J.; Lister, J.B.; Vaclavik, J.

    1990-01-01

    A number of features of Alfven wave heating make it potentially attractive for use in large tokamak reactors. Among them are the availability and relativity low cost of the power supplies, the potential ability to act selectively on the current profile, and the probable absence of operational limits in size, fields or density. The physics of Alfven wave heating in a large tokamak is assessed. Present theoretical understanding of mode coupling and antenna loading is extrapolated to a large machine. The problem of a recessed antenna is analysed. Calculations of loading and discussion of various heating scenarios for the particular case of NET are also presented. (author). 23 refs, 18 figs, 4 tabs

  2. Natural Convection Heat Transfer of Oxide Pool During In-Vessel Retention of Core Melts

    Energy Technology Data Exchange (ETDEWEB)

    Park, Hae-Kyun; Chung, Bum-Jin [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The integrity of reactor vessel may be threatened by the heat generation at the oxide pool and to the natural convection heat transfer to the reactor vessel by those two layers. Therefore, External Reactor Vessel Cooling (ERVC) is performed in order to secure the integrity of the reactor vessel. Whether the IVR(In-Vessel Retention) Strategy can be applicable to a larger reactor is the technical concern, which nourished the research interest for the natural convection heat transfer of metal and oxide pool and ERVC performance. Especially, it is hard to simulate oxide pool by experimentally due to the high level of buoyancy. Moreover, the volumetrically exothermic working fluid should be adopted to simulate the behavior of the core melts. Therefore, the volumetric heat sources that immersed in the working fluid have been adopted to simulate oxide pool by experiment. We investigated oxide pool with two different designs of the volumetric heat sources that adopted previous experiments. The investigation was performed by mass transfer experiment using analogy between heat and mass transfers. The results were compared to previous studies. We simulated the natural convection heat transfer of the oxide pool by mass transfer experiment. The isothermally cooled condition was established by limiting current technique firstly. The results were compared to previous studies under identical design of the volumetric heat sources. The average Nu's of the curvature and the top plate were close to the previous studies.

  3. Condensation heat transfer on natural convection at the high pressure

    International Nuclear Information System (INIS)

    Jong-Won, Kim; Hyoung-Kyoun, Ahn; Goon-Cherl, Park

    2007-01-01

    The Regional Energy Research Institute for the Next Generation is to develop a small scale electric power system driven by an environment-friendly and stable small nuclear reactor. REX-10 has been developed to assure high system safety in order to be placed in densely populated region and island. REX-10 adopts the steam-gas pressurizer to assure the inherent safety. The thermal-hydraulic phenomena in the steam-gas pressurizer are very complex. Especially, the condensation heat transfer with noncondensable gas on the natural convection is important to evaluate the pressurizer behavior. However, there have been few investigations on the condensation in the presence of noncondensable gas at the high pressure. In this study, the theoretical model is developed to estimate the condensation heat transfer at the high pressure using heat and mass transfer analogy. The analysis results show good agreement with correlations and experimental data. It is found that the condensation heat transfer coefficient increases as the total pressure increases or the mass fraction of the non-condensable gas decreases. In addition, the heat transfer coefficient no more increases over the specific pressure

  4. 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)

  5. 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

  6. A Heat Transfer Correlation in a Vertical Upward Flow of CO2 at Supercritical Pressures

    International Nuclear Information System (INIS)

    Kim, Hyung Rae; Bae, Yoon Yeong; Song, Jin Ho; Kim, Hwan Yeol

    2006-01-01

    Heat transfer data has been collected in the heat transfer test loop, named SPHINX (Supercritical Pressure Heat Transfer Investigation for NeXt generation), in KAERI. The facility primarily aims at the generation of heat transfer data in the flow conditions and geometries relevant to SCWR (SuperCritical Water-cooled Reactor). The produced data will aid the thermohydraulic design of a reactor core. The loop uses carbon dioxide, and later the results will be scaled to the water flows. The heat transfer data has been collected for a vertical upward flow in a circular tube with varying mass fluxes, heat fluxes, and operating pressures. The results are compared with the existing correlations and a new correlation is proposed by fine-tuning the one of the existing correlations

  7. Heat transfer studies in pool fire environment

    International Nuclear Information System (INIS)

    Nitsche, F.

    1993-01-01

    A Type B package has to withstand severe thermal accident conditions. To calculate the temperature behaviour of such a package in a real fire environment, heat transfer parameters simulating the effect of the fire are needed. For studying such heat transfer parameters, a systematic programme of experimental and theoretical investigations was performed which was part of the IAEA Coordinated Research Programme (Nitsche and Weib 1990). The studies were done by means of small, unfinned and finned, steel model containers of simplified design in hydrocarbon fuel open fire tests. By using various methods, flame and container temperatures were measured and also container surface absorptivity before and after the test to study the effect of sooting and surface painting on heat transfer. Based on all these experimental data and comparative calculations, simplified, effective heat transfer parameters could be derived, simulating the effect of the real fire on the model containers. (J.P.N.)

  8. Solid-Core Heat-Pipe Nuclear Batterly Type Reactor

    International Nuclear Information System (INIS)

    Ehud Greenspan

    2008-01-01

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP). Like the SAFE 400 space nuclear reactor core, the HPENHS core is comprised of fuel rods and HPs embedded in a solid structure arranged in a hexagonal lattice in a 3:1 ratio. The core is oriented horizontally and has a square rather cylindrical cross section for effective heat transfer. The HPs extend from the two axial reflectors in which the fission gas plena are embedded and transfer heat to an intermediate coolant that flows by natural-circulation. The HP-ENHS is designed to preserve many features of the ENHS including 20-year operation without refueling, very small excess reactivity throughout life, natural circulation cooling, walkaway passive safety, and robust proliferation resistance. The target power level and specific power of the HP-ENHS reactor are those of the reference ENHS reactor. Compared to previous ENHS reactor designs utilizing a lead or lead-bismuth alloy natural circulation cooling system, the HP-ENHS reactor offers a number of advantageous features including: (1) significantly enhanced passive decay heat removal capability; (2) no positive void reactivity coefficients; (3) relatively lower corrosion of the cladding (4) a core that is more robust for transportation; (5) higher temperature potentially offering higher efficiency and hydrogen production capability. This preliminary study focuses on five areas: material compatibility analysis, HP performance analysis, neutronic analysis, thermal-hydraulic analysis and safety analysis. Of the four high-temperature structural materials evaluated, Mo TZM alloy is the preferred choice; its upper estimated feasible operating temperature is 1350 K. HP performance is evaluated as a function of working fluid type, operating temperature, wick design and HP diameter and length. Sodium is the

  9. Burnout detector design for heat transfer experiments

    International Nuclear Information System (INIS)

    Dias, H.F.

    1992-01-01

    This paper describes the design of an burnout detector for heat transfer experiments, applied during tests for optimization of fuel elements for PWR reactors. The burnout detector avoids the fuel rods destruction during the experiments at the Centro de Desenvolvimento da Tecnologia Nuclear. The detector evaluates the temperature changes over the fuel rods in the temperature changes over the fuel rods in the area where the burnout phenomenon could be anticipated. As soon as the phenomenon appears, the system power supply is turned off. The thermal Circuit No. 1, during the experiments, had been composed by nine fuel rods feed parallelly by the same power supply. Fine copper wires had been attached at the centre and at the ends of the fuel rod to take two Wheat stone bridge arms. The detector had been applied across the bridge diagonals, which must be balanced the burnout excursion can be detected as a small but fast increase of the signal over the detector. Large scale experiments had been carried out to compare the resistance bridge performance against a thermocouple attached through the fuel rod wall. These experiments had been showed us the advantages of the first method over the last, because the bridge evaluates the whole fuel rod, while the thermocouple evaluates only the area where it had been attached. (author)

  10. Heat transfer from a high temperature condensable mixture

    International Nuclear Information System (INIS)

    Chan, S.H.; Cho, D.H.; Condiff, D.W.

    1978-01-01

    A new development in heat transfer is reported. It is concerned with heat transfer from a gaseous mixture that contains a condensable vapor and is at very high temperature. In the past, heat transfer associated with either a condensable mixture at low temperature or a noncondensable mixture at high temperature has been investigated. The former reduces to the classical problem of fog formation in, say, atmosphere where the rate of condensation is diffusion controlled (molecular or conductive diffusions). In the presence of noncondensable gases, heat transfer to a cooler boundary by this mechanism is known to be drastically reduced. In the latter case, where the high temperature mixture is noncondensable, radiative transfer may become dominant and a vast amount of existing literature exists on this class of problem. A fundamentally different type of problem of relevance to recent advances in open cycle MHD power plants and breeder reactor safety is considered. In the advanced coal-fired power plant using MHD as a topping cycle, a condensable mixture is encountered at temperatures of 2000 to 3000 0 . Condensation of the vaporized slag and seed materials at such a high temperature can take place in the MHD generator channel as well as in the radiant boiler. Similarly, in breeder reactor accident analyses involving hypothetical core disruptive accidents, a UO 2 vapor mixture at 400 0 K or higher is often considered. Since the saturation temperature of UO 2 at one atmosphere is close to 4000 0 K, condensation is also likely at a very high temperature. Accordingly, an objective of the present work is to provide an understanding of heat transfer and condensation mechanics insystems containing a high temperature condensable mixture. The results of the study show that, when a high temperature mixture is in contact with a cooler surface, a thermal boundary layer develops rapidly because of intensive radiative cooling from the mixture

  11. Rod Bundle Heat Transfer: Steady-State Steam Cooling Experiments

    International Nuclear Information System (INIS)

    Spring, J.P.; McLaughlin, D.M.

    2006-01-01

    Through the joint efforts of the Pennsylvania State University and the United States Nuclear Regulatory Commission, an experimental rod bundle heat transfer (RBHT) facility was designed and built. The rod bundle consists of a 7 x 7 square pitch array with spacer grids and geometry similar to that found in a modern pressurized water reactor. From this facility, a series of steady-state steam cooling experiments were performed. The bundle inlet Reynolds number was varied from 1 400 to 30 000 over a pressure range from 1.36 to 4 bars (20 to 60 psia). The bundle inlet steam temperature was controlled to be at saturation for the specified pressure and the fluid exit temperature exceeded 550 deg. C in the highest power tests. One important quantity of interest is the local convective heat transfer coefficient defined in terms of the local bulk mean temperature of the flow, local wall temperature, and heat flux. Steam temperatures were measured at the center of selected subchannels along the length of the bundle by traversing miniaturized thermocouples. Using an analogy between momentum and energy transport, a method was developed for relating the local subchannel centerline temperature measurement to the local bulk mean temperature. Wall temperatures were measured using internal thermocouples strategically placed along the length of each rod and the local wall heat flux was obtained from an inverse conduction program. The local heat transfer coefficient was calculated from the data at each rod thermocouple location. The local heat transfer coefficients calculated for locations where the flow was fully developed were compared against several published correlations. The Weisman and El-Genk correlations were found to agree best with the RBHT steam cooling data, especially over the range of turbulent Reynolds numbers. The effect of spacer grids on the heat transfer enhancement was also determined from instrumentation placed downstream of the spacer grid locations. The local

  12. 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)

  13. GAM-HEAT -- a computer code to compute heat transfer in complex enclosures

    International Nuclear Information System (INIS)

    Cooper, R.E.; Taylor, J.R.; Kielpinski, A.L.; Steimke, J.L.

