CPR1000核电厂一级管道应力分析

核级管道的应力分析是为了保证管道自身和与其相连的设备、支架的安全.分析内容包括3个方面:计算管道应力,并使之满足RCC-M规范规定的限值要求;计算管道对与其相连的机器、设备的作用力,并使之满足标准规范的要求,保证机器、设备的安全;计算管道对支吊架的作用力,为支吊架的设计提供依据.管道应力分析工作的步骤是:首先,对管道所在系统的功能和工况参数、管线的布置情况进行详细的了解,划分分析范围;其次,根据管道ISO图用软件建立分析管线部分的几何模型,并定义材料属性;然后,按照规范规定的载荷组合形式加载;最后,计算、评定并输出支反力,核级管道的应力分析不仅可保证管道、支架、设备的安全,而且可优化设计,在核电厂建造和运行中起到重要作用.     

某核电厂管道振动验收准则的计算和超标处理

某核电厂自动卸压系统（ADS）管道在热试过程中发生了振动超预期事件。为确定该管道振动超标的原因和处理措施，参考ASME核电厂运行和维修标准及导则第3篇，通过计算确定了该管道的振动验收准则。通过比例模型试验确定了管道振动超标的根本原因，并给出了针对性的解决管道振动超标的措施，可以作为核电厂管道振动超标问题处理的参考。     

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。     

核岛管道阻尼器低速摩擦阻力限值分析

核岛管道和设备上常设置液压阻尼器来保护管道或设备免受地震或其他突发震动而损坏,阻尼器的性能参数至关重要,低速摩擦阻力限值是其中的一个重要参数。但由于规范的缺乏,参数的确定无明确的依据,国内外的核电厂采用的数据不一。为了确定低速摩擦阻力限值,从阻尼器的工作原理分析摩擦阻力产生的原因,分析不同参数的阻尼器对管道应力的影响,并调研了国内外核电厂管道阻尼器的实际情况,提出了阻尼器摩擦阻力限值的推荐值和试验注意事项。对设计和运行阶段核电站阻尼器参数的提出具有参考价值。     

含内表面裂纹核级管道的初始塑性失效分析

我国在役和新建的大部分核电厂在主管道上应用了破前漏技术,针对该技术ASME采用净截面屈服准则对完全塑性断裂进行缺陷评定,大量研究表明,净截面屈服准则高估了结构的承载能力。本文采用有限元方法模拟了含内表面裂纹的核级管道在内压作用下的变形过程,并利用裂纹前沿J积分随内压变化的曲线特征确定了含裂纹管道的初始塑性失效载荷。随后,将初始失效载荷的计算值与ASME规范定义的理论值相比较,结果表明理论解高估了结构的承载能力。最后,评价了ASME-BPVC-XI规范中A级使用限制对应的允许薄膜应力的适用性。     

核电厂调试期间核级管道振动测量工作改进

美国机械工程师协会(ASME)OM-S/G-2000 Part 3导则存在操作性不足的缺陷,而国内核电厂核级管道振动测量工作均依此导则,采用的试验方法类似,但都无法全面、准确预测管道振动极大点。以某核电厂调试期间核级管道振动测量工作为例,从试验对象的筛选、关键设备或部件的选取、测点选择、现场试验及振动分析评价5个方面进行探讨,提出改进建议,最后以某核电厂安全壳喷淋系统(EAS)的某管段振动为研究对象,对测量方法、振动限值计算及评价等进行案例分析。     

某核电厂主蒸汽管道应力分析

本文研究了某核电厂中主蒸汽系统管道的计算和评定等典型内容。此系统管道运行中承受的载荷工况多样,管道应力状态复杂。为了保证系统管道能够正常运行,在设计上需保证该系统管道的应力能够满足相关规范要求。分析采用管道力学分析软件PIPESTRESS进行,计算模型包括主回路、主蒸汽系统及相关的管道和阀门,分析包含静力和动力计算等。对计算结果依据美国机械工程师学会的ASME及相关规范进行了应力评定,并包含了LBB评定,保证了回路运行的安全。     

