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

Study of condensation heat transfer following a main steam line break inside containment

Energy Technology Data Exchange (ETDEWEB)

Cho, J.H.; Elia, F.A. Jr.; Lischer, D.J. [Stone & Webster Engineering Corporation, Boston, MA (United States)

1995-09-01

An alternative model for calculating condensation heat transfer following a main stream line break (MSLB) accident is proposed. The proposed model predictions and the current regulatory model predictions are compared to the results of the Carolinas Virginia Tube Reactor (CVTR) test. The very conservative results predicted by the current regulatory model result from: (1) low estimate of the condensation heat transfer coefficient by the Uchida correlation and (2) neglecting the convective contribution to the overall heat transfer. Neglecting the convection overestimates the mass of steam being condensed and does not permit the calculation of additional convective heat transfer resulting from superheated conditions. In this study, the Uchida correlation is used, but correction factors for the effects of convection an superheat are derived. The proposed model uses heat and mass transfer analogy methods to estimate to convective fraction of the total heat transfer and bases the steam removal rate on the condensation heat transfer portion only. The results predicted by the proposed model are shown to be conservative and more accurate than those predicted by the current regulatory model when compared with the results of the CVTR test. Results for typical pressurized water reactors indicate that the proposed model provides a basis for lowering the equipment qualification temperature envelope, particularly at later times following the accident.

Survey of Thermal-Fluids Evaluation and Confirmatory Experimental Validation Requirements of Accident Tolerant Cladding Concepts with Focus on Boiling Heat Transfer Characteristics

Energy Technology Data Exchange (ETDEWEB)

Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Ali, Amir [Univ. of New Mexico, Albuquerque, NM (United States); Liu, Maolong [Univ. of New Mexico, Albuquerque, NM (United States); Blandford, Edward [Univ. of New Mexico, Albuquerque, NM (United States)

2016-06-01

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE) Advanced Fuels Campaign (AFC) is working closely with the nuclear industry to develop fuel and cladding candidates with potentially enhanced accident tolerance, also known as accident tolerant fuel (ATF). Thermal-fluids characteristics are a vital element of a holistic engineering evaluation of ATF concepts. One vital characteristic related to boiling heat transfer is the critical heat flux (CHF). CHF plays a vital role in determining safety margins during normal operation and also in the progression of potential transient or accident scenarios. This deliverable is a scoping survey of thermal-fluids evaluation and confirmatory experimental validation requirements of accident tolerant cladding concepts with a focus on boiling heat transfer characteristics. The key takeaway messages of this report are: 1. CHF prediction accuracy is important and the correlations may have significant uncertainty. 2. Surface conditions are important factors for CHF, primarily the wettability that is characterized by contact angle. Smaller contact angle indicates greater wettability, which increases the CHF. Surface roughness also impacts wettability. Results in the literature for pool boiling experiments indicate changes in CHF by up to 60% for several ATF cladding candidates. 3. The measured wettability of FeCrAl (i.e., contact angle and roughness) indicates that CHF should be investigated further through pool boiling and flow boiling experiments. 4. Initial measurements of static advancing contact angle and surface roughness indicate that FeCrAl is expected to have a higher CHF than Zircaloy. The measured contact angle of different FeCrAl alloy samples depends on oxide layer thickness and composition. The static advancing contact angle tends to decrease as the oxide layer thickness increases.

Natural convection heat transfer experiments of horizontal plates with fin arrays

Energy Technology Data Exchange (ETDEWEB)

Moon, Je Young; Chung, Bum Jin [Jeju National University 102 Jejudaehakno, Jeju (Korea, Republic of)

2012-10-15

Core melt in a severe accident condition, forms a molten pool in the reactor vessel lower head. The molten pool is divided by a metallic pool (top) and an oxide pool (bottom) by the density difference. The crust between the metallic layer and the oxide pool may be formed by solidification of the molten metallic materials. So the surface of the crust is formed irregularly. Experiments were performed to investigate the irregular crust as a preparatory study before an in-depth severe accident study. The natural convection heat transfer were investigated experimentally varying the height and spacing of fins, top plate of different kinds and the plate separation distance with/without the side walls. In order to simulate irregular crust surface condition, the finned plates was used. Using the analogy concept, heat transfer experiments were replaced by mass transfer experiments. A cupric acid.copper sulfate (H{sup 2S}O{sup 4-}CuSO{sup 4)} electroplating system was adopted as the mass transfer system and the electric currents were measured rather than the heat transfer rates.

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

International Nuclear Information System (INIS)

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

1995-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1995-09-01

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

Experimental analysis of heat transfer within the AP600 containment under postulated accident conditions

International Nuclear Information System (INIS)

Anderson, M.H.; Corradini, M.L.

1998-01-01

The new AP600 reactor designed by Westinghouse uses a passive safety system relying on heat removal by condensation to keep the containment within the design limits of pressure and temperature. Even though some research has been done so far in this regard, there are some uncertainties concerning the behavior of the system under postulated accident conditions. In this paper, steam condensation onto the internal surfaces of the AP600 containment walls has been investigated in two scaled vessels with similar aspect ratios to the actual AP600. The heat transfer degradation in the presence of noncondensable gas has been analyzed for different noncondensable mixtures of air and helium (hydrogen simulant). Molar fractions of noncondensables/steam ranged from (0.4-4.0) and helium concentrations in the noncondensable mixture were 0-50% by volume. In addition, the effects of the bulk temperatures, the mass fraction of noncondensable/steam, the cold wall surface temperature, the pressure, noncondensable composition, and the inclination of the condensing surface were studied. It was found that the heat transfer coefficients ranged from 50 to 800 J s -1 K -1 m -2 with the highest for high wall temperatures at high pressure and low noncondensable molar fractions. The effect of a light gas (helium) in the noncondensable mixture were found to be negligible for concentrations less than approximately 35 molar percent but could result in stratification at higher concentrations. The complete study gives a large and relatively complete data base on condensation within a scaled AP600 containment structure, providing an invaluable set of data against which to validate models. In addition, specific areas requiring further investigation are summarized. (orig.)

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

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

International Nuclear Information System (INIS)

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

2011-01-01

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

A study on the correlations development for film boiling heat transfer on spheres

Energy Technology Data Exchange (ETDEWEB)

Jeong, Yong Hoon; Baek, Won Pil; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1999-12-31

Film boiling is the heat transfer mechanism that can occurs when large temperature differences exist between a cold liquid and hot material. In the nuclear reactor safety analysis, film boiling has become an important issue in recent years. During severe accident, hot molten corium fall into relatively cool water, and fragment into spheres or sphere-like particles. If the steam explosion is triggered, the thermal energy of corlium is converted into the mechanical energy that can threaten the integrity of reactor vessel or reactor cavity. One of the important concerns in the heat transfer analysis during pre-mixing stage is the film boiling heat transfer between the corium and water/steam two-phase flow. Until now, considerable works on film boiling have been performed. However, there is no available correlation adequate for severe accident analysis. In this study, film boiling heat transfer correlations have been developed, and their applicable ranges have been enlarged and their prediction accuracy has been enhanced. 7 refs., 5 figs., 5 tabs. (Author)

A study on the correlations development for film boiling heat transfer on spheres

International Nuclear Information System (INIS)

Jeong, Yong Hoon; Baek, Won Pil; Chang, Soon Heung

1998-01-01

Film boiling is the heat transfer mechanism that can occurs when large temperature differences exist between a cold liquid and hot material. In the nuclear reactor safety analysis, film boiling has become an important issue in recent years. During severe accident, hot molten corium fall into relatively cool water, and fragment into spheres or sphere-like particles. If the steam explosion is triggered, the thermal energy of corium is converted into the mechanical energy that can threaten the integrity of reactor vessel or reactor cavity. One of the important concerns in the heat transfer analysis during pre-mixing stage is the film boiling heat transfer between the corium and water/steam two-phase flow. Until now, considerable works on film boiling have been performed. However, there is no available correlation adequate for severe accident analysis. In this study, film boiling heat transfer correlations have been developed, and their applicable ranges have been enlarged and their prediction accuracy has been enhanced

A study on the correlations development for film boiling heat transfer on spheres

Energy Technology Data Exchange (ETDEWEB)

Jeong, Yong Hoon; Baek, Won Pil; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1998-12-31

Film boiling is the heat transfer mechanism that can occurs when large temperature differences exist between a cold liquid and hot material. In the nuclear reactor safety analysis, film boiling has become an important issue in recent years. During severe accident, hot molten corium fall into relatively cool water, and fragment into spheres or sphere-like particles. If the steam explosion is triggered, the thermal energy of corlium is converted into the mechanical energy that can threaten the integrity of reactor vessel or reactor cavity. One of the important concerns in the heat transfer analysis during pre-mixing stage is the film boiling heat transfer between the corium and water/steam two-phase flow. Until now, considerable works on film boiling have been performed. However, there is no available correlation adequate for severe accident analysis. In this study, film boiling heat transfer correlations have been developed, and their applicable ranges have been enlarged and their prediction accuracy has been enhanced. 7 refs., 5 figs., 5 tabs. (Author)

Accident analysis of heat pipe cooled and AMTEC conversion space reactor system

International Nuclear Information System (INIS)

Yuan, Yuan; Shan, Jianqiang; Zhang, Bin; Gou, Junli; Bo, Zhang; Lu, Tianyu; Ge, Li; Yang, Zijiang

2016-01-01

Highlights: • A transient analysis code TAPIRS for HPS has been developed. • Three typical accidents are analyzed using TAPIRS. • The reactor system has the self-stabilization ability under accident conditions. - Abstract: A space power with high power density, light weight, low cost and high reliability is of crucial importance to future exploration of deep space. Space reactor is an excellent candidate because of its unique characteristics of high specific power, low cost, strong environment adaptability and so on. Among all types of space reactors, heat pipe cooled space reactor, which adopts the passive heat pipe (HP) as core cooling component, is considered as one of the most promising choices and is widely studied all over the world. This paper develops a transient analysis code (TAPIRS) for heat pipe cooled space reactor power system (HPS) based on point reactor kinetics model, lumped parameter core heat transfer model, combined HP model (self-diffusion model, flat-front startup model and network model), energy conversion model of Alkali Metal Thermal-to-Electric Conversion units (AMTEC), and HP radiator model. Three typical accidents, i.e., control drum failure, AMTEC failure and partial loss of the heat transfer area of radiator are then analyzed using TAPIRS. By comparing the simulation results of the models and steady state with those in the references, the rationality of the models and the solution method is validated. The results show the following. (1) After the failure of one set of control drums, the reactor power finally reaches a stable value after two local peaks under the temperature feedback. The fuel temperature rises rapidly, however it is still under safe limit. (2) The fuel temperature is below a safe limit under the AMTEC failure and partial loss of the heat transfer area of radiator. This demonstrates the rationality of the system design and the potential applicability of the TAPIRS code for the future engineering application of

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

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

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

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.)

Investigation and assessment of wall heat transfer correlations in SPACE code

International Nuclear Information System (INIS)

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

2010-06-01

SPACE, which is a safety analysis code for nuclear power plants, has been developed to analyze the multidimensional, two-component and three-field flow. This code can be applied to safety analysis for approval which is thermal-hydraulic analysis to support the nuclear power station design, establishment of accident ease strategy, development of operating guide line, experiment plan and analysis. To do so, SPACE code has 12 wall heat transfer mode and the corresponding models and correlations to deal with the physical heat transfer phenomenon in wall surface. In this report, the physical correlation models regarding the wall heat transfer are explained and their performance is assessed against several SET

An Experimental Study on the Heat Focusing of the Metallic layer in a Severe Accident

Energy Technology Data Exchange (ETDEWEB)

Moon, Je-Young; Chung, Bum-Jin [Kyung Hee Univ., Yongin (Korea, Republic of)

2014-10-15

By flooding the reactor cavity that surrounds the vessel, significant energy can be removed from relocated corium materials through the vessel wall. The aim of this study is to investigate the heat focusing depending on the aspect ratios and heat transfer characteristic of upper boundary for applications related to severe accident phenomena. Experiments were carried out for Rayleigh numbers and aspect ratio in the range of 8.49x10{sup 7}-5.49x10{sup 9}, 0.128-0.512 respectively. Also, the conditions of the top wall and the side wall are considered: (a) top plate cooling, side wall adiabatic, (b) top plate adiabatic, side wall cooling, (c) both walls cooling. In order to achieve high Rayleigh numbers, the heat transfer experiments were replaced by mass transfer experiments based on heat and mass transfer analogy concept. A sulfuric acid-copper sulfate (H{sub 2}SO{sub 4}-CuSO{sub 4}) electroplating system was adopted as the mass transfer system.An experimental study was performed to investigate the focusing effect appeared in the metallic layer in a severe accident condition. Mass transfer experiments, based on the analogy concept, carried out in order to achieve high Rayleigh number. The height of the side wall was varied for three different cooling conditions: top only, side only, and both top and side. The experimental results agreed well with the Rayleigh-Benard convection correlations of Dropkin and Somerscales and Globe and Dropkin. The test results for the three configurations (only top cooling, only side wall cooling and both walls cooling) presented that the heat transfer on side wall cooling condition without top cooling is biggest. Also, the heat transfer was enhanced by decreasing the aspect ratio (H/R)

A study on transient heat transfer of the EU-ABWR external core catcher using the phase-change effective convectivity model

International Nuclear Information System (INIS)

Tran Chi Thanh; Nguyen Viet Hung; Tahara, Mika; Kojima, Yoshihiro; Hamazaki, Ryoichi; Kudinov, Pavel

2015-01-01

In advanced designs of Nuclear Power Plants (NPPs), for mitigation of severe accident consequences, on the one hand, the In-Vessel Retention (IVR) concept has been implemented. On the other hand in other new NPP designs (Generation III and III+) with large power reactors, the External Core Catcher (ECC) has been widely adopted. Assessment of ECC design robustness is largely based on analysis of heat transfer of a melt pool formed in the ECC. Transient heat transfer analysis of an ECC is challenging due to (i) uncertainty in the in-vessel accident progression and subsequent vessel failure modes; (ii) long transient, (iii) high Rayleigh number and complex flows involving phase change of the melt pool formed in an ECC. The present paper is concerned with analysis of transient melt pool heat transfer in the ECC of new Advanced Boiling Water Reactor (ABWR) designed by Toshiba Corporation (Japan). According to the ABWR severe accident management strategy, the ECC is initially dry. In order to prevent steam explosion flooding is initiated after termination of melt relocation from the vessel. The ECC full of melt is cooled from the top directly by water and from the bottom through the ECC walls. In order to assess sustainability of the ECC, heat transfer simulation of a stratified melt pool formed in the ECC is carried out. The problem addressed in this work is heat flux distribution at ECC boundaries when cooling is applied (i) from the bottom, (ii) from the top and from the bottom. To perform melt pool heat transfer simulation, we employ Phase-change Effective Convectivity Model (PECM) which was originally developed as a computationally efficient, sufficiently accurate, 2D/3D accident analysis tools for simulation of transient melt pool heat transfer in the reactor lower plenum. Thermal loads from the melt pool to ECC boundaries are determined for selected ex-vessel accident scenarios. Performance of the ECC, efficiency of severe accident management (SAM) measures and

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.)

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

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.

Report from the Special Committee on Fukushima Nuclear Accident

International Nuclear Information System (INIS)

Ozawa, Mamoru

2012-01-01

The Special Committee on Fukushima Nuclear Accident was established in April 2011 under the Heat Transfer Society of Japan (HTSJ) and discussed (1) how had evolved heat transfer research in progress of nuclear technology, (2) role of expert group in the area of heat transfer academy and technology and (3) energy prospect in Japan after the Fukushima nuclear accident. This report was described by the chairman of the special committee summarizing one year discussions as (1) background of heat transfer research progress, (2) progression of Fukushima Daiichi Nuclear Power Plant accident, (3) energy problem in Japan after the Fukushima accident and (4) social role of the HTSJ. This HTSJ was a unique, nonprofit association in Japan of the people engaged in heat transfers research or in various engineering aspects related to heat transfer, which meant interdisciplinary or common platform of heat transfer as elementary technologies. Such actual complex problems could be discussed in the HTSJ from an overlooking viewpoint in order for the HTSJ to play a social role. (T. Tanaka)

Numerical Investigation of Turbulent Natural Convection Heat Transfer in an Internally-Heated Melt Pool and Metallic Layer

International Nuclear Information System (INIS)

Nourgaliev, R.R.; Dinh, A.T.; Dinh, T.N.; Sehgal, B.R.

1999-01-01

This paper presents results of numerical investigation of turbulent natural convection in an internally-heated oxidic pool, and in a metallic layer heated from below and cooled from top and sidewalls. Emphasis is placed upon applicability of the existing heat transfer correlations (obtained from simulant-material experiments) in assessments of a prototypic severe reactor accident. The objectives of this study are (i) to improve the current understanding of the physics of unstably stratified flows, and (ii) to reduce uncertainties associated with modeling and assessment of natural convection heat transfer in the above configuration. Prediction capabilities of different turbulence modeling approaches are first examined and discussed, based on extensive results of numerical investigations performed by present authors. Findings from numerical modeling of turbulent natural convection flow and heat transfer in melt pools and metallic layers are then described. (authors)

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

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

Radiation heat transfer model for the SCDAP code

International Nuclear Information System (INIS)

Sohal, M.S.

1984-01-01

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

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.

Radiation heat transfer model in a spent fuel pool by TRACE code

International Nuclear Information System (INIS)

Sanchez-Saez, F.; Carlos, S.; Villanueva, J.F.; Martorell, S.

2014-01-01

Nuclear policies have experienced an important change since Fukushima Daiichi nuclear plant accident and the safety of spent fuels has been in the spot issue among all the safety concerns. The work presented consists of the thermohydraulic simulation of spent fuel pool behavior after a loss of coolant throughout transfer channel with loss of cooling transient is produced. The simulation is done with the TRACE code. One of the most important variables that define the behavior of the pool is cladding temperature, which evolution depends on the heat emission. In this work convection and radiation heat transfer is considered. When both heat transfer models are considered, a clear delay in achieving the maximum peak cladding temperature (1477 K) is observed compared with the simulation in which only convection heat transfer is considered. (authors)

Analysis of heat transfer mechanism on in-vessel corium coolability in severe accidents

International Nuclear Information System (INIS)

Park, Rae Joon; Jeong, Ji Whan; Kim, Sang Baik; Kang, Kyung Ho; Kim, Jong Whan

1998-04-01

When the molten core material relocates to the lower plenum of the reactor vessel, the cooling process of corium and the related heat transfer mechanism have been analyzed. The critical heat flux in gap (CHFG) test is being performed as a part of simulation of naturally arrested thermal attack in (SONATA-IV) project and the state of art on CHF has been reviewed. A series of complex heat transfer mechanism of molten pool formation, natural convection in the molten pool, solidification and remelting of the corium, conduction in the solidified crust, and boiling heat transfer to surroundings can be occurred in the lower plenum. Many studies are needed to investigate the complex heat transfer mechanism in the lower plenum, because these phenomena have not been clearly understand until now. The SONATA-IV/CHFG experiments are being carried out to develop CHF correlation in a hemispherical gap, which is the upper limit of heat transfer. There is no experimental or analytical CHF correlation applicable to a hemispherical gap. So lots of analytical and experimental correlations developed using the similar experimental condition were gathered and compared with each other. According to the experimental work that was carried out with pool boiling condition, CHF in a parallel gap was reduced by 1/30 compared with the value measured without gap. A basic form of a CHF correlation has been developed to correlate measurements that will be made in the SONATA-IV/CHFG experiments. That correlation is based on the fact that the CHF in a hemispherical gap is enhanced by CCFL and a Kutateladze type CCFL correlation develops CCFL date will in geometry like this. The experimental facility consists of a heater, a pressure vessel, a heat exchanger and lots of sensors. The heater capacity is 40 kw and the maximum heat flux at the surface is 100 kw/m 2 . The experiments will be carried out in the range of 1 to 10 atm and the gap size of 0.5, 1, 2 mm. The CHF will be detected using 66 type

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.