    1991-02-01

    The GAM-HEAT code was developed for heat transfer analyses associated with postulated Double Ended Guillotine Break Loss Of Coolant Accidents (DEGB LOCA) resulting in a drained reactor vessel. In these analyses the gamma radiation resulting from fission product decay constitutes the primary source of energy as a function of time. This energy is deposited into the various reactor components and is re- radiated as thermal energy. The code accounts for all radiant heat exchanges within and leaving the reactor enclosure. The SRS reactors constitute complex radiant exchange enclosures since there are many assemblies of various types within the primary enclosure and most of the assemblies themselves constitute enclosures. GAM-HEAT accounts for this complexity by processing externally generated view factors and connectivity matrices, and also accounts for convective, conductive, and advective heat exchanges. The code is applicable for many situations involving heat exchange between surfaces within a radiatively passive medium. The GAM-HEAT code has been exercised extensively for computing transient temperatures in SRS reactors with specific charges and control components. Results from these computations have been used to establish the need for and to evaluate hardware modifications designed to mitigate results of postulated accident scenarios, and to assist in the specification of safe reactor operating power limits. The code utilizes temperature dependence on material properties. The efficiency of the code has been enhanced by the use of an iterative equation solver. Verification of the code to date consists of comparisons with parallel efforts at Los Alamos National Laboratory and with similar efforts at Westinghouse Science and Technology Center in Pittsburgh, PA, and benchmarked using problems with known analytical or iterated solutions. All comparisons and tests yield results that indicate the GAM-HEAT code performs as intended

  14. Nanoparticle enhanced ionic liquid heat transfer fluids

    Science.gov (United States)

    Fox, Elise B.; Visser, Ann E.; Bridges, Nicholas J.; Gray, Joshua R.; Garcia-Diaz, Brenda L.

    2014-08-12

    A heat transfer fluid created from nanoparticles that are dispersed into an ionic liquid is provided. Small volumes of nanoparticles are created from e.g., metals or metal oxides and/or alloys of such materials are dispersed into ionic liquids to create a heat transfer fluid. The nanoparticles can be dispersed directly into the ionic liquid during nanoparticle formation or the nanoparticles can be formed and then, in a subsequent step, dispersed into the ionic liquid using e.g., agitation.

  15. The future of the low temperature district heating reactor

    International Nuclear Information System (INIS)

    Lu Yingzhong; Wang Dazhong; Ma Changwen; Dong Duo; Tian Jiafu.

    1984-01-01

    In this paper, the role, development and situation of the low temperature district heating reactor (LTDHR) are briefly summarized. There are four types of LTDHR. They are PWR, reactor with boiling in the chimney, organic reactor and swimming pool reactor. The features of these reactors are introduced. The situation and role of the LTDHR in the future of the energy system are also discussed. The experiment on nuclear district heating with the swimming pool reactor in Qinghua Univ. is described briefly. (Author)

  16. Heat transfer analysis of short helical borehole heat exchangers

    International Nuclear Information System (INIS)

    Zarrella, Angelo; De Carli, Michele

    2013-01-01

    Highlights: ► Vertical ground heat exchanger with a helical shaped pipe is analyzed. ► The model considers the interaction between the ground and the environment. ► The results of the model are in good agreement with the experimental values. ► The weather conditions considerably affect the fluid heat carrier temperature. ► The pitch between the turns does not affect the behaviour of the heat exchanger. -- Abstract: In this paper a numerical model to analyze the thermal behaviour of vertical ground heat exchangers with a helical shaped pipe is presented. This type of configuration can be a suitable alternative to conventional ground heat exchangers, especially when the heating and cooling loads of the building are very low. The model describes the heat transfer problem by means of a network of interconnected thermal resistances and capacitances. Moreover, as the investigated ground heat exchanger is usually installed in shallow depth, the model takes into account the interaction between the ground and the ambient environment which affects the fluid heat carrier temperature into the heat exchanger and, as a consequence, the energy efficiency of the heat pump. After a sensitivity analysis on the mesh parameters, the presented model is compared with experimental data and the simulation results show good agreement with the measurements. Finally, analyses to investigate the influence of the weather conditions, of the axial heat transfer and of the pitch between the turns of the helical pipe for two types of ground are carried out.

  17. Assessment of the Use of Nitrogen Trifluoride for Purifying Coolant and Heat Transfer Salts in the Fluoride Salt-Cooled High-Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Scheele, Randall D.; Casella, Andrew M.

    2010-09-28

    This report provides an assessment of the use of nitrogen trifluoride for removing oxide and water-caused contaminants in the fluoride salts that will be used as coolants in a molten salt cooled reactor.

  18. Effect of radiation heat transfer on the performance of high temperature heat exchanger, (2)

    International Nuclear Information System (INIS)

    Yamada, Yukio; Mori, Yasuo; Hijikata, Kunio.

    1977-01-01

    In high temperature helium gas-cooled reactors, the nuclear energy can be utilized effectively, and the safety is excellent as compared with conventional reactors. They are advantageous also in view of environmental problems. In this report, the high temperature heat exchanger used for heating steam with the helium from a high temperature gas reactor is modeled, and the case that radiating gas flow between parallel plates is considered. Analysis was made on the case of one channel and constant heat flux and on the model for a counter-flow type heat exchanger with two channels, and the effect of radiation on the heat transfer in laminar flow and turbulent flow regions was clarified theoretically. The basic equations, the method of approximate solution and the results of calculation are explained. When one dimensional radiation was considered, the representative temperature Tr regarding fluid radiation was introduced, and its relation to mean mixing temperature Tm was determined. It was clarified that the large error in the result did not arise even if Tr was taken equally to Tm, especially in case of turbulent flow. The error was practically negligible when the rate of forced convection heat transfer in case of radiating medium flow was taken same as that in the case without radiation. (Kako, I.)

  19. 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

  20. 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...

  1. Nonlocal heat transfer in nanostructures

    International Nuclear Information System (INIS)

    Kanavin, A.P.; Uryupin, S.A.

    2008-01-01

    Kinetics of electrons in a degenerate conductor heated up by absorption of a high-frequency field localized in a region of about hundred nanometers has been studied. A new law for nonlocal electron thermal flux has been predicted

  2. Behavior study on Na heat pipe in passive heat removal system of new concept molten salt reactor

    International Nuclear Information System (INIS)

    Wang Chenglong; Tian Wenxi; Su Guanghui; Zhang Dalin; Wu Yingwei; Qiu Suizheng

    2013-01-01

    The high temperature Na heat pipe is an effective device for transporting heat, which is characterized by remarkable advantages in conductivity, isothermally and passively working. The application of Na heat pipe on passive heat removal system of new concept molten salt reactor (MSR) is significant. The transient performance of high temperature Na heat pipe was simulated by numerical method under the MSR accident. The model of the Na heat pipe was composed of three conjugate heat transfer zones, i.e. the vapor, wick and wall. Based on finite element method, the governing equations were solved by making use of FORTRAN to acquire the profiles of the temperature, velocity and pressure for the heat pipe transient operation. The results show that the high temperature Na heat pipe has a good performance on operating characteristics and high heat transfer efficiency from the frozen state. (authors)

  3. A passive emergency heat sink for water-cooled reactors with particular application to CANDU reactors

    International Nuclear Information System (INIS)

    Spinks, N.J.

    1996-01-01

    Water in an overhead pool can serve as a general-purpose passive emergency heat sink for water-cooled reactors. It can be used for containment cooling, for emergency depressurization of the heat transport-system, or to receive any other emergency heat, such as that from the CANDU moderator. The passive emergency water system provides in-containment depressurization of steam generators and no other provision is needed for supply of low-pressure emergency water to the steam generators. For containment cooling, the pool supplies water to the tube side of elevated tube banks inside containment. The elevation with respect to the reactor heat source maximizes heat transport, by natural convection, of hot containment gases. This effective heat transport combines with the large heat-transfer coefficients of tube banks, to reduce containment overpressure during accidents. Cooled air from the tube banks is directed past the break in the heat-transport system, to facilitate removal of hydrogen using passive catalytic recombiners. (author)

  4. Heat diffusion in cylindrical fuel elements of water cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    Randles, J [Technical Assessments and Services Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1961-09-15

    This report contains a theoretical study of heat diffusion in the cylindrical fuel elements of water reactors. After setting up appropriate boundary conditions on the temperature, the steady state Fourier equation is solved both for a flat and a tilted fission power source. It is shown that source tilting does not have an appreciable effect on the peak fuel temperature while the heat flux to the coolant suffers a circumferential variation of less than a half of that of the fission power. In the last section, the theory is extended to include the effect of a flat, time dependent fission power. The time dependent Fourier equation is solved by means of a Dini series of Bessel functions which is shown to be rapidly convergent. From this series is derived expressions for the fuel element transfer functions required in reactor servo-analysis. These have the form of a rapidly convergent series of time-lag terms. (author)

  5. Conjugate Heat Transfer Study in Hypersonic Flows

    Science.gov (United States)

    Sahoo, Niranjan; Kulkarni, Vinayak; Peetala, Ravi Kumar

    2018-04-01

    Coupled and decoupled conjugate heat transfer (CHT) studies are carried out to imitate experimental studies for heat transfer measurement in hypersonic flow regime. The finite volume based solvers are used for analyzing the heat interaction between fluid and solid domains. Temperature and surface heat flux signals are predicted by both coupled and decoupled CHT analysis techniques for hypersonic Mach numbers. These two methodologies are also used to study the effect of different wall materials on surface parameters. Effectiveness of these CHT solvers has been verified for the inverse problem of wall heat flux recovery using various techniques reported in the literature. Both coupled and decoupled CHT techniques are seen to be equally useful for prediction of local temperature and heat flux signals prior to the experiments in hypersonic flows.

  6. Theory of periodic conjugate heat transfer

    CERN Document Server

    Zudin, Yuri B

    2016-01-01

    This book presents the theory of periodic conjugate heat transfer in detail. It offers 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.

  7. Cornish heat transfer experiment - final report

    International Nuclear Information System (INIS)

    Bourke, P.J.; Hodgkinson, D.P.

    1985-01-01

    The transfer of heat released in an in-site heating experiment simulating high level radioactive waste packages in granite in Cornwall has been found to be mainly by conduction but some appreciable convection does occur. Interim analysis of the data suggests that the latter may account for about 20% of the total. (author)

  8. Interactive Heat Transfer Simulations for Everyone

    Science.gov (United States)

    Xie, Charles

    2012-01-01

    Heat transfer is widely taught in secondary Earth science and physics. Researchers have identified many misconceptions related to heat and temperature. These misconceptions primarily stem from hunches developed in everyday life (though the confusions in terminology often worsen them). Interactive computer simulations that visualize thermal energy,…

  9. Heat transfer of liquid-metal magnetohydrodynamic flow with internal heat generation

    International Nuclear Information System (INIS)

    Kumamaru, Hiroshige; Kurita, Kazuhisa; Kodama, Satoshi

    2000-01-01

    Numerical calculations on heat transfer of a magnetohydrodynamic (MHD) flow with internal heat generation in a rectangular channel have been performed for the cases of very-large Hartmann numbers, finite wall conductivities and small aspect ratio (i.e. small length ratios of the channel side perpendicular to the applied magnetic field and the side parallel to the field), simulating typical conditions for a fusion-reactor blanket. The Nusselt numbers of the MHD flow in rectangular channels with aspect ratios of 1/10 to 1/40 for Hartmann numbers of ∼5 x 10 5 become ∼10 times higher than those for the corresponding flow under no magnetic field. The Nusselt number becomes higher as the internal heat generation rate increases as far as the heat generation rates in a fusion reactor blanket are considered. (author)

  10. Investigation of the possibility of using residual heat reactor energy

    Science.gov (United States)

    Aminov, R. Z.; Yurin, V. E.; Bessonov, V. N.