钍基熔盐堆回路管道蠕变疲劳分析和评定研究

钍基熔盐堆(TMSR)回路管道的运行温度大于500℃,需要采用高温反应堆的评定准则来进行评定。对于高温管道需对其进行应力、应变及蠕变-疲劳分析来保证其完整性。传统的方法是采用ANSYS软件建立有限元模型来进行蠕变-疲劳分析,耗时费力。本文采用PIPESTRESS软件对TMSR回路管道进行评定。分析结果表明:通过引入应力指数,对PIPESTRESS软件的计算结果进行处理,可以完成回路管道蠕变-疲劳分析的快速评定。在TMSR回路管道分析中,该方法省时省力,是一种更加实用和有效的方法。     

钍基熔盐堆回路管道系统应力分析与评定

钍基熔盐堆(TMSR)管道设计温度可达700℃,设计标准采用美国机械工程师协会ASME-NH分卷。高温管道评定时除需要进行应力评定外,还需进行应变变形限值和蠕变疲劳限值等评定。利用通用有限元分析软件(ANSYS)对整体回路系统进行计算,并通过优化计算,使得管道应力达到ASME规范中限值要求。     

核电厂安全级数字化仪控系统通信隔离设计

核电厂仪表和控制系统被称为核电厂的“神经中枢”,对保障核电厂的安全稳定运行安全具有关键作用,是核电厂的重要组成部分。本文依据核电厂相关设计标准要求及参考核电厂的应用需求,提出一种核电厂安全级数字化仪控系统通信隔离设计方法,该方法针对安全级网络通信常见的两种通信方式——点对点通信和多节点通信,在安全级系统内部、安全级系统与非安全级系统之间分别设计独立于处理单元的通信模块,该通信模块本身属于安全级设备,采用异步通信、定制的双端口RAM及确定性的通信协议等方法;在多节点通信中采用双环路拓扑和节点旁路等机制来满足安全级通信隔离设计要求。通过搭建典型工程样机和专家独立工程评审,验证了本方案在工程应用中的正确性和可行性。     

基于ASME和RSE-M规范的含平面缺陷奥氏体不锈钢核级管道剩余寿命预测方法的数值对比研究

核级管道在加工和安装环节可能存在不同的缺陷。此外,由于核电厂运行条件的影响,管道中可能存在少量缺陷,如裂缝。需要合理预测评估含缺陷管道的剩余寿命,以便安排更换方案,避免对核电厂的效率造成严重影响。本文根据ASME和RSE-M规范,在应力强度因子计算、裂纹扩展分析和裂纹稳定性评价等环节,通过数值对比研究了含有平面缺陷的奥氏体不锈钢核级管道的剩余寿命评估方法,为类似工作提供参考。      

Probabilistic fracture failure analysis of nuclear piping containing defects using R6 method

Failure analysis of in-service nuclear piping containing defects is an important subject in the nuclear power plants. Considering the uncertainties in various internal operating loadings and external forces, including earthquake and wind, flaw sizes, material fracture toughness and flow stress, this paper presents a probabilistic assessment methodology for in-service nuclear piping containing defects, which is especially designed for programming. A general sampling computation method of the stress intensity factor (SIF), in the form of the relationship between the SIF and the axial force, bending moment and torsion, is adopted in the probabilistic assessment methodology. This relationship has been successfully used in developing the software, Safety Assessment System of In-service Pressure Piping Containing Flaws (SAPP-2003), based on a well-known engineering safety assessment procedure R6. A numerical example is given to show the application of the SAPP-2003 software. The failure probabilities of each defect and the whole piping can be obtained by this software.     

