蒸汽发生器传热管胀管过渡段水浸超声检查技术

近年来，核电厂提出了对蒸汽发生器传热管胀管过渡段进行检查的要求。由于采用常规涡流检查技术在传热管的胀管过渡段存在盲区，因此开展了传热管胀管过渡段水浸超声检查技术研究，并开发出一套完整的检查技术。通过试验结果分析，证明该技术完全满足蒸汽发生器传热管胀管过渡段检查要求，能够对传统传热管涡流检查形成补充，同时也能够应用于其他薄壁小径管道的检查。      

大亚湾核电站蒸汽发生器传热管的涡流检查

介绍了大亚湾核电站蒸汽恨生器传热管的涡流检查设备与技术,１号机组蒸汽发生器首闪在役检查的关键路径,技术改进及其检查结果。首次在役检查结果表明;３台蒸汽发生器没有发现传热管破损,保证了核电站的安全运行。     

蒸汽发生器的涡流检查——介绍ST98蒸汽发生器检查和修理管理系统

西屋公司已经开发了新的传热管涡流检查工艺，以满足目前对检查速度和精度的要求。ＳＴ９８蒸汽发生器检查和修理管理系统是目前最现代化的设备。     

核电厂蒸汽发生器传热管涡流检测噪声形成机制与测量

蒸汽发生器传热管涡流检测中,管材本底噪声的大小会影响缺陷显示的定性。在涡流检测理论中,对于本底噪声水平的计算方法有2种:1通过全管信号微分化后再通过均方根计算获取核电用蒸汽发生器传热管的真实本底噪声;2参考标定管基准信号获得检测系统信噪比(SNR)。本文对目前在役检查中几种检测方法的本底噪声水平进行定量研究,并讨论引起噪声的机理。     

蒸汽发生器传热管防振条位置涡流定位检验精度研究

蒸汽发生器传热管防振条的位置偏移会对蒸汽发生器的安全稳定运行带来隐患。利用涡流法测量防振条的位置,分析影响涡流检验精度的因素。分析结果表明:采集速度和导管长度对检验结果影响较小,导管状态及传热管曲率半径对检验结果的可靠性和稳定性影响较大。     

立式和卧式蒸汽发生器传热管在役检查技术

美国立式蒸汽发生器传热管在役检查技术与经验。在役检查要求美国核电站技术规格书(USNRC 1981)规定了压水堆核电站蒸汽发生器传热管的在役检查要求(取样规模和频率)，核电站在首次在役检查中接受检查的蒸汽发生器传热管的数量取决于该电站中蒸汽发生器的数目和是否对这些管子进行过役前检查．在随后的检查中每次检查一台待查的蒸汽发生器，进行轮流安排．     

核电站蒸汽发生器降质预防和在役检查

介绍了法国核电站蒸汽发生器在运行初期所发生的传热管降质现象,重点论述了大亚湾核电站１ 号机组蒸汽发生器对降质预防所采取的措施和在役检查,包括二回路水化学监控、泄漏率监测、传热管涡流检验、二次侧的机械清洗、清洁度检查和外来物取出等。实践证明,采取了上述降质预防措施和在役检查,对核电站的安全运行起到了重要作用     

蒸汽发生器传热管氦检漏定量定位技术分析及验证

蒸汽发生器传热管氦检漏是核电厂蒸汽发生器重要检查项目。本文针对蒸汽发生器传热管氦检漏漏点定量定位分析提出了一套算法理论,并通过试验平台进行试验,验证了算法的准确性,为漏点分析提供了可靠的理论依据。     

蒸汽发生器传热管结垢厚度的涡流检测方法与应用

蒸汽发生器二回路中有较多的沉积物存在并危害传热管安全,利用涡流检测方法可以对传热管二次侧泥渣进行有效检测。通过模拟传热管结垢的不同厚度并进行实验,可获得厚度与幅值的对应关系。本文描述了对蒸汽发生器传热管结垢的检测方法及幅值与厚度的对应关系,为统计蒸汽发生器传热管外壁结垢情况提供了较为有效的参考基准量。     

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。      

An experimental result on flaw sizing of steam generator tube at nuclear power plants using ultrasonic testing

A steam generator at a nuclear power plant consists of thousands of thin tubes, and is a highly important component in operation. Also, steam generator tubes play a critical role in maintaining pressure boundaries of the primary and secondary sides, and can be easily damaged due to operation conditions caused by high temperature and pressure. Therefore, considerable amount of efforts are being committed to evaluating structural integrity of steam generators during in-service inspection. Eddy current testing is the commonly used inspection technique to evaluate a steam generator tube's integrity, but it has limitations in accurately sizing flaws due to the nature of the technique which determines size based on the entire volume of a flaw. In this study, experiments were performed by using ultrasonic testing instead of eddy current testing for the inspection of steam generator tubes to detect various kinds of flaws and to see if the detected flaws can be sized accurately. Consequently, the ultrasonic testing technique could detect various types of flaw, and the flaw sizing results were reliable in length and depth.     