    2017-11-01

    The largest contribution to the probable frequency of core damage is blackout events. The main component of the heat capacity at each reactor within a few minutes following a blackout is the heat resulting from the braking of beta-particles and the transfer of gamma-ray energy by the fission fragments and their decay products, which is known as the residual heat. The power of the residual heat changes gradually over a long period of time and for a VVER-1000 reactor is about 15-20 MW of thermal power over 72 hours. Current cooldown systems increase the cost of the basic nuclear power plants (NPP) funds without changing the amount of electricity generated. Such systems remain on standby, accelerating the aging of the equipment and accordingly reducing its reliability. The probability of system failure increases with the duration of idle time. Furthermore, the reactor residual heat energy is not used. A proposed system for cooling nuclear power plants involves the use of residual thermal power to supply the station’s own needs in emergency situations accompanied by a complete blackout. The thermal power of residual heat can be converted to electrical energy through an additional low power steam turbine. In normal mode, the additional steam turbine generates electricity, which makes it possible to ensure spare NPP and a return on the investment in the reservation system. In this work, experimental data obtained from a Balakovo NPP was analyzed to determine the admissibility of cooldown of the reactors through the 2nd circuit over a long time period, while maintaining high-level parameters for the steam generated by the steam generators.

  11. Industrial furnace with improved heat transfer

    Energy Technology Data Exchange (ETDEWEB)

    Hoetzl, M.; Lingle, T.M.

    1993-07-20

    A method is described for effecting improved heat transfer with in an industrial furnace having a cylindrical furnace section, a door at one end of the furnace section, an end plate at the opposite end of the section a circular fan plate concentrically positioned within the furnace section to define a cylindrical fan chamber between the plate and the end section with a fan there between and a heat treat chamber between the plate and the door, the fan plate defining a non-orificing annular space extending between the interior of the cylindrical furnace section and the outer edge of the plate, the plate having a centrally located under-pressure opening extending there through and a plurality of circumferentially spaced tubular heating elements extending through the annular space into the heat treating chamber, the method comprising the steps of: (a) heating the heating elements to a temperature which is hotter that the temperature of the work within the heat treating chamber; (b) rotating the fan at a speed sufficient to form a portion of the furnace atmosphere as a wind mass swirling about the fan chamber; (c) propagating the wind mass through the annular space into the heat treating chamber as a swirling wind mass in the form of an annulus, the wind mass impinging the heating elements to establish heat transfer contact therewith while the mass retains its annulus shape until contacting the door and without any significant movement of the wind mass into the center of the heat treating chamber; (d) drawing the wind mass through the under-pressure zone after the wind mass comes into heat transfer contact with the work in the heat treating chamber; and (e) thereafter heating the work by radiation from the beating elements at high furnace temperatures in excess of about 1,600 F.

  12. A decay heat removal methodology for reuseable orbital transfer vehicles

    Science.gov (United States)

    McDaniel, Patrick J.; Perkins, David R.

    1992-07-01

    Operation of a nuclear thermal rocket(NTR) as the propulsion system for a reusable orbital transfer vehicle has been considered. This application is the most demanding in terms of designing a multiple restart capability for an NTR. The requirements on a NTR cooling system associated with the nuclear decay heat stored during operation have been evaluated, specifically for a Particle Bed Reactor(PBR) configuration. A three mode method of operation has been identified as required to adequately remove the nuclear decay heat.

  13. Local heat transfer where heated rods touch in axially flowing water

    International Nuclear Information System (INIS)

    Kast, S.J.

    1983-05-01

    An anlaytic model is developed to predict the azimuthal width of a stablesteam blanket region near the line of contact between two heated rods cooled by axially flowing water at high pressure. The model is intended to aid analysis of reduced surface heat transfer capability for the abnormal configuration of nuclear fuel rods bowed into contact in the core of a pressurized water nuclear reactor. The analytic model predicts the azimuthal width of the steam blanket zone having reduced surface heat transfer as a function of rod average heat flux, subchannel coolant conditions and rod dimensions. The analytic model is developed from a heat balance between the heat generated in the wall of a heated empty tube and the heat transported away by transverse mixing and axial convection in the coolant subchannel. The model is developed for seveal geometries including heated rods in line contact, a heated rod touching a short insulating plane and a heated rod touching the inside of a metal guide tube

  14. Pump/heat exchanger assembly for pool-type reactor

    International Nuclear Information System (INIS)

    Nathenson, R.D.; Slepian, R.M.

    1987-01-01

    A heat exchanger and pump assembly comprising a heat exchanger including a housing for defining an annularly shaped cavity and supporting therein a plurality of heat transfer tubes. A pump is disposed beneath the heat exchanger and is comprised of a plurality of flow couplers disposed in a circular array. Each flow coupler is comprised of a pump duct for receiving a first electrically conductive fluid, i.e. the primary liquid metal, from a pool thereof, and a generator duct for receiving a second electrically conductive fluid, i.e. the intermediate liquid metal. The primary liquid metal is introduced from the reactor pool into the top, inlet ends of the tubes, flowing downward therethrough to be discharged from the tubes' bottom ends directly into the reactor pool. The primary liquid metal is variously introduced into the pump ducts directly from the reactor pool, either from the bottom or top end of the flow coupler. The intermediate fluid introduced into the generator ducts via the inlet duct and inlet plenum and after leaving the generator ducts passes through the annular cavity of the exchanger to cool the primary liquid in the tubes. The annular magnetic field of the pump is produced by a circular array of electromagnets having hollow windings cooled by a flow of the intermediate metal. (author)

  15. Heat transfer enhancement in heat exchangers by longitudinal vortex generators

    International Nuclear Information System (INIS)

    Guntermann, T.; Fiebig, M.; Mitra, N.K.

    1990-01-01

    In this paper heat transfer enhancement and flow losses are computed for the interaction of a laminar channel flow with a pair of counterrotating longitudinal vortices generated by a pair of delta-winglets punched out of the channel wall. The geometry simulates an element of a fin-plate or fin-tube heat exchanger. The structure of the vortex flow and temperature distribution, the local heat transfer coefficients and the local flow losses are discussed for a sample case. For a Reynolds number of Re d = 1000 and a vortex generator angle of attack of β = 25 degrees heat transfer is enhanced locally by more than 300% and in the mean by 50%. These values increase further with Re and β

  16. Heat transfer with freezing and thawing

    CERN Document Server

    Lunardini, VJ

    1991-01-01

    This volume provides a comprehensive overview on the vast amount of literature on solidification heat transfer. Chapter one develops important basic equations and discusses the validity of considering only conductive heat transfer, while ignoring convection, in the large class of materials which make up the porous media. Chapters 2 to 4 deal with problems that can be expressed in plane (Cartesian) coordinates. These problems are further divided into boundary conditions of temperature, prescribed heat flux, and surface convection. Chapter 5 examines some plane geometries involving three-dime

  17. 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.

  18. Modelling of heat transfer to fluids at a supercritical pressure

    International Nuclear Information System (INIS)

    Shuisheng, He

    2014-01-01

    A key feature of Supercritical Water-cooled Reactor (SCWR) is that, by raising the pressure of the reactor coolant fluid above the critical value, a phase change crisis is avoided. However, the changes in water density as it flows through the core of an SCWR are actually much higher than in the current water-cooled reactors. In a typical design, the ratio of the density of water at the core inlet to that at exit is as high as 7:1. Other fluid properties also vary significantly, especially around the pseudo-critical temperature (at which the specific heat capacity peaks). As a result, turbulent flow and heat transfer behaviour in the core is extremely complex and under certain conditions, significant heat transfer deterioration can potentially occur. Consequently, understanding and being able to predict flow and heat transfer phenomena under normal steady operation conditions and in start-up and hypothetical fault conditions are fundamental to the design of SCWR. There have been intensive studies on flow and heat transfer to fluids at supercritical pressure recently and several excellent review papers have been published. In the talk, we will focus on some turbulence modelling issues encountered in CFD simulations. The talk will first discuss some flow and heat transfer issues related to fluids at supercritical pressures and their potential implications in SCWR, and some recent developments in the understanding and modelling techniques of such problems, which will be followed by an outlook for some future developments.Factors which have a major influence on the flow and will be discussed are buoyancy and flow acceleration due to thermal expansion (both are due to density variations but involve different mechanisms) and the nonuniformity of other fluid properties. In addition, laminar-turbulent flow transition coupled with buoyancy and flow acceleration plays an important role in heat transfer effectiveness and wall temperature in the entrance region but such

  19. Indirect evaporative coolers with enhanced heat transfer

    Science.gov (United States)

    Kozubal, Eric; Woods, Jason; Judkoff, Ron

    2015-09-22

    A separator plate assembly for use in an indirect evaporative cooler (IEC) with an air-to-air heat exchanger. The assembly includes a separator plate with a first surface defining a dry channel and a second surface defining a wet channel. The assembly includes heat transfer enhancements provided on the first surface for increasing heat transfer rates. The heat transfer enhancements may include slit fins with bodies extending outward from the first surface of separator plate or may take other forms including vortex generators, offset strip fins, and wavy fins. In slit fin implementations, the separator plate has holes proximate to each of the slit fins, and the separator plate assembly may include a sealing layer applied to the second surface of the separator plate to block air flow through the holes. The sealing layer can be a thickness of adhesive, and a layer of wicking material is applied to the adhesive.

  20. After-heat removing system in FBR type reactor

    International Nuclear Information System (INIS)

    Ohashi, Yukio.

    1990-01-01

    The after-heat removing system of the present invention removes the after heat generated in a reactor core without using dynamic equipments such as pumps or blowers. There are disposed a first heat exchanger for heating a heat medium by the heat in a reactor container and a second heat exchanger situated above the first heat exchanger for spontaneously air-cooling the heat medium. Recycling pipeways connect the first and the second heat exchangers to form a recycling path for the heat medium. Then, since the second heat exchanger for spontaneously air-cooling the heat medium is disposed above the first heat exchanger and they are connected by the recycling pipeways, the heat medium can be circulated spontaneously. Accordingly, dynamic equipments such as pumps or blowers are no more necessary. As a result, the after-heat removing system of the FBR type reactor of excellent safety and reliability can be obtained. (I.S.)

  1. Thermohydraulic behaviour and heat transfer in the molten core

    International Nuclear Information System (INIS)

    Reineke, H.H.

