卧式蒸汽发生器传热管涡流检查系统研制与应用

核电厂卧式蒸汽发生器由于集流管深度大，传热管弯道多、弯曲半径小、管数多以及材料均匀特性差等特点，给传热管涡流检查带来难度。研发了一套卧式蒸汽发生器传热管涡流检测系统（C-SMART），包括机械定位驱动装置、控制系统和控制软件、涡流数据采集和分析软件等。系统具备快速精准定位、单管标定、高效等特点，实现对传热管涡流的自动检查，具有显著的经济和社会效益。      

蒸汽发生器管板液压胀管数值模拟研究

针对蒸汽发生器管板材料和传热管材料进行常温力学性能测试,获得材料的真应力-真应变曲线。在此基础上,对蒸汽发生器管板开展液压胀管的数值模拟研究,研究胀接压力和胀接长度对残余接触压力和过渡区残余应力的影响,得出以下结论:残余接触压力随胀接压力的增大而增大,与胀接长度基本无关,过渡区段传热管内壁最可能发生应力腐蚀。     

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。     

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．     

蒸汽发生器二次侧清洁度的视频检查

立式平管板Ｕ形管束的压水堆核电站蒸汽发生器安装至瞬变工况期间，管板上表面的铁基金属残余物的堆积和外来物的存在，以及商业运行后管板、传热管、管子支撑板等的低流速区域里沉积物的堆积要求进行蒸汽发生器的清洁度检查。介绍了蒸汽发生器二次侧清洁度的视频检查技术。该技术适用于蒸汽发生器安装至瞬变工况和投入商业运行后可能发生污染的各阶段的清洁度检查。     

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。     

法国、日本和比利时立式蒸汽发生器传热管在役检查技术

法国核管理机构要求在核电站运行前进行役前检查，运行后至少每2年进行一次在役检查，每隔10年进行完全性检查(100％传热管的全长度)：法国电力公司(EDF)导则对具有敏感材料(600合金)传热管的蒸汽发生器要求对热侧胀管过渡区和第一排U形弯头区在每次大修时进行传热管的100％检查，每隔一一次大修时100％检查热侧支撑板和泥渣堆积区里的管子，     

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。     

蒸汽发生器管板超差管孔胀管的非线性有限元分析

本工作利用三维非线性有限元分析了某核蒸汽发生器管板的胀管效果,研究了尺寸偏差和材料特性变化等对胀管效果的影响,计算了几种典型传热管间隙和刚度下的接触压力,并与非超差管孔胀接触压力进行了对比研究,最后对合适胀管压力的选择提出了建议.     

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。     

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.     

Enhancements of eddy current testing techniques for quantitative nondestructive testing of key structural components of nuclear power plants

In this paper, studies on upgrade of eddy current testing (ECT) techniques for inspection of stress corrosion cracks (SCC) in key structural components of a nuclear power plant are reported. Access and scanning vehicle (robot), advanced probes for steam generator (SG) tube inspection, developments and evaluations of new ECT probes for welding joint, and ECT-based crack sizing technique are described, respectively. Based on these techniques, it is demonstrated that ECT can play as a supplement of ultrasonic testing (UT) for the quantitative inspection of welding zone. It is also proved in this work that new ECT sensors are efficient even for inspection of a stainless steel plate as thick as 15 mm.     

Development of probability of detection functions for evaluating integrity of axial outside diameter stress corrosion cracks in steam generator

The reliability of an eddy current testing (ECT) inspection system depends upon the inspection technique and quality of analyst. In evaluating the integrity of a steam generator (SG) tube, degradation detection and sizing accuracy are considered performance measures of the nondestructive evaluation (NDE) system. A probability of detection (POD) model serves as a functional measure of the ability of an NDE system to detect degradation. It is one of the inputs in the operational assessment, and it is used to estimate the degradation during service via ECT of the SG tube. In this study, the POD functions of the inspection technique and analyst were obtained to quantitatively analyze the ECT bobbin probe for axial outside diameter stress corrosion cracks in SG tubes. This should serve to evaluate the integrity of the SG tubes. The depth and amplitude of defects were used as parameters of the POD model. Hit (detection) and miss (no detection) binary data obtained from destructive and nondestructive inspection of cracked tubes were also used.     

Secondary side corrosion in steam generator tubes: lessons learned in France from the in-service inspection results

Non-destructive testing (NDT) has proved to be very important in the maintenance of steam generator tubing. This is particularly true in the case of secondary side corrosion, because this type of degradation leads to various morphologies which are often complex (intergrranular attack) (IGA), intergranular stress corrosion cracking (IGSCC), or a mixture of both. Their detection and characterization by the usual NDT techniques have been achieved through numerous laboratory studies, which were conducted in order to determine the performance and limitations of NDT. Pulled tube examination in a hot laboratory was very valuable, for both NDT and fracture mechanics aspects. The eddy current bobbin coil probe, used for multipurpose inspection of tubes, allows the detection of IGA-SCC at the tube support plate elevation. In France, the use of rotating probes is not required for that type of degradation, since the repair criterion is based on bobbin coil results only. The bobbin coil is also used for detection of IGSCC occurring in free spans, within sludge deposits. The eddy current rotating probe allows, in that case, characterization of main cracks. Concerning the outer diameter initiated circumferential cracks which occur at the top of the tube sheet, only the rotating probe is used. An ultrasonic (UT) inspection was performed several times, in order to obtain information on UT capabilities. The goal of tube inspection is obviously knowledge of the status of steam generators, but also to follow up degradations and to estimate their revolution, and to verify the beneficial effect of some corrective measures, e.g. boric acid injection.     

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。     

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.     

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     

Inspection techniques for primary pressurized water cooler tubes in the high temperature engineering test reactor

A primary pressurized water cooler (PPWC) with 136 inverse-U-tubes is installed in the primary cooling system of the high temperature engineering test reactor (HTTR). The HTTR is the first high temperature gas-cooled reactor in Japan with an outlet gas temperature of 950 °C and thermal power of 30 MW. The heat transfer tubes form the reactor pressure boundary of the primary coolant. Inspection techniques for the tubes should be established to carry out the in-service inspection efficiently. An automatic inspection system for the tubes uses probes for eddy current testing and ultrasonic testing. Defect detecting characteristics of the inspection probes and the application of the automatic inspection system to nondestructive test of the tubes were examined by a mockup test utilizing artificially degraded tubes. The automatic inspection system could smoothly insert and withdraw the probe at its moving velocity in the fixed positions of the defected tube. Nondestructive test of the tubes using the automatic inspection system was conducted during reactor shutdown period of the HTTR after test operation of about 55% of the full power. Through the nondestructive test, it was concluded that there was no defect on the outer surface of the heat transfer tubes of the PPWC inspected.