Application of advanced model of radiative heat transfer in a rod geometry to QUENCH and PARAMETER tests

International Nuclear Information System (INIS)

Vasiliev, A.D.; Kobelev, G.V.; Astafieva, V.O.

2007-01-01

Radiative heat transfer is very important in different fields of mechanical engineering and related technologies including nuclear reactors, heat transfer in furnaces, aerospace, different high-temperature assemblies. In particular, in the course of a hypothetical severe accident at PWR-type nuclear reactor the temperatures inside the reactor vessel reach high values at which taking into account of radiative heat exchange between the structures of reactor (including core and other reactor vessel elements) gets important. Radiative heat transfer dominates the late phase of severe accident because radiative heat fluxes (proportional to T4, where T is the temperature) are generally considerably higher than convective and conductive heat fluxes in a system. In particular, heat transfer due to radiation determines the heating and degradation of the core and surrounding steel in-vessel structures and finally influences the composition, temperature and mass of materials pouring out of the reactor vessel after its loss of integrity. Existing models of radiative heat exchange use many limitations and approximations: approximate estimation of view factors and beam lengths; the geometry change in the course of the accident is neglected; the database for emissivities of materials is not complete; absorption/emission by steam-noncondensable medium is taken into account approximately. The module MRAD was developed in this paper to model the radiative heat exchange in rod-like geometry typical of PWR-type reactor. Radiative heat exchange is computed using dividing on zones (zonal method) as in existing radiation models implemented to severe accident numerical codes such as ICARE, SCDAP/RELAP, MELCOR but improved in following aspects: new approach to evaluation of view factors and mean beam length; detailed evaluation of gas absorptivity and emissivity; account of effective radiative thermal conductivity for the large core; account of geometry modification in the course of severe

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)

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

International Nuclear Information System (INIS)

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

1976-12-01

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

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....

Identification of flow regimes and heat transfer modes in Angra-2 core during the simulation of the small break loss of coolant accident of 250 cm2 in the cold leg of primary loop using RELAP5 code

International Nuclear Information System (INIS)

Borges, Eduardo M.; Sabundjian, Gaiane

2017-01-01

The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2. gamma code in Angra-2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 250cm 2 of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of Angra-2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of Angra-2 during the postulated accident. The results obtained for Angra-2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core. (author)

Heat Transfer Analysis for a Fixed CST Column

International Nuclear Information System (INIS)

Lee, S.Y.

2004-01-01

In support of a small column ion exchange (SCIX) process for the Savannah River Site waste processing program, a transient two-dimensional heat transfer model that includes the conduction process neglecting the convection cooling mechanism inside the crystalline silicotitanate (CST) column has been constructed and heat transfer calculations made for the present design configurations. For this situation, a no process flow condition through the column was assumed as one of the reference conditions for the simulation of a loss-of-flow accident. A series of the modeling calculations has been performed using a computational heat transfer approach. Results for the baseline model indicate that transit times to reach 130 degrees Celsius maximum temperature of the CST-salt solution column are about 96 hours when the 20-in CST column with 300 Ci/liter heat generation source and 25 degrees Celsius initial column temperature is cooled by natural convection of external air as a primary heat transfer mechanism. The modeling results for the 28-in column equipped with water jacket systems on the external wall surface of the column and water coolant pipe at the center of the CST column demonstrate that the column loaded with 300 Ci/liter heat source can be maintained non-boiling indefinitely. Sensitivity calculations for several alternate column sizes, heat loads of the packed column, engineered cooling systems, and various ambient conditions at the exterior wall of the column have been performed under the reference conditions of the CST-salt solution to assess the impact of those parameters on the peak temperatures of the packed column for a given transient time. The results indicate that a water-coolant pipe at the center of the CST column filled with salt solution is the most effective one among the potential design parameters related to the thermal energy dissipation of decay heat load. It is noted that the cooling mechanism at the wall boundary of the column has significant

Study of heat and mass transfer phenomena in fuel assembly models under accident conditions

International Nuclear Information System (INIS)

Yefanov, A.D.; Kalyakin, C.G.; Loshchinin, V.M.; Pomet'ko, R.S.; Sergeev, V.V.; Shumsky, R.V.

1996-01-01

The majority of the material in support of the thermal - hydraulic safety of WWER core was obtained on single - assembly models containing a relatively small number of elements - heater rods. Upgrading the requirements to the reactor safety leads to the necessity for studying phenomena in channels representing the cross - sectional core dimensions and non - uniform radial power generation. Under such conditions, the contribution of natural convection can be significant in some core zones, including the occurrence of reverse flows and interchannel instability. These phenomena can have an important influence on heat transfer processes. Such influence is especially drastical under accident conditions associated with ceasing the forced circulation over the circuit. A number of urgent reactor safety problems at low operating parameters is related with the computer code verification and certification. One of the important trends in the reactor safety research is concerned with the rod bundle reflooding and verificational calculations of this phenomenon. To assess the water cooled reactor safety, the best fit computer codes are employed, which make it possible to simulate accident and transient operating conditions in a reactor installation. One of the most widely known computer codes is the RELAP5/MOD3 Code. The paper presents the comparison of the results calculated using this computer code with the test data on 4 - rod bundle quenching, which were obtained at the SSCRF-IPPE. Recently, the investigations on the steam - zirconium reaction kinetics have been performed at the SSCFR-IPPE and are being presently performed for the purpose of developing new and verifying available computer codes. (author). 3 refs, 6 figs

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

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

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.

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.

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

Nanofluids for power engineering: Emergency cooling of overheated heat transfer surfaces

Science.gov (United States)

Bondarenko, B. I.; Moraru, V. N.; Sidorenko, S. V.; Komysh, D. V.

2016-07-01

The possibility of emergency cooling of an overheated heat transfer surface using nanofluids in the case of a boiling crisis is explored by means of synchronous recording of changes of main heat transfer parameters of boiling water over time. Two nanofluids are tested, which are derived from a mixture of natural aluminosilicates (AlSi-7) and titanium dioxide (NF-8). It is found that the introduction of a small portions of nanofluid into a boiling coolant (distilled water) in a state of film boiling ( t heater > 500°C) can dramatically decrease the heat transfer surface temperature to 130-150°C, which corresponds to a transition to a safe nucleate boiling regime without affecting the specific heat flux. The fact that this regime is kept for a long time at a specific heat load exceeding the critical heat flux for water and t heater = 125-130°C is particularly important. This makes it possible to prevent a potential accident emergency (heater burnout and failure of the heat exchanger) and to ensure the smooth operation of the equipment.

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

Heat transfer from a plate cooled by a water film with countercurrent air flow

International Nuclear Information System (INIS)

Ambrosini, W.; Manfredini, A.; Mariotti, F.; Oriolo, F.; Vigni, P.

1995-01-01

An experimental program at the University of Pisa provides specific data for the evaluation of heat and mass transfer by falling film evaporation. The problem is addressed primarily because of its relevance to the study of the behavior of passive containment cooling systems in simplified pressurized water reactors. In these plants, after an accident that releases vapor from the primary circuit, the steel containment envelope is cooled either by an ascending stream of air in natural circulation or by the combination of air flow and falling film evaporation. To qualify models for the prediction of the heat transfer capabilities in postulated accident conditions, researchers have built an experimental facility consisting of a flat heated plate with water sprays and a fan to simulate a countercurrent air stream. The range of relevant parameters to be investigated has been determined on the basis of integral calculations performed for the AP600 reactor containment. The facility has enabled the collection of data that confirm the adequacy of the classical heat and mass transfer analogy in predicting evaporation phenomena. Further developments in the research are needed to confirm the first results and to extend the experimental database by considering more subtle aspects of the phenomenon such as the characteristics of surface waviness of the water film and its effect on heat transfer

Identification of flow regimes and heat transfer modes in Angra-2 core during the simulation of the small break loss of coolant accident of 250 cm{sup 2} in the cold leg of primary loop using RELAP5 code

Energy Technology Data Exchange (ETDEWEB)

Borges, Eduardo M.; Sabundjian, Gaiane, E-mail: borges.em@hotmail.com, E-mail: gdjian@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNE-SP), Sao Paulo, SP (Brazil)

2017-07-01

The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used by RELAP5/MOD3.2. gamma code in Angra-2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 250cm{sup 2} of rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of Angra-2 (FSAR-A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of Angra-2 during the postulated accident. The results obtained for Angra-2 nuclear reactor core during the postulated accident were satisfactory when compared with the FSAR-A2. Additionally, the results showed the correct actuation of the ECCS guaranteeing the integrity of the reactor core. (author)

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

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.)

Heat transfer in the core graphite structures of RBMK nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Knoglinger, E., E-mail: ernst.knoglinger@a1.net [Am Winklerwald 15, A 4020 Linz (Austria); Wölfl, H., E-mail: herbert.woelfl@tele2.at [Berg, Im Weideland 19, A 4060 Linz (Austria); Kaliatka, A., E-mail: algirdas.kaliatka@lei.lt [Laboratory of Nuclear Installation Safety, Lithuanian Energy Institute, Breslaujos 3, LT-44403 Kaunas (Lithuania)

2015-11-15

Highlights: • Proposed solution of heat transfer model from a hollow cylinder to a fluid through narrow duct. • Thermal conductance of annular gaps, filled by two component gas was discussed. • Xenon transient preceding the Chernobyl Accident was analyzed. • Reactivity balance during power manoeuvres and potenrial causes of the accident were discussed. - Abstract: Conductive and combined radiative/conductive gap conductance models are presented and discussed in great detail. The heat resistance concept and an exact solution to the one dimensional heat conduction equation for a 3-region composite hollow cylinder are used to calculate gap conductance in function of gap gas composition and fuel burn up. The study includes the back calculation of a reactor experiment performed at the Ignalina NPP Unit-1 which provides some insight in the function of the RBMK nitrogen supply and regulating device and an investigation of the role the graphite temperature played during the power manoeuvres preceding the Chernobyl Accident.

Effect of heat transfer in the fog region during core reflooding

International Nuclear Information System (INIS)

Rouai, N. M.; El-sawy, H. M.

1993-01-01

Core reflooding following a loss of coolant accident (LOCA) in a pressurized water reactor (PWR) received considerable attention during the past thirty years. In this paper a one dimensional model is used to study the effect of the heat transfer in the fog region ahead of the wet front reflooding rate of a cylindrical fuel element following a LOCA in a PWR. The heat conduction equation in the cladding is solved in coordinate system moving with the wet front under a variety of condition to investigate the effects of such parameters as the initial cladding surface temperature, the decay heat generation rate in the fuel and the mode of heat transfer prevailing. The cladding surface is divided into three axial regions according to the mechanism of heat transfer, namely, a boiling region behind the wet front, a fog region ahead of the wet front and a dry region further downstream of the wet front. The effect of changing the heat transfer coefficient in the fog region on the rewetting rate and on the fog length is investigated. The results of this simple model show that increasing the heat transfer in the fog region increases the rewetting velocity and consequently decreases the fog length. The results are in general agreement with a more accurate two-dimensional model and experimental data. (author)

Measurement of heat transfer effectiveness during collision of a Leidenfrost droplet with a heated wall - 15447

International Nuclear Information System (INIS)

Park, J.S.; Kim, H.; Bae, S.W.; Kim, K.D.

2015-01-01

Droplet-wall collision heat transfer during dispersed flow film boiling plays a role in predicting cooling rate and peak cladding temperature of overheated fuels during reflood following a LOCA accident in nuclear power plants. This study aims at experimentally studying effects of collision velocity and angle, as dynamic characteristics of the colliding droplet, on heat transfer. The experiments were performed by varying collision velocity from 0.2 to 1.5 m/s and collision angle between the droplet path and the wall in the range from 30 to 90 degrees under atmosphere condition. A single droplet was impinged on an infrared-opaque Pt film deposited on an infrared-transparent sapphire plate, which combination permits to measure temperature distribution of the collision surface using a high-speed infrared camera from below. The instantaneous local surface heat flux was obtained by solving transient heat conduction equation for the heated substrate using the measured surface temperature data as the boundary condition of the collision surface. Total heat transfer amount of a single droplet collision was calculated by integrating the local heat flux distribution on the effective heat transfer area during the collision time. The obtained results confirmed the finding from the previous studies that with increasing collision velocity, the heat transfer effectiveness increases due to the increase of the heat transfer area and the local heat flux value. Interestingly, it was found that as collision angle of a droplet with a constant collision velocity decreases from 90 to 50 degrees and thus the vertical velocity component of the collision decreases, the total heat transfer amount per a collision increases. It was observed that the droplet colliding with an angle less than 90 degrees slides on the surface during the collision and the resulting collision area is larger than that in the normal collision. On the other hand, further decrease of collision angle below 40 degrees

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

International Nuclear Information System (INIS)

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

1987-01-01

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

Heat transfer measurements of internally heated liquids in cylindrical convection cells

International Nuclear Information System (INIS)

Fieg, G.

1978-10-01

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

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.

Role of wall heat transfer and other system variables on fuel compaction and recriticality

International Nuclear Information System (INIS)

Dhir, V.K.; Castle, J.N.; Catton, I.; Kastenberg, W.E.; Doshi, J.B.

1976-01-01

The assessment of the molten fuel gaining recriticality after a hypothetical core disruptive accident in a fast reactor is an important safety consideration. Recriticality of the disrupted core can be envisioned to occur, if the fuel rearranges itself into a denser configuration either due to gravity slumping of the molten fuel or due to pressure or heat transfer driven compaction of the earlier dispersed fuel. In this paper the role played by wall heat transfer, internal radiation and the bottle pressure on the physical state of the molten fuel pool is discussed. It is suggested that in the absence of a solid crust the heat transfer process from the molten fuel to the surrounding steel will be very efficient because of melting and buoyancy driven removal of less dense steel through the pool of heavier UO 2 . The internal radiation at the high fuel temperature significantly increase the effective thermal conductivity of the molten fuel and lead to increased heat transfer in situations where a solid crust of UO 2 exists between molten UO 2 and molten steel. IN a boiled-up bottled pool, the pool pressure is shown to increase very rapidly with time and thus necessitate higher fission heating of the fuel to maintain it in a certain boiled up state. Finally, the results of the above discussion are applied to study the recriticality of a fuel pool formed during a hypothetical core disrupted accident in a fast reactor

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

Experimental investigation of natural convection heat transfer in volumetrically heated spherical segments. Final report

International Nuclear Information System (INIS)

Asfia, F.; Dhir, V.

1998-03-01

One strategy for preventing the failure of lower head of a nuclear reactor vessel is to flood the concrete cavity with subcooled water in accidents in which relocation of core material into the vessel lower head occurs. After the core material relocates into the vessel, a crust of solid material forms on the inner wall of the vessel, however, most of the pool remains molten and natural convection exists in the pool. At present, uncertainty exists with respect to natural convection heat transfer coefficients between the pool of molten core material and the reactor vessel wall. In the present work, experiments were conducted to examine natural convection heat transfer in internally heated partially filled spherical pools with external cooling. In the experiments, Freon-113 contained in a Pyrex bell jar was used as a test liquid. The pool was bounded with a spherical segment at the bottom, and was heated with magnetrons taken from a conventional microwave oven. The vessel was cooled from the outside with natural convection of water or with nucleate boiling of liquid nitrogen

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

Ftreign system studieo of hydrodynamics and heat-mass transfer at nuclear power plants

International Nuclear Information System (INIS)

Saltanov, G.A.

1981-01-01

Status and the main problems of system studies on hydrodynamics and heat-and-mass transfer at nuclear power plant transients and accidents are considered. Experimental benchmarks used for studying the loss of coolant accidents are described. The conclusion is made that contemporary level of measuring apparatus development and a large number of fast-response monitors of temperature, pressure and coolant level at most of described benchmarks permit to obtain sufficiently complete information of the behaviour of most important parts of a reactor unit during transients and accidents of different type [ru

A methodology for the transfer of probabilities between accident severity categories

International Nuclear Information System (INIS)

Whitlow, J.D.; Neuhauser, K.S.

1993-01-01

This paper will describe a methodology which has been developed to allow accident probabilities associated with one severity category scheme to be transferred to another severity category scheme, permitting some comparisons of different studies at the category level. In this methodology, the severity category schemes to be compared are mapped onto a common set of axes. The axes represent critical accident environments (e.g., impact, thermal, crush, puncture) and indicate the range of accident parameters from zero (no accident) to the most sever credible forces. The choice of critical accident environments for the axes depends on the package being transported and the mode of transportation. The accident probabilities associated with one scheme are then transferred to the other scheme. This transfer of category probabilities is based on the relationships of the critical accident parameters to probability of occurrence. The methodology can be employed to transfer any quantity between category schemes if the appropriate supporting information is available. (J.P.N.)

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

International Nuclear Information System (INIS)

Borges, Eduardo M.; Sabundjian, Gaiane

2015-01-01

Identifying the flow regimes and the heat transfer modes is important for the analysis of accidents such as the Loss-of-Coolant Accident (LOCA). The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used in the RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 100cm 2 -rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR - A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. (author)

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.

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.

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

Energy Technology Data Exchange (ETDEWEB)

Borges, Eduardo M.; Sabundjian, Gaiane, E-mail: gdgian@ipen.br, E-mail: borges.em@hotmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

Identifying the flow regimes and the heat transfer modes is important for the analysis of accidents such as the Loss-of-Coolant Accident (LOCA). The aim of this paper is to identify the flow regimes, the heat transfer modes, and the correlations used in the RELAP5/MOD3.2.gama code in ANGRA 2 during the Small-Break Loss-of-Coolant Accident (SBLOCA) with a 100cm{sup 2}-rupture area in the cold leg of primary loop. The Chapter 15 of the Final Safety Analysis Report of ANGRA 2 (FSAR - A2) reports this specific kind of accident. The results from this work demonstrated the several flow regimes and heat transfer modes that can be present in the core of ANGRA 2 during the postulated accident. (author)

Heat transfer analysis to investigate the core catcher plate assembly in SFR

International Nuclear Information System (INIS)

Patil, Swapnil; Sharma, Anil Kumar; Velusamy, K.; Nashine, B.K.; Selvaraj, P.