    1977-01-01

    Increasing the application of nuclear reactors to produce electrical power extremely unprobable accidents should be investigated too. In the Federal Republic of Germany, a research program is performed for some years engaged in accidents at light water reactors in which the melting of the reactor core is presumed. A part of this program is to investigate the thermohydraulic and the heat transfer behavior in an accumulation of molten core material. The knowledge of these events is necessary to analyse the accident exactly. Further on the results of this work are of great importance to build a catcher for the molten core material. As a result of the decay heat the molten material is heated up and the density differences induce a free convection motion. In this work the thermohydraulic behavior and the distribution of the escaping heat fluxes for several accumulations of molten core material were determined. The numerical methods for solving the system of partial differential equation were used to develop computer codes, able to compute the average and local heat fluxes at the walls enclosing the molten core material and the inside increase of the temperature. The numerical computations were confirmed and verified by experimental investigations. In these investigations the molten core material was always assumed as a homogeneous fluid. In this case, the results could be reproduced by simple power laws

  2. Solution of heat removal from nuclear reactors by natural convection

    Directory of Open Access Journals (Sweden)

    Zitek Pavel

    2014-03-01

    Full Text Available This paper summarizes the basis for the solution of heat removal by natural convection from both conventional nuclear reactors and reactors with fuel flowing coolant (such as reactors with molten fluoride salts MSR.The possibility of intensification of heat removal through gas lift is focused on. It might be used in an MSR (Molten Salt Reactor for cleaning the salt mixture of degassed fission products and therefore eliminating problems with iodine pitting. Heat removal by natural convection and its intensification increases significantly the safety of nuclear reactors. Simultaneously the heat removal also solves problems with lifetime of pumps in the primary circuit of high-temperature reactors.

  3. Heat transfer in the atmosphere

    NARCIS (Netherlands)

    Oerlemans, J.

    1982-01-01

    The atmosphere is almost transparent to solar radiation and almost opaque to terrestrial radiation. This implies that in the mean the atmosphere cools while the earth's surface is heated. Convection in the lower atmosphere must therefore occur. The upward flux of energy associated with it

  4. 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.

  5. 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.

  6. Development of heat transfer models for gap cooling

    Energy Technology Data Exchange (ETDEWEB)

    Kohriyama, Tamio; Murase, Michio; Tamaki, Tomohiko [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

    2001-09-01

    In a severe accident of a light water reactor (LWR), heat transfer models in a narrow annular gap between superheated core debris and a reactor pressure vessel (RPV) are important to evaluate the integrity of RPV and emergency procedures. This paper discusses the effects of superheat on the heat flux based on existing data. In low superheat conditions, the heat flux in the narrow gap is higher than the heat flux in pool nucleate boiling due to restricted flow area. It approaches the nucleate boiling heat flux as superheat increasing and reaches a critical value subject to the counter-current flow limiting (CCFL) at the top end of the gap. A heat transfer correlation was derived as a function of dimensionless superheat and a Kutateladze-type CCFL correlation was deduced for critical heat flux (CHF) restricted by CCFL, which gave good prediction for a wide range of the CHF data. Effect of an angle of inclination of the gap could also be incorporated in the CCFL correlation. In high superheat conditions, the heat flux in the narrow gap maintains a similar shape to the pool boiling curve but shifts the position to a higher superheated side than the pool boiling except film boiling, which could be expressed by the typical pool film boiling correlation. Incorporating quench test data, the heat flux correlation was derived as a function of dimensionless superheat using the same formula for the low superheat and the Kutateladze-type CCFL correlation was deduced for CHF. The CHF at the high superheat was 3-4 times as large as CHF at the low superheat and this difference was well predicted by different flow patterns in the gap and the balance of pressure gradients between gas and liquid phases. (author)

  7. Falling film flow, heat transfer and breakdown on horizontal tubes

    International Nuclear Information System (INIS)

    Rogers, J.T.

    1980-11-01

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

  8. Heating control system for nuclear reactor

    International Nuclear Information System (INIS)

    Shinohara, Kaoru.

    1981-01-01

    Purpose: To automatically control reactor heating while keeping the condition of temperature rising rate by determining the deviations based on the reactor water temperature, the aimed temperature and the aimed temperature rising rate and operating control rods. Constitution: Actual temperature in the reactor is measured by a temperature detector and compared with a value from a setter to determine the temperature deviation. While on the other hand, the rising rate for the measured temperature is calculated in a differentiator and compared with a value from a setter to determine the deviation, which is passed through an integrator to calculate the deviation for the temperature rising rate. The signals for the temperature deviation and the temperature rising rate deviation are selected in a lower value preference circuit and the operation amount for the control rod is judged in a control rod operation judging section depending on the deviation amount. The control rod to be operated is determined in a sequence control section for the selection of control rod. The control rod selected and the direction of the operation are displayed on a display and the selected control rod is automatically driven by a control rod drives to thereby carry our reactor heating. (Furukawa, Y.)

  9. Lunar ash flow with heat transfer.

    Science.gov (United States)

    Pai, S. I.; Hsieh, T.; O'Keefe, J. A.

    1972-01-01

    The most important heat-transfer process in the ash flow under consideration is heat convection. Besides the four important nondimensional parameters of isothermal ash flow (Pai et al., 1972), we have three additional important nondimensional parameters: the ratio of the specific heat of the gas, the ratio of the specific heat of the solid particles to that of gas, and the Prandtl number. We reexamine the one dimensional steady ash flow discussed by Pai et al. (1972) by including the effects of heat transfer. Numerical results for the pressure, temperature, density of the gas, velocities of gas and solid particles, and volume fraction of solid particles as function of altitude for various values of the Jeffreys number, initial velocity ratio, and two different gas species (steam and hydrogen) are presented.

  10. Heat transfer in a thermoacoustic process

    International Nuclear Information System (INIS)

    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 aimed at determining the stability–instability border of the thermoacoustic system. In this paper, we present a project type of physical examination and modelling task. We employed an electrically heated Rijke tube in our thermoacoustic project work. The aim of our project is to help our students enlarge their knowledge about thermodynamics, mainly about thermoacoustics, and develop their applied information technology and mathematical skills. (paper)

  11. SLOWPOKE: heating reactors in the urban environment

    International Nuclear Information System (INIS)

    Hilborn, J.W.; Lynch, G.F.

    1988-06-01

    Since global energy requirements are expected to double over the next 40 years, nuclear heating could become as important as nuclear electricity generation. To fill that need, AECL has designed a 10 MW nuclear heating plant for large buildings. Producing hot water at temperatures below 100 degrees Celsius, it incorporates a small pool-type reactor based on the successful SLOWPOKE Research Reactor. A 2 MW prototype is now being tested at the Whiteshell Nuclear Research Establishment in Manitoba, and the design of a 10 MW commercial unit is well advanced. With capital costs in the range $5 million to $7 million, unit energy costs could be as low as $0.02 per kWh, for a unit operating at 50% load factor over a 25-year period. By keeping the reactor power low and the water temperature below 100 degrees Celsius, much of the complexity of the large nuclear power plants can be avoided, thus allowing these small, safe, nuclear heating systems to be economically viable

  12. Laminar Mixed Convection Heat Transfer Correlation for Horizontal Pipes

    International Nuclear Information System (INIS)

    Chae, Myeong Seon; Chung, Bum Jin

    2013-01-01

    This study aimed at producing experimental results and developing a new heat transfer correlation based upon a semi-empirical buoyancy coefficient. Mixed convection mass transfers inside horizontal pipe were investigated for the pipe of various length-to-diameters with varying Re. Forced convection correlation was developed using a very short cathode. With the length of cathode increase and Re decrease, the heat transfer rates were enhanced and becomes higher than that of forced convection. An empirical buoyancy coefficient was derived from correlation of natural convection and forced convection with the addition of L/D. And the heat transfer correlation for laminar mixed convection was developed using the buoyancy coefficient, it describes not only current results, but also results of other studies. Mixed convection occurs when the driving forces of both forced and natural convections are of comparable magnitude (Gr/Re 2 ∼1). It is classical problem but is still an active area of research for various thermal applications such as flat plate solar collectors, nuclear reactors and heat exchangers. The effect of buoyancy on heat transfer in a forced flow is varied by the direction of the buoyancy force. In a horizontal pipe the direction of the forced and buoyancy forces are perpendicular. The studies on the mixed convections of the horizontal pipes were not investigated very much due to the lack of practical uses compared to those of vertical pipes. Even the definitions on the buoyancy coefficient that presents the relative influence of the forced and the natural convections, are different by scholars. And the proposed heat transfer correlations do not agree

  13. Heat transfer from thermal effluent

    International Nuclear Information System (INIS)

    Czapski, U.H.; Mumford, W.

    1975-01-01

    Measurements of the turbulent fluxes of sensible heat and momentum, together with profiles of horizontal wind, temperature, and humidity (wet bulb) have been conducted above the thermal plume of the Nine Mile Point Nuclear plant near Oswego, New York on Lake Ontario. The spectral analysis of the data, obtained with sonic anemometer and ultrafast thermocouples, reveals the importance of microthermals and similar features for the transport of heat. Temperature variance spectra and the cospectra wT and uw show distinct deviations from the -5/3 Kolmogorov law in the inertial subrange, suggesting a high input of energy in the eddy frequency range between 0.01 and 1 Hz. It is shown that microthermals in this frequency range are also responsible for a large portion of the momentum transport. 46 refs

  14. Heat Transfer Model for Hot Air Balloons

    OpenAIRE

    Lladó Gambín, Adriana

    2016-01-01

    A heat transfer model and analysis for hot air balloons is presented in this work, backed with a flow simulation using SolidWorks. The objective is to understand the major heat losses in the balloon and to identify the parameters that affect most its flight performance. Results show that more than 70% of the heat losses are due to the emitted radiation from the balloon envelope and that convection losses represent around 20% of the total. A simulated heating source is also included in the mod...

  15. Development of Off-take Model, Subcooled Boiling Model, and Radiation Heat Transfer Input Model into the MARS Code for a Regulatory Auditing of CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, C.; Rhee, B. W.; Chung, B. D. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ahn, S. H.; Kim, M. W. [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2009-05-15

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to a lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use a vendor's code for a regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed the RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of the existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of the RELAP5/MOD3/CANDU code to the MARS code including a quality assurance of the developed models.

  16. Development of Off-take Model, Subcooled Boiling Model, and Radiation Heat Transfer Input Model into the MARS Code for a Regulatory Auditing of CANDU Reactors

    International Nuclear Information System (INIS)

    Yoon, C.; Rhee, B. W.; Chung, B. D.; Ahn, S. H.; Kim, M. W.

    2009-01-01

    Korea currently has four operating units of the CANDU-6 type reactor in Wolsong. However, the safety assessment system for CANDU reactors has not been fully established due to a lack of self-reliance technology. Although the CATHENA code had been introduced from AECL, it is undesirable to use a vendor's code for a regulatory auditing analysis. In Korea, the MARS code has been developed for decades and is being considered by KINS as a thermal hydraulic regulatory auditing tool for nuclear power plants. Before this decision, KINS (Korea Institute of Nuclear Safety) had developed the RELAP5/MOD3/CANDU code for CANDU safety analyses by modifying the model of the existing PWR auditing tool, RELAP5/MOD3. The main purpose of this study is to transplant the CANDU models of the RELAP5/MOD3/CANDU code to the MARS code including a quality assurance of the developed models

  17. Evaluation of heat transfer enhancement in air-heating collectors

    Energy Technology Data Exchange (ETDEWEB)

    Mattox, D. L.

