An investigation of flow characteristics and critical heat flux in vertical upward round tube

Prediction of critical heat flux (CHF) in annular flow is important for the safety of once - through steam generator and the reactor core under accident conditions. The dryout in annular flow occurs at the point where the film is depleted due to entrainment, deposition, and evaporation. The film thickness, film mass flow rate along axial distribution, and CHF are calculated in vertical upward round tube on the basis of a separated flow model of annular flow. The theoretical CHF values are higher than those derived from experimental data, with error being within 30%.

Critical Heat Flux with Subcooled Flowing Water in Tubes for Pressures from Atmosphere to Near-Critical Point

During last 45 years, two groups of the experimental data on critical heat flux were obtained in bare tubes, covering the pressures from atmosphere to near-critical point. One group of the data were obtained in the inner diameter of 2.32, 5.16, 8.05, 10.0 and 16.0 mm, respectively, with the ranges of pressure of 0.1-1.92 MPa, velocity of 1.47-23.3 m/s, local subcooling of 3.7-108.7 ℃ and heat flux of up to 38.3 MW/m2. Another group of the data were obtained in the inner diameter of 4.62, 7.98 and 10.89 mm, respectively, with the ranges of pressure of 1.7-20.6 MPa, mass flux of 454-4,055 kg/（m2.s） and inlet subcooling of 53-361 ℃. The results showed complicated effects of the pressure, mass flux, subcooling and diameter on the critical heat flux. They were formulated by two empirical correlations. A mechanistic model on the limit of heat transfer capability from the bubbly layer to the subcooled core was also proposed for all the results.

Improvement of Critical Heat Flux Performance by Wire Spacer

In order to investigate the coolability of a tight lattice core in BWR （boiling water reactors）, an experiment of CHF （critical heat flux） was conducted using a single heater pin flow channel with and without a wire spacer with the mass flux ranged from 200 kg/（m2.s） to 600 kg/（m2·s）, the inlet temperature from 50 ℃ to 90 ℃ at the pressure of 0.1 MPa. The results of CHF values were compared between two different types of heater pin with and without the wire spacer. The heat removability of flow channel was improved by the wire spacer in comparison with the heater pin flow channel without the wire spacer. The CHF value was higher in the heater pin channel with the wire spacer than in that without the wire spacer.

Accurate prediction of the critical heat flux for pool boiling on the heater substrate

While the influence of liquid qualities,surface morphology,and operating circumstances on critical heat flux(CHF)in pool boiling has been extensively studied,the effect of the heater substrate has not.Based on the force balance analysis,a theoretical model has been developed to accurately predict the CHF in pool boiling on a heater substrate.An analytical expression for the CHF of a heater substrate is obtained in terms of the surface thermophysical property.It is indicated that the ratio of thermal conductivity(k)to the product of density(ρ)and specific heat(cp)is an essential substrate property that influences the CHF.By modifying the well-known force-balance-based CHF model(Kandlikar model),the thermal characteristics of the substrate are taken into consideration.The bias of predicted CHF values are within 5%compared with the experimental results.

Critical Heat Flux in Forced Convective Boiling with a Wall Jet

The critical heat flux (CHF) in the forced convective boiling with a wall jet has been investigated.The experiments of CHF with a wall jet have been performed over a wide range of ρ l/ρ g=6.6-1 603 and ΔT sub =0-60 K. The mechanism on CHF is discussed and a CHF model based on heat balance in sublayer can provide a good clue for analyzing and deriving CHF.Finally,a generalized correlation is presented, which can predict CHF for saturated and subcooled conditions.