我国核电厂气态流出物中惰性气体监测现状

核电厂流出物尤其是气态流出物中的放射性惰性气体监测多为低水平放射性核素,我国运行核电厂的环境监测结果均低于探测限,无法计算照射剂量.探测能力决定了放射性惰性气体排放评价的结果.本文分析了我国各运行核电厂流出物放射性惰性气体监测和排放评价的现状,比较欧盟的相关建议,研究我国核电厂流出物放射性惰性气体监测能力存在的问题,并提出了建议.     

核安全一级高温管道系统结构分析与安全评估方法研究

为解决复杂核安全一级高温管道系统结构分析与评定工程问题，在管道分析软件与核级高温评定规范ASME-NH之间建立了一座桥梁。首先，对管道结构（直管及弯管）在不同载荷作用下的应力状态解析解进行了详细推导分析，并且与有限元数值解进行了误差分析。结果显示，给出的直管及弯管结构应力状态解析解具有很好的准确性。随后，将一维管线力学分析模型与截面三维应力状态解析解相结合，给出了高温管道系统结构分析、评定方法及应用步骤，将ASME-NH-3650规范内容明确化。      

Experience with piping in German NPPs with respect to ageing-related aspects

Thirteen nuclear power plants (NPPs) with pressurized water reactors (PWRs) and six plants with boiling water reactors (BWRs) are currently in operation in Germany. For almost 25 years, GRS has been systematically evaluating the operating experience of these plants. In this paper, the operating experience relating to piping damage in safety-relevant systems of German plants with light water reactors (LWRs) is evaluated with respect to ageing-related effects. The experience with actions taken against piping degradation is illustrated by examples. The results of the evaluation confirm the conservativeness of the safety concept chosen for the design of German NPPs with LWRs, as well as the effectiveness of ageing management.     

Trends in the developments of new pipe systems for power plants

The necessity to save costs for building nuclear power plants whilst maintaining the high standard of safety requirements led, in the Federal Republic of Germany, through an increase in the stability of specifications to standardisation. As a result, the costs and schedules for nuclear power plant piping systems, which represent a schedule critical and thus a cost critical part of the nuclear power plant construction, are controllable and steerable. The principle of a basic, safe design offers the nuclear piping industry also for the future the opportunity for increasing efficiency, provided that sufficient pre-planning is available to ensure an orderly execution of piping, and that the available potential of experience of the specialized companies is used.     

Evaluation of piping damage in German nuclear power plants

Information relating to piping damage in safety-relevant systems in German nuclear power plants with light water reactors (both pressurized water reactors (PWRs) and boiling water reactors (BWRs)) were analyzed with respect to the modes and the causes of damages. In general, the total range of observed piping damage is low. The incidents (82) in plants with PWRs affected mainly pipes with small diameters. Almost all damaged piping showed wall-penetrating cracks combined with leakages, which revealed the damage. Initial cracks at piping with larger diameters were discovered in isolated cases during in-service inspections. With regard to the incidents (71) in plants with BWRs, piping with small as well as large diameters was affected to different degrees. Wall-penetrating cracks combined with leakages were detected at piping with small diameters. For large-diameters pipes, cracks were indicated during in-service inspections and supplementary examinations. The results of the incident evaluations confirm the conservativeness of the safety concept chosen for the design of German nuclear power plants with light water reactors.     

Progress in development of acceptance criteria for local thinned areas in pipe and piping components

