非稳态流体力作用下两相流管束流弹失稳研究

流弹失稳会引起传热管振幅过大而发生磨损破坏,是两相流作用下蒸汽发生器管束流致振动的重要机理。为了较为准确地预测两相流作用下圆柱管的失稳临界流速,对试验测量的两相流非稳态流体力进行参数拟合,建立了气-水两相流作用下单管的动力学模型。通过无量纲化,运用Galerkin方法对方程变量进行离散后,联立求解方程得到了不同空泡份额的临界流速。数值结果表明,数值解与试验测得失稳临界流速较为吻合,验证了该模型可用于两相流传热管临界流速的预测。      

垂直上升横掠水平管束的两相流空泡份额模型研究

与管内两相流空泡份额模型相比,垂直上升横掠水平管束的两相流空泡份额研究成果相对有限。利用垂直上升的气-水两相流横掠水平管束的实验数据,对现有的空泡份额计算模型进行对比分析,并对2种现有模型的拟合公式进行修正。采用其他实验结果对本文重新修正的空泡份额模型进行验证,结果表明:与原始模型相比,修正的空泡份额计算模型给出的空泡份额预测值更好。     

核电厂蒸汽发生器传热管束流弹失稳评价关键参数分析及实验研究

针对某压水型核电厂新研发蒸汽发生器的传热管，采用ASMEBPVC-Ⅲ推荐的半经验公式及相应参数取值，计算得到了悬臂传热管在空泡份额为0%（单相水）、10%、20%、50%、80%、90%下的流弹失稳临界流速。同时，设计开展了悬臂传热管阵在各空泡份额下的流致振动实验，测得了传热管流弹失稳临界流速、动水中的振动阻尼比及固有频率等关键参数。实验中测得的振动阻尼比主要包含了两相阻尼与粘滞阻尼，随空泡份额的变化而变化，范围为1.51%～3.98%，考虑测量不确定度后，该值可用于本文所述蒸汽发生器设计，且具有一定的保守性。分析结果表明，规范推荐的公式及参数计算所得传热管流弹失稳临界流速和实验结果趋势相同、规律一致，前者较后者有较大的保守性，安全系数在1.5以上；采用实验测得的阻尼比及固有频率重新计算得到的临界流速安全系数有所下降，但仍高于1.1。通过实验和分析，讨论了文中所述新研发核电厂蒸汽发生器传热管束流弹失稳评价关键参数取值及分析方法的合理性与保守性，可用于工程产品的设计及分析。     

支承板支撑多跨管束流致振动试验研究

本文针对支承板支撑4跨传热管直管束开展流致振动基础试验。试验件由49根旋转正三角形布置的模拟传热管组成,传热管两端固定,中间3处采用支承板支撑。试验测量获得了单向横流冲刷和双向横流冲刷下不同进口流速传热管束的振动特性,获得振幅、频率、临界雷诺数等关键信息。结果表明,双向横流冲刷下的传热管较单向横流冲刷下的在更低雷诺数下发生失稳,两种流动方式下传热管发生失稳时加速度峰值频率均为104Hz,该值与单跨两端固支模型的理论计算固有频率非常接近。研究结果可为传热管束流致振动数值模拟分析提供验证。     

振动幅度对两相流局部参数变化影响研究

为研究实验段振动对管内两相流局部参数变化的影响,利用电导探针技术对振动状态下局部两相流特性参数包括空泡份额、气泡直径和界面浓度进行了测量。实验首先在静态工况下进行,通过固定在实验段上方的偏心轮转动获得振动工况。实验段振动周期保持在0.5 s,偏心轮提供的振动幅度分别为4.8 mm、9.5mm和15.8 mm。实验结果表明,振动对环管内气-水两相流局部时均参数分布影响很小。但振动引起的附加惯性力作用使两相流局部参数径向分布在实验段振动周期中发生明显变化,而且局部参数的变化幅度随实验段振幅的增加而显著增大。在含气率较低的流动工况,当振幅增大到15.9 mm时振动工况下径向空泡份额峰值较静态工况下的空泡份额峰值的增量可以达到70%。但振动对局部流动参数的影响随气流量增大而降低。     