2015-01-01

Severe accident scenario in Sodium Cooled Fast Reactor (SFR) is the major concern for public acceptance. After severe accident, the molten core continuously generates substantial decay heat. However, an in-vessel core catcher plate is provided to remove the decay heat passively. The numerical investigation of pool hydraulics phenomena in sodium pool of typical Indian SFR has been carried out. The debris may form a heap with different angle over the core catcher plate due to molten fuel density and interaction force. Therefore, the debris bed with different heap angle has been analyzed for steady and transient state conditions. The governing equation of fluid flow and heat transfer are solved by finite volume method based solver with the k-ε turbulent model. The time period Δ for which temperature is exceeding above safety limit with different debris heap angle have been established. (author)

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

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.

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

A methodology for the transfer of probabilities between accident severity categories

International Nuclear Information System (INIS)

Whitlow, J.D.; Neuhauser, K.S.

1991-01-01

A methodology has been developed which allows the accident probabilities associated with one accident-severity category scheme to be transferred to another severity category scheme. The methodology requires that the schemes use a common set of parameters to define the categories. The transfer of accident probabilities is based on the relationships between probability of occurrence and each of the parameters used to define the categories. Because of the lack of historical data describing accident environments in engineering terms, these relationships may be difficult to obtain directly for some parameters. Numerical models or experienced judgement are often needed to obtain the relationships. These relationships, even if they are not exact, allow the accident probability associated with any severity category to be distributed within that category in a manner consistent with accident experience, which in turn will allow the accident probability to be appropriately transferred to a different category scheme

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

International Nuclear Information System (INIS)

Kondo, Masaya; Nishida, Ayumu; Sugimoto, Jun

2015-01-01

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

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;...

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-11-01

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

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

International Nuclear Information System (INIS)

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

2000-01-01

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

Heat up and potential failure of BWR upper internals during a severe accident

Energy Technology Data Exchange (ETDEWEB)

Robb, Kevin R [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-01-01

In boiling water reactors, the steam dome, steam separators, and dryers above the core are comprised of approximately 100 tons of stainless steel. During a severe accident in which the coolant boils away and exothermic oxidation of zirconium occurs, gases (steam and hydrogen) are superheated in the core region and pass through the upper internals. Historically, the upper internals have been modeled using severe accident codes with relatively simple approximations. The upper internals are typically modeled in MELCOR as two lumped volumes with simplified heat transfer characteristics, with no structural integrity considerations, and with limited ability to oxidize, melt, and relocate. The potential for and the subsequent impact of the upper internals to heat up, oxidize, fail, and relocate during a severe accident was investigated. A higher fidelity representation of the shroud dome, steam separators, and steam driers was developed in MELCOR v1.8.6 by extending the core region upwards. This modeling effort entailed adding 45 additional core cells and control volumes, 98 flow paths, and numerous control functions. The model accounts for the mechanical loading and structural integrity, oxidation, melting, flow area blockage, and relocation of the various components. The results indicate that the upper internals can reach high temperatures during a severe accident; they are predicted to reach a high enough temperature such that they lose their structural integrity and relocate. The additional 100 tons of stainless steel debris influences the subsequent in-vessel and ex-vessel accident progression.

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

International Nuclear Information System (INIS)

Zhang Tao; Liu Xiaohua; Zhang Lun; Jiang Yi

2012-01-01

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

Effect of heat transfer tube leak on dynamic characteristic of steam generator

International Nuclear Information System (INIS)

Sun Baozhi; Shi Jianxin; Li Na; Zheng Lusong; Liu Shanghua; Lei Yu

2015-01-01

Taking the steam generator of Daya Bay Nuclear Power Station as the research object, one-dimensional dynamic model of the steam generator based on drift flux theory and leak model of heat transfer tube were established. Steady simulation of steam generator under different conditions was carried out. Based on verifying the drift flux model and leak model of heat transfer tube, the effect of leak location and flow rate under different conditions on steam generator's key parameters was studied. The results show that the drift flux model and leak model can reflect the law of key parameter change accurately such as vapor mass fraction and steam pressure under different leak cases. The variation of the parameters is most apparent when the leak is at the entrance of boiling section and vapor mass fraction varies from 0.261 to 0.163 when leakage accounts for 5% of coolant flow rate. The successful prediction of the effect of heat transfer tube leak on dynamic characteristics of the steam generator based on drift flux theory supplies some references for monitoring and taking precautionary measures to prevent heat transfer tube leak accident. (authors)

Accident on the gas transfer system

International Nuclear Information System (INIS)

Heugel, J.

1991-10-01

An accident has happened on the Vivitron gas transfer system on the 7 th August 1991. This report presents the context, facts and inquiries, analyses the reasons and explains also how the repairing has been effected

Estimation of heat transfer and heat source in a molten pool

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-01

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

Estimation of heat transfer and heat source in a molten pool

International Nuclear Information System (INIS)

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

2001-01-01

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

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

International Nuclear Information System (INIS)

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

1994-09-01

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

Analysis of heat and mass transfers in two-phase flow by coupling optical diagnostic techniques

Energy Technology Data Exchange (ETDEWEB)

Lemaitre, P.; Porcheron, E. [Institut de Radioprotection et de Surete Nucleaire, Saclay (France)

2008-08-15

During the course of a hypothetical accident in a nuclear power plant, spraying might be actuated to reduce static pressure in the containment. To acquire a better understanding of the heat and mass transfers between a spray and the surrounding confined gas, non-intrusive optical measurements have to be carried out simultaneously on both phases. The coupling of global rainbow refractometry with out-of-focus imaging and spontaneous Raman scattering spectroscopy allows us to calculate the local Spalding parameter B{sub M}, which is useful in describing heat transfer associated with two-phase flow. (orig.)

Analysis of heat and mass transfers in two-phase flow by coupling optical diagnostic techniques

International Nuclear Information System (INIS)

Lemaitre, P.; Porcheron, E.

2008-01-01

During the course of a hypothetical accident in a nuclear power plant, spraying might be actuated to reduce static pressure in the containment. To acquire a better understanding of the heat and mass transfers between a spray and the surrounding confined gas, non-intrusive optical measurements have to be carried out simultaneously on both phases. The coupling of global rainbow refractometry with out-of-focus imaging and spontaneous Raman scattering spectroscopy allows us to calculate the local Spalding parameter B M , which is useful in describing heat transfer associated with two-phase flow. (orig.)

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

Energy Technology Data Exchange (ETDEWEB)

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

1999-11-01

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

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...

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.

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

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.)

Heat transfer on liquid-liquid interface of molten-metal and water

International Nuclear Information System (INIS)

Tanaka, T.; Saito, Yasushi; Mishima, Kaichiro

2001-01-01

Molten-core pool had been formed in the lower-head of TMI-2 pressure vessel at the severe accident. The lower head, however, didn't receive any damage by reactor core cooling. Heat transfer at outside of the lower head and boiling heat transfer at liquid-liquid interface of molten-metal and water, however, are important for initial cooling process of the molten-core pool. The heat transfer experiments for the liquid-liquid interface of molten-metal and water are carried out over the range of natural convection to film boiling region. Phenomenon on the heat transfer experiments are visualized by using of high speed video camera. Wood's metal and U-alloy 78 are used as molten-metal. The test section of the experiments consists of a copper block with heater, wood's metal, and water. Three thermocouple probes are used for temperature measurement of water side and the molten-metal side. Stability of the liquid-liquid interface is depended on the wetness of container wall for molten metal and the temperature distribution of the interface. Entrainment phenomena of molten-metal occurs by a fluctuation of the interface after boiling on the container wall surface. The boiling curves obtained from the liquid-liquid interface experiments are agree with the nucleate boiling and the film boiling correlations of solid-liquid system. (Suetake, M.)

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

Heat transfer effect of an extended surface in downward-facing subcooled flow boiling

Energy Technology Data Exchange (ETDEWEB)

Khan, Abdul R., E-mail: khan@vis.t.u-tokyo.ac.jp [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo, 7-3-1 Hongo, Bunkyo-ku, Tokyo 113-8656 (Japan); Erkan, Nejdet, E-mail: erkan@vis.t.u-tokyo.ac.jp [Nuclear Professional School, School of Engineering, The University of Tokyo, 2-22 Shirakata, Tokai-mura, Ibaraki, 319-1188 (Japan); Okamoto, Koji, E-mail: okamoto@n.t.u-tokyo.ac.jp [Nuclear Professional School, School of Engineering, The University of Tokyo, 2-22 Shirakata, Tokai-mura, Ibaraki, 319-1188 (Japan)

2015-12-15

Highlights: • Compare downward-facing flow boiling results from bare and extended surfaces. • Upstream and downstream temperatures were measured on the extended surface. • Downstream temperatures exceed upstream temperatures for all flow rates. • Bubble accumulation occurs downstream on extended surface. • Extended surface heat transfer lower than bare surface as flow rate reduced. - Abstract: New BWR containment designs are considering cavity flooding as an accident management strategy. Unlike the PWR, the BWR has many Control Rod Guide Tube (CRGT) penetrations in the lower head. During a severe accident scenario with core melt in the lower plenum along with cavity flooding, the penetrations may affect the heat transfer on the ex-vessel surface and disrupt fluid flow during the boiling process. A small-scale experiment was performed to investigate the issues existing in downward-facing boiling phenomenon with an extended surface. The results were compared with a bare (flat) surface. The mass flux of 244 kg/m{sup 2} s, 215 kg/m{sup 2} s, and 177 kg/m{sup 2} s were applied in this study. CHF conditions were observed only for the 177 kg/m{sup 2} s case. The boiling curves for both types of surfaces and all flow rates were obtained. The boiling curves for the highest flow rate showed lower surface temperatures for the extended surface experiments when compared to the bare surface. The downstream location on the extended surface yielded the highest surface temperatures as the flow rate was reduced. The bubble accumulation and low velocity in the wake produced by flow around the extended surface was believed to have caused the elevated temperatures in the downstream location. Although an extended surface may enhance the overall heat transfer, a reduction in the local heat transfer was observed in the current experiments.

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.

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

Flow visualization study of inverted annular flow of post dryout heat transfer region

International Nuclear Information System (INIS)

Ishii, M.; De Jarlais, G.

1985-01-01

The inverted annular flow is important in the area of LWR accident analysis in terms of the maximum cladding temperature and effectiveness of the emergency core cooling. However, the inverted annular flow thermal-hydraulics is not well understood due to its special heat transfer condition of film boiling. In view of this, the inverted flow is studied in detail experimentally. A new experimental apparatus has been constructed in which film boiling heat transfer can be established in a transparent test section. Data on liquid core stability, core break-up mechanism, and dispersed-core liquid slug and droplet sizes are obtained using F 113 as a test fluid. Both high speed movies and flash photographs are used

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.

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)

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

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

The Impact of Heat Waves on Occurrence and Severity of Construction Accidents.

Science.gov (United States)

Rameezdeen, Rameez; Elmualim, Abbas

2017-01-11

The impact of heat stress on human health has been extensively studied. Similarly, researchers have investigated the impact of heat stress on workers' health and safety. However, very little work has been done on the impact of heat stress on occupational accidents and their severity, particularly in South Australian construction. Construction workers are at high risk of injury due to heat stress as they often work outdoors, undertake hard manual work, and are often project based and sub-contracted. Little is known on how heat waves could impact on construction accidents and their severity. In order to provide more evidence for the currently limited number of empirical investigations on the impact of heat stress on accidents, this study analysed 29,438 compensation claims reported during 2002-2013 within the construction industry of South Australia. Claims reported during 29 heat waves in Adelaide were compared with control periods to elicit differences in the number of accidents reported and their severity. The results revealed that worker characteristics, type of work, work environment, and agency of accident mainly govern the severity. It is recommended that the implementation of adequate preventative measures in small-sized companies and civil engineering sites, targeting mainly old age workers could be a priority for Work, Health and Safety (WHS) policies.

The Impact of Heat Waves on Occurrence and Severity of Construction Accidents

Science.gov (United States)

Rameezdeen, Rameez; Elmualim, Abbas

2017-01-01

The impact of heat stress on human health has been extensively studied. Similarly, researchers have investigated the impact of heat stress on workers’ health and safety. However, very little work has been done on the impact of heat stress on occupational accidents and their severity, particularly in South Australian construction. Construction workers are at high risk of injury due to heat stress as they often work outdoors, undertake hard manual work, and are often project based and sub-contracted. Little is known on how heat waves could impact on construction accidents and their severity. In order to provide more evidence for the currently limited number of empirical investigations on the impact of heat stress on accidents, this study analysed 29,438 compensation claims reported during 2002–2013 within the construction industry of South Australia. Claims reported during 29 heat waves in Adelaide were compared with control periods to elicit differences in the number of accidents reported and their severity. The results revealed that worker characteristics, type of work, work environment, and agency of accident mainly govern the severity. It is recommended that the implementation of adequate preventative measures in small-sized companies and civil engineering sites, targeting mainly old age workers could be a priority for Work, Health and Safety (WHS) policies. PMID:28085067

Selection of Rational Heat Transfer Intensifiers in the Heat Exchanger

Directory of Open Access Journals (Sweden)

S. A. Burtsev

2016-01-01

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

Heat transfer in an asymmetrically heated duct, 2

International Nuclear Information System (INIS)

Satoh, Isao; Kurosaki, Yasuo

1986-01-01

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

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)

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.

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

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

Air temperature determination inside residual heat removal pump room of Angra-1 nuclear power plant after a design basic accident

International Nuclear Information System (INIS)

Siniscalchi, Marcio Rezende

2005-01-01

This work develops heat transfer theoretical models for determination of air temperature inside the Residual Heat Removal Pump Room of Angra 1 Nuclear Power Plant after a Design Basis Accident without forced ventilation. Two models had been developed. The differential equations are solved by analytical methods. A software in FORTRAN language are developed for simulations of temperature inside rooms for different geometries and materials. (author)

Analysis of relations for heat transfer at the post-CHF regime

Energy Technology Data Exchange (ETDEWEB)

Dorokhovich, S. L. [Obninsk State Techical Univ., Obninsk (Russian Federation); Kirillov, P. L. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

2003-07-01

Information about heat transfer rates in two-fluid flow at the post-CHF regime is important for analysis of accidents of water-cooled nuclear reactors as well as calculations of steam generators (liquid metal-water). It is complicate to create analytical methods because there is a variety of two-fluid flow regimes dictated by channel size, pressure, mass flow rate, heat flux, droplet spectrum, flow quality, other factors and, finally, by crisis type. At crisis in subcooled liquid or low quality two-phase flows Inverted Annular Film Boiling (IAFB) takes place when liquid flow is separated from the wall by a thin superheated vapor film. For dispersed flow crisis is usually related to drying a liquid film moving along the heated surface (wall). In both cases two-phase flow is thermodynamically nonequilibrium since the temperatures of phases (liquid and vapor) are different. The mean (at the rate of heat content) flow temperature is not the determining parameter. Different boundary conditions of experiment s uniform heat fluxes or 'hot' spots, are able to lead to different relations for the heat transfer coefficient. Last years the great number of semiempirical models were elaborated, that become more and more complicate. It is difficult to examine many parameters of the models. An agreement between final results and separate experimental data is not yet the evidence for the verity of extension while prerequisites taken in the models are often doubtful and hardly examined. Thus the correlations obtained from experimental data, for example, are used in practice. The analysis of relations for heat transfer at the Deteriorated Heat Transfer (DHT) regime, the comparison of relations with the data of look up tables made on the basis of the Institute of Physics and Power Engineering and Chalk River Laboratories experimental data banks are the objective of current report.

Analysis of relations for heat transfer at the post-CHF regime

International Nuclear Information System (INIS)

Dorokhovich, S. L.; Kirillov, P. L.

2003-01-01

Information about heat transfer rates in two-fluid flow at the post-CHF regime is important for analysis of accidents of water-cooled nuclear reactors as well as calculations of steam generators (liquid metal-water). It is complicate to create analytical methods because there is a variety of two-fluid flow regimes dictated by channel size, pressure, mass flow rate, heat flux, droplet spectrum, flow quality, other factors and, finally, by crisis type. At crisis in subcooled liquid or low quality two-phase flows Inverted Annular Film Boiling (IAFB) takes place when liquid flow is separated from the wall by a thin superheated vapor film. For dispersed flow crisis is usually related to drying a liquid film moving along the heated surface (wall). In both cases two-phase flow is thermodynamically nonequilibrium since the temperatures of phases (liquid and vapor) are different. The mean (at the rate of heat content) flow temperature is not the determining parameter. Different boundary conditions of experiment s uniform heat fluxes or 'hot' spots, are able to lead to different relations for the heat transfer coefficient. Last years the great number of semiempirical models were elaborated, that become more and more complicate. It is difficult to examine many parameters of the models. An agreement between final results and separate experimental data is not yet the evidence for the verity of extension while prerequisites taken in the models are often doubtful and hardly examined. Thus the correlations obtained from experimental data, for example, are used in practice. The analysis of relations for heat transfer at the Deteriorated Heat Transfer (DHT) regime, the comparison of relations with the data of look up tables made on the basis of the Institute of Physics and Power Engineering and Chalk River Laboratories experimental data banks are the objective of current report

Analysis of the dual phase lag bio-heat transfer equation with constant and time-dependent heat flux conditions on skin surface

Directory of Open Access Journals (Sweden)

Ziaei Poor Hamed

2016-01-01

Full Text Available This article focuses on temperature response of skin tissue due to time-dependent surface heat fluxes. Analytical solution is constructed for DPL bio-heat transfer equation with constant, periodic and pulse train heat flux conditions on skin surface. Separation of variables and Duhamel’s theorem for a skin tissue as a finite domain are employed. The transient temperature responses for constant and time-dependent boundary conditions are obtained and discussed. The results show that there is major discrepancy between the predicted temperature of parabolic (Pennes bio-heat transfer, hyperbolic (thermal wave and DPL bio-heat transfer models when high heat flux accidents on the skin surface with a short duration or propagation speed of thermal wave is finite. The results illustrate that the DPL model reduces to the hyperbolic model when τT approaches zero and the classic Fourier model when both thermal relaxations approach zero. However for τq = τT the DPL model anticipates different temperature distribution with that predicted by the Pennes model. Such discrepancy is due to the blood perfusion term in energy equation. It is in contrast to results from the literature for pure conduction material, where the DPL model approaches the Fourier heat conduction model when τq = τT . The burn injury is also investigated.

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

Containment condensing heat transfer

International Nuclear Information System (INIS)

Gido, R.G.; Koestel, A.

1983-01-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

The Impact of Heat Waves on Occurrence and Severity of Construction Accidents

Directory of Open Access Journals (Sweden)

Rameez Rameezdeen

2017-01-01

Full Text Available The impact of heat stress on human health has been extensively studied. Similarly, researchers have investigated the impact of heat stress on workers’ health and safety. However, very little work has been done on the impact of heat stress on occupational accidents and their severity, particularly in South Australian construction. Construction workers are at high risk of injury due to heat stress as they often work outdoors, undertake hard manual work, and are often project based and sub-contracted. Little is known on how heat waves could impact on construction accidents and their severity. In order to provide more evidence for the currently limited number of empirical investigations on the impact of heat stress on accidents, this study analysed 29,438 compensation claims reported during 2002–2013 within the construction industry of South Australia. Claims reported during 29 heat waves in Adelaide were compared with control periods to elicit differences in the number of accidents reported and their severity. The results revealed that worker characteristics, type of work, work environment, and agency of accident mainly govern the severity. It is recommended that the implementation of adequate preventative measures in small-sized companies and civil engineering sites, targeting mainly old age workers could be a priority for Work, Health and Safety (WHS policies.

