换热器管束排列角对流体力的影响研究

排列角通过改变换热管束的排布方式来影响流场分布,进而影响流体力的幅频特性。采用大涡模型(LES)对工程中常见排列角管束的流场分布和所受流体力进行了数值计算及相应的试验验证。计算结果表明:管束排列角明显影响漩涡脱落形成和湍流扰动,特别是对错排管束和上游管束的流体力影响尤为复杂。数值计算结果与试验结果吻合较好。研究结果为流体诱导振动分析提供了参考和依据。     

含均匀减薄缺陷传热管的爆破压力预测研究

蒸汽发生器传热管微振磨损等体积型缺陷是导致其承压能力降低而破裂的主要原因.本文针对含均匀减薄缺陷Inconel690蒸汽发生器传热管的极限承载能力进行了研究.通过液压爆破实验获得了含缺陷传热管的爆破压力,确定了缺陷尺寸对传热管爆破压力的影响规律;采用数值模拟方法对含缺陷传热管的爆破压力进行了预测,在此基础上建立了爆破压力的预测公式.将预测公式与已有工程方法进行对比,结果表明该公式是合理且保守的.     

基于内流场数值模拟的微湍流发生器机械结构优化设计方法

根据微湍流发生机理,探讨微湍流发生器内浆料悬浮液的微湍流特性和衡量指标.以纸浆为介质,通过引入合理的纸浆纤维流动假设来建立数学模型和控制方程.将微湍流强度和湍流尺度等作为目标值,提出基于计算流体动力学(CFD)数值模拟的微湍流发生器机械结构优化设计方法,并将其应用于圆变方管束型和阶梯扩散型两类结构形式,进行性能比较研究.研究结果表明:阶梯扩散型微湍流发生器的性能要优于圆变方管束型结构形式,算法和结论可用于指导微湍流发生器的具体结构优化设计.     

直流过热刺刀管式蒸汽发生器设计分析

采用直流过热刺刀管式蒸汽发生器作为自然循环池式铅铋快堆的主蒸汽发生器。刺刀管式蒸汽发生器传热管由内外两根管同心组成，使得环形通道内工质与内管工质及壳侧工质同时换热，造成设计困难。提出了刺刀管式蒸汽发生器离散式设计方法，,并利用该方法设计了热功率为37.5 MW的直流过热式刺刀管式蒸汽发生器。从换热管数量、环形通道尺寸、内管保温层的导热系数三个方面对刺刀管式蒸汽发生器进行了设计分析，同时采用JF因子对蒸汽发生器的综合性能进行评价。研究表明，当换热管数量由244根增加到550根，一次侧流阻由40 kPa下降至5 kPa,JF因子增加0.4,蒸汽发生器性能提升；环形通道宽度增加0.5 mm,换热段长度增加10%左右，总传热系数和JF因子分别下降13%和14%,蒸汽发生器性能下降；内管保温层导热系数与换热段长度成线性关系，保温导热系数增加，蒸汽发生器的综合性能下降。因此，刺刀管式蒸汽发生器设计宜采用较多的换热管数量、导热系数较小的内管保层材料，同时选择合适的环形通道尺寸，获得最佳的JF因子值。     

发夹式换热器管束流致振动数值模拟研究

运用计算流体力学方法,采用ANSYS CFX软件对发夹式换热器的壳侧流场进行了三维数值模拟。流场计算中采用多孔介质模型对管束区域进行简化,分析了壳侧流场的速度分布,结果表明:直管段部分的流体湍流强度大于弯管段,且外层管束所在区域为高流速区,受流体冲刷严重。结合流场信息,通过功率谱生成随机激振力,采用ABAQUS软件模拟计算了湍流激振下管束的振动响应,结果显示管束的面外均方根位移远大于面内位移,且弯管部分的振动位移最大。该研究结果可为发夹式换热器的性能分析和优化设计提供参考和依据。     