蒸汽发生器传热管间隙对传热管动态特性的影响分析

传热管流致振动是核电厂蒸汽发生器传热管失效的主要原因之一,在核电厂设计蒸汽发生器时,需对蒸汽发生器传热管流致振动问题进行分析。传热管与支撑板及抗振条之间存在小尺度间隙,这类间隙具有非线性效应,在进行流致振动线性分析时应考虑对间隙进行线性化等效处理。本文从理论研究和模拟分析两方面出发,对传热管与支撑板及抗振条之间间隙对传热管动态特性的影响进行分析。理论和模拟分析可知,传热管间隙对传热管整体振动的作用接近于简支。在进行流致振动分析时,可采用简支代替间隙进行线性分析。     

窄矩形通道内搅混流和环状流相界面参数的计算方法

空泡份额和界面浓度是两相流动中重要的相界面参数,准确获取窄矩形通道内搅混流和环状流工况下空泡份额和界面浓度是构建和完善两流体模型的关键。本文针对横截面为65 mm×2 mm的矩形通道开展了气液两相流动特性可视化实验研究,气相折算速度j_g＝1～9 m/s,液相折算速度j_f＝0.1～1.5 m/s,流型包含搅混流和环状流。提出了基于高速摄像法获取搅混流和环状流下空泡份额和界面浓度的分析计算方法,利用该方法所得空泡份额与窄矩形通道内经验关系式计算值的相对偏差约在10%以内。此计算方法可为研究复杂流型下窄矩形通道内的相界面参数提供理论依据。     

传热管流致振动设计准则比较

蒸汽发生器是核反应堆中将一回路热量传递给二回路（水-蒸汽动力回路）的重要设备。二次侧流体冲刷引起的传热管振动是管壁疲劳、磨损直至破裂的主要原因之一。为确保传热管的结构完整，有必要对传热管开展流致振动设计与评价。针对传热管这样的圆柱形结构的流致振动设计与评价，有3种设计规范：GB/T151、TEMA、ASME。本文对3种设计规范中传热管流致振动的设计准则进行了比较，指出了相关准则和参数在计算上的差异以及对流致振动相关机制判定上的差异。选取某典型热交换器为算例，具体展现了3种设计规范对传热管流致振动评价结果的异同，给出了不同规范设计准则的选取对于表征流致振动的特征参数（如附加质量、固有频率、漩涡脱落频率、振幅等）计算结果带来的偏差，并对3种设计规范在工程应用中的保守性提出了参考性建议。结果表明：对不同参数，设计规范保守性不相同，TEMA计算频率相对更为保守，其他参数GB/T151较为保守，TEMA和ASME整体相对GB/T151更加宽容。     

基于ANSYS的蒸汽发生器传热管流致振动分析程序

基于通用有限元软件ANSYS的APDL语言编写蒸汽发生器传热管流致振动分析程序。采用三维梁单元建立传热管有限元模型,对传热管进行模态分析,计算传热管的流弹不稳定率和湍流激励响应,并与专用流致振动计算软件分析结果进行对比。结果表明,模态分析以及流弹不稳定率计算结果与流致振动专用计算软件分析结果一致,湍流激励响应更偏于保守。计算程序基于通用有限元软件,较专用软件建模方便、可读性强、适用范围广泛,可大大提高实际工程分析效率。     

摇摆状态下矩形通道内弹状流压力模型

本文通过可视化方法对竖直与倾斜条件下矩形通道内弹状流单元的参数进行研究,尝试给出摇摆状态下矩形通道内弹状流压力模型。通过图像处理给出气弹段空泡份额以及两相速度的计算关系式,并验证漂移流模型在液弹段的适用性,给出弹状流单元的长度份额以及空泡份额的计算关系式。根据实验结果给出摇摆条件下矩形通道内弹状流压力组分的模型,并重点分析摩擦压降模型的适用程度。结果表明,弹状流压力模型可很好地预测摇摆条件下矩形通道内的压力。     