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

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

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

International Nuclear Information System (INIS)

Kim, Jae Kwang

2000-02-01

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

Mass transfer experiments for the heat load during in-vessel retention of core melt

Energy Technology Data Exchange (ETDEWEB)

Park, Hae Kyun; Chung, Bum Jin [Dept. of Nuclear Engineering, Kyung Hee University, Seoul (Korea, Republic of)

2016-08-15

We investigated the heat load imposed on the lower head of a reactor vessel by the natural convection of the oxide pool in a severe accident. Mass transfer experiments using a CuSO{sub 4}–H{sub 2}SO{sub 4} electroplating system were performed based on the analogy between heat and mass transfer. The Ra′{sub H} of 10{sup 14} order was achieved with a facility height of only 0.1 m. Three different volumetric heat sources were compared; two had identical configurations to those previously reported, and the other was designed by the authors. The measured Nu's of the lower head were about 30% lower than those previously reported. The measured angular heat flux ratios were similar to those reported in existing studies except for the peaks appearing near the top. The volumetric heat sources did not affect the Nu of the lower head but affected the Nu of the top plate by obstructing the rising flow from the bottom.

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

Assessment of Two-Phase Flow Heat Transfer Correlations for Molten Core-Concrete Interaction Study

International Nuclear Information System (INIS)

Tourniaire, B.; Varo, O.

2006-01-01

The prediction of heat transfer between corium pool and concrete basemat is of particular significance in the framework of the study of PWR's severe accident. Heat transfer directly governs the ablation velocity of concrete in case of molten core-concrete interaction (MCCI) and, consequently, the time delay when the reactor cavity may fail. From a restricted hydrodynamic point of view, this issue is related to heat transfer between a heated bubbling pool and a porous wall with gas injection. Several experimental studies have been performed with simulant materials and many correlations have been provided to address this issue. The main purpose of this paper is to assess these correlations from comparisons against the available experimental data. After a review of these data, the different correlations are presented. Attention focuses here on the correlations generally used in MCCI study: Kutateladze-Malenkov, Konsetov and BALI correlations. The Deckwer's correlation is also included in this review. The comparisons between the results of these correlations and the experimental data are then discussed. (authors)

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

International Nuclear Information System (INIS)

Chen Lingen; Li Jun; Sun Fengrui

2008-01-01

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

Deformation and Heat Transfer on Three Sides Protected Beams under Fire Accident

Science.gov (United States)

Imran, M.; Liew, M. S.; Garcia, E. M.; Nasif, M. S.; Yassin, A. Y. M.; Niazi, U. M.

2018-04-01

Fire accidents are common in oil and gas industry. The application of passive fire protection (PFP) is a costly solution. The PFP is applied only on critical structural members to optimise project cost. In some cases, beams cannot be protected from the top flange in order to accommodate for the placement of pipe supports and grating. It is important to understand the thermal and mechanical response of beam under such condition. This paper discusses the response of steel beam under ISO 834 fire protected, unprotected and three sides protected beams. The model validated against an experimental study. The experimental study has shown good agreement with FE model. The study revealed that the beams protected from three sides heat-up faster compare to fully protected beam showing different temperature gradient. However, the affects load carrying capacity are insignificant under ISO 834 fire.

Convective heat transfer the molten metal pool heated from below and cooled by two-phase flow

International Nuclear Information System (INIS)

Cho, J. S.; Suh, K. Y.; Chung, C. H.; Park, R. J.; Kim, S. B.

1998-01-01

During a hypothetical servere accident in the nuclear power plant, a molten core material may form stratified fluid layers. These layers may be composed of high temperature molten debris pool and water coolant in the lower plenum of the reactor vessel or in the reactor cavity. This study is concerned with the experimental test and numerical analysis on the heat transfer and solidification of the molten metal pool with overlying coolant with boiling. This work examines the crust formation and the heat transfer characteristics of the molten metal pool immersed in the boiling coolant. The metal pool is heated from the bottom surface and coolant is injected onto the molten metal pool. The simulant molten pool material is tin (Sn) with the melting temperature of 232 .deg. C. Demineralized water is used as the working coolant. Tests were performed under the condition of the bottom surface heating in the test section and the forced convection of the coolant being injected onto the molten metal pool. The constant temperature and constant heat flux conditions are adopted for the bottom heating. The test parameters included the heated bottom surface temperature of the molten metal pool, the input power to the heated bottom surface of the test section, and the coolant injection rate. Numerical analyses were simultaneously performed in a two-dimensional rectangular domain of the molten metal pool to check on the measured data. The numerical program has been developed using the enthalpy method, the finite volume method and the SIMPLER algorithm. The experimental results of the heat transfer show general agreement with the calculated values. In this study, the relationship between the Nusselt number and Rayleigh number in the molten metal pool region was estimated and compared with the dry experiment without coolant nor solidification of the molten metal pool, and with the crust formation experiment with subcooled coolant, and against other correlations. In the experiments, the

Phase change heat transfer device for process heat applications

International Nuclear Information System (INIS)

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

2010-01-01

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

Heat transfer study under supercritical pressure conditions

International Nuclear Information System (INIS)

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

2003-01-01

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

Aerosol formation from heat and mass transfer in vapour-gas mixtures

International Nuclear Information System (INIS)

Clement, C.F.

1985-01-01

Heat and mass transfer equations and their coupling to the equation for the aerosol size distribution are examined for mixtures in which pressure changes are slow. Specific results in terms of Cn (the condensation number) and Le (the Lewis number - the ratio of the relative rates of evaporation and condensation) are obtained for the proportion of vapour condensing as a aerosol during the cooling and heating of a mixture in a well-mixed cavity. The assumption of allowing no supersaturations, the validity of which is examined, is shown to lead to maximum aerosol formation. For water vapour-air mixtures predictions are made as to temperature regions in which aerosols will evaporate or not form in cooling processes. The results are also qualitatively applied to some atmospheric effects as well as to water aerosols formed in the containment of a pressurized water reactor following a possible accident. In this context, the present conclusion that the whereabouts of vapour condensation is controlled by heat and mass transfer, contrasts with previous assumptions that the controlling factor is relative surface areas. (U.K.)

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

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

Research on the fundamental process of thermal-hydraulic behaviors in severe accident. Heat transfer on the liquid-liquid interface between molten core pool and coolant. JAERI's nuclear research promotion program, H10-027-6. Contract research

International Nuclear Information System (INIS)

Mishima, Kaichiro; Saito, Yasushi

2002-03-01

Heat transfer experiments under steady and transient conditions were performed using molten Wood's metal and distilled water to study heat transfer on the liquid-liquid interface between molten fuel pool and coolant under severe accident conditions. In the steady state experiment, boiling curve was measured over the range from natural convection region to film boiling region. The boiling behavior was observed using a high-speed video camera. In the transient experiment, distilled water was poured onto the hot molten metal surface, and the boiling curve was obtained in the cooling process. Comparing the measured boiling curve with existing correlations and experimental data for solid-liquid and liquid-liquid systems, the following conclusions were drawn: (a) When the interface surge is negligible and oxide layer is formed on the interface, the boiling curve at the liquid-liquid surface could be approximately reproduced by the heat transfer correlations for nucleate boiling and film boiling regions and the critical heat flux correlation for a liquid-solid system. (b) When no oxide layer is formed on the interface, the boiling curve at the liquid-liquid surface moved towards higher wall superheat than that at the liquid-solid surface, as Novakovic et al. observed in their experiment using mercury. (c) Transient heat transfer coefficient for film boiling at the liquid-liquid surface was about 100% higher than that predicted by the heat transfer correlation for a solid-liquid system. (author)

Stokes flow heat transfer in an annular, rotating heat exchanger

International Nuclear Information System (INIS)

Saatdjian, E.; Rodrigo, A.J.S.; Mota, J.P.B.

2011-01-01

The heat transfer rate into highly viscous, low thermal-conductivity fluids can be enhanced significantly by chaotic advection in three-dimensional flows dominated by viscous forces. The physical effect of chaotic advection is to render the cross-sectional temperature field uniform, thus increasing both the wall temperature gradient and the heat flux into the fluid. A method of analysis for one such flow-the flow in the eccentric, annular, rotating heat exchanger-and a procedure to determine the best heat transfer conditions, namely the optimal values of the eccentricity ratio and time-periodic rotating protocol, are discussed. It is shown that in continuous flows, such as the one under consideration, there exists an optimum frequency of the rotation protocol for which the heat transfer rate is a maximum. - Highlights: → The eccentric, annular, rotating heat exchanger is studied for periodic Stokes flow. → Counter-rotating the inner tube with a periodic velocity enhances the heat transfer. → The heat-transfer enhancement under such conditions is due to chaotic advection. → For a given axial flow rate there is a frequency that maximizes the heat transfer. → There is also an optimum value of the eccentricity ratio.

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Fluid dynamics and heat transfer methods for the TRAC code

International Nuclear Information System (INIS)

Reed, W.H.; Kirchner, W.L.

1977-01-01

A computer code called TRAC is being developed for analysis of loss-of-coolant accidents and other transients in light water reactors. This code involves a detailed, multidimensional description of two-phase flow coupled implicitly through appropriate heat transfer coefficients with a simulation of the temperature field in fuel and structural material. Because TRAC utilizes about 1000 fluid mesh cells to describe an LWR system, whereas existing lumped parameter codes typically involve fewer than 100 fluid cells, new highly implicit difference techniques are developed that yield acceptable computing times on modern computers. Several test problems for which experimental data are available, including blowdown of single pipe and loop configurations with and without heated walls, have been computed with TRAC. Excellent agreement with experimental results has been obtained

Natural convection heat transfer in a rectangular pool with volumetric heat sources

International Nuclear Information System (INIS)

Lee, Seung Dong; Lee, Kang Hee; Suh, Kune Y.

2003-01-01

Natural convection plays an important role in determining the thermal load from debris accumulated in the reactor vessel lower head during a severe accident. The heat transfer within the molten core material can be characterized by buoyancy-induced flows resulting from internal heating due to decay of fission products. The thermo-fluid dynamic characteristics of the molten pool depend strongly on the thermal boundary conditions. The spatial and temporal variation of heat flux on the pool wall boundaries and the pool superheat are mainly characterized by the natural convection flow inside the molten pool. In general, natural convection involving internal heat generation is delineated in terms of the modified Rayleigh number, Ra', which quantifies the internal heat source and hence the strength of buoyancy. The test section is of rectangular cavity whose length, width, and height are 500 mm, 80 mm, and 250 mm, respectively. A total of twenty-four T-type thermocouples were installed in the test loop to measure temperature distribution. Four T-type thermocouples were utilized to measure temperatures on the boundary. A direct heating method was adopted in this test to simulate the uniform heat generation. The experiments covered a range of Rayleigh number, Ra, between 4.87x10 7 and 2.32x10 14 and Prandtl number, Pr, between 0.7 and 3.98. Tests were conducted with water and air as simulant. The upper and lower boundary conditions were maintained at a uniform temperature of 10degC. (author)

A literature survey on numerical heat transfer

Science.gov (United States)

Shih, T. M.

1982-12-01

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

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.)

Study of sodium film-boiling heat transfer from a high-temperature sphere

International Nuclear Information System (INIS)

Le-Belguet, A.

2013-01-01

During a severe accident in a sodium-cooled fast reactor, molten fuel may come into contact with the surrounding liquid sodium, resulting in a so-called Fuel-Coolant Interaction. This work aims at providing a better understanding and knowledge of the associated heat transfer, likely to be in the film-boiling regime and required to study the risks related to a vapor explosion. Scarce literature has been found on sodium film boiling, both from an experimental and a theoretical point of view. Only one experiment has been conducted to investigate sodium pool film-boiling heat transfer. In our analysis of the experiment, two film-boiling regimes have been identified: a stable film boiling regime, without liquid-solid contact, and an unstable film-boiling regime, with contacts. Besides, the only theoretical model dedicated to sodium film boiling has shown some weaknesses. First, a scaling analysis of the problem has been proposed for free and forced convection, considering the two extreme cases of saturated and highly subcooled liquid. This simplified approach, which shows a good agreement with the experimental data, provides the dimensionless numbers which should be used to build correlations. A theoretical model has been developed to describe sodium film-boiling heat transfer from a hot sphere in free and forced convection, whatever the liquid subcooling. It is based on a two-phase laminar boundary layer integral method and includes the inertial and convective terms in the vapor momentum and energy equations, usually neglected. The radiation has been taken into account in the interfacial energy balance and contributes directly to produce vapor. This model enables to predict the heat lost from a hot body within an acceptable error compared to the tests results especially when the experimental uncertainties are considered. The heat partition between liquid heating and vaporization, essential to study the vapor explosion phenomenon, is also estimated. The influence of

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.

Molten corium concrete interaction: investigation of heat transfer in two-phase flow

International Nuclear Information System (INIS)

Amizic, Milan

2014-01-01

In the context of severe accident research for the second and the third generation of nuclear power plants, there are still open issues concerning some aspects of the concrete cavity ablation during the molten corium - concrete interaction (MCCI). The determination of heat transfer along the interfacial region between the molten corium pool and the ablating basemat concrete is crucial for the assessment of concrete ablation progression and eventually the basemat melt through. For the purpose of experimental investigation of thermal hydraulics inside a liquid pool agitated by gas bubbles, the CLARA project has been launched. The CLARA experiments are performed using simulant materials and they reveal the influence of superficial gas velocity, liquid viscosity and pool geometry on the heat transfer coefficient between the internally heated liquid pool and vertical and horizontal pool walls maintained at uniform temperature. The first test campaign has been conducted with the small pool configuration (50 cm * 25 cm * 25 cm). The tests have been performed with liquids covering a wide range of dynamic viscosity from approximately 1 mPa s to 10000 mPa s and the superficial gas velocity is varied up to 8 cm/s. This thesis comprises a brief description of MCCI phenomenology, literature reviews on the existing heat transfer correlations for two phase flow and the void fraction, a description of CLARA setup, experimental results and their interpretation. The experimental results are compared with existing models and some new models for the assessment of heat transfer coefficient in two-phase flow. (author) [fr

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

NARCIS (Netherlands)

Brouwers, H.J.H.

1996-01-01

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

Modeling of pipe break accident in a district heating system using RELAP5 computer code

International Nuclear Information System (INIS)

Kaliatka, A.; Valinčius, M.

2012-01-01

Reliability of a district heat supply system is a very important factor. However, accidents are inevitable and they occur due to various reasons, therefore it is necessary to have possibility to evaluate the consequences of possible accidents. This paper demonstrated the capabilities of developed district heating network model (for RELAP5 code) to analyze dynamic processes taking place in the network. A pipe break in a water supply line accident scenario in Kaunas city (Lithuania) heating network is presented in this paper. The results of this case study were used to demonstrate a possibility of the break location identification by pressure decrease propagation in the network. -- Highlights: ► Nuclear reactor accident analysis code RELAP5 was applied for accident analysis in a district heating network. ► Pipe break accident scenario in Kaunas city (Lithuania) district heating network has been analyzed. ► An innovative method of pipe break location identification by pressure-time data is proposed.

Boiling Heat Transfer to Halogenated Hydrocarbon Refrigerants

Science.gov (United States)

Yoshida, Suguru; Fujita, Yasunobu

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

Heat exchanger network retrofit optimization involving heat transfer enhancement

International Nuclear Information System (INIS)

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

2012-01-01

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

Approximate model for calculating overall heat transfer between overlying immiscible liquid layers with bubble-induced liquid entrainment

International Nuclear Information System (INIS)

Greene, G.A.; Schwarz, C.E.

1982-01-01

In the event a commercial power reactor is subjected to a Class 9 accident resulting in gross core melting and reactor pressure vessel penetration, it has been shown that the containment integrity may subsequently be threatened by steam overpressurization, combustible gas reactions, and basemat penetration. A major contributor to these events would be the interaction of molten core debris with the structural concrete. Modeling of core-concrete interactions involves many poorly understood and complicated heat transfer phenomena for which there exists a sparse data base. One of these phenomena, which has been shown to have significant impact upon code calculations of core-concrete interactions, is the rate of heat transfer between overlying immiscible layers of core oxides and molten metals whose interface is agitated by transverse gas flow. A mathematical model is developed to analyze this heat transfer

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

A CANDU Severe Accident Analysis

International Nuclear Information System (INIS)

Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie

2006-01-01

As interest in severe accident studies has increased in the last years, we have developed a set of simple models to analyze severe accidents for CANDU reactors that should be integrated in the EU codes. The CANDU600 reactor uses natural uranium fuel and heavy water (D2O) as both moderator and coolant, with the moderator and coolant in separate systems. We chose to analyze accident development for a LOCA with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 10000 deg C, a contact between pressure tube and calandria tube occurs and the residual heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) will be uncovered, then will disintegrate and fall down to the calandria vessel bottom. After all the quantity of moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which normally surrounds the calandria vessel. The phenomena described above are modelled, analyzed and compared with the existing data. The results are encouraging. (authors)

The Natural Convection Heat Transfer inside Vertical Pipe: Characteristic of Pipe Flow according to the Boundary layer

Energy Technology Data Exchange (ETDEWEB)

Ohk, Seung Min; Chung, Bum Jin [Kyunghee University, Yongin (Korea, Republic of)

2016-05-15

The Passive Cooling System (PCS) driven by natural forces drew research attention since Fukushima nuclear power plant accident. This study investigated the natural convection heat transfer inside of vertical pipe with emphasis on the phenomena regarding the boundary layer interaction. Numerical calculations were carried out using FLUENT 6.3. Experiments were performed for the parts of the cases to explore the accuracy of calculation. Based on the analogy, heat transfer experiment is replaced by mass transfer experiment using sulfuric acid copper sulfate (CuSO{sub 4}. H{sub 2}SO{sub 4}) electroplating system. The natural convection heat transfer inside a vertical pipe is studied experimentally and numerically. Experiments were carried out using sulfuric acid-copper sulfate (H{sub 2}SO{sub 4}-CuSO{sub 4}) based on the analogy concept between heat and mass transfer system. Numerical analysis was carried out using FLUENT 6.3. It is concluded that the boundary layer interaction along the flow passage influences the heat transfer, which is affected by the length, diameter, and Prandtl number. For the large diameter and high Prandtl number cases, where the thermal boundary layers do not interfered along the pipe, the heat transfer agreed with vertical flat plate for laminar and turbulent natural convection correlation within 8%. When the flow becomes steady state, the forced convective flow appears in the bottom of the vertical pipe and natural convection flow appears near the exit. It is different behavior from the flow on the parallel vertical flat plates. Nevertheless, the heat transfer was not different greatly compared with those of vertical plate.

Transfer coefficients in a four-cusp duct simulating a typical nuclear reactor channel degraded by accident

International Nuclear Information System (INIS)

Souza Dutra, A. de.

1985-01-01

An experimental study on forced convection in a four-cusp duct simulating a typical nuclear reactor channel degraded by accident is presented. Transfer coefficients were obtained by using the analogy between heat and mass tranfer, with the naphtalene sublimation technique. The experiment consisted in forcing air past a four-cusp naphthalene moulded duct. Mass transfer coefficients were determined in nondimensional form as Sherwood number. Experimental curves correlating the Sherwood number with a nondimensional length, x + , were obtained for Reynolds number varying from 891 to 30.374. This range covers typical flow rates that are expected to exist in a degraded nuclear reactor core. (Author) [pt

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...

