秦山核电厂化容系统振动测量

本文简要地介绍了秦山核电厂化容系统在调试期间的振动测量。测量与调试同步进行；主要使用便携式测振仪对事先确定的测点进行现场实测。测量结果表明，泵的振动级别大部分达到Ａ级，管道的振动满足振动鉴定准则的要求。     

非能动核电厂1E级阀门电动装置的鉴定与试验

介绍了三代非能动核电厂1E级阀门电动装置的设计及其鉴定要求,阐述了鉴定试验方案,包括基于IEEE 382-2006标准的代表性样机选型和鉴定试验序列,以及电磁兼容性、热老化、热循环、辐照老化、磨损老化、正常循环加压、振动老化、抗震和设计基准事故模拟等一系列的鉴定试验方法和结果。对抗震试验、设计基准事故试验中的技术问题进行了探讨,指出了相应的解决方案和措施。通过对国内自主研制的阀门电动装置样机的鉴定试验,最终验证电动装置在核电厂服役过程中能够达到规范书的安全功能性能要求,并具有60年鉴定寿命。     

孔板汽蚀诱发管道振动问题研究

针对核电厂安全壳喷淋系统(EAS)出现的强烈管道振动问题,采用现场振动试验和数值计算相结合的方法进行研究。研究发现,管道节流孔板过度节流,导致在孔板下游出现汽蚀是诱发管道强烈振动的根本原因;通过数值计算方法对孔板压降、级数、孔径和结构形式等进行一系列优化设计,给出采用三级孔板消除气蚀的减振改造方案,并对改造方案开展完整性评估。通过对改造后的管道进行再鉴定试验表明,采用本文的优化设计分析方法设计的工程改造方案很好地解决了孔板汽蚀诱发的管道强烈振动问题,管道振动和噪声均大幅降低,可以确保EAS管道系统的长期安全运行。     

HTR-10备用柴油发电机组的抗震鉴定

备用柴油发电机组是10MW高温气冷实验堆的重要I类抗震设备，采用在振动台上模拟地震试验方法进行抗震鉴定。本文详细介绍了抗震试验的方法和要求，并对试验结果进行了研究与分析。试验表明，机组能够满足HTR－10工程抗震的要求。     

某三代核电设备安全壳喷淋泵力学分析和试验研究

以某三代核电设备安全壳喷淋泵为对象,利用有限元对整个系统进行建模,完成了该系统抗震性能、各部件模态和应力、结构屈曲以及连接螺栓应力的分析和评定,并进行了现场振动频率测试。结果表明,此安全壳淋喷泵满足规范要求,与设备模态试验情况符合较好。本研究可为后续安全评审和设备鉴定提供研究支持。      

主泵试验台架振动原因分析与修改方案

反应堆冷却剂泵（简称主泵）在试验台架进行试验时出现振动偏大的现象,振幅超出样机规范书的要求,振动数据的频谱显示为低频振动。通过转子动力学的有限元方法分析了主泵的振动特性,对比振动频谱排除了主泵转子振动和轴承等自身振动的原因。提出了主泵振动特性分析应包括试验台架整体。将主泵、试验回路和试验台架作为整体,分析了整体的振动特性。结合敲击试验推断振动的原因是试验回路中流体压力脉动的宽频激励引发了主泵和试验台架整体振动,并在低速运行试验中进行验证。在此基础上提出了修改方案,包括增加吸能阻尼器、提前投入缓冲罐和滤波等,重启试验后主泵振动明显减小。     

核电厂管壳式换热器管束振动研究

管束流致振动是管壳式换热器管束发生机械损坏的重要因素,已引起换热器设计人员的高度重视。文章基于核电厂中管壳式换热器,介绍了管束振动造成的主要损坏形式、振动区域以及引起管束振动的主要机理。给出了HTRI软件中分析管束振动的判断依据,并结合振动机理给出了在换热器设计中影响管束振动的结构参数,旨在为设计人员提供一定的参考,能在实际设计过程中采取合理的防振措施,避免管束振动的发生。     