    1979-06-01

    The present research effort was initiated for the purpose of increasing the thermal efficiency of air heating solar collectors through identification and development of optimum design and operation criteria for solar absorber-to-air heat exchangers. Initially this effort took the form of a solar collector systems analysis to evaluate the impact of various techniques for enhancing the heat transfer between the absorber and air stream on overall thermal performance of the entire solar collector. This systems analysis resulted in the selection of solar collector designs providing ducted cooling air on the absorber shaded side as a base line. A transient heat transfer analysis of a complete solar air heating collector was used to demonstrate that an optimum absorber-to-air heat exchanger design could be provided with several interrupted fin configurations. Additional analyses were performed to establish that the maximum solar collector thermal performance to required pumping power was realized for a Reynolds number range of 1000 to 2000. This Reynolds number range was used to establish a theoretical design limit curve for maximum thermal performance versus required pumping power for all interrupted fin designs as published in the open literature. Heat and momentum transfer empirical relationships were defined for scaling the state-of-the-art high conductance fin designs identified from a compact configuration to the less compact designs needed for solar collectors.

  18. Heat transfer in a sodium-to-sodium heat exchanger under conditions of combined force and free convection

    International Nuclear Information System (INIS)

    Jackson, J.D.; Axcell, B.P.; Johnston, S.E.

    1987-01-01

    A combined experimental and theoretical investigation of heat transfer in a vertical tube and annulus, countercurrent flow heat exchanger is reported. The working fluid was liquid sodium. Included in the range of conditions covered were those which are of interest in connection with the low flow rate operation of fast reactor intermediate heat exchanger systems. The heat transfer process ranged from that of pure forced convection to combined forced and free convection. By changing the direction of fluid flow or the direction of heat flow four different configurations were studied. In two cases the convection process was buoyancy aided and in the other two it was buoyancy opposed. Results are presented showing the influence of flow rate and temperature difference on overall heat transfer coefficient for each case. A theoretical model of turbulent flow and heat transfer incorporating influences of buoyancy was used to produce results for the range of conditions covered in the experiments. The predictions of overall heat transfer coefficient were found to be in reasonable general agreement with the measurements. It was clear from these calculations that the influence of buoyancy on heat transfer stemmed largely, under the conditions of the present experiment, from the modification of the convection process due to the distortion of the velocity field. This led to an enhancement of the heat transfer for the buoyancy-aided process and an impairment for the buoyancy-opposed process. The contribution of the turbulent diffusion of heat was relatively small. (author)

  19. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ehud Greenspan

    2008-09-30

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

  20. 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

  1. Convective heat transfer characteristics in the turbulent region of molten salt in concentric tube

    International Nuclear Information System (INIS)

    Chen, Y.S.; Wang, Y.; Zhang, J.H.; Yuan, X.F.; Tian, J.; Tang, Z.F.; Zhu, H.H.; Fu, Y.; Wang, N.X.

    2016-01-01

    In order to better understand the heat transfer behavior and characteristics of molten salt in heat exchanger, the convective heat transfer characteristics of molten salt in salt-to-oil concentric tube are studied. Overall heat transfer coefficients of the heat exchanger are calculated using Wilson plots. Heat transfer coefficients of tube side molten salt with the range of Reynolds number from 10,000 to 50,000 and the Prandtl number from 11 to 27 are evaluated invoking the calculated overall heat transfer coefficients. The effects of velocity and temperature on the convective heat transfer in the turbulent region of molten salt are studied by comparing with the traditional correlations. The results show that the heat transfer characteristics of molten salt are in line with the empirical heat transfer correlation; however, Dittus–Boelter, Gnielinski, Sieder–Tate and Hausen correlations all give a larger deviation for the experimental data. Finally, based on the experimental data and Sieder–Tate correlation, a modified heat transfer correlation is proposed and good agreement is observed between the experimental data and the modified correlation. The results will also provide an important reference for the design of the heat exchangers in the Thorium-based Molten Salt Reactor.

  2. 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

  3. Effect of Tube Diameter on Heat Transfer to Vertically Upward Flowing Supercritical CO2

    International Nuclear Information System (INIS)

    Kang, Deog Ji; Kim, Sin; Bae, Yoon Yeong; Kim, Hwan Yeol; Kim, Hyung Rae

    2007-01-01

    Heat transfer characteristics of supercritical carbon dioxide are being investigated experimentally in the test loop named as SPHINX(Supercritical Pressure Heat Transfer Investigation for NeXt generation) at KAERI. The main purpose of the experiment is to provide a reliable heat transfer database for a SCWR (SuperCritical Water-cooled Reactor) by a prudent extension of the carbon dioxide test results to the estimation of a heat transfer for water. The produced data will be used in the thermo-hydraulic design of core and safety analysis for SCWR. The aim of the present paper is to study the influence of a tube diameter on a heat transfer. The experiments were completed for tubes of an inside diameter of 4.4mm and 9.0mm, respectively. The heat transfer characteristics from the two tubes of different diameters were compared and discussed

  4. Effect of Tube Diameter on Heat Transfer to Vertically Upward Flowing Supercritical CO{sub 2}

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Deog Ji; Kim, Sin [Cheju National University, Jeju (Korea, Republic of); Bae, Yoon Yeong; Kim, Hwan Yeol; Kim, Hyung Rae [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2007-07-01

    Heat transfer characteristics of supercritical carbon dioxide are being investigated experimentally in the test loop named as SPHINX(Supercritical Pressure Heat Transfer Investigation for NeXt generation) at KAERI. The main purpose of the experiment is to provide a reliable heat transfer database for a SCWR (SuperCritical Water-cooled Reactor) by a prudent extension of the carbon dioxide test results to the estimation of a heat transfer for water. The produced data will be used in the thermo-hydraulic design of core and safety analysis for SCWR. The aim of the present paper is to study the influence of a tube diameter on a heat transfer. The experiments were completed for tubes of an inside diameter of 4.4mm and 9.0mm, respectively. The heat transfer characteristics from the two tubes of different diameters were compared and discussed.

  5. Calculations of combined radiation and convection heat transfer in rod bundles under emergency cooling conditions

    International Nuclear Information System (INIS)

    Sun, K.H.; Gonzalez-Santalo, J.M.; Tien, C.L.

    1976-01-01

    A model has been developed to calculate the heat transfer coefficients from the fuel rods to the steam-droplet mixture typical of Boiling Water Reactors under Emergency Core Cooling System (ECCS) operation conditions during a postulated loss-of-coolant accident. The model includes the heat transfer by convection to the vapor, the radiation from the surfaces to both the water droplets and the vapor, and the effects of droplet evaporation. The combined convection and radiation heat transfer coefficient can be evaluated with respect to the characteristic droplet size. Calculations of the heat transfer coefficient based on the droplet sizes obtained from the existing literature are consistent with those determined empirically from the Full-Length-Emergency-Cooling-Heat-Transfer (FLECHT) program. The present model can also be used to assess the effects of geometrical distortions (or deviations from nominal dimensions) on the heat transfer to the cooling medium in a rod bundle

  6. Condensing heat transfer following a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Krotiuk, W.J.; Rubin, M.B.

    1978-01-01

    A new method for calculating the steam mass condensation energy removal rates on cold surfaces in contact with an air-steam mixture has been developed. This method is based on the principles of mass diffusion of steam from an area of high concentration to the condensing surface, which is an area of low steam concentration. This new method of calculating mass condensation has been programmed into the CONTEMPT-LT Mod 26 computer code, which calculates the pressure and temperature transients inside a light water reactor containment following a loss-of-coolant accident. The condensing heat transfer coefficient predicted by the mass diffusion method is compared to existing semi-empirical correlations and to the experimental results of the Carolinas Virginia Tube Reactor Containment natural decay test. Closer agreement with test results is shown in the calculation of containment pressure, temperature, and heat sink surface temperature using the mass diffusion condensation method than when using any existing semi-empirical correlation

  7. 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.

  8. 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

  9. Experimental study of natural convective heat transfer in a vertical hexagonal sub channel

    International Nuclear Information System (INIS)

    Tandian, Nathanael P.; Umar, Efrizon; Hardianto, Toto; Febriyanto, Catur

    2012-01-01

    The development of new practices in nuclear reactor safety aspects and optimization of recent nuclear reactors, including the APWR and the PHWR reactors, needs a knowledge on natural convective heat transfer within sub-channels formed among several nuclear fuel rods or heat exchanger tubes. Unfortunately, the currently available empirical correlation equations for such heat transfer modes are limited and researches on convective heat transfer within a bundle of vertical cylinders (especially within the natural convection modes) are scarcely done. Although boundary layers around the heat exchanger cylinders or fuel rods may be dominated by their entry regions, most of available convection correlation equations are for fully developed boundary layers. Recently, an experimental study on natural convective heat transfer in a subchannel formed by several heated parallel cylinders that arranged in a hexagonal configuration has been being done. The study seeks for a new convection correlation for the natural convective heat transfer in the sub-channel formed among the hexagonal vertical cylinders. A new convective heat transfer correlation equation has been obtained from the study and compared to several similar equations in literatures.

  10. Chemical Kinetics, Heat Transfer, and Sensor Dynamics Revisited in a Simple Experiment

    Science.gov (United States)

    Sad, Maria E.; Sad, Mario R.; Castro, Alberto A.; Garetto, Teresita F.

    2008-01-01

    A simple experiment about thermal effects in chemical reactors is described, which can be used to illustrate chemical reactor models, the determination and validation of their parameters, and some simple principles of heat transfer and sensor dynamics. It is based in the exothermic reaction between aqueous solutions of sodium thiosulfate and…

  11. Design Guideline for Primary Heat Exchanger in a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sunil; Seo, Kyoung-Woo; Kim, Seong-Hoon; Chi, Dae-Young; Park, Cheol [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    In this paper, analytical study is conducted to track the variation of the PCS outlet temperature in conditions of the constant core power and constant SCS inlet temperature. The PCS circulates demineralized water to remove the heat generated in reactor core. The heat is transferred to the cold water of the SCS through the primary heat exchanger. In JRTR, Plate-type Heat Exchanger (PHE) was used as the primary heat exchanger. The cooling tower automatically sets the SCS inlet temperature constant by fan speed control. The flow rate of SCS is adjusted to be identical with the PCS flow rate. To design the PHE, the inlet and outlet temperatures and the flow rates for both systems should be determined. The flow rate has the allowable band for the safe operation from the lower limit to upper limit resulting in different temperature distribution in the PHE. Specially, the PCS outlet temperature which is the core inlet temperature is used for a safety parameter for the reactor shutdown. Therefore, we need to figure out which limit for the flow rate should be used from the conservative point of view. At 200 kg/s of PCS and SCS flow rates, the inlet and outlet temperatures are 41.3℃and 34℃, respectively. With increase of the flow rate, both of PCS inlet and outlet temperatures decrease to 33.6℃ and 39.9℃. This result means the low limit of the allowable flow band should be used for the conservative design of primary heat exchanger. If the upper limit of the allowable flow band is used, the PCS outlet temperature which is the safety parameter used for the reactor shutdown increases with decrease of the flow rate.

