蒸汽发生器传热管的微动磨损损伤及预测模型

蒸汽发生器是核电厂中能量转换的关键装备,内部高速流经的高温、高压流体引起传热管流激振动,造成传热管微动磨损损伤,严重时发生管道破裂。文章介绍了传热管典型的微动磨损失效案例,相应的模拟实验研究结果,以及机械磨损与冲蚀-腐蚀共同作用的损伤机制。采用工作率模型可对传热管的磨损失效进行合理的寿命预测评估,该预测模型已经在核电厂安全评估方面应用。     

世界核电设备与结构将长期面临的一个问题--微动损伤

核电设备至仍面临许多不易解决的问题,经过对核电和结构部件多例事故的分析可知,核电设备中微动是不可避免的现象,核能工程中的相当一部分结构损伤事故与和微动损伤有着直接的关系;在反应力集中,腐蚀部位,微动又是许多核电设备提前损伤失效的直接原因。     

侧压对传热管用690合金微动疲劳寿命影响的计算分析与估算公式

微动是蒸汽发生器传热管失效的一个主要原因,揭示传热管用690合金的微动疲劳十分重要。本文通过有限元模型和自编程序计算分析了690合金与抗震条间平-平面接触副微动疲劳裂纹萌生寿命,重点研究了侧压对微动疲劳寿命的影响。结果表明,侧压下的裂纹萌生寿命远低于其标准疲劳寿命,降低程度与微动接触状态和微动磨损均有关。在此基础上提出了一个考虑侧压影响的微动疲劳寿命估算公式。该经验公式具有较简单的解析表达式,且对疲劳寿命的计算较为保守,可方便地用于工程设计和寿命预估     

锆合金包壳和GH4169镍基合金的微动摩擦磨损性能研究

燃料棒在冷却剂流过时易受到扰动而发生微振动,导致在格架弹簧与包壳管接触点附近产生微动摩擦磨损,严重时会导致燃料棒破损,放射性产物泄漏,从而影响核电厂安全运行,因而需要对燃料包壳的微动摩擦磨损性能进行充分研究。本研究旨在比较分析2种牌号、2种状态的锆合金（Zr-4）和N36与格架材料GH4169镍基合金在不同环境条件下的微动摩擦磨损性能,分析载荷、循环次数、环境条件对其摩擦磨损性能的影响,并结合磨损表面的形貌、成分分析结果,揭示其微动摩擦磨损机理。研究结果表明,微动摩擦磨损时摩擦系数随载荷的增加呈线性增加趋势;相同条件下,Zr-4/Zr-4摩擦副组合的微动摩擦系数最大,GH4169/N36摩擦副组合的微动摩擦系数最小;预氧化对材料的微动摩擦系数影响显著,预氧化态样品的摩擦系数均高于非预氧化态的样品。      

TiAlZr合金及其热喷涂Ni-Cr合金涂层高温微动磨损的研究

对TiAlZr合金及其热喷涂Ni—Cr合金涂层的高温微动磨损性能进行了对比研究。不同温度下的微动磨损实验结果表明：在滑移区,温度对微动磨损的影响十分显著,TiAlZr合金抗微动磨损能力随着温度上升而急剧下降,热喷涂Ni-Cr合金涂层抗高温微动磨损能力随着温度上升而增强。在400℃时,由于氧化膜快速生成的原因．热喷涂Ni-Cr合金涂层具有优异的抗高温微动磨损特性。     

核电厂常规岛蒸汽系统液压阻尼器漏油原因分析及对策

近年来核电厂常规岛蒸汽系统液压阻尼器液压油泄漏故障多有发生,个别位置液压油泄漏缺陷频发,造成较大的经济损失和设备运维成本.阻尼器液压油泄漏有多方面的原因,本文针对常规岛蒸汽系统高周低幅振动工况下发生的液压油泄漏缺陷阻尼器进行了解体检查,对活塞杆单一位置持续磨损造成液压油密封失效的原因进行了分析.从运行工况、密封部位结构...     