船用离心泵板梁支承动力特性改造分析及实验研究

离心泵是船舶常用动力设备,其板梁支承结构动力特性兲系着设备总体振动水平及振动特征。借助结构有限元仿真分析、设计及振动试验方法,对振动水平较高的某船用离心泵板梁支承结构迚行振动改造设计及试验验证研究。基于ANSYS结构动力学分析程序建立板梁支承结构振动分析模型,分析结构的动态特性,幵与振动试验结果迚行对比验证;基于验证后的振动分析模型,对结构迚行振动优化设计及分析,获得改造后的离心泵板梁支承结构的振动特性;依据改造方案对离心泵板梁支承结构迚行改造,通过振动测试获得结构振动特性;测试结果表明,结构动力特性改造方案可行,优化了结构动力特性,降低了结构总体振动水平。     

核动力主泵振动探测与积分处理方法研究

由于核动力主泵结构与运行环境的特殊性,采用特殊的加速度计进行振动探测,为了对其进行振动烈度评价,必须对振动加速度信号进行积分处理。在分析振动信号软件积分误差的基础上,提出基于FFT变换的波形修正积分算法,并通过设置合理的积分频率下限,消除了振动信号的积分趋势,提高了振动信号积分的精度,并在核动力主泵振动监测系统中的得到成功的应用。     

起伏振动下倾斜管内单相流流动阻力特性分析

起伏振动状态下单相流流动阻力的正确计算对漂浮核电站的安全性有显著影响。实验研究了不同起伏振动工况和流动工况对倾斜圆管通道内单相水摩擦压降的影响,提出了方便计算的振动摩擦阻力系数。结果表明,振动摩擦压降大于稳定状态的,并呈周期性波动,波动周期与振动频率一致。振动摩擦阻力系数平均值随雷诺数和倾角的增大而减小,随管径和振动频率的增大先增大后减小,随振幅的增大而增大。通过实验数据拟合得到起伏振动下倾斜管内单相水振动摩擦阻力系数计算经验关系式,计算结果与实验结果吻合较好,为起伏振动单相水流动阻力的计算提供了新思路。     

核电厂设备抗震鉴定的审评

介绍了核电厂设备抗震鉴定的标准和要求，鉴定的范围、步骤、方法和程序，以及抗震鉴定审评所依据的准则。详细论述了抗震试验鉴定中所采用的具体实施程序，并通过审评实践对核电厂设备抗震鉴定中常见的一些问题也进行了探讨。     

Qualification of equipment for nuclear power generating stations

The seismic qualification of equipment is demonstrated by test or analysis or combined test and analysis. Equipment generally have strong coupling of responses in different directions. In the qualification by analysis, the three components of the specified vibratory motions are considered. In the case of qualification by test, the qualification can be incomplete, as generally only two components of the vibratory motion are applied simultaneously. In the tests, concurrent thermal pressure, mechanical and vibratory loads are not applied simultaneously. As the flexibility of the floor can have a significant influence on the response to vibratory motions, such flexibilities should be considered in the development of the floor response spectra, used in the qualification of equipment.     

核电厂在役检查技术验证方法对比

以基于ASME规范的验证方法和ENIQ验证方法为对象,讨论2种在役检查技术验证方法的特点及其各自存在的问题。ASME规范要求的验证方法侧重于某一类部件检测技术的验证,强调盲测试验和判定验证的统计学证据的重要性;对于某些特定部件检测技术的验证,ASME规范XI卷附录Ⅷ中的相关规定不够明确。ENIQ验证方法强调通过技术论证和实际测试的结合进行验证,适应性较强。通过比较2种验证方法的特点,对我国的在役检查技术验证提出建议。     

核电厂1E级充电器及逆变器设备鉴定方案的研究

基于IEEE、IEC、RCC-E等标准及文献,阐述了1E级电气设备的质量鉴定方法,进而结合国内外设备鉴定的实践经验,制定了1E级充电器、逆变器的鉴定方案,对元器件评估、性能及应力试验、EMC试验、抗地震试验、软件鉴定等环节进行重点剖析。可为1E级充电器、逆变器设备国产化过程中的质量鉴定提供参考。     