  12. Design Guideline for Primary Heat Exchanger in a Research Reactor

    International Nuclear Information System (INIS)

    Lee, Sunil; Seo, Kyoung-Woo; Kim, Seong-Hoon; Chi, Dae-Young; Park, Cheol

    2016-01-01

    In this paper, analytical study is conducted to track the variation of the PCS outlet temperature in conditions of the constant core power and constant SCS inlet temperature. The PCS circulates demineralized water to remove the heat generated in reactor core. The heat is transferred to the cold water of the SCS through the primary heat exchanger. In JRTR, Plate-type Heat Exchanger (PHE) was used as the primary heat exchanger. The cooling tower automatically sets the SCS inlet temperature constant by fan speed control. The flow rate of SCS is adjusted to be identical with the PCS flow rate. To design the PHE, the inlet and outlet temperatures and the flow rates for both systems should be determined. The flow rate has the allowable band for the safe operation from the lower limit to upper limit resulting in different temperature distribution in the PHE. Specially, the PCS outlet temperature which is the core inlet temperature is used for a safety parameter for the reactor shutdown. Therefore, we need to figure out which limit for the flow rate should be used from the conservative point of view. At 200 kg/s of PCS and SCS flow rates, the inlet and outlet temperatures are 41.3℃and 34℃, respectively. With increase of the flow rate, both of PCS inlet and outlet temperatures decrease to 33.6℃ and 39.9℃. This result means the low limit of the allowable flow band should be used for the conservative design of primary heat exchanger. If the upper limit of the allowable flow band is used, the PCS outlet temperature which is the safety parameter used for the reactor shutdown increases with decrease of the flow rate

  13. Pressure drop and heat transfer characteristics of a high-temperature printed circuit heat exchanger

    International Nuclear Information System (INIS)

    Chen, Minghui; Sun, Xiaodong; Christensen, Richard N.; Skavdahl, Isaac; Utgikar, Vivek; Sabharwall, Piyush

    2016-01-01

    Highlights: • Pressure drop and heat transfer characteristics of a high-temperature printed circuit heat exchanger have been obtained. • Comparisons of experimental data and available correlations have been performed. • New Fanning friction factor and heat transfer correlations for the test PCHE are developed. - Abstract: Printed circuit heat exchanger (PCHE) is one of the leading intermediate heat exchanger (IHX) candidates to be employed in the very-high-temperature gas-cooled reactors (VHTRs) due to its capability for high-temperature, high-pressure applications. In the current study, a reduced-scale zigzag-channel PCHE was fabricated using Alloy 617 plates for the heat exchanger core and Alloy 800H pipes for the headers. The pressure drop and heat transfer characteristics of the PCHE were investigated experimentally in a high-temperature helium test facility (HTHF) at The Ohio State University. The PCHE helium inlet temperatures and pressures were varied up to 464 °C/2.7 MPa for the cold side and 802 °C/2.7 MPa for the hot side, respectively, while the maximum helium mass flow rates on both sides of the PCHE reached 39 kg/h. The corresponding maximum channel Reynolds number was approximately 3558, covering the laminar flow and laminar-to-turbulent flow transition regimes. New pressure drop and heat transfer correlations for the current zigzag channels with rounded bends were developed based on the experimental data. Comparisons between the experimental data and the results obtained from the available PCHE and straight circular pipe correlations were conducted. Compared to the heat transfer performance in straight circular pipes, the zigzag channels provided little advantage in the laminar flow regime but significant advantage near the transition flow regime.

  14. 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.)

  15. A model for dispersed flow heat transfer in rod bundles during reflood

    International Nuclear Information System (INIS)

    Wong, S.

    1980-01-01

    The present model calculates the heat transfer characteristics of the non-equilibrium dispersed droplet flow regime above the quench front during reflood by solving simultaneously the wall-to-vapor interactions, wall-to-droplet interactions and vapor-to-droplet interactions by an iterative numerical method. The unique feature in the present study is various heat transfer mechanisms are combined in an overall energy balance equation, and the convective heat transfer to vapor is obtained by calculating the vapor temperature distributions at the heated walls. The reactor rod bundle geometry, axial variations of vapor temperature and flow properties, radiative heat transfers, and enhancement of heat transfer due to turbulence are considered carefully, so that the present model could be used to predict PWR (Pressurized Water Reactor) reflood heat transfers, and hence the fuel cladding wall temperature transients. In order to achieve closure of the problem formulations, the droplet size and its motion are determined from the FLECHT (Full Length Emergency Cooling Heat Transfer Program) low flooding rate series consine axial power shape test data. The model is then verified by comparing the heat transfer predictions with FLECHT low flooding rate series skewed axial power shape test data. Comparisons of predictions with data show satisfactory agreements

  16. Heat Transfer Characteristics for an Upward Flowing Supercritical Pressure CO2 in a Vertical Annulus Passage

    International Nuclear Information System (INIS)

    Kang, Deog Ji; Kim, Sin; Kim, Hwan Yeol; Bae, Yoon Yeong

    2007-01-01

    Heat transfer experiments at a vertical annulus passage were carried out in the SPHINX(Supercritical Pressure Heat Transfer Investigation for NeXt Generation) to investigate the heat transfer behaviors of supercritical CO 2 . The collected test data are to be used for the reactor core design of the SCWR (SuperCritical Water-cooled Reactor). The mass flux was in the range of 400 ∼1200 kg/m 2 s and the heat flux was chosen up to 150 kW/m 2 . The selected pressures were 7.75 and 8.12 MPa. The heat transfer data were analyzed and compared with the previous tube test data. The test results showed that the heat transfer characteristics were similar to those of the tube in case of a normal heat transfer mode and degree of heat transfer deterioration became smaller than that in the tube. Comparison of the experimental heat transfer coefficients with the predicted ones by the existing correlations showed that there was not a distinct difference between the correlations

  17. Atomic reactor thermal engineering

    International Nuclear Information System (INIS)

    Kim, Gwang Ryong

    1983-02-01

    This book starts the introduction of atomic reactor thermal engineering including atomic reaction, chemical reaction, nuclear reaction neutron energy and soon. It explains heat transfer, heat production in the atomic reactor, heat transfer of fuel element in atomic reactor, heat transfer and flow of cooler, thermal design of atomic reactor, design of thermodynamics of atomic reactor and various. This deals with the basic knowledge of thermal engineering for atomic reactor.

  18. Two phase nonequilibrium heat transfer in the TRAC-PD2 code

    International Nuclear Information System (INIS)

    Mandell, D.A.; Liles, D.R.

    1980-01-01

    TRAC is a best-estimate, multidimensional, nonequilibrium computer code intended for the analysis of loss-of-coolant accidents (LOCA's) in light water reactors. TRAC-PD2 is the third, detailed, pressurized water reactor version of the code. The TRAC code is modular both by components and by function. That is, vessels, pipes, pumps, etc. can be coupled together in any manner in order to simulate a reactor or a particular experimental facility. Individual physical phenomena are also coded in separate subroutines. This paper discusses the wall to fluid heat transfer coefficient correlations, the interfacial heat transfer models, and presents results for several experimental facilities

  19. Heat transfer with a split water channel

    International Nuclear Information System (INIS)

    Krinsky, S.

    1978-01-01

    The heat transfer problem associated with the incidence of synchrotron radiation upon a vacuum chamber wall cooled by a single water channel was previously studied, and a numerical solution to the potential problem was found using the two-dimensional magnet program POISSON. Calculations were extended to consider the case of a split water channel using POISSON to solve the potential problem for a given choice of parameters. By optimizing the dimensions, boiling of the water can be avoided. A copper chamber is a viable solution to the heat transfer problem at a beam port

  20. Natural Convective Heat Transfer from Narrow Plates

    CERN Document Server

    Oosthuizen, Patrick H

    2013-01-01

    Natural Convective Heat Transfer from Narrow Plates deals with a heat transfer situation that is of significant practical importance but which is not adequately dealt with in any existing textbooks or in any widely available review papers. The aim of the book is to introduce the reader to recent studies of natural convection from narrow plates including the effects of plate edge conditions, plate inclination, thermal conditions at the plate surface and interaction of the flows over adjacent plates. Both numerical and experimental studies are discussed and correlation equations based on the results of these studies are reviewed.

  1. Heat transfer applications for the practicing engineer

    CERN Document Server

    Theodore, Louis

    2011-01-01

    This book serves as a training tool for individuals in industry and academia involved with heat transfer applications. Although the literature is inundated with texts emphasizing theory and theoretical derivations, the goal of this book is to present the subject of heat transfer from a strictly pragmatic point of view. The book is divided into four Parts: Introduction, Principles, Equipment Design Procedures and Applications, and ABET-related Topics. The first Part provides a series of chapters concerned with introductory topics that are required when solving most engineering problems, inclu

  2. Microscale surface modifications for heat transfer enhancement.

    Science.gov (United States)

    Bostanci, Huseyin; Singh, Virendra; Kizito, John P; Rini, Daniel P; Seal, Sudipta; Chow, Louis C

    2013-10-09

    In this experimental study, two surface modification techniques were investigated for their effect on heat transfer enhancement. One of the methods employed the particle (grit) blasting to create microscale indentations, while the other used plasma spray coating to create microscale protrusions on Al 6061 (aluminum alloy 6061) samples. The test surfaces were characterized using scanning electron microscopy (SEM) and confocal scanning laser microscopy. Because of the surface modifications, the actual surface area was increased up to 2.8× compared to the projected base area, and the arithmetic mean roughness value (Ra) was determined to vary from 0.3 μm for the reference smooth surface to 19.5 μm for the modified surfaces. Selected samples with modified surfaces along with the reference smooth surface were then evaluated for their heat transfer performance in spray cooling tests. The cooling system had vapor-atomizing nozzles and used anhydrous ammonia as the coolant in order to achieve heat fluxes up to 500 W/cm(2) representing a thermal management setting for high power systems. Experimental results showed that the microscale surface modifications enhanced heat transfer coefficients up to 76% at 500 W/cm(2) compared to the smooth surface and demonstrated the benefits of these practical surface modification techniques to enhance two-phase heat transfer process.

  3. Turbulent Heat Transfer in Curved Pipe Flow

    Science.gov (United States)

    Kang, Changwoo; Yang, Kyung-Soo

    2013-11-01

    In the present investigation, turbulent heat transfer in fully-developed curved pipe flow with axially uniform wall heat flux has been numerically studied. The Reynolds numbers under consideration are Reτ = 210 (DNS) and 1,000 (LES) based on the mean friction velocity and the pipe radius, and the Prandtl number (Pr) is 0.71. For Reτ = 210 , the pipe curvature (κ) was fixed as 1/18.2, whereas three cases of κ (0.01, 0.05, 0.1) were computed in the case of Reτ = 1,000. The mean velocity, turbulent intensities and heat transfer rates obtained from the present calculations are in good agreement with the previous numerical and experimental results. To elucidate the secondary flow structures due to the pipe curvature, the mean quantities and rms fluctuations of the flow and temperature fields are presented on the pipe cross-sections, and compared with those of the straight pipe flow. To study turbulence structures and their influence on turbulent heat transfer, turbulence statistics including but not limited to skewness and flatness of velocity fluctuations, cross-correlation coefficients, an Octant analysis, and turbulence budgets are presented and discussed. Based on our results, we attempt to clarify the effects of Reynolds number and the pipe curvature on turbulent heat transfer. This research was supported by Basic Science Research Program through the National Research Foundation of Korea (NRF) funded by the Ministry of Education, Science and Technology (2010-0008457).