微动摩擦对聚合物自由体积的影响

采用正电子湮没方法对聚四氟乙烯,尼龙以及聚对苯二甲酸丁二醇酯材料微动摩擦后自由体积缺陷的变经规律进行了研究,结果表明,微动摩擦造成了材料内部平均自由体积尺寸增大,导致了材料的磨损及失效。     

核级承压设备密封结构的有限元分析

在AP1000反应堆系统中,很多设备具有承压的功能,其密封性能直接关系到系统能否正常运行,因而密封失效是较之弹塑性失效、疲劳失效等更为基本的失效形式。在ASME规范中采用的密封结构设计方法是华脱尔斯法,此方法采用了一些保守的经验和假设,无法对密封结构处的变形和应力进行细致的计算。本文采用ANSYS有限元分析软件对核承压设备典型的密封结构进行了建模计算,提出了在有限元模型中螺栓预紧力和垫片的等效处理方法,能够对密封结构处垫片的回弹量、法兰的变形及应力分布进行预测。模型分析了采用华脱尔斯法进行密封设计时的设计余量,得到了垫片回弹量与设备内压之间的关系,对于核级承压设备密封结构的设计具有一定的借鉴意义。     

核电设备中螺纹联接结构的松动,损伤机理

根据核电设备结构中的动态力学环境,从宏观和微观力学分析上提出了在振动环境中螺纹联接结构的松动机理,并用微动损伤理论解释了螺纹面的咬死、粘着磨损、产生微裂纹、螺牙断裂等现象,得出了在振动环境中,螺纹联接结构必然存在松动的趋势,即螺纹副自锁性将失效的结论。     

1Cr13不锈钢部分滑移区和滑移区的高温微动磨损特性

采用Deltalab—Nene7型电液伺服式高温微动磨损试验机,研究了核反应堆蒸汽发生器管支撑部件材料1Cr13不锈钢从室温到400℃的微动磨损行为。通过动力学特性分析,结合显微观测．结果发现：1Cr13不锈钢的高温微动行为与微动区域特性密切相关。在滑移区内,随着温度的升高,摩擦系数与磨损降低。微动磨损机理为氧化与剥层：高温氧化效应以及氧化膜的形成随温度的升高而加剧。在微动工况下．氧化膜萌生裂纹,裂纹扩展折向表面,直至氧化膜剥落,形成磨屑。然而。在部分滑移区内,温度对微动行为影响很小。     

The study on grid-to-rod fretting wear models for PWR fuel

The fretting wear is found to be generated at grid-to-rod contact areas by flow-induced vibration. This flow-induced grid-to-rod fretting wear may be initiated at a certain critical grid-to-rod gap that strongly depends on the extent of flow-induced vibration and grid spring designs. Three fretting wear excitation mechanisms acting on the grid-to-rod fretting wear are summarized. In order to examine the impact of grid spring designs on the fretting wear rate, the fretting wear tests for three kinds of grid spring designs were carried out for 500 h, simulating the reactor flow conditions. In parallel, three grid-to-rod fretting wear models that include constant work rate model, constant work density rate model and linear work density rate model have been developed. The three fretting wear models were used to predict the fuel rod perforation times with the use of the fretting wear test results. It is said that the constant work density rate model or the linear work density rate model is quite effective in predicting the grid-to-rod fretting-induced rod failure time observed in commercial nuclear power plants.     

基于核主元分析法的核电厂设备状态监测技术研究

In order to solve the limitations of the traditional monitoring methods for nuclear power plants, this paper proposes to introduce Kernel Principal Component Analysis (KPCA) into the online monitoring field of nuclear power plant equipment, and design the monitoring method and online monitoring strategy. In order to verify the effectiveness of the algorithm, it has been applied in the real monitoring case of the motor driven main feed water pump a nuclear power plant in China. The simulation results show that the KPCA algorithm can adapt to the requirements of nuclear power plant equipment monitoring, and can provide earlier warnings of failure than the existing threshold monitoring methods. At the same time, compared with the conventional PCA algorithm, the KPCA algorithm can extract the nonlinear relationship between variables, identify different operating modes of the device, and effectively reduce false alarms.     

基于核主元分析法的核电厂设备状态监测技术研究

为解决核电厂传统监测手段的局限性,提出将核主元分析法(KPCA)引入核电厂设备在线监测领域中,并设计了监测模型建设方法以及在线监测策略。为验证算法的有效性,将其应用在国内某核电机组电动主给水泵的真实监测案例中。仿真结果表明,KPCA算法可适应核电厂设备监测的要求,能比现有阈值监测手段提供更为早期的故障预警。同时,相比于常规的主元分析法(PCA),KPCA算法能够提取各变量之间的非线性关系,识别出设备不同的运行模式,有效减少误报警。     

On the geometry of the fuel rod supports concerning a fretting wear failure

Geometrical conditions of spacer grid springs and dimples of a light water reactor fuel assembly are studied in this paper concerning a fuel rod’s fretting wear failure. In this framework, the springs/dimples are categorized from the aspects of their orientation with respect to the fuel axis and the contact types. Possible motions on the contacts between the springs/dimples and fuel rods are estimated by conducting a flow-induced vibration test. Features of the wear scar and depth are investigated by independent fretting wear tests carried out with spring and dimple specimens of typical contact geometries. It is also attempted here to apply the contact mechanics theory to a fuel fretting wear analysis such as the prediction of a wear depth profile and its rate, which is influenced by the contact shape of the springs/dimples. It is shown that the theory can be applied to a dimensional control of a coining for the springs/dimples, which is usually carried out in a thin plate fabrication. From the results, the necessary conditions for a spring/dimple geometry for restraining a fretting wear failure are discussed.     