Status and progress of research on a removed-from-service steam generator

Under an NRC directed group sponsored project (including French, Italian, Japanese, and EPRI participation) a steam generator removed from service is the subject of extensive research. The generator now serves as a vehicle for studies involving validation of the accuracy and reliability of current nondestructive examination (NDE) characterization during inservice inspections, determination of remaining integrity of service defected steam generator tubes, determination of failure consequences (leak rate) of defects, and demonstration of cleaning and decontamination techniques. Program objectives are to provide inputs to regulatory guides on inservice inspection and tube plugging criteria.During the past year dilute chemical reagent decontamination of the steam generator channelhead using modified LOMI and Candecon processes, each on one side, has been completed. In addition to decontamination effectiveness other factors such as corrosivity during the process, methods for waste handling, and potential for affecting return to service of the component were evaluated. Following decontamination a subcontracted effort was conducted to remove a large number of plugs placed in the generator during its service life. In under three weeks 969 explosive type plugs were removed. This provided a new level of experience in large scale plug removal. The plug removal was to optimize access to defected steam generator tubes for nondestructive primary side characterization. The first of these nondestructive examinations was conducted toward fiscal year end, employing state-of-the-art eddy current technology. A map of the generator condition, as obtained from 100% eddy current examination, is being formed. In parallel with the above endeavors, extensive efforts have been made toward characterizing the secondary side of the generator. The tubesheet surface and the inner row U-bend regions have been extensively examined. Innovative photographic approaches have provided success in documenting generator conditions in the sludge pile area.     

Development of a hybrid ECT sensor for JSFR SG double-wall tubes

A new design has been adopted for the steam generator (SG) tubes of the Japan Sodium-cooled Fast Reactor (JSFR) using double-wall tubes. This paper estimates and assesses the effectiveness of detecting defects in SG double-wall tubes of the JSFR by using combined high-frequency eddy current testing (ECT) and low-frequency remote field eddy current sensors. We confirm that the proposed hybrid ECT sensor is highly sensitive to small defects, fatigue cracks, and other defects even when located under support plates of tubes. The parameters of the hybrid ECT sensor are designed and optimized to detect small defects using accurate numerical simulations based on the finite element method, using an in-house developed code. The sensitivity and high performance of the hybrid ECT sensor was validated with experimental measurements.     

In-service inspection of PWR steam generators by eddy currents; French achievements in defect detection and evaluation

In a nuclear plant the steam generator tubes must be efficiently inspected. The highest possible detection sensitivity is necessary to get a clear decision for plugging the rejected tubes. For this reason multi-frequency eddy current examinations have been developed at the Commissariat à l'Energie Atomique and industrialized by Intercontrôle. Defects are characterised case by case as the needs arose on the site: sludge height determination; measurement of tube deformation under pressure of oxides; determination of tube degradation due to the pressure of foreign bodies; detection of cracks under coatings; detections of cracks at the end of flanging. Problems unsolved by standard probes were dealt with by rotating-head machines. This report sums up part of the research undertaken in the field of eddy current testing.     

Effect of shot peening on primary water stress corrosion cracking of Alloy 600 steam generator tubes in an operating PWR plant

The effect of shot peening on the primary stress corrosion cracking behavior of thermally treated Alloy 600 steam generator tubes in an operating pressurized water reactor (PWR) plant was analyzed based on pulled tube examinations and in-service inspection eddy current test (ISI-ECT) data. The evaluation was focused on the shape of crack, evolution of the number of new cracks and cracked tube fraction, and variation of crack length and the corresponding eddy current amplitude before and after shot peening. The shape of the crack was changed from a half-elliptical type before shot peening, to an elliptical one with bulging after peening. It was concluded that the shot peening was not effective for retarding both crack initiation and growth for this plant.     

Surry steam generator - examination & evaluation

This report summarizes research conducted during the fourth year of the five year Steam Generator Group Project. During this period the project conducted numerous nondestructive examination (NDE) round robin inspections of the original Surry 2A steam generator. They included data acquisition/analysis and analysis-only round robins using multifrequency bobbin coil eddy current tests. In addition, the generator was nondestructively examined by alternate or advanced techniques including ultrasonics, optical fiber, profilometry and special eddy current instrumentation. The round robin interpretation data were compared. To validate the NDE results and for tube integrity testing, a selection of tubing samples, determined to be representative of the generator, was designated for removal. Initial sample removals from the generator included three sections of tube sheet, two sections of support plate and encompassed tubes, and a number of straight and U-bend tubing sections. Metallographic examination of these sections was initiated. Details of significant results are presented in the following paper.     

快堆钠-水直流蒸汽发生器七管样机热工水力特性分析

钠-水直流蒸汽发生器是钠冷快堆主热传输系统的关键设备之一,其结构及内部的传热现象是十分复杂的。管内外侧的介质及压力不同,管内侧为高温高压的水/蒸汽,存在复杂的两相流动传热传质现象;管外侧为高温液态金属钠,沿换热管高度方向存在较大的钠温变化。本文以钠-水直流蒸汽发生器七管样机为研究对象,对其热工水力特性进行了CFD分析和实验研究,CFD分析结果和实验结果吻合较好,验证了CFD分析所采用的数学模型和数值方法的可靠性。结果表明,钠-水直流蒸汽发生器七管样机的传热面积是足够的,达到了设计指标要求,其界限质量含汽率约为0.42,临界热流密度约为451.98 kW/m²,从而确定了蒸干点的位置。