一回路核级设备磨损的监测模型

核级设备的结构失效与微动损伤有直接关系,在应力集中部位,微动又是许多核电设备提前失效的直接原因。本文以一回路核级设备为对象,研究其磨损失效的原理和特征,并针对不同的磨损失效情况建立监测模型,针对难以避免的典型磨损,构建监测模型,在线监测敏感部位的磨损,并提取信号进行分析,确定磨损部位和磨损程度。监测模型可以通过至少两种监测手段监测易发生磨损的部位,同时,通过不同的定位方法找到磨损发生的位置,并发出警报,做到事故前预防。     

TA16钛合金微动磨损特性

采用PLINT微动磨损试验机,进行TA16钛合金传热管与0Cr18Ni9不锈钢实心圆柱体配副件的微动磨损特性试验.试验条件为:常温、法向载荷为50 N和80 N、位移幅值为80～200μm、频率为2Hz.结果表明,法向载荷和位移幅值对材料损伤程度和损伤机制产生显著影响.材料损伤程度随位移幅值、载荷的增加显著增加,而微动...     

外来物对蒸汽发生器传热管微动磨损的分析研究

以蒸汽发生器传热管固定不动为前提,研究管板表面外来物由于流体绕流引起的振动或运动造成的传热管微动磨损.分析过程中选取一种典型的外来物形状和位置作为分析对象,以Archard计算模型为基础,借助瞬态动力学分析方法,对外来物造成传热管的微动磨损进行了分析研究.结果表明,质量不大且一端固定的外来物在周期性的脉动载荷下对传热管...     

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     

压力诱发SGTR的概率计算方法探讨

诱发蒸汽发生器传热管破裂现象对核电厂堆芯损伤和放射性早期大量释放风险有非常大的影响。准确地对诱发蒸汽发生器传热管断裂概率进行计算和定值对正确认知核电厂的核安全风险非常重要。文章调研了已有压力诱发蒸汽发生器传热管破裂概率的取值计算方法,通过对不同取值计算方法的对比分析,结合国内实际情况,提出了一套较为合理可行的压力诱发蒸汽发生器传热管破裂概率的计算方法。文章推荐的诱发蒸汽发生器传热管破裂数据采集和分析计算方法为后续国内核电厂概率安全分析应用和安全监管提供了参考。     

蒸汽发生器传热管堵头有效性分析

蒸汽发生器传热管是核电厂—回路压力边界的薄弱环节,传热管的完整性直接影响到整个一次侧的安全。当传热管出现裂纹、腐蚀或磨损等缺陷时,在评定确认可能会发生一次侧流体进入二次侧情况下,需要对传热管进行堵管。利用有限元法对某蒸汽发生器传热管的滚压堵头进行分析评定,模拟计算在堵管时以及堵管后堵头、传热管接触力情况,通过计算及分析确认堵头在极限运行工况的有效性,计算显示此堵头满足强度要求。     

蒸汽发生器传热管弯管区最小间隙分析

在蒸汽发生器传热管的制造、穿管等环节中,有可能出现传热管弯管区局部区域的变形,导致传热管弯管区间隙过小甚至接触。依据传热管的设计原则,以CPR1000蒸汽发生器(55/19B型)传热管弯管区为例,通过流致振动分析、磨损分析、蒸干评估以及应力分析,对传热管弯管区间隙进行分析及评估。分析结果表明,传热管弯管区最小间隙应大于其湍流振幅,以避免弯管区发生不可接受的流致振动、磨损、蒸干等问题,设计过程中须考虑足够的安全裕量。     

蒸汽发生器存在的主要问题与对策

蒸汽发生器的可靠性对核电厂的安全，可靠性和经济效益有重大影响。国外ＳＧ上前存在的主要问题是传热管遭受二次侧晶腐蚀与晶间应力腐蚀，一次侧应力腐蚀和微振磨损。致使其可靠性较低。文中分析了传热管破损的主要原因，提出了国内新ＳＧ设计应采取的对策。     

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。      

蒸汽发生器传热管的微振磨损及其防护

蒸汽发生器二次侧U形管防振条处经常会发生微振磨损,外来物对传热管的磨损也时有发生。本文介绍了传热管微振磨损及外来物磨损的概况,传热管微振磨损的机理,预测传热管的微振磨损量,垂直接触力和滑动距离。最后论述了美国西屋公司、法国法马通公司、德国西门子KWU公司和加拿大B&W公司对传热管微振磨损的防护措施。     