非均匀两相流传热管流致振动中阻尼模拟研究

In the flow-induced vibration analysis of the heat exchanger tube of the stream generator, the damping of each position on the tube is different, since the secondary side of the tube is the two-phase flow (stream-water) and the void-fraction is gradually increased from bottom to top. It is necessary to study the simulation of the tube damping in non-uniform two-phase flow. The study is based on the Pettigrew's damping formula of tube in two-phase flow and the void-fraction distribution along the tube in the typical example. For the two-phase damping component of the tube damping, the disadvantage of damping overestimation of void-fraction processing method in common engineering software are analyzed. The reason is the nonlinearity of the void-fraction influence coefficient. The method of segmentation weighting void-fraction is developed. The effects of different segment lengths are studied, indicating that the section lengths should be minimized. For the subsequent segmentation damping weighting problem, the effects of different weighting factors in engineering methods and standards are compared, and the difference is small. The results of flow-induced vibration analysis with different damping inputs were compared to judge the applicability of Pettigrew's damping formula. From the above four researches, the recommended simulation method of the tube damping in non-uniform two-phase flow is given to more accurately carry out the flow-induced vibration analysis of heat exchanger tube.     

核承压热交换器两相流流致振动现象研究

为保障核承压热交换器的安全运行,采用数值模拟以及软件计算相结合的方法,对核承压热交换器两相流流致振动现象及减振措施进行了探究。研究结果表明:基于流致振动发生机理,热交换器横流速度、固有频率、卡门旋涡脱落频率以及紊流抖振频率为重点分析因素;由公式得出流量、换热管直径、换热管壁厚、管束排列等对流致振动有直接影响,无支撑跨距是影响管束流致振动较大因素;最易发生流致振动的部位包括入口区域、出口区域、折流板缺口区域以及无支撑跨距大管束;设计中,应在流量、换热管直径、壁厚、无支撑跨距、管束排列及入口防冲挡板设置等方面优化,以减小流致振动危害。     

A review of heat exchanger tube bundle vibrations in two-phase cross-flow

Flow-induced vibration is an important concern to the designers of heat exchangers subjected to high flows of gases or liquids. Two-phase cross-flow occurs in industrial heat exchangers, such as nuclear steam generators, condensers, and boilers, etc. Under certain flow regimes and fluid velocities, the fluid forces result in tube vibration and damage due to fretting and fatigue. Prediction of these forces requires an understanding of the flow regimes found in heat exchanger tube bundles. Excessive vibrations under normal operating conditions can lead to tube failure.

Relatively little information exists on two-phase vibration. This is not surprising as single-phase flow induced vibration; a simpler topic is not yet fully understood. Vibration in two-phase is much more complex because it depends upon two-phase flow regime, i.e. characteristics of two-phase mixture and involves an important consideration, which is the void fraction. The effect of characteristics of two-phase mixture on flow-induced vibration is still largely unknown. Two-phase flow experiments are much more expensive and difficult to carry out as they usually require pressurized loops with the ability to produce two-phase mixtures. Although convenient from an experimental point of view, air–water mixture if used as a simulation fluid, is quite different from high-pressure steam–water. A reasonable compromise between experimental convenience and simulation of steam–water two-phase flow is desired.

This paper reviews known models and experimental research on two-phase cross-flow induced vibration in tube bundles. Despite the considerable differences in the models, there is some agreement in the general conclusions. The effect of tube bundle geometry, random turbulence excitations, hydrodynamic mass and damping ratio on tube response has also been reviewed. Fluid–structure interaction, void fraction modeling/measurements and finally Tubular Exchanger Manufacturers Association (TEMA) considerations have also been highlighted.     