Heat transfer characteristics of horizontal steam generators under natural circulation conditions

International Nuclear Information System (INIS)

Hyvaerinen, J.

1996-01-01

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

Heat transfer II essentials

CERN Document Server

REA, The Editors of

1988-01-01

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

Transfer of heat to fluidized-solids beds

Energy Technology Data Exchange (ETDEWEB)

1952-10-16

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

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

Simulation and uncertainties of the heat transfer from a heat-generating DEBRIS bed in the lower plenum

International Nuclear Information System (INIS)

Schaaf, K.; Trambauer, K.

1999-01-01

The findings of the TMI-2 post-accident analyses indicated that internal cooling mechanisms may have a considerable potential to sustain the vessel integrity after a relocation of core material to the lower plenum, provided that water is continuously available in the RPV. Numerous analytical and experimental research activities are currently underway in this respect. This paper illustrates some major findings of the experimental work on internal cooling mechanisms and describes the limitations and the uncertainties in the simulation of the heat transfer processes. Reference is made especially to the joint German DEBRIS/ RPV research program, which encompasses the experimental investigation of the thermal-hydraulics in gaps, of the heat transfer within a particulate debris bed, and of the high temperature performance of vessel steel, as well as the development of simulation models for the heat transfer in the lower head and the structural response of the RPV. In particular, the results of uncertainty and sensitivity analyses are presented, which have been carried out at GRS using an integral model that describes the major phenomena governing the long-term integrity of the reactor vessel. The investigation of a large-scale relocation indicated that the verification of a gap cooling mechanism as an inherent mechanism is questionable in terms of a stringent probabilistic uncertainty criterion, as long as the formation of a large molten pool cannot be excluded. (author)

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.

Post-accident fuel relocation and heat removal in the LMFBR

International Nuclear Information System (INIS)

Kazimi, M.S.; Tsai, S.S.; Gasser, R.D.

1976-08-01

Assessment of the dynamics of post-accident fuel relocation and heat removal is an important aspect of the evaluation of the consequences of a hypothetical accident in an LMFBR. Such an assessment is of particular importance in the evaluation of the post-accident radiological doses around the reactor site. In the present evaluation particular attention is given to the design features of the Clinch River Breeder Reactor Plant (CRBR). Fuel relocation and heat removal, assuming certain conditions have resulted in core disruption, are discussed. The discussion of events and phenomena involved in the relocation processes is centered around the resulting patterns of heat source distribution. The factors influencing fuel relocation and distribution in the inlet and outlet plena of the reactor vessel are discussed. The current technology of in-vessel heat removal is applied to the design of the CRBR reactor. Both fuel debris cooling limits and overall coolant flow in the reactor under natural convection conditions are explored. Some of the uncertainties in ex-vessel fuel behavior are addressed. In particular, the effect of melting the cavity bed on the rate of growth of a molten fuel pool is investigated

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

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.

Flow visualization study of inverted annular flow of post dryout heat transfer region

International Nuclear Information System (INIS)

Ishii, M.; De Jarlais, G.

1985-01-01

The inverted annular flow is important in the area of LWR accident analysis in terms of the maximum cladding temperature and effectiveness of the emergency core cooling. However, the inverted annular flow thermal-hydraulics is not well understood due to its special heat transfer condition of film boiling. The review of existing data indicates further research is needed in the areas of basic hydrodynamics related to liquid core disintegration mechanisms, slug and droplet formation, entrainment, and droplet size distributions. In view of this, the inverted flow is studied in detail experimentally. A new experimental apparatus has been constructed in which film boiling heat transfer can be established in a transparent test section. The test section consists of two coaxial quartz tubes. The annular gap between these two tubes is filled with a hot, clear fluid (syltherm 800) so as to maintain film boiling temperatures and heat transfer rates at the inner quartz tube wall. Data on liquid core stability, core break-up mechanism, and dispersed-core liquid slug and droplet sizes are obtained using F 113 as a test fluid. Both high speed movies and flash photographs (3 μsec) are used

Boiling heat transfer on horizontal tube bundles

International Nuclear Information System (INIS)

Anon.

1987-01-01

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

Heat transfer and fire spread

Science.gov (United States)

Hal E. Anderson

1969-01-01

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

Heat transfer in heterogeneous propellant combustion systems

International Nuclear Information System (INIS)

Brewster, M.Q.

1992-01-01

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

Condensation heat transfer in plate heat exchangers

International Nuclear Information System (INIS)

Panchal, C.B.

1985-01-01

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

Fluid dynamics and heat transfer methods for the TRAC code

International Nuclear Information System (INIS)

Reed, W.H.; Kirchner, W.L.

1977-01-01

A computer code called TRAC is being developed for analysis of loss-of-coolant accidents and other transients in light water reactors. This code involves a detailed, multidimensional description of two-phase flow coupled implicitly through appropriate heat transfer coefficients with a simulation of the temperature field in fuel and structural material. Because TRAC utilizes about 1000 fluid mesh cells to describe an LWR system, whereas existing lumped parameter codes typically involve fewer than 100 fluid cells, we have developed new highly implicit difference techniques that yield acceptable computing times on modern computers. Several test problems for which experimental data are available, including blowdown of single pipe and loop configurations with and without heated walls, have been computed with TRAC. Excellent agreement with experimental results has been obtained. (author)

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

International Nuclear Information System (INIS)

Gunes, M.

1998-01-01

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

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

Science.gov (United States)

Nadolny, Zbigniew; Gościński, Przemysław; Bródka, Bolesław

2017-10-01

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

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

Directory of Open Access Journals (Sweden)

Nadolny Zbigniew

2017-01-01

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

Heat transfer pipe shielding device for heat exchanger

International Nuclear Information System (INIS)

Hanawa, Jun.

1991-01-01

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

Heat transfer in two-phase flow of helium

International Nuclear Information System (INIS)

Subbotin, V.I.; Deev, V.I.; Solodovnikov, V.V.; Arkhipov, V.V.

1986-01-01

The results of experimental study of heat transfer in two-phase helium flow are presented. The effect of operating parameters (pressure, mass velocity, heat flux and quality) on boiling heat transfer intensity was investigated. A significant influence of boiling process prehistory on heat transfer coefficients was demonstrated. On the basis of experimental data obtained three typical regimes of flow boiling heat transfer were found. Analogy of heat transfer in flow boiling and pool boiling of helium and noncryogenic liquids was established. Correlations were developed which are in close agreement with available heat transfer data

Heat transfer from internally-heated molten UO2 pools

International Nuclear Information System (INIS)

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

1978-01-01

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

Theory of Periodic Conjugate Heat Transfer

CERN Document Server

Zudin, Yuri B

2012-01-01

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

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

International Nuclear Information System (INIS)

Rosenfeld, J.H.

1987-01-01

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

Heat exchanger device and method for heat removal or transfer

Science.gov (United States)

Koplow, Jeffrey P

2013-12-10

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

Reflooding phase of the LOCA - state of the art I. Heat transfer and fluid flow during reflooding

International Nuclear Information System (INIS)

Yadigaroglu, G.

1977-01-01

Complex heat transfer processes take place during the reflooding phase of the Loss-of-Coolant Accident in Light-Water Reactors. Reflooding experiments conducted with simple single-channel geometries (round tubes and annuli) and with rod bundles are reviewed. The experimental findings and various parametric trends are critically discussed, explained, and summarized. Analytical methods that are in use in safety analysis and features of advanced models that have been proposed are outlined. These advanced models attempt to solve the conservation equations in the core channels in order to find the local coolant conditions. The values of the heat transfer coefficients are related to local parameters to arrive at cladding temperature predictions

Preliminary Numerical Analysis of Convective Heat Transfer Loop Using MARS Code

Energy Technology Data Exchange (ETDEWEB)

Lee, Yongjae; Seo, Gwang Hyeok; Jeun, Gyoodong; Kim, Sung Joong [Hanyang Univ., Seoul (Korea, Republic of)

2014-05-15

The MARS has been developed adopting two major modules: RELAP5/MOD3 (USA) for one-dimensional (1D) two-fluid model for two-phase flows and COBRA-TF code for a three-dimensional (3D), two-fluid, and three-field model. In addition to the MARS code, TRACE (USA) is a modernized thermal-hydraulics code designed to consolidate and extend the capabilities of NRC's 3 legacy safety code: TRAC-P, TRAC-B and RELAP. CATHARE (French) is also thermal-hydraulic system analysis code for Pressurized Water Reactor (PWR) safety. There are several researches on comparing experimental data with simulation results by the MARS code. Kang et al. conducted natural convection heat transfer experiments of liquid gallium loop, and the experimental data were compared to MARS simulations. Bang et al. examined the capability of the MARS code to predict condensation heat transfer experiments with a vertical tube containing a non-condensable gas. Moreover, Lee et al. adopted MELCOR, which is one of the severe accident analysis codes, to evaluate several strategies for the severe accident mitigation. The objective of this study is to conduct the preliminary numerical analysis for the experimental loop at HYU using the MARS code, especially in order to provide relevant information on upcoming experiments for the undergraduate students. In this study, the preliminary numerical analysis for the convective heat transfer loop was carried out using the MARS Code. The major findings from the numerical simulations can be summarized as follows. In the calculations of the outlet and surface temperatures, the several limitations were suggested for the upcoming single-phase flow experiments. The comparison work for the HTCs shows validity for the prepared input model. This input could give useful information on the experiments. Furthermore, the undergraduate students in department of nuclear engineering, who are going to be taken part in the experiments, could prepare the program with the input, and will

Preliminary Numerical Analysis of Convective Heat Transfer Loop Using MARS Code

International Nuclear Information System (INIS)

Lee, Yongjae; Seo, Gwang Hyeok; Jeun, Gyoodong; Kim, Sung Joong

2014-01-01

The MARS has been developed adopting two major modules: RELAP5/MOD3 (USA) for one-dimensional (1D) two-fluid model for two-phase flows and COBRA-TF code for a three-dimensional (3D), two-fluid, and three-field model. In addition to the MARS code, TRACE (USA) is a modernized thermal-hydraulics code designed to consolidate and extend the capabilities of NRC's 3 legacy safety code: TRAC-P, TRAC-B and RELAP. CATHARE (French) is also thermal-hydraulic system analysis code for Pressurized Water Reactor (PWR) safety. There are several researches on comparing experimental data with simulation results by the MARS code. Kang et al. conducted natural convection heat transfer experiments of liquid gallium loop, and the experimental data were compared to MARS simulations. Bang et al. examined the capability of the MARS code to predict condensation heat transfer experiments with a vertical tube containing a non-condensable gas. Moreover, Lee et al. adopted MELCOR, which is one of the severe accident analysis codes, to evaluate several strategies for the severe accident mitigation. The objective of this study is to conduct the preliminary numerical analysis for the experimental loop at HYU using the MARS code, especially in order to provide relevant information on upcoming experiments for the undergraduate students. In this study, the preliminary numerical analysis for the convective heat transfer loop was carried out using the MARS Code. The major findings from the numerical simulations can be summarized as follows. In the calculations of the outlet and surface temperatures, the several limitations were suggested for the upcoming single-phase flow experiments. The comparison work for the HTCs shows validity for the prepared input model. This input could give useful information on the experiments. Furthermore, the undergraduate students in department of nuclear engineering, who are going to be taken part in the experiments, could prepare the program with the input, and will

Visualisation of heat transfer in laminar flows

NARCIS (Netherlands)

Speetjens, M.F.M.; Steenhoven, van A.A.

2009-01-01

Heat transfer in fluid flows traditionally is examined in terms of temperature field and heat-transfer coefficients at non-adiabatic walls. However, heat transfer may alternatively be considered as the transport of thermal energy by the total convective-conductive heat flux in a way analogous to the

A heat transfer textbook

CERN Document Server

Lienhard, John H

2011-01-01

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

Elementary heat transfer analysis

CERN Document Server

Whitaker, Stephen; Hartnett, James P

1976-01-01

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

Heat and mass transfer

CERN Document Server

Karwa, Rajendra

2017-01-01

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

Heat Transfer Basics and Practice

CERN Document Server

Böckh, Peter

2012-01-01

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

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)

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

International Nuclear Information System (INIS)

Stafford, Jason; Walsh, Ed; Egan, Vanessa

2009-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-03-15

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

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.

ORNL rod-bundle heat-transfer test data. Volume 7. Thermal-Hydraulic Test Facility experimental data report for test series 3.07.9 - steady-state film boiling in upflow

International Nuclear Information System (INIS)

Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

1982-05-01

Thermal-Hydraulic Test Facility (THTF) test series 3.07.9 was conducted by members of the Oak Ridge National Laboratory Pressurized-Water Reactor (ORNL-PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on September 11, September 18, and October 1, 1980. The objective of the program is to investigate heat transfer phenomena believed to occur in PWRs during accidents, including small- and large-break loss-of-coolant accidents. Test series 3.07.9 was designed to provide steady-state film boiling data in rod bundle geometry under reactor accident-type conditions. This report presents the reduced instrument responses for THTF test series 3.07.9. Also included are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers

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…

"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

Fission products transport in CANDU Primary Heat Transport System in a severe accident

International Nuclear Information System (INIS)

Constantin, M.; Rizoiu, A.; Turcu, I.; Negut, Gh.

2005-01-01

Full text: The paper is intended to analyse the distribution of the fission products (FPs) in CANDU Primary Heat Transport (PHT) System by using the ASTEC code (Accident Source Term Evaluation Code). The complexity of the data required by ASTEC and the complexity of CANDU PHT were strong motivation to begin with a simplified geometry in order to avoid the introducing of unmanageable errors at the level of input deck. Thus only 1/4 of the PHT circuit was simulated, an simplified FPs inventory and some simplifications in the feeders geometry were also used. The circuit consists of 95 horizontal fuel channels connected to 95 horizontal out-feeders, then through vertical feeders to the outlet-header (a big pipe that collects the water from feeders); the circuit continues from the outlet-header with a riser and then with the steam generator and a pump. After this pump, the circuit was broken; in this point the FPs are transferred to the containment. The data related to the nodes' definitions, temperatures and pressure conditions were chosen as possible as real data from CANDU NPP loss of coolant accident sequence. Temperature and pressure conditions in the time of the accident were calculated by CATHENA code and the source term of FPs introduced into the PHT was estimated by ORIGEN code. The results consist of mass distributions in the nodes of the circuit and the mass transfer to the containment through the break for different species (FPs and chemical species). The study is completed by sensitivity analysis for the parameters with important uncertainties. (authors)

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

International Nuclear Information System (INIS)

Cheng, XueTao; Liang, XinGang

2013-01-01

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

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...

ORNL rod-bundle heat-transfer test data. Volume 3. Thermal-hydraulic test facility experimental data report for test 3.06.6B - transient film boiling in upflow

International Nuclear Information System (INIS)

Mullins, C.B.; Felde, D.K.; Sutton, A.G.; Gould, S.S.; Morris, D.G.; Robinson, J.J.

1982-05-01

Reduced instrument responses are presented for Thermal-Hyraulic Test Facility (THTF) Test 3.06.6B. This test was conducted by members of the Oak Ridge National Laboratory Pressurized-Water-Reactor (PWR) Blowdown Heat Transfer (BDHT) Separate-Effects Program on August 29, 1980. The objective of the program was to investigate heat transfer phenomena believed to occur in PWR's during accidents, including small and large break loss-of-coolant accidents. Test 3.06.6B was conducted to obtain transient film boiling data in rod bundle geometry under reactor accident-type conditions. The primary purpose of this report is to make the reduced instrument responses for THTF Test 3.06.6B available. Included in the report are uncertainties in the instrument responses, calculated mass flows, and calculated rod powers

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

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

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...

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...

Key Characteristics of Combined Accident including TLOFW accident for PSA Modeling

Energy Technology Data Exchange (ETDEWEB)

Kim, Bo Gyung; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [Khalifa University of Science, Technology and Research, Abu Dhabi (United Arab Emirates)

2015-05-15

The conventional PSA techniques cannot adequately evaluate all events. The conventional PSA models usually focus on single internal events such as DBAs, the external hazards such as fire, seismic. However, the Fukushima accident of Japan in 2011 reveals that very rare event is necessary to be considered in the PSA model to prevent the radioactive release to environment caused by poor treatment based on lack of the information, and to improve the emergency operation procedure. Especially, the results from PSA can be used to decision making for regulators. Moreover, designers can consider the weakness of plant safety based on the quantified results and understand accident sequence based on human actions and system availability. This study is for PSA modeling of combined accidents including total loss of feedwater (TLOFW) accident. The TLOFW accident is a representative accident involving the failure of cooling through secondary side. If the amount of heat transfer is not enough due to the failure of secondary side, the heat will be accumulated to the primary side by continuous core decay heat. Transients with loss of feedwater include total loss of feedwater accident, loss of condenser vacuum accident, and closure of all MSIVs. When residual heat removal by the secondary side is terminated, the safety injection into the RCS with direct primary depressurization would provide alternative heat removal. This operation is called feed and bleed (F and B) operation. Combined accidents including TLOFW accident are very rare event and partially considered in conventional PSA model. Since the necessity of F and B operation is related to plant conditions, the PSA modeling for combined accidents including TLOFW accident is necessary to identify the design and operational vulnerabilities.The PSA is significant to assess the risk of NPPs, and to identify the design and operational vulnerabilities. Even though the combined accident is very rare event, the consequence of combined

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

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

Science.gov (United States)

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

2017-11-01

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

Effect of surface etching on condensing heat transfer

Energy Technology Data Exchange (ETDEWEB)

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

2016-02-15

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

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

Science.gov (United States)

Dobran, Flavio

1989-08-01

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

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

International Nuclear Information System (INIS)

Shepherd, I.M.

1982-01-01

The computer code BUSH has been developed for the calculation of steady state heat transfer in a rod bundle. For a given power, flow and geometry it can calculate the temperatures in the rods, coolant and shroud assuming that at any axial level each rod can be described by one temperature and the coolant fluid is also radially uniform at this level. Heat transfer by convection and radiation are handled and the geometry is flexible enough to model nearly all types of envisaged shroud design for the SUPERSARA test series. The modular way in which BUSH has been written makes it suitable for future development, either within the present BUSH framework or as part of a more advanced code

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.

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.)

