核电站冷却剂系统测温旁路支吊架失效原因分析及解决措施

采用Bentley AutoPIPE软件对在役检查经常出现支吊架失效的核电站冷却剂系统测温旁路管线进行静态分析。结果表明:弹簧吊架热态失载是由于主管道实际热位移偏离设计值,弹簧选型余量不足引起的。通过对弹簧吊架重新选型,其热态位移量在量程范围内,载荷变化率符合设计要求,且更改后的测温旁路一次应力、二次应力均符合规范要求。在随后的机组大修检查中,该条管线未再出现弹簧失效故障。     

管道弹簧支吊架的选型分析

采用理论分析与实践案例相结合的方法,分析管道弹簧支吊架的选型。简述弹簧支吊架的应用和分类,指出2类弹簧支吊架的主要特点。分析弹簧支吊架在不同工作状态下对管道的作用力,提出管系吊零原则,指出刚度校验的理论依据,研究地震时弹簧支吊架对管道的作用。最后以某核电厂主蒸汽管道的弹吊选型作为工程实例,描述了弹簧支吊架的选型过程。     

基于功率流的管道系统振动控制及优化设计

基于有限元法对管道系统在泵激励下的振动特性进行分析,以管系传至基础的振动功率流最小为目标,采用基于Matlab开发的优化分析(遗传算法)平台对管道系统支吊架的位置以及刚度进行优化设计,在保证系统抗冲击能力的前提下,达到降低管系振动的目的。通过分析得知,在管道系统的振动控制中,支吊架位置参数的调整及刚度特性的变化是降低管道振动的有效方法。     

管道系统金属橡胶支吊架性能实验研究

管道支吊架是管道系统振动传递的重要途径,针对核动力管道支吊架多为刚性连接的现状,设计了一种金属橡胶支吊架,对多支吊架性能进行了实验分析研究,通过对实验数据的比较与分析,掌握了该型支吊架的力学性能,并通过隔振实验说明该结极对管道振动具有一定的抑制作用,该研究对一回路管道系统的振动控制具有一定的适用性。     

CEFR事故余热排放系统优化设计

为提高快堆管道设计的经济性,根据支吊架优化原则,对事故余热排放系统进行优化设计,使管道系统在各种预期的载荷工况下,都能满足ASME设计规范规定的应力限值,并减少阻尼器、弹簧支吊架的使用量。通过对CEFR的事故余热排放系统的优化设计,积累了高温管道支吊架的布置经验,为以后快堆的管道设计打下基础。     

百万千瓦级核电站核级支吊架的国产化开发

本文简介了核电站管道支吊架的国内外研究现状和发展方向以及核级标准管道支吊架研究的目的、意义，从设计和制造的角度出发，阐述了如何进行核级标准管道支吊架的国产化工作。     

核级工艺管道支吊架系列的研制

核级工艺管道支吊架是核电站一回路工艺系统核级管道的一个重要组成部分,它直接影响核电站的运行可靠性,影响核电站的建设周期和投资。按ASME-ⅢNF篇规定进行设计、研制出能满足核级管道设计要求的支吊架系列,在PC工程(巴基斯坦CHASNUPP核电站)设计中是一项必不可少的工作,也是我国今后核电事业发展中必须解决的一项重要设计。经两年多的工作,已建立起拥有57种型式、2460多种规格的部件所组成的核级工艺管道支吊架,系列完善,能满足核电站核级管道施工设计的需要。核级工艺管道支吊架主要用于安全1、2、3级工艺管道,也用于其他核级管道和有抗震要求的非核级管道。     

高温气冷堆支承冷却系统管道应力分析

本文采用管道有限元方法对高温气冷堆反应堆压力容器承重支承的冷却系统进行了管道应力分析,分析了管道的特征及应力分布特点,讨论了关键点支吊架的设置方案对应力计算结果的影响,提出了管道支吊架布置的优化方案。分析结果表明,该管道的应力计算结果满足ASME标准。     