The Corrected Simulation Method of Critical Heat Flux Prediction for Water-Cooled Divertor Based on Euler Homogeneous Model

An accurate critical heat flux（CHF） prediction method is the key factor for realizing the steady-state operation of a water-cooled divertor that works under one-sided high heating flux conditions.An improved CHF prediction method based on Euler＇s homogeneous model for flow boiling combined with realizable k-ε model for single-phase flow is adopted in this paper in which time relaxation coefficients are corrected by the Hertz-Knudsen formula in order to improve the calculation accuracy of vapor-liquid conversion efficiency under high heating flux conditions.Moreover,local large differences of liquid physical properties due to the extreme nonuniform heating flux on cooling wall along the circumference direction are revised by formula IAPWSIF97.Therefore,this method can improve the calculation accuracy of heat and mass transfer between liquid phase and vapor phase in a CHF prediction simulation of water-cooled divertors under the one-sided high heating condition.An experimental example is simulated based on the improved and the uncorrected methods.The simulation results,such as temperature,void fraction and heat transfer coefficient,are analyzed to achieve the CHF prediction.The results show that the maximum error of CHF based on the improved method is 23.7%,while that of CHF based on uncorrected method is up to 188%,as compared with the experiment results of Ref.[12].Finally,this method is verified by comparison with the experimental data obtained by International Thermonuclear Experimental Reactor（ITER）,with a maximum error of 6% only.This method provides an efficient tool for the CHF prediction of water-cooled divertors.

Experimental Studies on Critical Heat Flux in a Uniformly Heated Vertical Tube at Low Pressure and Flowrates

Investigations into critical beat flux at low flow and pressure conditions are of particular interest when predicting the nuclear reactor core behavior during Loss of Coolant accident （LOCA）. Therefore, critical heat flux （CHF） has been investigated in a uniformly heated vertical round tube at two low system pressures and six low water flowrates. The results have been compared with two correlations which have different approaches and CHF look-up table. Good agreements have been obtained for the three comparisons at the lower sets of mass fluxes. The Bowring correlation was found to be the best to correlate the experimental results with Root Mean Square Error RMSE of 0.54% and 0.56% for the 5 bar and 15 bar system pressure respectively. A comparisons with the Shim and Lee correlation yielded RMSE of 0.23% and 5.74% for the two system pressure respectively. When the look-up table of Groeneveld et al. was used, RMES of 0.55% and 25.2% was obtained for the two system pressure respectively.

Enhancement of Critical Heat Flux in Tight Rod Bundle with Wire Spacer

The experiment of CHF （critical heat flux） was conducted for water boiling two-phase flow in three-pin tight rod bundle. The effects of with and without wire spacers and the pitch to diameter ratio p/d on CHF were investigated under the conditions of mass flux range 250-430 kg/（m2·s）, inlet temperature from 70 ℃ to 92 ℃ and the pressure of 0.1 MPa. The CHF was enhanced by wire spacers in comparison with the results of CHF without wire spacers. The CHF was enhanced by reducing thep/dfrom 1.18 to 1.10 under the same flow rate condition, although it did not change appreciably with the change ofp/d under the same mass flux condition.

Heat Flux Analysis in Electrical Transformers through Integral Operators of Mechanics

Considering the continuous functioning of a power transformer under charge of high capacity of 50 MVA, predicted studies are proposed to be performed of their thermal behavior under perma nent and variable regimens of flow of charge, using noninvasive methods based in integral trans forms that measure and determine parameters of geometrical, analytical and physical type of the transformer. In before works, we have studied a basic geometry of a winding composed of high and low voltage sections with a uniform heat generation and heat convection boundary conditions. The heat conduction equation representing the phenomena of heat generation in a cylindrical structure was solved by using an integral transform. In this sense, this new study considers the basic geometry composed of a three cylindrical windings (high and low voltage turns) and a rec tangular core. Thus it is proposed to solve magnetic flow equations using integral transforms (Han kel transforms and Bessel integrals) in order to obtain the heat source distribution in the core due to the magnetization currents which are developed in function of the magnetic field flow equations. Based on this, it is proposed as a second step to use this heat source distribution to obtain the corresponding temperature distribution in the core by solving the cylindrical heat conduction equation for the core (cylindrical). Bearing this in mind, it is proposed finally to solve the 3D cylindrical heat conduction equation for the one winding using the calculated heat convection coefficients, the conductivity of the winding, behavior of the mineral oil and the non uniform winding heat generation predicted in recent researches. This equation will be solved by using integral methods (Radon, Hankel and Fourier transforms). This methodology will be useful to establish a new design of a power transformer based on the values of their integrals and the results that throw the inverse methods for this case. Finally if possible we will use the programs of Fluent and/or Phoenics for the validation of functional proposed models of prediction and prevention of heat flow and charge based on the obtained results.