Determination of equivalent single crack based on coalescence criterion of collinear axial cracks

In a nuclear power plant the steam generator tubes cover a major portion of the primary pressure-retaining boundary. Thus, very conservative approaches have been taken in the light of steam generator tube integrity. According to the present criteria, tubes wall-thinned in excess of 40% should be plugged whatever the cause. However, many analytical and experimental results have shown that no safety problems exist even with thickness reductions greater than 40%. The present criterion was developed about 20 years ago when wear and pitting were dominant causes for steam generator tube degradation, and it is based on tubes with single cracks regardless of the fact that the appearance of multiple cracks is more common in general. The objective of this study is to review the conservatism of the present plugging criteria of steam generator tubes and to propose a new coalescence model for two adjacent through-wall cracks existing in steam generator tubes. Using the existing failure models and experimental results, we reviewed the conservatism of the present plugging criteria. In order to verify the usefulness of the proposed new coalescence model, we performed finite element analysis and some parametric studies. Then, we developed a coalescence evaluation diagram.     

Restraining effect of support plates on the limit loads for circumferential cracks in the steam generator tube

The steam generator in a nuclear power plant is a large heat exchanger that uses heat from a reactor to generate steam to drive the turbine generator. Rupture of a steam generator tube can result in release of fission products to environment outside. Therefore, an accurate integrity assessment of the steam generator tubes with cracks is of great importance for maintaining the safety of a nuclear power plant. The steam generator tubes are supported at regular intervals by tube support plates and rotations of the tubes are restrained. Although it has been reported that the limit load for a circumferential crack is significantly affected by boundary condition of the tube, existing limit load solutions do not consider the restraining effect of tube support plate correctly. In addition, there are no limit load solutions for circumferential cracks in U-bend region with the effect of tube support plate. This paper provides detailed limit load solutions for circumferential cracks in top of tube sheet and the U-bend regions of the steam generator tube with the actual boundary conditions to simulate the restraining effect of the tube support plate. Such solutions are developed based on three-dimensional (3-D) finite element analyses. The resulting limit load solutions are given in a polynomial form, and thus can be simply used in practical integrity assessment of the steam generator tubes.     

An experimental study on fretting wear behavior of cross-contacting Inconel 690 tubes

A fretting wear test rig employing a piezoelectric actuator has been developed, which is equipped with a heating and water circulation system. The fretting wear tests of cross-contacting Inconel 690 tubes, which is widely used for power plant steam generator, have been carried out in room temperature ambient and 80 °C in-water conditions. Maximum normal load was 55 N, and the sliding amplitude was below 50 μm. Scars of the mixed-slip and the gross-slip fretting wear have been measured in terms of scar diameter and wear volume. From the relationship between the work rate and the wear rate, a threshold of work rate has been defined, and this is found to be closely related with fretting wear regimes. The wear coefficients have been evaluated in the gross-slip regime. Distinct fretting wear mechanisms have been observed for the two different test conditions from SEM microphotographs. The crack formation, large particle separation and resulting third body effect were significant in room temperature ambient condition. The protective nature of the tribologically transformed layers coupled with non-uniform contact results in the lower wear coefficient while smooth wear scar and extensive abrasion produces higher wear volume in the other condition.     

A study on fretting fatigue life for the Inconel alloy 600 at high temperature

Fatigue tests of the Inconel 600, a type of nickel-chromium based heat resistant alloy used for steam generator tubes in nuclear power plants, were carried out. Temperature increase to 320 °C did not change the fatigue strength much, but the fretting condition caused a significant reduction in the fatigue strength. The reduction at 10⁷ cycles was about 70% for both of room and the high temperatures. An apparatus to realize the fretting condition has been developed and instrumented to measure the normal and friction forces. The bridge type of contact pad was fabricated of SUS 409 stainless steel. Fracture surfaces and wear scars were observed by electron microscope and the profiles of wear scar were measured by non-contact 3D-profiler.     