  4. Natural convective heat transfer from square cylinder

    Energy Technology Data Exchange (ETDEWEB)

    Novomestský, Marcel, E-mail: marcel.novomestsky@fstroj.uniza.sk; Smatanová, Helena, E-mail: helena.smatanova@fstroj.uniza.sk; Kapjor, Andrej, E-mail: andrej.kapjor@fstroj.uniza.sk [University of Žilina, Faculty of Mechanical Engineering, Department of Power Engineering, Univerzitná 1, 010 26 Žilina (Slovakia)

    2016-06-30

    This article is concerned with natural convective heat transfer from square cylinder mounted on a plane adiabatic base, the cylinders having an exposed cylinder surface according to different horizontal angle. The cylinder receives heat from a radiating heater which results in a buoyant flow. There are many industrial applications, including refrigeration, ventilation and the cooling of electrical components, for which the present study may be applicable.

  5. Heat and mass transfer in buildings

    International Nuclear Information System (INIS)

    Kristoffersen, Astrid Rusaas

    2005-01-01

    This thesis has presented four journal papers about ventilation and heat transfer in buildings. Ventilation and heat transfer in buildings are elements that decide our indoor air quality, thermal comfort and energy use in buildings. Models and experiments are tools to understand the complex physics of heat and air transfer in buildings. As computers are, getting cheaper and more powerful, there is a need to develop reliable models that can predict heat and air transfer in buildings. The first paper in this thesis addressed the widely used multizone model. This model is mainly used to find the airflows between zones in a building. A multizone model is often coupled to an energy analysis program, and affects therefore the calculated energy use in a building. The first paper in this thesis, titled ''Effect of room air recirculation delay on the decay rate of tracer gas concentration'' discussed the impact of a recirculating ventilation system on the decay of the tracer gas concentration in the room. The delay of the tracer gas through the ventilation system affects the concentration in the room, and must be accounted for when calculating the amount of fresh air that the ventilation system supplies. The second paper titled ''CFD Investigation of Room Ventilation for Improved Operation of a Downdraft Table: Novel Concepts'' investigated the performance of a downdraft table by changing the ventilation configuration in the room by use of Computational Fluid Dynamics (CFD). CFD can provide a microscopic description of the airflow and the behavior of pollutants and temperature distribution in a room. This paper calculated the airflow pattern in the room without influence of thermal effects, and demonstrated the usage of CFD. It was found that the total airflow could be reduced compared to an existing configuration (and hence reduce energy costs), and at the same time increasing the performance of the downdraft table (increasing the indoor air quality). A room with a

  6. 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)

  7. Heat Transfer in Directional Water Transport Fabrics

    Directory of Open Access Journals (Sweden)

    Chao Zeng

    2016-10-01

    Full Text Available Directional water transport fabrics can proactively transfer moisture from the body. They show great potential in making sportswear and summer clothing. While moisture transfer has been previously reported, heat transfer in directional water transport fabrics has been little reported in research literature. In this study, a directional water transport fabric was prepared using an electrospraying technique and its heat transfer properties under dry and wet states were evaluated, and compared with untreated control fabric and the one pre-treated with NaOH. All the fabric samples showed similar heat transfer features in the dry state, and the equilibrium temperature in the dry state was higher than for the wet state. Wetting considerably enhanced the thermal conductivity of the fabrics. Our studies indicate that directional water transport treatment assists in moving water toward one side of the fabric, but has little effect on thermal transfer performance. This study may be useful for development of “smart” textiles for various applications.

  8. Heat Flux Sensors for Infrared Thermography in Convective Heat Transfer

    Science.gov (United States)

    Carlomagno, Giovanni Maria; de Luca, Luigi; Cardone, Gennaro; Astarita, Tommaso

    2014-01-01

    This paper reviews the most dependable heat flux sensors, which can be used with InfraRed (IR) thermography to measure convective heat transfer coefficient distributions, and some of their applications performed by the authors' research group at the University of Naples Federico II. After recalling the basic principles that make IR thermography work, the various heat flux sensors to be used with it are presented and discussed, describing their capability to investigate complex thermo-fluid-dynamic flows. Several applications to streams, which range from natural convection to hypersonic flows, are also described. PMID:25386758

  9. Heat Flux Sensors for Infrared Thermography in Convective Heat Transfer

    Directory of Open Access Journals (Sweden)

    Giovanni Maria Carlomagno

    2014-11-01

    Full Text Available This paper reviews the most dependable heat flux sensors, which can be used with InfraRed (IR thermography to measure convective heat transfer coefficient distributions, and some of their applications performed by the authors’ research group at the University of Naples Federico II. After recalling the basic principles that make IR thermography work, the various heat flux sensors to be used with it are presented and discussed, describing their capability to investigate complex thermo-fluid-dynamic flows. Several applications to streams, which range from natural convection to hypersonic flows, are also described.

  10. Enhancement of heat transfer in HPLWR fuel assemblies

    International Nuclear Information System (INIS)

    Bastron, A.; Hofmeister, J.; Meyer, L.; Schulenberg, T.

    2005-01-01

    A study on different methods for enhancement of heat transfer in fuel assemblies for a High Performance Light Water Reactor has been performed to indicate the potential for a further increase of core outlet temperature at given cladding temperatures, or for reduction of peak cladding temperatures at the envisaged core outlet temperature. As a result, the introduction of an artificial surface roughness or the use of a staircase type grid spacer should increase the heat transfer coefficient of the coolant at the cladding surface by more than a factor of two, which will reduce the peak cladding temperature by at least 50 degC. The paper provides further details for realization of these measures. (author)

  11. Development of heat transfer package for core thermal-hydraulic design and analysis of upgraded JRR-3

    International Nuclear Information System (INIS)

    Sudo, Yukio; Ikawa, Hiromasa; Kaminaga, Masanori

    1985-01-01

    A heat transfer package was developed for the core thermal-hydraulic design and analysis of the Japan Research Reactor-3 (JRR-3) which is to be remodeled to a 20 MWt pool-type, light water-cooled reactor with 20 % low enriched uranium (LEU) plate-type fuel. This paper presents the constitution of the developed heat transfer package and the applicability of the heat transfer correlations adopted in it, based on the heat transfer experiments in which thermal-hydraulic features of the new JRR-3 core were properly reflected. (author)

  12. The problems of calculation of heat transfer crisis in fuel assemblies of PW reactors based on modern versions of thermohydraulic codes

    International Nuclear Information System (INIS)

    Fialko, N.M.; Sharaevskij, G.I.; Sharaevskaya, E.I.; Babak, E.I.

    2014-01-01

    The article gives an analysis of the adequacy of computer software systems FASCICLE BM-DF and COBRA, which are designed to calculate the main parameters of the safety of water-cooled nuclear reactors. This calculation is based on determining the local thermal-hydraulic parameters of the flow of coolant in the fuel rod assembled elements. In this article introduced the results of the comparison of experiments performed to determine the distribution of the main thermal-hydraulic flow parameters characteristic of subchannels of fuel rod assembled elements with the data for calculating these parameters on the basis of declared computer codes. Particular attention is paid to the analysis of experimental and calculation data, by definition, burnout in rod fuel assembled elements

  13. The log mean heat transfer rate method of heat exchanger considering the influence of heat radiation

    International Nuclear Information System (INIS)

    Wong, K.-L.; Ke, M.-T.; Ku, S.-S.

    2009-01-01

    The log mean temperature difference (LMTD) method is conventionally used to calculate the total heat transfer rate of heat exchangers. Because the heat radiation equation contains the 4th order exponential of temperature which is very complicate in calculations, thus LMTD method neglects the influence of heat radiation. From the recent investigation of a circular duct in some practical situations, it is found that even in the situation of the temperature difference between outer duct surface and surrounding is low to 1 deg. C, the heat radiation effect can not be ignored in the situations of lower ambient convective heat coefficient and greater surface emissivities. In this investigation, the log mean heat transfer rate (LMHTR) method which considering the influence of heat radiation, is developed to calculate the total heat transfer rate of heat exchangers.

  14. Characteristics of convective heat transport in a packed pebble-bed reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abdulmohsin, Rahman S., E-mail: rsar62@mst.edu [Department of Chemical and Biochemical Engineering, Missouri University of Science and Technology, 400 West 11th Street/231 Schrenk Hall, Rolla, MO 65409-1230 (United States); Al-Dahhan, Muthanna H., E-mail: aldahhanm@mst.edu [Department of Chemical and Biochemical Engineering, Missouri University of Science and Technology, 400 West 11th Street/231 Schrenk Hall, Rolla, MO 65409-1230 (United States); Department of Nuclear Engineering, 301 W. 14th St./222 Fulton Hall (United States)

    2015-04-01

    Highlights: • A fast-response heat transfer probe has been developed and used in this work. • Heat transport has been quantified in terms of local heat transfer coefficients. • The method of the electrically heated single sphere in packing has been applied. • The heat transfer coefficient increases from the center to the wall of packed bed. • This work advancing the knowledge of heat transport in the studied packed bed. - Abstract: Obtaining more precise results and a better understanding of the heat transport mechanism in the dynamic core of packed pebble-bed reactors is needed because this mechanism poses extreme challenges to the reliable design and efficient operation of these reactors. This mechanism can be quantified in terms of a solid-to-gas convective heat transfer coefficient. Therefore, in this work, the local convective heat transfer coefficients and their radial profiles were measured experimentally in a separate effect pilot-plant scale and cold-flow experimental setup of 0.3 m in diameter, using a sophisticated noninvasive heat transfer probe of spherical type. The effect of gas velocity on the heat transfer coefficient was investigated over a wide range of Reynolds numbers of practical importance. The experimental investigations of this work include various radial locations along the height of the bed. It was found that an increase in coolant gas flow velocity causes an increase in the heat transfer coefficient and that effect of the gas flow rate varies from laminar to turbulent flow regimes at all radial positions of the studied packed pebble-bed reactor. The results show that the local heat transfer coefficient increases from the bed center to the wall due to the change in the bed structure, and hence, in the flow pattern of the coolant gas. The findings clearly indicate that one value of an overall heat transfer coefficient cannot represent the local heat transfer coefficients within the bed; therefore, correlations are needed to

  15. Fuel assembly transfer and storage system for nuclear reactors

    International Nuclear Information System (INIS)

    Allain, Albert; Thomas, Claude.

    1981-01-01

    Transfer and storage system on a site comprising several reactors and at least one building housing the installations common to all these reactors. The system includes: transfer and storage modules for the fuel assemblies comprising a containment capable of containing several assemblies carried on a transport vehicle, a set of tracks for the modules between the reactors and the common installations, handling facilities associated with each reactor for moving the irradiated assemblies from the reactor to a transfer module placed in loading position on a track serving the reactor and conversely to move the new assemblies from the transfer module to the reactor, and at least one handling facility located in the common installation building for loading the modules with new assemblies [fr

  16. Cooperative heat transfer and ground coupled storage system

    Science.gov (United States)

    Metz, P.D.

    A cooperative heat transfer and ground coupled storage system wherein collected solar heat energy is ground stored and permitted to radiate into the adjacent ground for storage therein over an extended period of time when such heat energy is seasonally maximally available. Thereafter, when said heat energy is seasonally minimally available and has propagated through the adjacent ground a substantial distance, the stored heat energy may be retrieved by a circumferentially arranged heat transfer means having a high rate of heat transfer.

  17. Numerical simulation of heat transfer process in automotive brakes

    OpenAIRE

    Gonzalo Voltas, David

    2013-01-01

    This master thesis concerns the theoretical investigations of the heat transfer process in automotive brakes. The process of heat generation and heat transfer to ambient air in automotive brake was presented. The two–dimensional, axi-symmetrical model of transient heat conduction for the brake was applied. The relevant boundary conditions, that describe the heat generated in the brake and the heat transferred to ambient air, were used. The unsteady heat conduction problem was solved by the...