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

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

International Nuclear Information System (INIS)

Wang, Y; Ding, G-F

2008-01-01

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

Heat transfer capability analysis of heat pipe for space reactor

International Nuclear Information System (INIS)

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

2015-01-01

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

Study on heat transfer and fluid flow in the stand pipe rupture accident

International Nuclear Information System (INIS)

Fumizawa, Motoo; Hishida, Makoto

1991-09-01

This paper deals with an experimental investigation of the buoyancy driven exchange flow which takes place through a narrow cylindrical channel, during the stand pipe rupture accident in a high temperature gas-cooled reactor (HTGR). The velocity distribution through the cylindrical channel is measured by a laser Doppler velocimeter, in order to evaluate the air ingress flow rate. The experiments are performed under atmospheric pressure with nitrogen as a working fluid. Rayleigh number ranges from 1.3 x 10 7 to 7.0 x 10 7 . The following conclusions were obtained: (1) The laser Doppler velocimeter was found a good method for the measurement of the velocity of the exchange flow. (2) When the temperature of the hemisphere and the bottom heated plate, which simulate the top cover of the reactor, was kept uniform, the volumetric exchange flow rate agreed well with Epstein's result. (3) The exchange flow through a narrow cylindrical channel fluctuated irregularly with time and space. (author)

Summary of flow and heat transfer in RPV under PTS

International Nuclear Information System (INIS)

Lu Donghua; Wang Haijun; Chen Tingkuan; Luo Yushan

2003-01-01

PTS under loss of coolant accident (LOCA) has great effect on the safety of RPV. Many research works focusing on flow and heat transfer in RPV under PTS have been done in developed countries for many years, and a lot of results have been got both on experiment and numerical simulation. The safety of nuclear power plant is enhanced greatly by these research works. With the developing of nuclear power technology in China, RPV integration under PTS has been studied. The author summarizes research works at home and abroad in recent years. The problems existed in present work and research direction in the future are discussed

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)

Conjugate heat and mass transfer in heat mass exchanger ducts

CERN Document Server

Zhang, Li-Zhi

2013-01-01

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

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

International Nuclear Information System (INIS)

Green, G.A.

1982-08-01

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

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

Science.gov (United States)

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

2016-08-01

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

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.

On the heat transfer correlation for membrane distillation

International Nuclear Information System (INIS)

Wang, Chi-Chuan

2011-01-01

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

Simulation of LOF accidents with directly electrical heated UO2 pins

International Nuclear Information System (INIS)

Alexas, A.

1976-01-01

The behavior of directly electrical heated UO 2 pins has been investigated under loss of coolant conditions. Two types of hypothetical accidents have been simulated, first, a LOF accident without power excursion (LOF accident) and second, a LOF accident with subsequent power excursion (LOF-TOP accident). A high-speed film shows the sequence of events for two characteristic experiments. In consequence of the high-speed film analysis as well as the metallographical evaluation statements are given in respect to the cladding meltdown process, the fuel melt fraction and the energy input from the beginning of a power transient to the beginning of the molten fuel ejections

Finite-elements modeling of radiant heat transfers between mobile surfaces; Modelisation par elements finis de transferts radiatifs entre surfaces mobiles

Energy Technology Data Exchange (ETDEWEB)

Daurelle, J V; Cadene, V; Occelli, R [Universite de Provence, 13 - Marseille (France)

1997-12-31

In the numerical modeling of thermal industrial problems, radiant heat transfers remain difficult to take into account and require important computer memory and long computing time. These difficulties are enhanced when radiant heat transfers are coupled with finite-elements diffusive heat transfers because finite-elements architecture is complex and requires a lot of memory. In the case of radiant heat transfers along mobile boundaries, the methods must be optimized. The model described in this paper concerns the radiant heat transfers between diffuse grey surfaces. These transfers are coupled with conduction transfers in the limits of the diffusive opaque domain. 2-D and 3-D geometries are analyzed and two configurations of mobile boundaries are considered. In the first configuration, the boundary follows the deformation of the mesh, while in the second, the boundary moves along the fixed mesh. Matter displacement is taken into account in the term of transport of the energy equation, and an appropriate variation of the thermophysical properties of the transition elements between the opaque and transparent media is used. After a description of the introduction of radiative limit conditions in a finite-elements thermal model, the original methods used to optimize calculation time are explained. Two examples of application illustrate the approach used. The first concerns the modeling of radiant heat transfers between fuel rods during a reactor cooling accident, and the second concerns the study of heat transfers inside the air-gap of an electric motor. The method of identification of the mobile surface on the fixed mesh is described. (J.S.) 12 refs.

Finite-elements modeling of radiant heat transfers between mobile surfaces; Modelisation par elements finis de transferts radiatifs entre surfaces mobiles

Energy Technology Data Exchange (ETDEWEB)

Daurelle, J.V.; Cadene, V.; Occelli, R. [Universite de Provence, 13 - Marseille (France)

1996-12-31

In the numerical modeling of thermal industrial problems, radiant heat transfers remain difficult to take into account and require important computer memory and long computing time. These difficulties are enhanced when radiant heat transfers are coupled with finite-elements diffusive heat transfers because finite-elements architecture is complex and requires a lot of memory. In the case of radiant heat transfers along mobile boundaries, the methods must be optimized. The model described in this paper concerns the radiant heat transfers between diffuse grey surfaces. These transfers are coupled with conduction transfers in the limits of the diffusive opaque domain. 2-D and 3-D geometries are analyzed and two configurations of mobile boundaries are considered. In the first configuration, the boundary follows the deformation of the mesh, while in the second, the boundary moves along the fixed mesh. Matter displacement is taken into account in the term of transport of the energy equation, and an appropriate variation of the thermophysical properties of the transition elements between the opaque and transparent media is used. After a description of the introduction of radiative limit conditions in a finite-elements thermal model, the original methods used to optimize calculation time are explained. Two examples of application illustrate the approach used. The first concerns the modeling of radiant heat transfers between fuel rods during a reactor cooling accident, and the second concerns the study of heat transfers inside the air-gap of an electric motor. The method of identification of the mobile surface on the fixed mesh is described. (J.S.) 12 refs.

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

Subcooled boiling heat transfer on a finned surface

International Nuclear Information System (INIS)

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

1992-01-01

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

Engineering calculations in radiative heat transfer

CERN Document Server

Gray, W A; Hopkins, D W

1974-01-01

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

Industrial furnace with improved heat transfer

Energy Technology Data Exchange (ETDEWEB)

Hoetzl, M.; Lingle, T.M.

1992-07-07

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

Liquid metal heat transfer in heat exchangers under low flow rate conditions

International Nuclear Information System (INIS)

Mochizuki, Hiroyasu

2015-01-01

The present paper describes the liquid metal heat transfer in heat exchangers under low flow rate conditions. Measured data from some experiments indicate that heat transfer coefficients of liquid metals at very low Péclet number are much lower than what are predicted by the well-known empirical relations. The cause of this phenomenon was not fully understood for many years. In the present study, one countercurrent-type heat exchanger is analyzed using three, separated countercurrent heat exchanger models: one is a heat exchanger model in the tube bank region, while the upper and lower plena are modeled as two heat exchangers with a single heat transfer tube. In all three heat exchangers, the same empirical correlation is used in the heat transfer calculation on the tube and the shell sides. The Nusselt number, as a function of the Péclet number, calculated from measured temperature and flow rate data in a 50 MW experimental facility was correctly reproduced by the calculation result, when the calculated result is processed in the same way as the experiment. Finally, it is clarified that the deviation is a superficial phenomenon which is caused by the heat transfer in the plena of the heat exchanger. (author)

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

Thermal hydraulics of CANDU severe accident analysis

International Nuclear Information System (INIS)

Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie; Dupleac, Daniel

2007-01-01

As interest in severe accident studies has increased in the last years, we have developed a set of simple models to analyze severe accidents in CANDU reactors that should be integrated in the EU codes. The CANDU600 reactor uses natural uranium fuel and heavy water (D 2 O) as both moderator and coolant, with the moderator and coolant in separate systems. We chose to analyze accident development for a LOCA with simultaneous loss of moderator cooling and the loss of emergency core cooling system (ECCS). This type of accident is likely to modify the reactor geometry and will lead to a severe accident development. When the coolant temperatures inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the residual heat is transferred to the moderator. Due to the lack of cooling, the moderator eventually begins to boil and is expelled, through the calandria vessel relief ducts, into the containment. Therefore the calandria tubes (fuel channels) will be uncovered, then will disintegrate and fall down to the calandria vessel bottom. After all the quantity of moderator is vaporized and expelled, the debris will heat up and eventually boil. The heat accumulated in the molten debris will be transferred through the calandria vessel wall to the shield tank water, which normally surrounds the calandria vessel. The phenomena described above are modelled, analyzed and compared with the available data. The results are encouraging. (authors)

Numerical simulation of fluid flow and heat transfer in a concentric tube heat exchanger

International Nuclear Information System (INIS)

Mokamati, S.V.; Prasad, R.C.

2003-01-01

In this paper, numerical simulation of a concentric tube heat exchanger is presented to determine the convective heat transfer coefficient and friction factor in a smooth tube. Increasing the convective heat transfer coefficient can increase heat transfer rate in a concentric tube heat exchanger from a given tubular surface area. This can be achieved by using heat transfer augmentation devices. This work constitutes the initial phase of the numerical simulation of heat transfer from tubes employing augmentation devices, such as twisted tapes, wire-coil inserts, for heat transfer enhancement. A computational fluid dynamics (CFD) simulation tool was developed with CFX software and the results obtained from the simulations are validated with the empirical correlations for a smooth tube heat exchanger. The difficulties associated with the simulation of a heat exchanger augmented with wire-coil inserts are discussed. (author)

Visualisation of heat transfer in unsteady laminar flows

NARCIS (Netherlands)

Speetjens, M.F.M.; Steenhoven, van A.A.

2011-01-01

Heat transfer in fluid flows traditionally is examined in terms of temperature fields and heat-transfer coefficients. However, heat transfer may alternatively be considered as the transport of thermal energy by the total convective-conductive heat flux in a way analogous to the transport of fluid by

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.

Heat transfer analysis of parabolic trough solar receiver

International Nuclear Information System (INIS)

Padilla, Ricardo Vasquez; Demirkaya, Gokmen; Goswami, D. Yogi; Stefanakos, Elias; Rahman, Muhammad M.

2011-01-01

Highlights: → In this paper a detailed one dimensional numerical heat transfer analysis of a PTC is performed. → The receiver and envelope were divided into several segments and mass and energy balance were applied in each segment. → Improvements either in the heat transfer correlations or radiative heat transfer analysis are presented. → The proposed heat transfer model was validated with experimental data obtained from Sandia National Laboratory. → Our results showed a better agreement with experimental data compared to other models. -- Abstract: Solar Parabolic Trough Collectors (PTCs) are currently used for the production of electricity and applications with relatively higher temperatures. A heat transfer fluid circulates through a metal tube (receiver) with an external selective surface that absorbs solar radiation reflected from the mirror surfaces of the PTC. In order to reduce the heat losses, the receiver is covered by an envelope and the enclosure is usually kept under vacuum pressure. The heat transfer and optical analysis of the PTC is essential to optimize and understand its performance under different operating conditions. In this paper a detailed one dimensional numerical heat transfer analysis of a PTC is performed. The receiver and envelope were divided into several segments and mass and energy balance were applied in each segment. Improvements either in the heat transfer correlations or radiative heat transfer analysis are presented as well. The partial differential equations were discretized and the nonlinear algebraic equations were solved simultaneously. Finally, to validate the numerical results, the model was compared with experimental data obtained from Sandia National Laboratory (SNL) and other one dimensional heat transfer models. Our results showed a better agreement with experimental data compared to other models.

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....

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).

Main results of assessing integrity of RNPP-3 steam generator heat exchange tubes in accident management

International Nuclear Information System (INIS)

Shugajlo, Al-j P.; Mustafin, M.A.; Shugajlo, Al-r P.; Ryzhov, D.I.; Zhabin, O.I.

2017-01-01

Tubes integrity evaluation under accident conditions considering drain of SG and current technical state of steam exchange tubes is an important question regarding SG long-term operation and improvement of accident management strategy.The main investigation results prepared for heat exchange surface of RNPP-3 steam generator are presented in this research aimed at assessing integrity of heat exchange tubes under accident conditions, which lead to full or partial drain of heat exchange surface, in particular during station blackout.

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)

Heat transfer to a dispersed two phase flow and detailed quench front velocity research

International Nuclear Information System (INIS)

De Boer, T.C.; Van der Molen, S.B.

1985-01-01

During the blow-down phase of a loss-off coolant accident (LOCA) in a pressurized water reactor the core will heat up dramatically. Water will be injected in the system, and by bottom flooding the core will be cooled. The use of one-dimensional computer models for the calculation of the reflood process in a bundle needs a better justification. The influence of an unheated shroud on prequench heat transfer is investigated in a tube, an annulus and a 4 rod bundle. By using a glass shroud for the annulus, optical analysis of the dispersed two-phase flow regime has been performed. The ECN 36-rod bundle tests as performed with axial uniform power profile are reflood and boil-down at 0.2 MPa pressure executed for different conditions. The experiment yield a data base suitable for code validation and development. Better understanding is obtained for the influence of the radial non-uniform temperature and/or power distributions on the reflood process. Heat transfer improvement induced by the presence of spacer grids is observed. 72 refs.; 220 figs.

Heat transfer characteristics and limitations analysis of heat-pipe-cooled thermal protection structure

International Nuclear Information System (INIS)

Guangming, Xiao; Yanxia, Du; Yewei, Gui; Lei, Liu; Xiaofeng, Yang; Dong, Wei

2014-01-01

The theories of heat transfer, thermodynamics and fluid dynamics are employed to develop the coupled heat transfer analytical methods for the heat-pipe-cooled thermal protection structure (HPC TPS), and a three-dimensional numerical method considering the sonic limit of heat pipe is proposed. To verify the calculation correctness, computations are carried out for a typical heat pipe and the results agree well with experimental data. Then, the heat transfer characteristics and limitations of HPC TPS are mainly studied. The studies indicate that the use of heat pipe can reduce the temperature at high heat flux region of structure efficiently. However, there is a frozen startup period before the heat pipe reaching a steady operating state, and the sonic limit will be a restriction on the heat transfer capability. Thus, the effects of frozen startup must be considered for the design of HPC TPS. The simulation model and numerical method proposed in this paper can predict the heat transfer characteristics of HPC TPS quickly and exactly, and the results will provide important references for the design or performance evaluation of HPC TPS. - Highlights: • Numerical methods for the heat-pipe-cooled thermal protection structure are studied. • Three-dimensional simulation model considering sonic limit of heat pipe is proposed. • The frozen startup process of the embedded heat pipe can be predicted exactly. • Heat transfer characteristics of TPS and limitations of heat pipe are discussed

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Post-accident heat removal research: A state of the art review

International Nuclear Information System (INIS)

Mueller, U.; Schulenberg, T.

1983-11-01

For a realistic assessment of the consequence of extremely unlikely reactor accidents resulting in core degradation or core meltdown key questions are how to remove the decay heat from the reactor system and how to retain the radioactive core debris within the containment. Usually, this complex of questions is referred to as Post-Accident Heat Removal (PAHR). In this article the research work on PAHR performed by various institutions during the last decade has been reviewed. The main results have been summarized under the chapter headings ''Accident Scenarios,'' - ''Core Debris Accommodation Concepts,'' and ''PAHR Topics.'' Particular emphasis has been placed on the presentation of the following problems: characteristics and coolability of solid core debris in the vector vessel, heat removal from molten pools of core material, and core-melt interaction with structural materials. Some unresolved or insufficiently answered questions relating to special ''PAHR Topics'' have been mentioned or discussed at the end of the particular Chapter. Problem areas of major uncertainty have been identified and listed at the end of the review article. They include the following subjects: formation of debris beds and bed characteristics, post dryout behaviour of particle beds, long-term availability and proper location of heat sinks, creep rupture of structures under high thermal loads. (orig.) [de

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.

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

Flow and heat transfer in a curved channel

Science.gov (United States)

Brinich, P. F.; Graham, R. W.

1977-01-01

Flow and heat transfer in a curved channel of aspect ratio 6 and inner- to outer-wall radius ratio 0.96 were studied. Secondary currents and large longitudinal vortices were found. The heat-transfer rates of the outer and inner walls were independently controlled to maintain a constant wall temperature. Heating the inner wall increased the pressure drop along the channel length, whereas heating the outer wall had little effect. Outer-wall heat transfer was as much as 40 percent greater than the straight-channel correlation, and inner-wall heat transfer was 22 percent greater than the straight-channel correlation.

Non-Uniform Heat Transfer in Thermal Regenerators

DEFF Research Database (Denmark)

Jensen, Jesper Buch

, a numerical model, which simulates a single-blow operation in a parallel-plate regenerator, was developed and used to model the heat transfer under various conditions. In addition to the modeling of the heat transfer, a series of experiments on passive regenerators with non-uniform, but precisely controlled....... Additionally, the experiments gave real comparative results, whereas the model to a certain degree more served to provide insight to the heat transfer processes taking place inside the regenera- tors, something that would be - if not impossible - then highly impractical to do experimentally. It has been found......This thesis presents investigations on the heat transfer in complex heat ex- changers in general and in regenerative heat exchangers (regenerators) in par- ticular. The motivation for this work is a result of inconsistencies obeserved in the results from a series of experiments on active magnetic...

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.

Aerodynamics, heat and mass transfer in steam-aerosol turbulent flows in containment

Energy Technology Data Exchange (ETDEWEB)

Nigmatulin, B.I.; Pershukov, V.A.; Ris, V.V. [Research & Engineering Centre of Nuclear Plants Safety, Moscow (Russian Federation)] [and others

1995-09-01

In this report an analysis of aerodynamic and heat transfer processes at the blowdown of gas-dispersed mixture into the containment volume is presented. A few models for description of the volume averaged and local characteristics are analyzed. The mathematical model for description of the local characteristics of the turbulent gas-dispersed flows was developed. The calculation of aerodynamic, heat and mass transfer characteristics was based on the Navier-Stokes, energy and gas mass fractions conservation equations. For calculation of dynamics and deposition of the aerosols the original diffusion-inertia model is developed. The pulsating characteristics of the gaseous phase were calculated on the base (k-{xi}) model of turbulence with modification to account thermogravitational force action and influence of particle mass loading. The appropriate boundary conditions using the {open_quotes}near-wall function{close_quotes} approach was obtained. Testing of the mathematical models and boundary conditions has shown a good agreement between computation and data of comparison. The described mathematical models were applied to two- and three dimensional calculations of the turbulent flow in containment at the various stages of the accident.

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

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

Heat transfer in Rockwool modelling and method of measurement. The effect of natural convection on 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 large thickness dimensions the resulting heat transfer through the

Heat transfer unit and method for prefabricated vessel

Science.gov (United States)

Tamburello, David A.; Kesterson, Matthew R; Hardy, Bruce J.

2017-11-07

Vessel assemblies, heat transfer units for prefabricated vessels, and methods for heat transfer prefabricated vessel are provided. A heat transfer unit includes a central rod, and a plurality of peripheral rods surrounding the central rod and connected to the central rod. The plurality of peripheral rods are movable between a first collapsed position and a second bowed position, wherein in the second bowed position a midpoint of each of the plurality of peripheral rods is spaced from the central rod relative to in the first position. The heat transfer unit further includes a heat transfer element connected to one of the plurality of peripheral rods.

Numerical simulation of heat transfer in metal foams

Science.gov (United States)

Gangapatnam, Priyatham; Kurian, Renju; Venkateshan, S. P.