Over the last 30 years there has been a considerable amount of research conducted on the effect of corrosion on the burst strength of buried gas and oil transmission pipelines. The results of numerous burst tests on artificial flaws and corroded pipe removed from service were used to validate an empirical analysis that was essentially the limit–load solution for an axial crack in a pipe under pressure loading. This basic concept led to acceptance standards in ANSI B31G, and a more recent modified B31G criterion using the RSTRENG computer program developed at Battelle. This program takes into account variable flaw depths rather than the parabolic flaw shape assumed in the original B31G criterion. Since that time, more fundamental research has been conducted to develop a more accurate and theoretically based failure criterion. The Battelle/Pipeline Research Committee International PCORR computer program is an example of a special purpose shell-element based, finite element, PC criterion for the evaluation of local thinned area (LTA) flaws. This program has evolved with time from linear-elastic to elastic-plastic stress with provisions for axial as well as hoop stresses. The development and new insights into blunt flaw behavior resulting from this program will be one aspect covered in this paper. In the nuclear industry erosion-corrosion, or flow-accelerated corrosion, in single-phase liquid lines has become a major problem. Computer programs, such as the EPRI Checworks program, have been developed to assist the plant operators with deciding where to focus their inspections. However, to date no generally validated acceptance criteria have been developed for the plant piping. Plant piping, whether in nuclear power plants, fossil power plants, or petrochemical plants, have several differences from buried pipelines which need to be considered. The buried pipelines typically have low longitudinal stresses that frequently are compressive, and have no pipe fittings such as tees, elbows, and reducers except at compressor stations. Plant piping needs to consider hoop stresses and axial tension loads from the pressure, as well as, bending stresses from dead-weight loads, thermal expansion stresses, and seismic loads. In an effort to develop flaw acceptance criteria for Section XI of the ASME Boiler and Pressure Vessel Code, the criteria in Code Case N-480 have been revised and implemented into a new code case (the number has not yet been assigned). These criteria essentially use either the ANSI B31G approach for axial flaws, or the ANSI B31.1 or ASME Section III stress analysis rules to show that the residual strength of the thinned region meets the initial design stress limits. This paper presents some of the validation efforts recently undertaken to determine the inherent margins in the design stress equation approach compared with the applied safety factors in the axial and circumferential flaw limit–load solutions in: (i) the gas and oil pipeline industries; (ii) the proposed criteria in Belgium for the nuclear industry and other criteria, and (iii) the preliminary criteria from a recently proposed ASME Code Case on erosion/corrosion acceptance criteria and the ASME Appendix H criteria for flawed ferritic nuclear pipe.     

Philosophy and practice of the aseismic design of nuclear power plants — Summary of the guidelines in Japan

Aseismic design is considered to be one of the most important factors for the safety of the nuclear power plants built in zones of high seismicity such as Japan. All structures, equipment and piping are classified in accordance with the importance of their radioactive safety to the plant, and the dynamic analysis and/or factored seismic coefficient analysis are applied accordingly. Site and ground conditions, as well as seismicity, should be studied thoroughly in order to estimate the intensities of the design earthquake and the safety margin check earthquake. Dynamic analyses of buildings and structures are performed using the multi-lumped-mass-system supported by soil springs with time history analysis conceptions. This idea is also applied to the design of equipment and piping by coupled system to the major structure or by the floor response spectra criteria. Tolerances are applied to damping factors although some experiments show more realistic results. Allowable stresses of ferrous metals for equipment and piping during earthquakes are more scientifically precise.

This report summarizes a guideline for aseismic design of nuclear power plants. The guideline was prepared by the Japan Electric Association in April, 1970, after three years laborious work.

In sect. 1, the philosophy and criteria are described. All components of a plant should be classified into three classes in accordance with their contributions to reactor safeties. Design to earthquake loadings should be based on “design basis earthquake” which is decided in consideration of local seismicity.

In sect 2, site selection and review for ground are described in the sense of seismic aspects.

In sect 3, deciding the earthquake motion for design is discussed. In Japan, semi-statistical approaches are used in normal practice.

In sect. 4, design philosophy and practice of building structures and containment vessels are described. They are designed under statical seismic forces, and the design of the class “A” structures should be checked by a dynamic response technique.

In sect. 5, design philosophy and practice of piping, vesels and equipment are described. Those which belong to class “A” items should be designed in a dynamic sense. Several programs for dynamic analyses of these items are prepared. Allowable stress under earthquake conditions is discussed in relation to other codes, for example, ASME Section III.

The greater part of the philosophy and design criteria have been adopted to all nuclear power plants which have been and are currently being built in Japan.