Forced Convection Boiling Heat Transfer and Dryout Characteristics in Helical Coiled Tubes with Various Axial Angles

对水和水蒸汽汽液两相流体在螺旋轴呈各种倾角放置的螺旋管内强制对流沸腾传热与烧毁特性进行了系统地试验研究，试验中系统及结构参数范围如下：压力：Ｐ＝０．４～３．０ＭＰａ质量流速：Ｇ＝１００～２４００ｋｇ／ｍ２·ｓ进口水温：Ｔ＝３０～８０℃出口干度：ｘ＝－０．０５～１．２管内壁面热负荷：ｑ＝０～５４０ｋＥ／ｍ２试验段结构参数：总长Ｌ＝６４４８ｍｍ，管内径ｄ＝１１ｍｍ，螺旋直径Ｄ＝２５６ｍｍ，螺旋升角β＝４．２７°螺旋管轴向放置倾角：水平位置（０°）、向上倾斜４５°（＋４５°）、垂直向上（＋９０°）、向下倾斜４５°（－４５°）．共进行了１０５０个工况的试验．试验结果表明，螺旋管内汽液两相流强制对流沸腾传热可以划分为核态沸腾区、两相流强制对流区、烧毁及烧毁后传热区等３种区域．通过数据处理和分析总结，给出了３区域间转变的边界方程，和３个区域内两相流传热系数的计算公式．根据试验观察和数据结果，对烧毁现象及烧毁点或区域发生的条件及机理进行了深入的分析研究，发现一系列有关现象的规律和特点，指出了其主要影响因素，并给出了烧毁点临界质量干度的预报公式．

Nanofluid Heat Transfer Enhancement for Nuclear Reactor Applications

Colloidal dispersions of nanoparticles are known as ＇nanofluids＇. Such engineered fluids offer the potential for enhancing heat transfer, particularly boiling heat transfer, while avoiding the drawbacks （e.g., erosion, settling, clogging） that hindered the use of particle-laden fluids in the past. At Massachusetts Institute of Technology （MIT）, the authors have been studying the heat transfer characteristics of nanofluids for the past five years, with the goal of evaluating their benefits for and applicability to nuclear power systems （e.g., primary coolant, safety systems, severe accident mitigation strategies）. This paper summarizes the MIT research in this area with particular emphasis to boiling behavior, including, prominently, the Critical Heat Flux limit and quenching phenomena.

基于平板热管技术的电子器件热管理研究进展

随着电子器件不断趋于小型化和高度集成化发展,电子器件功率过高引发的散热问题亟待解决。平板热管由于具有均温性好、散热效率高等优势已为许多大功率电子元器件的散热问题提供了有效的热管理解决方案。但面对如今越来越多元化的散热需求,开发适应新型电子设备散热需求的高效平板热管仍是当前研究的热点。基于此,在对平板热管工作原理及结构特点进行阐述的基础上,对平板热管的传热性能、影响传热性能的因素(毛细芯结构、腔体厚度、工质、充液率、倾角和热源)、强化平板热管传热性能的措施(不同润湿性表面和支撑柱结构等)及其在小型化电子设备、IGBT晶体管、LED和电池组等电子器件散热应用等方面进行了系统性综述,指出研究不同毛细芯结构与腔体厚度下平板热管内部流动传热特性、散热性能强化的机制和综合优化方法,以及如何在不同电子器件热管理需求下设计出空间适应性强而散热性能高效的平板热管等仍是当前平板热管有待进一步突破的关键技术,将为平板热管在电子器件散热领域的进一步深入研究与优化提供一定参考。

Liquid film dryout model for predicting critical heat flux in annular two-phase flow

Gas-liquid two-phase flow and heat transfer can be encountered in numerous fields, such as chemical engineering, refrigeration, nuclear power reactor, metallurgical industry, spaceflight. Its critical heat flux (CHF) is one of the most important factors for the system security of engineering applications. Since annular flow is the most common flow pattern in gas-liquid two-phase flow, predicting CHF of annular two-phase flow is more significant. Many studies have shown that the liquid film dryout model is successful for that prediction, and determining the following parameters will exert predominant effects on the accuracy of this model: onset of annular flow, inception criterion for droplets entrainment, entrainment fraction, droplets deposition and entrainment rates. The main theoretical results achieved on the above five parameters are reviewed; also, limitations in the existing studies and problems for further research are discussed.