基于局部离群因子和神经网络模型的设备状态在线监测方法研究

核电厂重大设备状态在线监测是保障核电厂安全和经济运行的重要技术,针对传统阈值监测的固有缺陷,提出一种基于局部离群因子（LOF）和神经网络模型的设备状态在线监测方法。此方法属于多参数动态阈值监测方法,首先分析监测对象的故障模式和故障现象,选择一组可覆盖故障现象的传感器测点;根据设备运行特点采集足够长时间的历史运行数据,筛除异常数据;计算历史运行数据的LOF,以历史运行数据为输入、LOF为输出,建立并训练得到神经网络模型;最后基于神经网络模型和传感器测点实时数据计算设备健康指数,监控当前设备健康状态。将本文的监测方法用于循环水泵泵体健康状态的监测,并采集了一段时间的正常数据和异常数据以验证其监测效果,验证结果表明,本文提出的监测方法可以提前10d进行预警,降低误报率,大幅提升监控效能。       

预测性维修在重水堆设备管理中的应用

秦山重水堆核电厂设备数量多、结构复杂，有些设备故障与运行时间无直接关系，并不能简单通过定期维修来避免这些设备失效。为了提高设备可靠性、增加设备可用率并降低维修成本，对于电厂核安全、机组发电具有重要作用的设备，应开展设备状态监测和预测性维修工作。按照各类型设备特点，采用合适的设备状态监测方法、内容和频度，发现一些设备潜在故障，主动安排这些设备进行预测性维修，优化设备维修管理。     

Friction and wear of Inconel 690 and Inconel 600 for steam generator tube in room temperature water

In the steam generators of nuclear power plants, the flow of cooling water can cause the tubes to vibrate, resulting in fretting wear damage due to contacts between these tubes and their supports. The tubes are made of Inconel 690 and Inconel 600 and the supports are made of STS 304. In this paper, fretting wear tests in water were performed using the materials Inconel 690 and Inconel 600 in contact with STS 304. Fretting tests using a cross-cylinder type set up were conducted under various vibrating amplitudes and applied normal loads in order to measure friction forces and wear volumes. Also, conventional sliding tests using a pin-on-disk type set up were carried out to compare these test results.In the fretting tests, friction force was found to be strongly dependent on normal load and vibrating amplitude. Coefficients of friction decreased with an increase in the normal load and a decrease in the vibrating amplitude applied. Also, the wear of Inconel 600 and Inconel 690 was predicted using a work rate model. Depending on the normal load and vibrating amplitude applied, distinctively different wear mechanisms and often drastically different wear rates occurred. It was found that the fretting wear coefficients for Inconel 600 and Inconel 690 were 9.3×10⁻¹⁵ and 16.2×10⁻¹⁵ Pa⁻¹, respectively. This study shows that Inconel 690 can result in lesser friction forces and exhibits less wear resistance than Inconel 600 in room temperature water.     

The effect of fuel rod supporting conditions on fuel rod vibration characteristics and grid-to-rod fretting wear

The grid-to-rod fretting wear-induced fuel rod failure observed in PWRs may be caused by excessive fluid-induced vibration and inadequate fuel rod support by the spacer grid spring. In order to simulate in-reactor grid-to-rod fretting wear behaviors, the grid-to-rod fuel rod supporting conditions as a function of time were predicted by taking into account cladding creep rate, initial spacer grid spring deflection, spacer grid spring force relaxation, etc. Based on these grid-to-rod supporting conditions, the fuel rod vibration modes and natural frequencies were calculated with the help of the ANSYS code, while the fuel rod vibration amplitudes were estimated by the Paidoussis’ empirical formula. With these vibration characteristics that depend upon the grid-to-rod supporting conditions, the in-reactor fretting wear axial profile observed on the fuel rod surface are found to be simulated quite well. In addition, key design guidelines for the fuel assembly and the spacer grid are proposed to minimize the grid-to-rod fretting wear that may be utilized to develop an advanced fuel design against fretting wear.