蒸汽发生器单根传热管振动特性试验和分析

掌握蒸汽发生器单根传热管的振动特性是开展管束区流致振动分析评价的基础,针对某型蒸汽发生器,选取弯管半径最大的局部单根传热管,通过试验方法研究传热管管内外附加水质量、管内压力、防振条支撑间隙以及防振条支撑位置传热管正常磨损等因素对振动特性的影响,并通过试验结果与分析结果进行对比,论证设计间隙情况下分析模型和边界条件的合理性。研究结果表明,附加水质量会造成传热管频率降低,管内压力和传热管正常磨损不会影响传热管振动特性,连续两个防振条支撑点间隙大于0.13 mm会造成传热管弯管区开始出现支撑失效,处于设计名义间隙内的传热管在防振条位置的边界条件可以近似简化为简支。     

压水堆燃料棒湍流激励流致振动响应计算方法研究

当湍流流过燃料棒表面时,湍流速度分量产生表面随机压力脉动。采用随机振动响应的模态分析技术,得出圆柱体横向振动位移的均方值一般表达形式。根据压水堆燃料棒的结构和流场分布特征,将作用在燃料棒上的流体力等效为脉动随机载荷,引入湍流功率谱密度函数,结合相关功率谱密度试验参数,推导每阶模态的振动位移均方值。分别讨论轴向和横向流引起的各阶模态的振动幅值,并给出各阶模态的对数衰减阻尼,轴向和横向流引起的振动幅值采用线性组合方法,对各阶模态的采用10%的模态组合方法得到所有模态的燃料棒振动幅值,给出了完整的燃料棒轴向和横向湍流激励振动总幅值的通用计算方法和分析流程。     

流致振动差压式测振装置的研究

为解决管束流致振动试验中,现有接触式振动测量方法的应用局限性,提出一种基于测量受激对象换热管两侧压差脉动频率原理的测振装置,即差压式测振装置。差压式测振装置解决了传统接触式传感器对换热管外部流场的干扰问题以及安装空间受限的问题。为了验证差压式测振装置的有效性,分别进行了单管和换热管束两类流致振动测试试验。试验结果表明,差压式测振装置能在复杂的流动环境中准确捕捉到换热管受到的各类激振形式的频率,而且单管流致振动试验结果表明差压式测振装置所测得的周期性涡激振动频率与理论值偏差在5%以内。因此,差压式测振装置的研制与应用对于流致振动试验的测试以及换热管振动的长期监测均有很重要的意义。     

压水堆蒸汽发生器三维两相热工水力分析程序开发与验证

为解决我国核电厂蒸汽发生器二次侧三维两相热工水力分析程序缺乏的问题,针对压水堆核电厂蒸汽发生器结构特点,以CFD软件中多孔介质模型为基础,采用UDF对CFD软件进行了二次开发,完成了蒸汽发生器二次侧三维两相热工水力分析程序的开发。基于模化分析采用小规模管束进行流动传热试验,模拟蒸汽发生器一、二次侧流场,获得试验体试验工况下的一、二次侧流场数据。采用开发的程序对试验件建模分析,结果表明:该程序能够实现对蒸汽发生器试验件的模拟,预测结果与试验测量结果符合良好,证明了该程序计算结果的可靠性。     

气泡发生器内三相流动数值模拟

本文运用计算流体力学方法并采用非结构化网格技术对流动区域进行了网格划分,应用三相流混合模型对自行设计的射流微泡发生器进行了气、液、固三相流的流动仿真,得到了其内部三相流的流场分布,结果为射流微泡发生器的设计研究提供依据.     

Experimental investigation on sliding and fretting wear of steam generator tube materials

In nuclear power steam generators, high flow rates can induce vibration of the tubes resulting in fretting wear damage due to contacts between the tubes and their supports. In this paper, the sliding and fretting wear tests were performed using Inconel 600HTMA and 690TT against STS 304, which are the steam generator tube materials. The sliding wear tests with a pin-on-disk type tribometer were carried out under various applied loads and sliding speeds at air environment. The fretting wear tests were carried out under various vibrating amplitudes and applied normal loads.

