秦山核电二期工程反应堆吊篮结构的流致振动响应计算

采用实验与理论相结合的方法,计算吊篮在动水中随机响应.首先,利用15吊篮结构模型试验所获得的脉动压力功率谱,经归一化折算为原型堆的脉动压力功率谱,作为计算的输入载荷--力函数;其次,采用静水中吊篮结构振动特性计算所获得的特征值(频率,和特征向量,振型),得到结构的力-位移传递函数矩阵;最后,根据随机振动响应的计算方法,由力函数与传递函数矩阵算出吊篮的均方根位移.     

用大涡模拟计算流致振动的流体激励力

本文对湍流大涡模拟理论的脉动过滤和Smagorinsky亚格子模型进行了介绍,并用大涡模拟对秦山II期反应堆1∶5模型压力容器和吊篮的环腔内流场进行了数值模拟,计算了吊篮表面压力的时间分布和空间分布,其压力脉动的功率谱密度计算值和实验值在同一个量级内.因此,可以用计算所得的压力作为吊篮振动的激励载荷.     

反应堆吊篮在泵致脉动压力载荷下的响应研究

反应堆结构的流致振动问题一直受到核工程界的广泛关注。主泵的泵致脉动压力是一个重要激励源,其将导致反应堆吊篮等部件周期性振动,长期运行会导致结构的疲劳损坏。为研究新设计的“华龙一号”反应堆吊篮在泵致脉动压力作用下的振动响应,本文首先分析反应堆吊篮所受的泵致脉动压力,而后建立吊篮有限元模型,对其在泵致脉动压力载荷下的动力学响应进行研究,并综合考虑湍流激励,评价吊篮在堆内构件流体作用下的整体影响。应力分析表明,吊篮各位置流致振动的最大应力强度小于疲劳应力限值,结构是安全的。但对于新设计的反应堆,或反应堆冷却剂系统更换新的主泵,则反应堆吊篮及堆内构件的泵致振动需受到重视。     

反应堆吊篮水流振动研究和分析

应用流-固耦合理论推导相似准则和模型设计。在反应堆1:10模型上进行了吊篮结构的空气和静水中模态分析试验,获得动态特性。在反应堆1:5水力回路上进行吊篮水力振动试验,测得了流体对吊篮表面脉动压力、加速度、应变等信号与流量关系。经信号处理后获得了流体载荷谱及各种响应参数。在这两个试验基础上对流-固耦合与流致振动定量分析,导出吊篮水流振动理论计算方程。计算了秦山核电厂300MW的吊篮水流振动的响应与试验相符合。试验分析结果表明吊篮水流振动在寿期内是安全的。     

反应堆吊篮水流振动研究和分析

应用流-固耦合理论推导相似准则和模型设计。在反应堆1:10模型上进行了堆芯吊篮结构的空气和静水中模态分析试验,获得动态特性。在反应堆1:5水力回路上进行吊篮水力振动试 验,测得了流体对吊篮表面脉动压力、加速度、应变等信号与流量关系。经信号处理后获得了流体载荷谱及各种响应参数。在这两个试验基础上对流-固耦合与流致振动定量分析,导出吊篮水流振动理论计算方程。计算了秦山核电厂300MW的吊篮水流振动的响应,结果与试验相符合。试验分析结果表明吊篮水流振动在寿期内是安全的。     

秦山核电二期工程反应堆堆内构件模型流致振动试验研究

秦山核电二期工程是我国自行设计的第一个600MW级核电站,有必要进行反应堆堆内构件流致振动试验研究.本文介绍了按相似准则设计的实堆15的堆内构件试验模型,进行流致振动试验采用的试验方法,完成的试验内容以及试验数据的分析和处理.测得了吊篮结构在冷却剂流动冲刷下的脉动压力和各种响应参数.试验结果可用于秦山核电二期工程安全评审,并提供了吊篮流致振动响应计算的载荷谱和实堆振动监测、故障诊断的参考样本.     