  18. 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

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

    International Nuclear Information System (INIS)

    Boucher, T.J.

    1986-01-01

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

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

    International Nuclear Information System (INIS)

    Boucher, T.J.

    1987-01-01

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

  1. Heat Transfer in a Paper Cup

    Science.gov (United States)

    Ribeiro, Carla

    2017-01-01

    The double-wall paper cup is an everyday object that can be used in the laboratory to study heat transfer. The experiment described here has been done by physics students aged 12-13 years; it can also be used in a different context to prompt debate about environmental issues.

  2. Free convection film flows and heat transfer

    CERN Document Server

    Shang, Deyi

    2010-01-01

    Presents development of systematic studies for hydrodynamics and heat and mass transfer in laminar free convection, accelerating film boiling and condensation of Newtonian fluids, and accelerating film flow of non-Newtonian power-law fluids. This book provides a system of analysis models with a developed velocity component method.

  3. Conjugate problems in convective heat transfer

    CERN Document Server

    Dorfman, Abram S

    2009-01-01

    The conjugate heat transfer (CHT) problem takes into account the thermal interaction between a body and fluid flowing over or through it, a key consideration in both mechanical and aerospace engineering. Presenting more than 100 solutions of non-isothermal and CHT problems, this title considers the approximate solutions of CHT problems.

  4. CENTRIFUGAL COMPRESSOR EFFICIENCY CALCULATION WITH HEAT TRANSFER

    Directory of Open Access Journals (Sweden)

    Valeriu Dragan

    2017-12-01

    and manner under which the efficiency itself is calculated. The paper  presents a more robust approach to measuring efficiency, regardless of the heat transfer within the turbomachinery itself. Possible applications of the study may range from cold-start regime simulation to the optimization of inter-cooling setup or even flow angle control without mechanically actuated OGV

  5. A five MW nuclear heating reactor

    International Nuclear Information System (INIS)

    Zhang Dafang; Don Duo; Su Quingshan

    1997-01-01

    The 5 MW Nuclear Heating Reactor (NHR-5) developed and designed by the Institute of Nuclear Energy and Technology (INET) and has been operated for four winter seasons since 1989. During the time of commissioning and operation a number of experiments including self-stability, self-regulation and simulation of ATWS etc. were carried out. Some operating experiences such as water chemistry, radiation protection, and environmental impacts and so on, were also obtained at the same time. All of these demonstrate that the design of NHR-5 is successful. (author)

  6. Evaporative Heat Transfer Mechanisms within a Heat Melt Compactor

    Science.gov (United States)

    Golliher, Eric L.; Gotti, Daniel J.; Rymut, Joseph Edward; Nguyen, Brian K; Owens, Jay C.; Pace, Gregory S.; Fisher, John W.; Hong, Andrew E.

    2013-01-01

    This paper will discuss the status of microgravity analysis and testing for the development of a Heat Melt Compactor (HMC). Since fluids behave completely differently in microgravity, the evaporation process for the HMC is expected to be different than in 1-g. A thermal model is developed to support the design and operation of the HMC. Also, low-gravity aircraft flight data is described to assess the point at which water may be squeezed out of the HMC during microgravity operation. For optimum heat transfer operation of the HMC, the compaction process should stop prior to any water exiting the HMC, but nevertheless seek to compact as much as possible to cause high heat transfer and therefore shorter evaporation times.

  7. Effects of ocean conditions upon the passive residual heat removal system (PRHRS) of ship reactor

    International Nuclear Information System (INIS)

    Su Guanghui; Zhang Jinling; Guo Yujun; Qiu Suizheng; Yu Zhenwan; Jia Dounan

    1996-01-01

    The authors investigate the influence of ocean conditions (heaving, listing, rolling) on the natural circulation flow and the ability of heat transfer of the ship reactor's PRHRS, and develops a mathematical model. A program, MISAP 02, is compiled with the structured FORTRAN 77 using the advanced Gear method. the program is used to calculate the above influence. The results show that the ocean conditions have some effects on the natural circulation flow and the ability of heat transfer

  8. Average Natural Convective Heat Transfer of Air-cooled Condensing Heat Exchanger of Emergency Cooldown Tank - Effect of Tube Banks

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Seon Jeong; Lee, Hee Joon [Kookmin University, Seoul (Korea, Republic of); Kim, Myoung Jun; Moon, Joo Hyung; Bae, Youngmin; Kim, Young-In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Recently emergency cooldown tank(ECT) is a great concern of passive cooling system for the safety of nuclear reactor. After the operation of a conventional passive cooling system for an extended period, however, the water level falls as a result of the evaporation from the ECT, as steam is emitted from the open top of the tank. In this study, the effect of heat transfer area at the air cooled condensing heat exchanger was investigated by changing 5×5 tube banks into 4×4 and 3×3. Moreover, each of air-side natural convective heat transfer coefficient of tube banks was compared to existing correlations. This study presents the effect of heat transfer area at air-cooled condensing heat exchanger. As heat transfer area decreased, the temperature of outlet increased. In other words, the cooling performance got lower with the decrease of heat transfer area. In addition, the average natural convective heat transfer coefficient was 15.3 W/m{sup 2}/K from the 4×4 tube banks, and 4.92 W/m{sup 2}/K from the 3×3 tube banks, which had quite a large error more than 46% especially with the value of 4×4 tube banks compared to the value from correlation equation. Therefore, according to this result, it is needed to measure the local heat transfer coefficient of vertical cylinder more elaborately in further study.

  9. Average Natural Convective Heat Transfer of Air-cooled Condensing Heat Exchanger of Emergency Cooldown Tank - Effect of Tube Banks

    International Nuclear Information System (INIS)

    Huh, Seon Jeong; Lee, Hee Joon; Kim, Myoung Jun; Moon, Joo Hyung; Bae, Youngmin; Kim, Young-In

    2016-01-01

    Recently emergency cooldown tank(ECT) is a great concern of passive cooling system for the safety of nuclear reactor. After the operation of a conventional passive cooling system for an extended period, however, the water level falls as a result of the evaporation from the ECT, as steam is emitted from the open top of the tank. In this study, the effect of heat transfer area at the air cooled condensing heat exchanger was investigated by changing 5×5 tube banks into 4×4 and 3×3. Moreover, each of air-side natural convective heat transfer coefficient of tube banks was compared to existing correlations. This study presents the effect of heat transfer area at air-cooled condensing heat exchanger. As heat transfer area decreased, the temperature of outlet increased. In other words, the cooling performance got lower with the decrease of heat transfer area. In addition, the average natural convective heat transfer coefficient was 15.3 W/m"2/K from the 4×4 tube banks, and 4.92 W/m"2/K from the 3×3 tube banks, which had quite a large error more than 46% especially with the value of 4×4 tube banks compared to the value from correlation equation. Therefore, according to this result, it is needed to measure the local heat transfer coefficient of vertical cylinder more elaborately in further study

  10. Liquid-metal-gas heat exchanger for HTGR type reactors

    International Nuclear Information System (INIS)

    Werth, G.

    1980-01-01

    The aim of this study is to investigate the heat transfer characteristics of a liquid metal heat exchanger (HE) for a helium-cooled high temperature reactor. A tube-type heat exchanger is considered as well as two direct exchangers: a bubble-type heat exchanger and a heat exchanger according to the spray principle. Experiments are made in order to determine the gas content of bubble-type heat exchangers, the dependence of the droplet diameter on the nozzle diameter, the falling speed of the droplets, the velocity of the liquid jet, and the temperature variation of liquid jets. The computer codes developed for HE calculation are structured so that they may be used for gas/liquid HE, too. Each type of HE that is dealt with is designed by accousting for a technical and an economic assessment. The liquid-lead jet spray is preferred to all other types because of its small space occupied and its simple design. It shall be used in near future in the HTR by the name of lead/helium HE. (GL) [de

  11. 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

  12. Wall heat transfer coefficient in a molten salt bubble column: testing the experimental setup

    CSIR Research Space (South Africa)

    Skosana, PJ

    2014-10-01

    Full Text Available reactors that are highly exothermic or endothermic. This paper presents the design and operation of experimental setup used for measurement of the heat transfer coefficient in molten salt media. The experimental setup was operated with tap water, heat...

  13. 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)

  14. Joule-Heated Molten Regolith Electrolysis Reactor Concepts for Oxygen and Metals Production on the Moon and Mars

    Science.gov (United States)

    Sibille, Laurent; Dominques, Jesus A.

    2012-01-01

    The maturation of Molten Regolith Electrolysis (MRE) as a viable technology for oxygen and metals production on explored planets relies on the realization of the self-heating mode for the reactor. Joule heat generated during regolith electrolysis creates thermal energy that should be able to maintain the molten phase (similar to electrolytic Hall-Heroult process for aluminum production). Self-heating via Joule heating offers many advantages: (1) The regolith itself is the crucible material, it protects the vessel walls (2) Simplifies the engineering of the reactor (3) Reduces power consumption (no external heating) (4) Extends the longevity of the reactor. Predictive modeling is a tool chosen to perform dimensional analysis of a self-heating reactor: (1) Multiphysics modeling (COMSOL) was selected for Joule heat generation and heat transfer (2) Objective is to identify critical dimensions for first reactor prototype.

  15. Control of reactor coolant flow path during reactor decay heat removal

    Science.gov (United States)

    Hunsbedt, Anstein N.

    1988-01-01

    An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

  16. 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.

  17. On the thermoeconomics of heat transfer

    International Nuclear Information System (INIS)

    El-Sayed, Y.M.

    1991-01-01

    The cost effectiveness of improving the thermodynamics of heat transfer in an energy system is investigated by considering steam power systems bottoming a given gas turbine. Higher efficiencies are basically achieved by improving the temperature match of the heat addition process using both structural and parametric modes of change. The heat transfer surfaces, when expressed solely in terms of efficiency, indicate the existence of an envelope bounding them. The envelope can be approximated by a simple continuous function. Minimum surface for a given efficiency is on or closest to the envelope. Similar features apply to capital cost and to the cost objective function. In this paper the generalization and the limitations of the envelopment concept are discussed as well as the relevance to artificial intelligence

  18. 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

  19. Open Channel Natural Convection Heat Transfer on a Vertical Finned Plate

    International Nuclear Information System (INIS)

    Park, Joo Hyun; Heo, Jeong Hwan; Chung, Bum Jin

    2013-01-01

    The natural convection heat transfer of vertical plate fin was investigated experimentally. Heat transfer systems were replaced by mass-transfer systems, based on the analogy concept. The experimental results lie within the predictions of the existing heat transfer correlations of plate-fin for the natural convections. An overlapped thermal boundary layers caused increasing heat transfer, and an overlapped momentum boundary layers caused decreasing heat transfer. As the fin height increases, heat transfer was enhanced due to increased inflow from the open side of the fin spacing. When fin spacing and fin height are large, heat transfer was unaffected by the fin spacing and fin height. Passive cooling by natural convection becomes more and more important for the nuclear systems as the station black out really happened at the Fukushima NPPs. In the RCCS (Reactor Cavity Cooling System) of a VHTR (Very High Temperature Reactor), natural convection cooling through duct sys