The result of sliding and fretting wear tests show that the heat-treated Inconel 690TT has better wear resistance than Inconel 600HTMA in air. The fretting wear regimes were plotted using the test results and the wear coefficient was calculated also. From the results, it was observed that the wear and tear by stick-slip has very strong effect on the fretting wear behavior.     

Impact Fretting Wear Behavior of Alloy 690 Tubes in Dry and Deionized Water Conditions

The impact fretting wear has largely occurred at nuclear power device induced by the flow-induced vibration, and it will take potential hazards to the service of the equipment. However, the present study focuses on the tangential fretting wear of alloy 690 tubes. Research on impact fretting wear of alloy 690 tubes is limited and the related research is imminent. Therefore, impact fretting wear behavior of alloy 690 tubes against 304 stainless steels is investigated. Deionized water is used to simulate the flow environment of the equipment, and the dry environment is used for comparison. Varied analytical techniques are employed to characterize the wear and tribochemical behavior during impact fretting wear. Characterization results indicate that cracks occur at high impact load in both water and dry equipment; however, the water as a medium can significantly delay the cracking time. The crack propagation behavior shows a jagged shape in the water, but crack extended disorderly in dry equipment because the water changed the stress distribution and retarded the friction heat during the wear process. The SEM and XPS analysis shows that the main failure mechanisms of the tube under impact fretting are fatigue wear and friction oxidation. The effect of medium(water) on fretting wear is revealed, which plays a potential and promising role in the service of nuclear power device and other flow equipments.     

Fretting wear of laterally supported tube

The fretting wear of a tube, which is in contact with a lateral support, is examined experimentally. A fretting wear tester is specifically designed. Elastic springs are used as the support, which can simulate the contact between a spacer grid and a fuel rod in pressurized water reactor fuel. The tubes and the springs are made of Zircaloy-4. The experiments are conducted in air at room temperature. The experimental conditions, i.e. the normal and shear forces on the contact, the slip range and the number of cycles, are set to be the same. To investigate the influence of the contact geometry on the wear, the spring supports have a concave, a flat or a convex contour. The influence on the axial and transverse slip directions is investigated to incorporate the actual tube motion caused by such a flow-induced vibration in the reactor. The wear on the tube is examined by the surface roughness tester, which measures the depth, and the contour of the worn surface of the tube. Since the shape and the distribution of wear are found arbitrary, a method for evaluating the wear volume is proposed using the signal processing technique. It is found that wear can be restrained when the slip direction is transverse, and if the support has a concave contour.     

Fretting wear mechanisms of Zircaloy-4 and Inconel 600 contact in air

The fretting wear behavior of the contact between Zircaloy-4 tube and Inconel 600, which are used as the fuel rod cladding and grid, respectively, in PWR nuclear power plants was investigated in air. In this study, number of cycles, slip amplitude and normal load were selected as the main factors of fretting wear. The results indicated that wear increased with load, slip amplitude and number of cycles but was affected mainly by the slip amplitude. SEM micrographs revealed the characteristics of fretting wear features on the surface of the specimens such as stick, partial slip and gross slip which depended on the slip amplitude. It was found that fretting wear was caused by the crack generation along the stick-slip boundaries due to the accumulation of plastic flow at small slip amplitudes and by abrasive wear in the entire contact area at high slip amplitudes.     

Three dimensional fretting wear analysis by finite element substructure method

In steam generator of nuclear power plant, flow induced vibration in a U-tube bundle could cause wear and fatigue failure. This vibration causes fretting-wear between the supporting plates and tubes by generating infinitesimal friction. In this study, a substructure method is developed for three dimensional finite element models of fretting wear problems and its feasibility is also verified. The substructure method can reduce large amount of computation time required by conventional finite element analysis.     