蒸汽发生器管板孔桥超差情况下的结构安全性分析

蒸汽发生器制造过程中对管板进行深孔钻时，发生管板孔桥超差。管板二次侧的3个管孔C165-R59、C167-R59、C168-R58不能满足设计要求，管板一次侧的这些管孔满足设计要求。针对该不符合项，核审评单位联合蒸汽发生器制造单位和设计单位，从管板的强度、管板孔桥超差不符合项对流致振动的影响、堵管后的传热管应力分析、传热管堵管的压差对孔桥强度的影响、孔桥超差导致的传热管接触磨损等角度进行了结构安全性分析。分析结果表明，目前的堵管方案合理可行，但需加强在役阶段的跟踪检查，以保证修复的可靠性和质量。     

传热管与支撑件随机碰撞实验研究

针对蒸汽发生器中传热管与支撑件的碰撞行为,对悬臂梁固定的传热管在不同支撑条件下开展了激振实验,获得了传热管均方根位移与接触率,分析了传热管与支撑件磨损功率的变化规律,并探究了传热管固有频率对振动特性的影响。结果表明,防振条支撑与波纹带支撑时传热管的法向均方根位移均随激振力增加逐渐放缓,而防振条支撑对应的切向位移呈线性增长。防振条支撑与波纹带支撑时的接触率均表现为随激振力增大趋于稳定,其中间隙对防振条支撑的接触率影响更明显。在以冲击为主导的激励方式下,激振力与磨损功率表现为明显的正相关。支撑间隙对磨损功率的影响相对复杂,防振条支撑下磨损功率在0.1 mm和0.25 mm间隙存在极值,而波纹带支撑磨损功率仅在0.2 mm间隙存在极值。传热管固有频率对振动响应结果的影响很小。     

蒸汽发生器传热管在泵致脉动压力载荷下的动力学响应研究

泵致脉动压力是核电站中引起主设备部件疲劳失效的主要原因之一。本文建立了蒸汽发生器传热管的泵致脉动压力载荷表达式，并建立不同弯曲半径的传热管有限元模型，对蒸汽发生器传热管在泵致脉动压力载荷下的动力学响应进行了研究。结果表明：34、64、94、114、124、144排传热管附近的频率、振型对泵致脉动压力最为敏感；包络泵致脉动压力作用下，最大应力出现在32排传热管上；传热管在泵致脉动压力载荷作用下，泵致脉动压力载荷的轴频频率对结构响应的贡献最大。本文分析结果为蒸汽发生器传热管在泵致脉动压力载荷下的磨损分析提供了参考。     

The method to predict wear depth and wear volume of tube at near-distant future from measured or prescribed depth at present time

The existing results of the impact model for a cylindrical projectile are modified. The material constant associated with the impact model is inversely determined from the given wear depth and number of impacts. The wear depth or wear volume in the near-distant future is estimated from the material constant using the present wear status. The results are compared with measured depths of a steam generator tube in a nuclear power plant. The results show that predicted wear depth envelops the measured depth. It can be concluded that the methods employed in the present paper can be applied to the prediction of wear depth from a given status of wear at the present time.     

核电站蒸汽发生器泄漏监测——氮-16辐射监测仪概述

蒸汽发生器是压水堆核电站核蒸汽供应系统的主要设备之一,对蒸汽发生器传热管进行泄漏监测关系到核电站的安全和经济运行。介绍了用于蒸汽发生器泄漏监测的氮-16辐射监测仪的概况、工作原理、系统组成等。     

基于时间序列神经网络的蒸汽发生器传热管泄漏程度诊断研究

针对蒸汽发生器U形传热管泄漏，本文提出了一种基于时间序列神经网络对蒸汽发生器传热管泄漏程度进行诊断研究的方法。首先，对核电厂蒸汽发生器U型传热管泄漏进行机理分析，构建其数学模型，提取其泄漏的直接特征参数，再依据Fisher得分法，提取其间接特征参数；其次，通过滑动时间窗口法从预处理后的时间序列数据中生成数据样本，作为时间序列神经网络的输入，并以蒸汽发生器U形传热管泄漏程度信息为标注，基于反向传播（BP）算法对五层神经网络系统进行训练，得到蒸汽发生器U形传热管泄漏的时间序列神经网络模型；最后，模拟核电厂运行过程蒸汽发生器U形传热管泄漏时的时间序列测试数据。仿真结果表明，时间序列神经网络对演变事件的处理具有较好的有效性和较高的泛化能力，对故障程度的诊断研究具有参考价值。