2018-02-01

This paper reports a numerical study of forced convection heat transfer in high porosity aluminum foams. Numerical modeling is done considering both local thermal equilibrium and non local thermal equilibrium conditions in ANSYS-Fluent. The results of the numerical model were validated with experimental results, where air was forced through aluminum foams in a vertical duct at different heat fluxes and velocities. It is observed that while the LTE model highly under predicts the heat transfer in these foams, LTNE model predicts the Nusselt number accurately. The novelty of this study is that once hydrodynamic experiments are conducted the permeability and porosity values obtained experimentally can be used to numerically simulate heat transfer in metal foams. The simulation of heat transfer in foams is further extended to find the effect of foam thickness on heat transfer in metal foams. The numerical results indicate that though larger foam thicknesses resulted in higher heat transfer coefficient, this effect weakens with thickness and is negligible in thick foams.

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)

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

Heat transfer between relocated materials and the RPV lower head

International Nuclear Information System (INIS)

Rempe, J.L.; Knudson, D.L.; Kohriyama, T.

2001-01-01

Questions about the coolability of a continuous mass of relocated corium were raised during the Three Mile Island Unit 2 (TMI-2) Vessel Investigation Project (VIP) Post-accident examinations indicate that nearly half of the material that relocated to the vessel lower head during the TMI-2 accident formed a cohesive or ''continuous'' layer. TMI-2 VIP results and other evidence suggest that conduction through this continuous layer of solidified corium materials was assisted by other cooling mechanisms. Because increased knowledge about in-vessel coolability of corium materials may assist reactor designers in demonstrating that their concepts are passively safe, there is international interest in this topic. However, data are needed to identify what cooling mechanism(s) occurred and to develop a validated model for predicting this cooling. Corium cooling models significantly impact predictions for subsequent accident progression, such as the estimated time and mode of vessel failure. Hence, improved cooling models will provide a much needed, missing component of severe accident analyses. This paper provides a critical review of research investigating the coolability of corium relocating to a water-filled lower head. Where possible, existing models and data for predicting cooling are quantitatively compared; and governing relationships are identified. Key phenomena that should be incorporated into models for predicting this heat transfer are discussed, and deficiencies in current models and available data for predicting cooling are noted. Recommendations for improving these models and for obtaining data to validate these models are also provided. (author)

Proceedings of the 33rd national heat transfer conference NHTC'99

International Nuclear Information System (INIS)

Jensen, M.K.; Di Marzo, M.

1999-01-01

The papers in this conference were divided into the following sections: Radiation Heat Transfer in Fires; Computational Fluid Dynamics Methods in Two-Phase Flow; Heat Transfer in Microchannels; Thin Film Heat Transfer; Thermal Design of Electronics; Enhanced Heat Transfer I; Porous Media Convection; Contact Resistance Heat Transfer; Materials Processing in Solidification and Crystal Growth; Fundamentals of Combustion; Challenging Modeling Aspects of Radiative Transfer; Fundamentals of Microscale Transport; Laser Processing and Diagnostics for Manufacturing and Materials Processing; Experimental Studies of Multiphase Flow; Enhanced Heat Transfer II; Heat and Mass Transfer in Porous Media; Heat Transfer in Turbomachinery and Gas Turbine Systems; Conduction Heat Transfer; General Papers; Open Forum on Combustion; Combustion and Instrumentation and Diagnostics I; Radiative Heat Transfer and Interactions in Participating and Nonparticipating Media; Applications of Computational Heat Transfer; Heat Transfer and Fluid Aspects of Heat Exchangers; Two-Phase Flow and Heat Transfer Phenomena; Fundamentals of Natural and Mixed Convection Heat Transfer I; Fundamental of Natural and Mixed Convection Heat Transfer II; Combustion and Instrumentation and Diagnostics II; Computational Methods for Multidimensional Radiative Transfer; Process Heat Transfer; Advances in Computational Heat and Mass Transfer; Numerical Methods for Porous Media; Transport Phenomena in Manufacturing and Materials Processing; Practical Combustion; Melting and Solidification Heat Transfer; Transients in Dynamics of Two-Phase Flow; Basic Aspects of Two-Phase Flow; Turbulent Heat Transfer; Convective Heat Transfer in Electronics; Thermal Problems in Radioactive and Mixed Waste Management; and Transport Phenomena in Oscillatory Flows. Separate abstracts were prepared for most papers in this conference

A study on the heat transfer characteristics of a self-oscillating heat pipe

International Nuclear Information System (INIS)

Yoon, Seok Hun; Oh, Cheol; Choi, Jae Hyuk

2002-01-01

In this paper, the heat transfer characteristics of a self-oscillating heat pipe are experimentally investigated for the effect of various working fluid fill charge ratios and heat loads. The characteristics of temperature oscillations of the working fluid are also analysed based on chaotic dynamics. The heat pipe is composed of a heating section, a cooling section and an adiabatic section, and has a 0.002m internal diameter, a 0.34m length in each turn and consists of 19 turns. The heating and the cooling portion of each turn has a length of 70mm. A series of experiments was carried out to measure the temperature distributions and the pressure variations of the heat pipe. Furthermore, heat transfer performance, effective thermal conductivity, boiling heat transfer and condensation heat transfer coefficients are calculated for various operating conditions. Experimental results show the efficacy of this type of heat pipe

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

Accident prevention ordinance 2.0 Thermal Power Plants

International Nuclear Information System (INIS)

Egyptien, H.H.; Fischermann, E.

This accident prevention ordinance is to cover primarily the very section of a power station where fossil or nuclear energy is converted into thermal energy, e.g. by heating or vaporization of a heat source. In paragraph 1, 40 GJ/h are stipulated as the lower limit of capacity corresponding to about 11 MW. Therefore, the accident prevention ordinance does not only marshal the operation of steam generators in electricity supply utilities but also covers smaller industrial power stations which partly do only meet the company's own requirements. Pipes are only covered as far as they are operated in conjunction with a heat generator. The same applies to coal handling and ash removal facilities. This means that for heat release e.g. in the framework of a district heating grid, the transfer station to the distribution grid is regarded as being a border of the power station and thus a border to the area of application of the accident prevention ordinance. (orig./HP) [de

PWR blowdown heat transfer separate-effects program: thermal-hydraulic test facility experimental data report for test 104

International Nuclear Information System (INIS)

Leon, D.M.; White, M.D.; Moore, P.A.; Hedrick, R.A.

1978-01-01

Reduced instrument responses are presented for Thermal-Hydraulic Test Facility (THTF) test 104, which is part of the ORNL Pressurized-Water Reactor (PWR) Blowdown Heat Transfer Separate-Effects Program. The objective of the program is to investigate the thermal-hydraulic phenomenon governing the energy transfer and transport processes that occur during a loss-of-coolant accident in the PWR system. Test 104 was conducted to obtain CHF in bundle 1 under blowdown conditions. The primary purpose of this report is to make the reduced instrument responses during test 104 available

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)

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

NARCIS (Netherlands)

de Jong, Anne; Wijnant, Ysbrand H.; de Boer, Andries

2014-01-01

A one-dimensional (1D) laminar oscillating flow heat transfer model is derived and applied to parallel-plate thermoacoustic heat exchangers. The model can be used to estimate the heat transfer from the solid wall to the acoustic medium, which is required for the heat input/output of thermoacoustic

Enhancement of heat and mass transfer by cavitation

International Nuclear Information System (INIS)

Zhang, Y N; Du, X Z; Xian, H Z; Zhang, Y N

2015-01-01

In this paper, a brief summary of effects of cavitation on the heat and mass transfer are given. The fundamental studies of cavitation bubbles, including its nonlinearity, rectified heat and mass diffusion, are initially introduced. Then selected topics of cavitation enhanced heat and mass transfer were discussed in details including whales stranding caused by active sonar activity, pool boiling heat transfer, oscillating heat pipe and high intensity focused ultrasound treatment

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

Experimental study on local heat transfer characteristics of porous media with internal heat source

International Nuclear Information System (INIS)

Zan Yuanfeng; Wang Taotao; Xiao Zejun; Wang Fei; Huang Yanping

2008-01-01

Model of porous media with internal heat source is established. The model uses water as flowing media, and the stainless steel test section is packed with steel spheres in manner of regular triangle, respectively. The armoured resistance wire is inserted inside the steel sphere. On the basis of the experimental model, many parameters of the local heat transfer characteristics including current velocity and wall temperature of steel sphere are measured. The experimental results show that the coefficient of heat transfer scarcely changes with pressure. The coefficient of heat transfer increases with the surface heat flux of steel sphere. When raising the inlet temperature of the cooling water, the coefficient of heat transfer presents the descending trend. In addition, the influence of entrance effect on heat transfer is discovered in the experiment, which is much less than the liquid flow in the light tube. After experiment data are analyzed and processed, the relation model of heat transfer on local heat transfer characteristic of porous media with internal heat source was described with a power-law-equation. The deviations between calculation and experimental values are within ±10%. (authors)

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

A Preliminary Study of Transverse Curvature Effects on Condensation Heat Transfer on Vertical Tube in the Presence of Non-condensable Gas

Energy Technology Data Exchange (ETDEWEB)

Lee, Yeon Gun; Kim, Sin [Jeju National Univ., Jeju (Korea, Republic of); Jerng, Dong Wook [Chung Ang Univ., Seoul (Korea, Republic of)

2013-10-15

In this study, the effect of the transverse curvature on the condensation HTC on a vertical tube in the presence of air is preliminarily investigated by using the analysis of boundary layer for free convective heat transfer. The results indicate that the heat transfer performance can be enhanced as the outer diameter of condenser tubes is small. To confirm this curvature effect, an experimental program to obtain the condensation heat transfer data for various values of tube diameter is indispensable. Currently, by a joint research project of Jeju National University and Chung-Ang University, a condensation test facility is being designed and constructed to acquire the condensation HTC data as shown in Fig. 3. From a series of experiment on a single vertical tube, the effects of not only the tube diameter but the inclination, the existence of fins and the local velocity of a bulk mixture by natural circulation will be evaluated precisely. An empirical correlation for the condensation heat transfer of a steam-air mixture will also be developed for design optimization and performance evaluation of the PCCS. The Passive Containment Cooling System (PCCS) provides passive means to remove the decay heat and protect the integrity of the containment during severe accidents. Korea, in which all the NPPs employ the concrete containment, may adopt a PCCS using internal condensers. In the event of the loss-of-coolant accident (LOCA), steam released from the reactor coolant system is mixed with air inside the containment and condensed on the outer surface of inclined condenser tubes. It is noted that, among previous theoretical and empirical models for condensation on outer wall in the presence of non-condensable gas, no one took into account the effect of a tube diameter. Though the condensation heat transfer coefficient may vary with transverse curvature of condenser tubes, such a curvature effect has not been reported so far. In this study, a preliminary analysis is conducted

A Preliminary Study of Transverse Curvature Effects on Condensation Heat Transfer on Vertical Tube in the Presence of Non-condensable Gas

International Nuclear Information System (INIS)

Lee, Yeon Gun; Kim, Sin; Jerng, Dong Wook

2013-01-01

In this study, the effect of the transverse curvature on the condensation HTC on a vertical tube in the presence of air is preliminarily investigated by using the analysis of boundary layer for free convective heat transfer. The results indicate that the heat transfer performance can be enhanced as the outer diameter of condenser tubes is small. To confirm this curvature effect, an experimental program to obtain the condensation heat transfer data for various values of tube diameter is indispensable. Currently, by a joint research project of Jeju National University and Chung-Ang University, a condensation test facility is being designed and constructed to acquire the condensation HTC data as shown in Fig. 3. From a series of experiment on a single vertical tube, the effects of not only the tube diameter but the inclination, the existence of fins and the local velocity of a bulk mixture by natural circulation will be evaluated precisely. An empirical correlation for the condensation heat transfer of a steam-air mixture will also be developed for design optimization and performance evaluation of the PCCS. The Passive Containment Cooling System (PCCS) provides passive means to remove the decay heat and protect the integrity of the containment during severe accidents. Korea, in which all the NPPs employ the concrete containment, may adopt a PCCS using internal condensers. In the event of the loss-of-coolant accident (LOCA), steam released from the reactor coolant system is mixed with air inside the containment and condensed on the outer surface of inclined condenser tubes. It is noted that, among previous theoretical and empirical models for condensation on outer wall in the presence of non-condensable gas, no one took into account the effect of a tube diameter. Though the condensation heat transfer coefficient may vary with transverse curvature of condenser tubes, such a curvature effect has not been reported so far. In this study, a preliminary analysis is conducted

Application of intensified heat transfer for the retrofit of heat exchanger network

International Nuclear Information System (INIS)

Wang, Yufei; Pan, Ming; Bulatov, Igor; Smith, Robin; Kim, Jin-Kuk

2012-01-01

Highlights: → Novel design approach for the retrofit of HEN based on intensified heat transfer. → Development of a mathematical model to evaluate shell-and-tube heat exchanger performances. → Identification of the most appropriate heat exchangers requiring heat transfer enhancements in the heat exchanger network. -- Abstract: A number of design methods have been proposed for the retrofit of heat exchanger networks (HEN) during the last three decades. Although considerable potential for energy savings can be identified from conventional retrofit approaches, the proposed solutions have rarely been adopted in practice, due to significant topology modifications required and resulting engineering complexities during implementation. The intensification of heat transfer for conventional shell-and-tube heat exchangers can eliminate the difficulties of implementing retrofit in HEN which are commonly restricted by topology, safety and maintenance constraints, and includes high capital costs for replacing equipment and pipelines. This paper presents a novel design approach to solve HEN retrofit problems based on heat transfer enhancement. A mathematical model has been developed to evaluate shell-and-tube heat exchanger performances, with which heat-transfer coefficients and pressure drops for both fluids in tube and shell sides are obtained. The developed models have been compared with the Bell-Delaware, simplified Tinker and Wills-Johnston methods and tested with the HTRI (registered) and HEXTRAN (registered) software packages. This demonstrates that the new model is much simpler but can give reliable results in most cases. For the debottlenecking of HEN, four heuristic rules are proposed to identify the most appropriate heat exchangers requiring heat transfer enhancements in the HEN. The application of this new design approach allows a significant improvement in energy recovery without fundamental structural modifications to the network.

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.)

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

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

The contact heat transfer between the heating plate and granular materials in rotary heat exchanger under overloaded condition

Directory of Open Access Journals (Sweden)

Luanfang Duan

2018-03-01

Full Text Available In the present work, the contact heat transfer between the granular materials and heating plates inside plate rotary heat exchanger (PRHE was investigated. The heat transfer coefficient is dominated by the contact heat transfer coefficient at hot wall surface of the heating plates and the heat penetration inside the solid bed. A plot scale PRHE with a diameter of Do = 273 mm and a length of L = 1000 mm has been established. Quartz sand with dp = 2 mm was employed as the experimental material. The operational parameters were in the range of ω = 1 – 8 rpm, and F = 15, 20, 25, 30%, and the effect of these parameters on the time-average contact heat transfer coefficient was analyzed. The time-average contact heat transfer coefficient increases with the increase of rotary speed, but decreases with the increase of the filling degree. The measured data of time-average heat transfer coefficients were compared with theoretical calculations from Schlünder’s model, a good agreement between the measurements and the model could be achieved, especially at a lower rotary speed and filling degree level. The maximum deviation between the calculated data and the experimental data is approximate 10%. Keywords: Rotary heat exchanger, Contact heat transfer, Granular material, Heating plate, Overloaded

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

Science.gov (United States)

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

2014-03-01

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

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)

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.

Numerical study on heat transfer characteristics of thermosyphon heat pipes using nanofluids

International Nuclear Information System (INIS)

Huminic, Gabriela; Huminic, Angel

2013-01-01

Highlights: • Numerical study of nanofluid heat transfer in thermosyphon heat pipes is performed. • Effect of nanoparticle concentration and operating temperature are studied. • Fe 2 O 3 –water nanofluid with 5.3% volume concentration shows the best performance. • Results show the improvement the thermal performances of thermosyphon heat pipe with nanofluids. - Abstract: In this work, a three-dimensional analysis is used to investigate the heat transfer of thermosyphon heat pipe using water and nanofluids as the working fluid. The study focused mainly on the effects of volume concentrations of nanoparticles and the operating temperature on the heat transfer performance of the thermosyphon heat pipe using the nanofluids. The analysis was performed for water and γ-Fe 2 O 3 nanoparticles, three volume concentrations of nanoparticles (0 vol.%, 2 vol.% and 5.3 vol.%) and four operating temperatures (60, 70, 80 and 90 °C). The numerical results show that the volume concentration of nanoparticles had a significant effect in reducing the temperature difference between the evaporator and condenser. Experimental and numerical results show qualitatively that the thermosyphon heat pipe using the nanofluid has better heat transfer characteristics than the thermosyphon heat pipe using water

Convective heat transfer enhancement using Carbon nanofibers (CNFs): influence of amorphous carbon layer on heat transfer performance

NARCIS (Netherlands)

Taha, T.J.; Lefferts, Leonardus; van der Meer, Theodorus H.

2013-01-01

In this work, an experimental heat transfer investigation was carried out to investigate the combined influence of both amorphous carbon (a-C) layer thickness and carbon nanofibers (CNFs) on the convective heat transfer behavior. Synthesis of these carbon nano structures was achieved using catalytic

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.

Heat transport and surface heat transfer with helium in rotating channels

International Nuclear Information System (INIS)

Schnapper, C.

1978-06-01

Heat transport and surface heat transfer with helium in rotating radially arranged channels were experimentally studied with regard to cooling of large turbogenerators with superconducting windings. Measurements with thermosiphon and thermosiphon loops of different channel diameters were performed, and results are presented. The thermodynamic state of the helium in a rotating thermosiphon and the mass flow rate in a thermosiphon loop is characterized by formulas. Heat transport by directed convection in thermosiphon loops is found to be more efficient 12 cm internal convection in thermosiphons. Steady state is reached sooner in thermosiphon loops than in thermosiphons, when heat load suddenly changes. In a very large centrifugal field single-phase heat transfer with natural and forced convection is described by similar formulas which are also applicable 10 thermosiphons in gravitation field or to heat transfer to non-rotating helium. (orig.) [de

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Evaluation of empirical heat transfer models using TFG heat flux sensors

International Nuclear Information System (INIS)

De Cuyper, T.; Broekaert, S.; Chana, K.; De Paepe, M.; Verhelst, S.

2017-01-01

Thermodynamic engine cycle models are used to support the development of the internal combustion engine (ICE) in a cost and time effective manner. The sub model which describes the in-cylinder heat transfer from the working gases to the combustion chamber walls plays an important role in the accuracy of these simulation tools. The heat transfer affects the power output, engine efficiency and emissions of the engine. The most common heat transfer models in engine research are the models of Annand and Woschni. These models provide an instantaneous spatial averaged heat flux. In this research, prototype thin film gauge (TFG) heat flux sensors are used to capture the transient in-cylinder heat flux behavior within a production spark ignition (SI) engine as they are small, robust and able to capture the highly transient temperature swings. An inlet valve and two different zones of the cylinder head are instrumented with multiple TFG sensors. The heat flux traces are used to calculate the convection coefficient which includes all information of the convective heat transfer phenomena inside the combustion chamber. The implementation of TFG sensors inside the combustion chamber and the signal processing technique are discussed. The heat transfer measurements are used to analyze the spatial variation in heat flux under motored and fired operation. Spatial variation in peak heat flux was observed even under motored operation. Under fired operation the observed spatial variation is mainly driven by the flame propagation. Next, the paper evaluates the models of Annand and Woschni. These models fail to predict the total heat loss even with calibration of the models coefficients using a reference motored operating condition. The effect of engine speed and inlet pressure is analyzed under motored operation after calibration of the models. The models are able to predict the trend in peak heat flux value for a varying engine speed and inlet pressure. Next, the accuracy of the

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.