Experimental Investigation on Critical Heat Flux for Water Flowing in a Vertical Uniformly Heated Rifled Tube under Near-critical Pressures

This study experimentally investigated the critical heat flux(CHF) of departure from nucleate boiling(DNB) of water flowing under near-critical pressures in a 2 m-long vertical upward rifled tube with the size of Φ35 × 5.67 mm. Operating conditions included pressures of 18–21 MPa, mass fluxes of 475–1000 kg·m^(-2)·s^(-1), inlet subcooling temperatures of 3–5°C, and wall heat fluxes of 40–960 kW·m^(-2). Tube wall temperature distribution and heat transfer performance in different test conditions were obtained. The effects of the operating parameters on CHF were analyzed. Four heat transfer coefficient correlations were evaluated against our experimental data for further investigation of the two-phase heat transfer characteristics. A heat transfer correlation based on Martinelli number utilized in two-phase region and two empirical correlations used to predict the CHF in rifled tube at near-critical pressures were proposed. Meanwhile, experimental CHF data in rifled tube were compared with six widely used correlations and a CHF look-up table. The CHF enhancement effect in rifled tube is obvious as compared with the CHF data in smooth tube. Results show that DNB occurs at low vapor quality and subcooled region in the rifled tube at near-critical pressures. The increase in pressure leads to the early occurrence of DNB and the decrease in CHF, whereas the increase in mass flux delays the occurrence of DNB and results in the increase in CHF. DNB presents a tendency to move toward the inlet of the rifled tube. At individual operating conditions, DNB and dryout coexist at different sections of the rifled tube.

Critical Heat Flux(CHF)Correlations for Subcooled Water Flow Boiling at High Pressure and High Heat Flux

The subcooled water flow boiling is beneficial for removing the high heat flux from the divertor in the fusion reactor,for which an accurate critical heat flux(CHF)correlation is necessary.Up to now,there are many CHF correlations mentioned for subcooled water flow boiling in the open literatures.However,the CHF correlations’accuracies for the prediction of subcooled water flow boiling are not satisfactory at high heat flux and high pressure for reactor divertor.The present paper compiled 1356 CHF experimental data points from 15 independent open literatures and evaluated 10 existing CHF correlations in subcooled water flow boiling.From the evaluation,the W-2 CHF correlation performs best for the experimental CHF data in all existing critical heat flux correlations.However,the predicted mean absolute error(MAE)of the W-2 correlation is not very ideal for all database and the MAE of the W-2 correlation is from 30%to 50%for some database.In order to enhance the CHF prediction accuracy in subcooled water flow boiling at high heat flux and high pressure,the present paper developed a new CHF correlation.Compared with other existing CHF correlations,the new CHF correlation greatly enhances the prediction accuracy over a broad range of pressures and heat fluxes which are desired in the cooling of high heat flux devices,such as those in the fusion reactor divertor.The validation results show that the new correlation has a MAE of 10.05%and a root mean squared error(RMSE)of 16.61%,predicting 68.1%of the entire database within±10%and 81.5%within±15%.The MAE of the new CHF correlation is 7.4%less than that of the best existing one(W-2 correlation),further confirming its superior prediction accuracy and reliability.Besides,the new CHF correlation works well not only for a uniform power profile but also for a non-uniform power profile in subcooled water flow boiling at high pressure and high heat flux.