The experiment of flow induced vibration in PWR RCCAs

Recently, severe wear on the shutdown rod cladding of Ulchin Nuclear Power Plant #1, #2 were observed by the Eddy Current Test (E.C.T.). In particular, the wear at the sixth card location was up to 75%. The test results indicated that the Flow Induced Vibration (F.I.V.) might be the cause of the fretting wear resulting from the contact between Rod Cluster Control Assemblies (RCCAs) and their spacing cards (guide plates) arranged in the guide tube. From reviewing RCCAs fretting wear reports and analyzing the general characteristics of F. I. V. mechanism in the reactor, geometric layout and flow conditions arround the control rod, it is concluded that the turbulence excitation is the most probable vibration mechanism of RCCA. To identify the governing mechanism of RCCA vibration, an experiment was performed for a representative rod position in which the most serious fretting wear was experienced among the six rod positions. The experimental rig was designed and set up to satisfy the governing nondimensional numbers which are Reynolds number and mass damping parameter. The vibration amplitude measurement by the non-contact laser displacement sensor showed good agreements in the frequency and the maximum wearing (vibration) location with Ulchin E. C. T. results and Framatome report, respectively. The sudden increase in the vibration amplitude was sensed around the 6th guide plate with mass flow rate variation. Comparing the similitude rod behaviour with the idealized response of a cylinder in flow induced vibration, it was found that the dominant mechanism of vibration was transferred from turbulence excitation to periodic shedding at the mass flow rate 90l/min. Also the critical velocity of the vibration in RCCAs was determined and the vibration can be prevented by reducing the bypass flow rate below the critical velocity.     

Experimental Study of the Fretting Wear Behavior of Incoloy 800 Alloy at High Temperature

The fretting wear behavior of the nuclear power material Incoloy 800 was investigated in this study. A PLINT high-temperature fretting tester was used on an Incoloy 800 cylinder against a 304SS cylinder at vertical cross contact under different temperatures (25, 300, and 400°C). During testing, a normal load of 80 N was applied, and the displacement amplitudes ranged from 2 to 40 µm. The fretting wear mechanism at high temperatures and the kinetic character of the materials of the Incoloy 800 steam generator tube were analyzed. Results showed that the fretting running regimes varied little with ncreasing temperature, and some microcracks were observed in both the mixed fretting regime (MFR) and the partial slip regime (PSR) at high temperatures. Slight abrasive wear and microcracks were the main wear mechanisms of the Incoloy 800 alloy in PSR, whereas those in the MFR and the gross slip regime were oxidative wear, abrasive wear, and delamination.     

Friction and wear of Inconel 690 for steam generator tube in elevated temperature water under fretting condition

Recently, material of Inconel 690TT (thermal treatment) for the steam generator tubes in a nuclear power plant was substituted for the existing material of Inconel 600HTMA (high temperature mill-annealing). Inconel 690TT has more chromium than Inconel 600HTMA in order to improve the corrosion resistance. In this study, to evaluate the friction and wear characteristics of Inconel 690TT under fretting condition, the fretting tests as well as sliding tests were carried out in air and in elevated temperature water environment, respectively. Fretting tests of the cross-cylinder type were done under various applied normal loads, and sliding tests of pin-on-disk type were also carried out to compare with the results of the fretting test. In summary, the results of the fretting tests correlated with the results of the sliding tests. The wear mechanism of Inconel 690TT in air was delamination wear and the mechanism in water was affected by micro-pitting. Also, it was found that the fretting wear coefficients in water were increased with increase in the temperature of water.     

恒定动能作用下薄壁管的冲击微动磨损行为研究

在新型冲击微动磨损试验机上对四种常见材料的薄壁管(不锈钢、铜合金、纯钛和铝合金)进行了冲击磨损试验,考察了材料属性、冲击能量对薄壁管损伤行为的影响。对其冲击动力学行为、磨损行为进行了分析。研究结果表明,不同材料金属管的能量吸收率、冲击接触力和冲击管变形有显著差异;同一种材料,随着初始冲击动能的增大,冲击过程中接触力、冲击管变形和冲击吸收能也在增大。通过分析磨痕微观形貌和磨痕轮廓,发现薄壁管的冲击磨损抵抗性能与材料属性密切相关;随着初始冲击动能的增加,材料损伤加剧,其损伤机制为疲劳磨损。