核动力管道抗震完整性试验及分析研究

对于偶发性地震载荷过分保守的处理导致核电厂管道系统使用大量的阻尼器、支吊架而使管道系统刚性过大,使核电厂的制造、安装、在役检查及维修等费用增加。依托主管道先进设计技术和试验验证项目,完成了核动力管道系统的抗震极限承载能力试验。将试验和计算分析结果与现行规范对比,明确了当前管道设计标准的安全裕量,提出了管道系统阻尼比值与应力评价准则等参数的取值建议。     

基于PDMS的管道支吊架结构设计软件开发

为提高核电工程设计中工艺系统管道支吊架的设计效率,管部和根部的接口准确性,并减少支吊架与其他专业的干涉碰撞,基于核电三维布置设计平台(PDMS)开发管道支吊架结构设计软件NPHS。NPHS软件采用了嵌入式开发模式,实现与PDMS平台的无缝集成,减少程序代码规模,提高程序执行效率;采用独特的数据库结构和机理,简化零部件库的三维建模和信息存储,提高支吊架数据库的开放性和可维护性,通过零部件的连接关键点的设置,提高支吊架组装建模效率。经实际核电工程验证:NPHS软件执行速度快、性能稳定、数据准确,操作、维护方便,输出结果满足工程设计要求。     

中国先进研究堆主回路出堆管接管载荷优化分析

在设备管嘴载荷超值而工况温度又需要提高的情况下,结合系统结构特点,采用先静力工况后动力工况的原则,找到接管载荷超值的主要原因.当温度适当提高后,通过合理调整支架反复对比分析,得到工程可实现的约束方案,使管嘴载荷满足了要求,从而保护了设备,提高了系统安全性能.     

Steam generator small bore piping socket weld inspection using the phased array ultrasonic technique

As nuclear power plants age, the likelihood of failures in the small bore piping used in those plants caused by exposure to mechanical vibrations during plant operations increases. While small bore piping failures rarely cause plant shutdown, the management of small piping has been a keen area of interest since their repair or maintenance requires a reactor trip. Steam generator (SG) drain pipe socket welds are small diameter piping connections (nominal pipe schedule 3–4 inches) susceptible to mechanical vibration. SG drain pipe socket weld failures have caused coolant leakage. Therefore, more reliable inspection technologies for small bore piping need to be developed to detect problems at an early stage and prevent pipe failures. This research aims to improve the reliability and accuracy of small bore piping inspections through the design, manufacture and application of a new phased array ultrasonic testing technique and inspection system for SG drain line socket welds.     

CARR二次水管道系统应力分析与评定

使用PIPSTRESS软件对中国先进研究堆(CARR)二次水管道系统进行分析计算,针对管道系统及所连接设备的不同特点,采取不同的措施,对支吊架类型、安放位置和方向等进行了优化配置,得到适当的管道系统应力数值和接管载荷数值。结果表明,二次水系统的应力分析与评定符合规范要求。     

The evaluation of erosion/corrosion problems of carbon steel piping in Taiwan PWR nuclear power plant

Taiwan Pressurized Water Reactor (PWR), Maanshan nuclear power plant Units 1 and 2 implemented measurements of the wall thinning of the carbon steel piping under the request of regulation authority to prevent the events due to erosion/corrosion since 1989. At first, the licensee established the comprehensive inspection program by itself. Over 2000 components were inspected per unit and 300–500 pipe components were examined by ultrasonic testing per scheduled outage. The simple predictive methodology determined the operability of each individual piping component in the next fuel cycle. Based on the inspection results, the susceptible pipe components were established. Implementation of effective correction management and an improved inspection program can improve the safety, as well as the efficiency, of long-term reactor operations.     

Development of an ISI Robot for the Fast Breeder Reactor MONJU Primary Heat Transfer System Piping

This paper describes the development of a new inspection robot for the In-Service Inspection of the heat transfer system of the Fast Breeder Reactor MONJU. The inspection was carried out using a tire-type ultrasonic sensor for volumetric tests at high temperature (atmosphere, 55°C; piping surface, 80°C) and radiation exposure condition (dose rate, 10 mGy/h; piping surface dose rate, 15 mGy/h). An inspection robot using a new tire type for the ultrasonic testing sensor and a new control method was developed. A signal-to-noise ratio S/N over 2 was obtained during the functional test for a calibration defect with a depth of 50%t (from the tube wall thickness). In the automatic inspection test, an EDM slit with a depth of 9% from the pipe thickness was detectable and with an S/N ratio = 4.0 (12.0 dB).     