Critical Heat Flux of Blast Furnace Hearth in China

The critical heat flux surveys of thirteen Chinese blast furnaces were carried out. The mathematical model of hearth bottom was established and the temperature field was simulated by utilizing the method of inverse problem based on the collected parameters and temperature data. The critical heat flux and dangerous critical heat flux of hearth were defined and analyzed as well as the initial and investigative critical heat flux of hearth, and the influences of thermal conductivity and residual thickness of carbon bricks on critical heat flux were discussed. The relationships between critical heat flux of stave and hearth bricks were also compared. It is found that the dangerous critical heat flux of these blast furnaces ranged from 9.38 to 57 kW/mz. Therefore, there was no uniform critical heat flux of hearth due to the structure design, refractory materials selection, construction quality of hearth and other factors. The heat flux should be lower than the critical heat flux with corresponding thickness of carbon bricks to control the erosion of hearth. The critical heat flux of stave would be much lower than that of hearth bricks with the air gap. However, the critical heat flux of stave should be higher than that of hearth bricks when gas existed between furnace shell and staves.

Critical Heat Flux of Boiling Heat Transfer in aModerate Narrow Space

Boiling heat transfer process is analyzed in a moderate narrow space consisted of two horizontal plates.The main difference between this process and the conventional unconfined pool boilillg is the liquidsupply mechanism which is absolutely prevented by the growth of coalescence bubble along with theheated surface in the narrow space. As a result, the macrolayer becomes thinner due to the evaportion of the individual bubbles within the macrolayer during the period of bubble coalescence, with orwithout dryout that depends on both the gap size of narrow space and the size of heated surface. Fora specified size of the heated surface, the initial thickness of the liquid layer has a critical value whichapproaches a constant while the space height is larger than a critical value. The individual bubblebehaviors and tlie heat transfer can be considered as the same as that in the unconfined pool boiling, ifthe space gap is large. However, the individual bubbles do not generate in the last period of the bubblecoalescence and a lower maximum heat fiux will be resulted if the space gap is reduced. In such a case,the macrolayer is dryout.

High-Quality Critical Heat Flux in Horizontally Coiled Tubes

An investigation on the high-quality dryout in two electrically heated coiled tubes with horizontally helix axes is reported. The temperature profiles both along the tube and around the circumference are measured, and it is found that the temperature profiles around the circumference are not idelltical for the cross-sections at different parts of the coil. The 'local condition hypothesis' seems applicable under present conditions, and the critical heat flux qcr decreases with increasing critical quality xcr The CHF increases as mass velocity and ratio of tube diameter to coil diameter (d/D) increases, and it seems not to be effected by the system pressure. The CHF is larger with coils than that with straight tubes, and the difference increases with increasing mass velocity and d/D.

铜表面-氧化石墨烯纳米涂层池沸腾换热特性实验

石墨烯材料因其优异的导热性能和化学惰性,在核反应堆先进换热设备领域有很大的应用前景。为探索石墨烯材料的强化换热性能,本文采用沸腾沉积法制备氧化石墨烯纳米涂层,开展了氧化石墨烯涂层-去离子水池沸腾换热特性实验。氧化石墨烯纳米流体工质浓度为0.001、0.003、0.005 mg/mL,加热时间控制为1 h和2 h,通过沸腾沉积在裸露铜表面制备6种氧化石墨烯涂层。以此为基础开展池沸腾实验,结果发现氧化石墨烯涂层对池沸腾换热系数的影响很小,但可显著增强池沸腾临界热流密度。从涂层表面润湿性、热导率及形貌结构变化等角度分析了氧化石墨烯涂层对池沸腾换热特性的影响机理。

老化效应对涂层表面池沸腾传热性能的影响

压力容器外部冷却(ERVC)是第三代压水堆核电站严重事故缓解策略的关键技术,其表面临界热流密度(CHF)决定了压力容器的完整性。为探究压力容器真实材料SA508钢表面老化对涂层表面CHF特性的影响,开展了SA508钢光表面和基于SA508钢的TC4钛合金多孔涂层表面在不同老化状态下的沸腾换热特性实验研究。结果表明:SA508钢光表面在水中会快速氧化,随着老化的加深,其CHF呈上升趋势,上升幅度在20%以内;CHF增强的原因在于SA508钢氧化生成了能改善表面润湿性的Fe3O4磁性微纳米颗粒层。随着氧化的加深,TC4涂层表面由超亲水特性变为疏水特性,基体SA508钢生成的氧化物易堵塞涂层孔隙,CHF呈恶化趋势。