换热器管束流体激振研究的新思路

提出了一种快速简便地求解非定常粘性流方程的方法、以得到换热器管束复杂流道内作用于振动管子上的流体激振力。简述了如何将管束流动的流体弹性稳定性分析与工程实际的振动疲劳破坏相联系，以及怎样在求解流动基本方程的基础上分析管束复杂流道内的振荡压力传播。将求解复杂流动的快速方法—参数多项式方法与流体激振的全功能分析及振荡压力传播理论结合起来，为管束流体激振研究开拓了一个新的途径。     

Three-dimensional study on steady thermohydraulics characteristics in secondary side of steam generator

Steam generator (SG), as the primary-to-secondary heat exchanger and pressure boundary of primary loop, should be integrated and perform well in heat transfer ability. Flow characteristics of the secondary side fluid of SG are essential to analyze U-tube wastage caused by the flow-induced vibration and thermal stress. In this paper, secondary side two-phase flow was simulated based on the porous media model. Additional momentum and energy source terms were appended to the momentum and energy equations for porous media region, respectively. The additional momentum source contained the resistances of downcomer, tube bundle, support plate and separator. The additional energy source included the heat transfer from primary side to secondary side fluid. Solving the governing equations by ANSYS FLUENT solver yielded the distributions of velocity, temperature, pressure, density and quality, which can be used in the analysis of flow-induced vibration and separators. The thermal-hydraulic characteristics of hot side differed from these of cold side considerably. The minimum flow quality of cold side was 0.07, while the maximum one of hot side was 0.71; the average flow quality of outlet was 0.272. The flow rate in the gap of the hot side was 1.02 times of that of the cold side.     

Modeling and analysis of thermal damping in heat exchanger tube bundles

Most structures and equipment used in nuclear power plant and process plant, such as reactor internals, fuel rods, steam generator tubes bundles, and process heat exchanger tube bundles, are subjected to flow-induced vibrations (FIV). Costly plant shutdowns have been the source of motivation for continuing studies on cross-flow-induced vibration in these structures. Damping has been the target of various research attempts related to FIV in tube bundles. A recent research attempt has shown the usefulness of a phenomenon termed as ‘thermal damping’. The current paper focuses on the modeling and analysis of thermal damping in tube bundles subjected to cross-flow. It is expected that the present attempt will help in establishing improved design guidelines with respect to damping in tube bundles.     

Vibration analysis of heat exchanger and steam generator designs

A thorough flow-induced vibration analysis of nuclear components such as heat exchangers and steam generators is essential at the design stage to ensure good performance and reliability. This paper presents our approach and techniques in this respect. In a steam generator, for example, the flow may be liquid or two-phase. In general, parallel and cross-flow exist in the tube bundles of heat exchange components. In cross-flow three basic vibration excitation mechanisms are considered, namely fluidelastic instability, periodic wake shedding resonance, and forced response to random flow turbulence. The latter may need to be considered in parallel flow. These vibration excitation mechanisms and the dynamics of multispan tubes are formulated in a computer model which is used to predict the vibration response of the tubes. The computer model and the parameters required to formulate the vibration excitation mechanism are discussed. Examples of vibration analysis of steam generators and heat exchangers are outlined. It is concluded that most flow-induced vibration problems may be avoided by proper analysis at the design stage.     

Flow-induced vibration: recent findings and open questions

Component failures due to excessive flow-induced vibration are still affecting the performance and reliability of nuclear power stations. Tube failures due to fretting-wear in nuclear steam generators, and vibration related damage of reactor internals are of particular concern. The purpose of this paper is to review some of the recent findings in the area of flow-induced vibration and to discuss some of the remaining questions. Vibration excitation mechanisms and damping mechanisms are described with particular emphasis on fluidelastic instability and damping in two-phase flows. The need for a better understanding of two-phase flow regimes, particularly in cross flow, is outlined. The dynamic characteristics of nuclear structures are explained. The statistical nature of some parameters, in particular support conditions, is discussed. The prediction of fretting-wear damage is approached from several points of view. An energy approach to formulate fretting-wear damage is proposed.     

双跨换热器传热管在静水中阻尼特性的研究

本文对双跨换热器传热管在静水中的结构阻尼与粘滞阻尼作了简介,并对在总阻尼中占主导地位的挤压膜阻尼作了较深入的讨论。Mulcahy 于80年代初提出的挤压膜阻尼数学模型是较为完善的,其缺陷是未计入管子振幅的影响,本文对它进行了修正。为了验证理论分析的结果,笔者对双跨传热管在静水中的挤压膜阻尼作了实验研究,得出挤压膜阻尼与中间支承板厚度、管子-中间支承板间隙以及管子振幅之间的关系。