Screening method for piping wall loss by flow accelerated corrosion

Flow accelerated corrosion (FAC) phenomenon has persisted its impact on plant reliability and personnel safety. Unless we change the operation condition drastically, most parameters affecting FAC will not be effectively controlled. In order to help expand piping inspection coverage, we have developed a screening approach to monitor the wall thinning by direct current potential drop (DCPD) technique. To improve the applicability to the complex piping network such as the secondary cooling water system in PWR's, we devised the equipotential control method that can eliminate undesired leakage currents outside a measurement section. In this paper, we present Wide Range Monitoring (WiRM) and Narrow Range Monitoring (NaRM) with Equipotential Switching Direct Current Potential Drop (ES-DCPD) method to rapidly monitor the thinning of piping. Based on the WiRM results, susceptible locations can be identified for further inspection by ultrasound technique (UT). On-line monitoring of a thinned location can be made by NaRM. Finite element analysis results and a closed-form resistance model are developed for the comparison with measured wall thinning by the developed DCPD technique. Verification experiments were conducted using UT as the reference. The result shows that model predictions and the experimental results agree well to confirm that both WiRM and NaRM based on ES-DCPD can be applicable to FAC management efforts.     

Reliability analysis of nuclear piping system using semi-Markov process model

The paper presents a general model for evaluating the rupture frequencies and reliability of the piping system in nuclear power plant based on the theory of semi-Markov process. The proposed model is able to incorporate the effect of aging related degradation of pipes. Time dependent rupture frequencies are computed and compared against those obtained from the homogeneous Markov process model. The influence of flaw and leak inspection intervals on the piping rupture probability is investigated.     

Development of Ultrasonic Testing Equipment Incorporating Electromagnetic Acoustic Transducer

An ultrasonic testing equipment for use in in-service inspection of nuclear power plant piping has been developed, which comprises an angle-beam electromagnetic acoustic transducer mounted on a vehicle for scanning the piping surface to be inspected. The transducer functions without direct contact with the piping surface through couplant, and the vehicle does not require a guide track installed on the piping surface, being equipped with magnetic wheels that adhere to the piping material, permitting it to travel along the circumferential weld joint of a carbon steel pipe. The equipment thus dispenses with the laborious manual work involved in preparing the piping for inspection, such as removal of protective coating, surface polishing and installation of guide track and thereby considerably reduces the duration of inspection. The functioning principle and structural features of the transducer and vehicle are described, together with the results of trial operation of a prototype unit, which proved a 1 mm deep notch cut on a test piece of 25 mm thick carbon steel plate to be locatable with an accuracy of ±2 mm.     

Diagnostics of piping by ambient vibration analysis

The presented examples show the possibility to obtain actual information on the stiffness distribution and constraint parameters of a piping system by a non-destructive method without demounting hangers. A strategy for model updating by means of ambient vibration analysis is being developed to determine the real constraints of a piping system quantitatively. Hereby the amplitude dependency of the hanger stiffness must be investigated carefully.     

Case studies of risk-informed inservice inspection of nuclear piping systems

In the present paper, a probabilistic failure analysis is used to find failure probabilities of piping segments, and a probabilistic risk assessment model is employed to obtain risks to a nuclear power plant should these failures occur. The multiplication of the piping failure probability and the consequence for that particular failure results in the risk contribution of the pipe. The degrees of risk for different piping segments can then be ranked, and their results can be used as the basis for planning a risk-informed inservice inspection program. Numerical studies are offered with special emphases on: (1) the status and experience with RI-ISI applications in Taiwan; (2) the comparison of risk-rankings performed with three different methods developed in the US; (3) aspects of the probabilistic fracture mechanics calculation including the flaw size distributions and stress corrosion cracking model. The results indicate the proposed method can indeed be adopted for planning a cost effective inservice inspection program.