相关系数法评估核安全设备鉴定样机代表性

核电厂安全级设备在质量鉴定过程中的样机选型代表性直接关系到将样机鉴定结论外推至电厂实际安装设备的合理性。围绕IEEE标准中的安全级电气设备鉴定样机代表性准则和选型方法进行讨论,重点讨论了选型方法的优化和改进,在标准选型方法的基础上提出了一种基于相关系数的样机代表性计算方法,应用该计算方法所得的鉴定样机选型结果具有符合标准要求、代表性可量化且选型结果准确等特点。     

百万千瓦级核电站蒸汽发生器大锻件工艺评定

文章介绍了百万千瓦级核电站蒸汽发生器大锻件工艺评定的背景、依据、目的、技术指标和评定方法。根据蒸汽发生器锻件的结构特点和制造工艺,形成了一整套评定试验方案,可以对锻件各部位的化学成分、力学性能、金相组织和内部质量进行全面验证。该方案已在国内福清、方家山等多个核电工程中的蒸汽发生器锻件评定中成功应用。     

Realistic seismic qualification using the updated finite element model for in-core components of reactors

It is now mandatory to seismically qualify the safety-related structures and components used in the nuclear power plants. Among several qualification approaches the qualification by the analysis using finite element (FE) method is the most common approach used in practice. However, the estimated dynamic behaviour by FE model of a structure is known to show significant deviations from the dynamic behaviour of the ‘as-installed’ structure in many cases. Considering such limitations, few researchers have advocated re-qualification of such structures after their installation at site to enhance the confidence in qualification vis-à-vis plant safety. For such an exercise, validation of FE model with experimental modal data is important. A validated FE model can be obtained by the model updating methods in conjugation with the in situ experimental modal data. Such a model can then be used for qualification. However, for the reactor in-core components such a modal testing and FE model updating may not be straightforward. Hence, the complication involved in the reliable seismic qualification of in-core components and the advantage of using the FE model updating has been brought out in the paper through an example of a typical in-core component—a perforated horizontal tube recently installed in a nuclear reactor in India.     

Performance demonstration for an automated ultrasonic testing system for piping welds

This paper addresses the implementation of an automated ultrasonic testing (AUT) system qualification by performance demonstration (PD) as imposed by the ASME Boiler and Pressure Vessel Code Section XI. To improve the reliability of the ultrasonic testing results for nuclear power plant (NPP) components, almost all engineering codes related to NPP inspection require the ultrasonic inspection systems to be qualified by passing a PD examination. In this study, an AUT system developed to inspect pipe welding parts in NPPs is introduced. To acquire a Korean Performance Demonstration (KPD) qualification, the developed system had a KPD. System obtained the qualification for flaw detection, length, and depth sizing from KPD.     

Dynamic qualification of complex structural components of nuclear power plants

The safety requirements and the lack of accessibility for any future repair, impose the design requirement that the integrity of reactor components of nuclear power plants be assured for the lifetime of the plant. To meet this design requirement it is essential to qualify the component, i.e. prove its capability to perform the design function for the design life. In performing its design function, the component is subjected to both static and dynamic loads. The qualification for static loads is rather simple and reliable, but qualification for dynamic loads is complex and often uncertain. This is because analytical tools are often inadequate for a realistic dynamic qualification and exact structurally simulated experimental models are almost always difficult to build. In such a situation, methods using tests on simple experimental set-ups supplemented by conservative analytical back-ups must be evolved. This paper highlights the intricacies involved in the conservative dynamic qualification of the complex components by considering the example of the moderator sparger tube. This component is a perforated tube submerged in water and excited by flow. For such a case, a completely analytical or a totally experimental qualification is not possible. This paper describes a procedure by which the required dynamic characteristics such as added mass, damping and fluid forces are generated from simple experiments and the component is qualified by analysis using these data.     

压水堆核电站严重事故下的设备鉴定

本文给出了压水堆核电站严重事故下设备鉴定研究的基本内容和研究结果。对需鉴定设备的确定原则,鉴定环境条件包括压力、温度及辐照剂量的确定方法,以及设备鉴定建立的过程进行了详细的讨论和分析。