Impact of bulk atmospheric motion on local and global containment heat transfer

International Nuclear Information System (INIS)

Green, J.A.; Almenas, K.

1995-01-01

Local and global correlations for condensing energy transfer in the presence of noncondensable gases in a containment facility have been evaluated. The database employed stems from the E11.2 and E11.4 tests conducted at the German HDR facility. The HDR containment is a 11060-ml, 60-m-high decommissioned light water reactor. The tests simulated long-term (up to 56 h) accident conditions. Numerous instrumented structural blocks (concrete and lead) were located throughout the containment to provide detailed local heat transfer measurements. These data represent what is probably the most extensive database of integral energy transfer measurements available. It is well established that the major resistance to condensation heat transfer in the presence of noncondensable gases is a gaseous boundary layer that builds up in front of the condensing surface. Correlations that seek to model heat transfer for these conditions should depend on parameters that most strongly determine the buildup and thickness of this boundary layer. Two of the most important parameters are the vapor/noncondensable concentration ratio and the local atmospheric motion. Secondary parameters include the atmosphere-to-surface temperature difference, the pressure, and condensing surface properties. The HDR tests are unique in terms of the quantity and variety of instrumentation employed. However, one of the most important parameters, the local bulk atmospheric velocity, is inherently difficult to measure, and only fragmentary measurements are available even in the HDR data-base. A detailed analysis of these data is presented by Green. This study uses statistical methods to evaluate local and global empirical correlations that do not include the atmospheric velocity. The magnitude of the differences between the correlations emphasizes the importance of the local atmospheric velocity and serves to illustrate the accuracy limits of correlations that neglect this essential parameter

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

An introduction to heat transfer principles and calculations

CERN Document Server

Ede, A J; Ower, E

1967-01-01

An Introduction to Heat Transfer Principles and Calculations is an introductory text to the principles and calculations of heat transfer. The theory underlying heat transfer is described, and the principal results and formulae are presented. Available techniques for obtaining rapid, approximate solutions to complicated problems are also considered. This book is comprised of 12 chapters and begins with a brief account of some of the concepts, methods, nomenclature, and other relevant information about heat transfer. The reader is then introduced to radiation, conduction, convection, and boiling

ANL ITER high-heat-flux blanket-module heat transfer experiments

International Nuclear Information System (INIS)

Kasza, K.E.

1992-02-01

An Argonne National Laboratory facility for conducting tests on multilayered slab models of fusion blanket designs is being developed; some of its features are described. This facility will allow testing under prototypic high heat fluxes, high temperatures, thermal gradients, and variable mechanical loadings in a helium gas environment. Steady and transient heat flux tests are possible. Electrical heating by a two-sided, thin stainless steel (SS) plate electrical resistance heater and SS water-cooled cold panels placed symmetrically on both sides of the heater allow achievement of global one-dimensional heat transfer across blanket specimen layers sandwiched between the hot and cold plates. The heat transfer characteristics at interfaces, as well as macroscale and microscale thermomechanical interactions between layers, can be studied in support of the ITER engineering design effort. The engineering design of the test apparatus has shown that it is important to use multidimensional thermomechanical analysis of sandwich-type composites to adequately analyze heat transfer. This fact will also be true for the engineering design of ITER

Heat transfer performance of heat pipe for passive cooling of spent fuel pool

International Nuclear Information System (INIS)

Wang Minglu; Xiong Zhengqin; Gu Hanyang; Ye Cheng; Cheng Xu

2014-01-01

A large-scale loop heat pipe has no electricity driven component and high efficiency of heat transfer. It can be used for the passive cooling of the SFP after SBO to improve the safety performance of nuclear power plants. In this paper, such a large-scale loop heat pipe is studied experimentally. The heat transfer rate, evaporator average heat transfer coefficient operating temperature, operating pressure and ammonia flow rate have been obtained with the water flow ranging from 0.007 m/s to 0.02 m/s outside the evaporator section, heating water temperature in the range of 50 to 90℃, air velocity outside the condensation section ranging from 0.5 to 2.5 m/s. It is found that the heat transfer rate reaches as high as 20.1 kW. Parametric analysis indicates that, the heat transfer rate and ammonia flow rate are influenced significantly by hot water inlet temperature and velocity, while beyond 1.5 m/s, the effect of air velocity outside the condensation section is minor. (authors)

Integrated CFD investigation of heat transfer enhancement using multi-tray core catcher in SFR

International Nuclear Information System (INIS)

Rakhi; Sharma, Anil Kumar; Velusamy, K.

2017-01-01

Highlights: • Heat transfer enhancement using multi-tray core catcher for SFR is investigated. • The capability of a single core collector tray is estimated. • Double and triple collector trays with innovative designs is discussed. • Provision of openings in the trays contributed to enhanced natural circulation. - Abstract: To render future SFR more robust and safe, certain BDBE have been considered in the recent years. A Core Disruptive Accident leading to a whole core meltdown scenario has gained the interest of researchers. Various design concepts and safety measures have been suggested and incorporated in design to address such a low probability scenario. A core catcher concept, in particular, has proved to be inevitable as an in-vessel core retention device in SFR for safe retention of core debris arising out after the severe accident. This study aims to analyse the cooling capability of the innovative design concept of core catcher to remove decay heat of degraded core after the accident. First, the capability of single collection tray is established and then the study is extended to two and three collection trays with different design concepts. Transient forms of governing equations of mass, momentum and energy conservations along with k-ε turbulence model are solved by finite volume based CFD solver. Boussinesq approximation is invoked to model buoyancy in sodium. The study shows that a single collection tray is capable of removing up to 20 MW decay heat load in a typical 500 MWe pool type SFR. Further, studies are carried out to improve the natural circulation of sodium around the source, in the lower plenum and to distribute core debris of the whole core to multiple collection trays. It is found that the double and triple collection trays can accommodate decay loads up to 29 MW. Provision of openings in the collection trays has proved to be effective in improving the heat transfer and sodium flow as well as in distributing the core debris to the

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.)

Overall conductance and heat transfer area minimization of refrigerators and heat pumps with finite heat reservoirs

International Nuclear Information System (INIS)

Sarkar, J.; Bhattacharyya, Souvik

2007-01-01

In the present study, the overall conductance and the overall heat transfer area per unit capacity of refrigeration and heat pump systems have been minimized analytically considering both internal and external irreversibilities with variable temperature (finite capacity) heat reservoirs. Hot and cold side refrigerant temperatures, conductance and heat transfer area ratios have been optimized to attain this goal. The results have been verified with the more elaborate numerical optimization results obtained for ammonia based vapour compression refrigeration and heat pump systems working with variable temperature reservoirs. It is observed that the analytical results for optimum refrigerant temperatures, minimum overall conductance and heat transfer area deviate marginally from the numerically optimized results (within 1%), if one assumes a constant heat rejection temperature. The deviation of minimum overall conductance and heat transfer area is more (about 20%), if one considers both the desuperheating and condensation regions separately. However, in the absence of complex and elaborate numerical models, the simple analytical results obtained here can be used as reasonably accurate preliminary guidelines for optimization of refrigeration and heat pump systems

Turbulent heat transfer for heating of water in a short vertical tube

International Nuclear Information System (INIS)

Hata, Koichi; Noda, Nobuaki

2008-01-01

The turbulent heat transfer coefficients for the flow velocities (u=4.0 to 21 m/s), the inlet liquid temperatures (T in =296.5 to 353.4 K), the inlet pressures (P in =810 to 1014 kPa) and the increasing heat inputs (Q 0 exp(t/τ), τ=10, 20 and 33.3 s) are systematically measured by an experimental water loop. The Platinum test tubes of test tube inner diameters (d=3, 6 and 9 mm), heated lengths (L=32.7 to 100 mm), ratios of heated length to inner diameter (L/d=5.51 to 33.3) and wall thickness (δ=0.3, 0.4 and 0.5 mm) with surface roughness (Ra=0.40 to 0.78 μm) are used in this work. The turbulent heat transfer data for Platinum test tubes were compared with the values calculated by other workers' correlations for the turbulent heat transfer. The influence of Reynolds number (Re), Prandtl number (Pr), Dynamic viscosity (μ) and L/d on the turbulent heat transfer is investigated into details and, the widely and precisely predictable correlation of the turbulent heat transfer for heating of water in a short vertical tube is given based on the experimental data. The correlation can describe the turbulent heat transfer coefficients obtained in this work for the wide range of the temperature difference between heater inner surface temperature and average bulk liquid temperature (ΔT L =5 to 140 K) with d=3, 6 and 9 mm, L=32.7 to 100 mm and u=4.0 to 21 m/s within ±15%, difference. (author)

Turbulent heat transfer for heating of water in a short vertical tube

International Nuclear Information System (INIS)

Hata, Koichi; Noda, Nobuaki

2007-01-01

The turbulent heat transfer coefficients for the flow velocities (u=4.0 to 21 m/s), the inlet liquid temperatures (T in =296.5 to 353.4 K), the inlet pressures (P in =810 to 1014 kPa) and the increasing heat inputs (Q 0 exp(t/τ), τ=10, 20 and 33.3 s) are systematically measured by the experimental water loop. The Platinum test tubes of test tube inner diameters (d=3, 6 and 9 mm), heated lengths (L=32.7 to 100 mm), ratios of heated length to inner diameter (L/d=5.51 to 33.3) and wall thicknesses (δ=0.3, 0.4 and 0.5 mm) with surface roughness (Ra=0.40 to 0.78 μm) are used in this work. The turbulent heat transfer data for Platinum test tubes were compared with the values calculated by other workers' correlations for the turbulent heat transfer. The influences of Reynolds number (Re), Prandtl number (Pr), Dynamic viscosity (μ) and L/d on the turbulent heat transfer are investigated into details and, the widely and precisely predictable correlation of the turbulent heat transfer for heating of water in a short vertical tube is given based on the experimental data. The correlation can describe the turbulent heat transfer coefficients obtained in this work for wide range of the temperature difference between heater inner surface temperature and average bulk liquid temperature (ΔT L =5 to 140 K) with d=3, 6 and 9 mm, L=32.7 to 100 mm and u=4.0 to 21 m/s within ±15% difference. (author)

Subcooled boiling heat transfer correlation to calculate the effects of dissolved gas in a liquid

International Nuclear Information System (INIS)

Zarkasi, Amin S.; Chao, W.W.; Kunze, Jay F.

2004-01-01

The water coolant in most operating power reactor systems is kept free of dissolved gas, so as to minimize corrosion. However, in most research reactors, which operate at temperatures below 70 deg. C, and between 1 and 5 atm. pressure, the dissolved gas remains present in the water coolant system during operation. This dissolved gas can have a significant effect during accident conditions (i.e. a LOCA), when the fluid quickly reaches boiling, coincident with flow stagnation and subsequent flow reversal. A benchmark experiment was conducted, with an electrically heated, closed loop channel, modeling a research reactor fuel coolant channels (2 mm thick). The results showed 'boiling (bubble) noise' occurring before wall temperatures reached saturation, and a significant increase (up to 50%) in the heat transfer coefficient in the subcooled boiling region when in the presence of dissolved gas, compared to degassed water. Since power reactors do not involve dissolved gas, the RELAP safety analysis code does not include any provisions for the effect of dissolved gas on heat transfer. In this work, the effects of the dissolved gas are evaluated for inclusion in the RELAP code, including provision for initiating 'nucleate boiling' at a lower temperature, and a provision for enhancing the heat transfer coefficient during the subcooled boiling region. Instead of relying on Chen's correlation alone, a modification of the superposition method of Bjorge was adopted. (author)

Loop heat pipes - highly efficient heat-transfer devices for systems of sun heat supply

Energy Technology Data Exchange (ETDEWEB)

Maydanik, Yu. [Ural Branch of the Russian Academy of Sciences, Ekaterinburg (Russian Federation). Inst. of Thermophysics

2004-07-01

Loop heat pipes (LHPs) are hermetic heat-transfer devices operating on a closed evaporation-condensation cycle with the use of capillary pressure for pumping the working fluid [1]. In accordance with this, they possess all the main advantages of conventional heat pipes, but, as distinct from the latter, have a considerably higher heat-transfer capacity, especially when operating in the ''antigravity'' regime, when heat is transferred from above downwards. Besides, LHPs possess a higher functional versatility, are adaptable to different operating conditions and provide great scope for various design embodiments. This is achieved at the expense of both the original design of the device and the properties of the wick - a special capillary structure used for the creation of capillary pressure. The LHP schematic diagram is given in Fig. 1. The device contains an evaporator and a condenser - heat exchanger connected by means of smooth-walled pipe-lines with a relatively small diameter intended for separate motion of vapor and liquid. At present loop heat pipes are most extensively employed in thermoregulation systems of spacecrafts. Miniature LHPs are used for cooling electronics and computers. At the same time there exists a considerable potential of using these devices for the recovery of low-grade (waste) heat from different sources, and also in systems of sun heat supply. In the latter case LHPs may serve as an efficient heat-transfer link between a sun collector and a heat accumulator, which has a low thermal resistance and does not consume any additional energy for pumping the working fluid between them. (orig.)

Heat transfer enhancement in cross-flow heat exchanger using vortex generator

International Nuclear Information System (INIS)

Yoo, S. Y.; Kwon, H. K.; Kim, B. C.; Park, D. S.; Lee, S. S.

2003-01-01

Fouling is very serious problem in heat exchanger because it rapidly deteriorates the performance of heat exchanger. Cross-flow heat exchanger with vortex generators is developed, which enhance heat transfer and reduce fouling. In the present heat exchanger, shell and baffle are removed from the conventional shell-and-tube heat exchanger. The naphthalene sublimation technique is employed to measure the local heat transfer coefficients. The experiments are performed for single circular tube, staggered array tube bank and in-line array tube bank with and without vortex generators. Local and average Nusselt numbers of single tube and tube bank with vortex generator are investigated and compared to those of without vortex generator

Influence of short heat pulses on the helium boiling heat transfer rate

International Nuclear Information System (INIS)

Andreev, V.K.; Deev, V.I.; Savin, A.N.; Kutsenko, K.V.

1987-01-01

Investigation results on heat transfer in the process of helium boiling on a heated wall under conditions of pulsed heat effect are described. Results of the given study point to one of possible ways of heat exchange intensification in boiling helium by supplying short heat pulse to the heater. Even short-time noncontrolled or incidental increase in the heater capacity during experiment with boiling helium can result in a considerable disagreement of experimental data on heat transfer

Natural convection heat transfer of fluid with temperature-dependent specific heat

International Nuclear Information System (INIS)

Tanaka, Amane; Kubo, Shinji; Akino, Norio

1998-01-01

The present study investigates natural convection from a heated vertical plate of fluid with temperature-dependent specific heat, which is introduced as a model of microencapsulated phase change material slurries (MCPCM slurries). The temperature dependence of specific heat is represented by Gauss function with three physical parameters (peak temperature, width of phase change temperature and latent heat). Boundary layer equations are solved numerically, and the velocity and temperature fields of the flow are obtained. The relation between the heat transfer coefficients and the physical parameters of specific heat is discussed. The results show that the velocities and temperatures are smaller, and the heat transfer coefficients are larger comparing with those of the fluid with constant specific heat. (author)

Evaluation of food chain transfer data for use in accident consequence assessment

International Nuclear Information System (INIS)

Coughtrey, P.J.; Kirton, J.A.; Mitchell, N.G.

1991-01-01

Input data for the food chain transport component of radiological assessment models are summarised in the context of the sources of information available prior to the Chernobyl accident and those derived after the accident. Particular attention is devoted to interception and retention soil-to-plant, and plant-to-animal transfer, and to the applicability of environmental data to both equilibrium and time-dependent models. It is argued that much of the current uncertainty in parameter values for use in radiological assessment models reflects lack of understanding of processes involved in the various stages of transfer of radionuclides to man. The Chernobyl accident highlighted this lack of fundamental knowledge and illustrated a number of areas where further research and model development is justified. These areas are identified and suggestions given for appropriate research to support model development

Proceedings of the twenty third national heat and mass transfer conference and first international ISHMT-ASTFE heat and mass transfer conference: souvenir and book of abstracts

International Nuclear Information System (INIS)

2015-01-01

The conference covered various aspects of heat and mass transfer like Aero-thermodynamics, Atmospheric flows, Biological heat and mass transfer, Combustion and reactive flows, Cryogenics, Electronic and photonic cooling, Energy engineering, Environmental engineering, Experimental techniques, Heat transfer enhancement, Heat transfer equipment's, Heat transfer in nuclear applications, Mass transfer, Materials processing and manufacturing, Microscale and nanoscale transport, Multiphase transport and phase change, Multi mode heat transfer, Numerical methods, Refrigeration and air conditioning, Space heat transfer, Transport phenomena in porous media, and Turbulent transport. Papers relevant to INIS are indexed separately

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

Heat Transfer and Entropy Generation Analysis of an Intermediate Heat Exchanger in ADS

Science.gov (United States)

Wang, Yongwei; Huai, Xiulan

2018-04-01

The intermediate heat exchanger for enhancement heat transfer is the important equipment in the usage of nuclear energy. In the present work, heat transfer and entropy generation of an intermediate heat exchanger (IHX) in the accelerator driven subcritical system (ADS) are investigated experimentally. The variation of entropy generation number with performance parameters of the IHX is analyzed, and effects of inlet conditions of the IHX on entropy generation number and heat transfer are discussed. Compared with the results at two working conditions of the constant mass flow rates of liquid lead-bismuth eutectic (LBE) and helium gas, the total pumping power all tends to reduce with the decreasing entropy generation number, but the variations of the effectiveness, number of transfer units and thermal capacity rate ratio are inconsistent, and need to analyze respectively. With the increasing inlet mass flow rate or LBE inlet temperature, the entropy generation number increases and the heat transfer is enhanced, while the opposite trend occurs with the increasing helium gas inlet temperature. The further study is necessary for obtaining the optimized operation parameters of the IHX to minimize entropy generation and enhance heat transfer.

Heat Transfer and Pressure Drop Characteristics in Straight Microchannel of Printed Circuit Heat Exchangers

Directory of Open Access Journals (Sweden)

Jang-Won Seo

2015-05-01

Full Text Available Performance tests were carried out for a microchannel printed circuit heat exchanger (PCHE, which was fabricated with micro photo-etching and diffusion bonding technologies. The microchannel PCHE was tested for Reynolds numbers in the range of 100‒850 varying the hot-side inlet temperature between 40 °C–50 °C while keeping the cold-side temperature fixed at 20 °C. It was found that the average heat transfer rate and heat transfer performance of the countercurrrent configuration were 6.8% and 10%‒15% higher, respectively, than those of the parallel flow. The average heat transfer rate, heat transfer performance and pressure drop increased with increasing Reynolds number in all experiments. Increasing inlet temperature did not affect the heat transfer performance while it slightly decreased the pressure drop in the experimental range considered. Empirical correlations have been developed for the heat transfer coefficient and pressure drop factor as functions of the Reynolds number.

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

International Nuclear Information System (INIS)

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

2013-01-01

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