基于中子噪声分析的某核电厂堆芯吊篮梁型振动特征研究

研究了基于堆外电离室中子噪声信号监测压水堆核电厂反应堆吊篮的方法,通过计算电离室中子噪声的互功率密度谱、相干和相位,分析得到了堆芯吊篮梁型振动的频率;利用该方法,计算获得了某正常运行状态下压水堆核电厂换料周期内堆芯吊篮梁型振动频率和中子噪声功率谱幅度的变化趋势,结果说明了在反应堆正常运行状态下,随着堆芯燃耗的增加,吊篮梁型振动频率发生了微小漂移,频率变小,该频率处中子噪声功率谱幅度变大。     

中子噪声在核反应堆吊篮振动监测中的应用研究

通过对临界装置堆芯吊篮激励振动引发中子噪声实验得到的功率谱密度(PSD)进行分析,证实了从中子噪声PSD中获得吊篮振动特性(各阶特征频率)是可行的,并给出了中子噪声探测器PSD幅度与吊篮振动幅度之间的比例因子(刻度因子)的计算方法.针对临界装置测量获得的中子噪声PSD和吊篮振动PSD,实际计算了对应吊篮各阶振型的刻度因子.本文证实,可以通过中子噪声分析,给出吊篮结构的振动频率和振动位移,证实了中子噪声在堆内构件振动监测领域的有效性.     

摇摆条件下水平圆管内湍流压力脉动特性分析

一回路设备在设计时需考虑压力脉动的影响,而在海洋条件下,船体的运动会影响压力脉动,因此需要考虑海洋条件对压力脉动的影响。本文建立了海洋条件下的附加惯性力模型,通过自定义函数将该模型应用于Fluent求解器中,采用大涡数值模拟方法对细长管道进行数值计算,得到不同摇摆频率及雷诺数下的流场结果,并对监测点进行时域与频域分析,获取到监测点处的压力均方根及功率谱密度曲线。分析结果表明：在湍流充分发展区,压力脉动量相对于平均压力而言为小值;湍流压力脉动随着船体摇摆频率/雷诺数的增加而增加;随着雷诺数的增加,其对湍流压力脉动的影响减小,但摇摆条件增大了雷诺数对压力脉动的影响。     

基于压力容器振动信号分析的堆芯吊篮壳型振动特性研究

核反应堆堆芯吊篮的振动状态直接关系到堆芯的安全运行,但堆芯吊篮处于高温和强辐照环境下,无法直接在吊篮上布置传感器测量其振动。本文利用安装在压力容器上的加速度计间接监测吊篮的振动,通过对多核电机组压力容器振动信号相干谱、自功率谱和互功率谱进行分析,获得吊篮壳型振动频率和振幅,并将分析结果与秦山核电厂二期1号机组试验实测值进行比较,分析结果与试验结果相近。研究表明通过对压力容器振动信号的监测与分析,能够有效识别堆芯吊篮壳型振动特性,为吊篮状态评价提供基础。      

Mathematical modelling of dynamic deformations in a cylindrical shell excited by turbulent flow

In this paper the mathematical model of the vibration response of structures to random exciting forces is applied to explore the influence of fluid flow parameters on the dynamic deformations and behaviour of a cylindrical shell. A physical model of the PWR core barrel, considered as a cylindrical shell supported at both ends, is used to demonstrate by numerical experiments the mentioned influence. The mean velocity vector of the fluid flow is the basis for the aero-hydrodynamic excitation expressed by a coherence function of fluctuating surface pressure both in the axial and circumferential directions, stated as the dependence on the values of the correlation lengths. The results of the courses of the generalized spectral loadings, of the ms amplitudes of the displacement and stress distributions are shown at the dependence on the flow parameters.     

格架松弛对燃料棒湍流激励及微振磨损的影响研究

假设所有支承有效,基于燃料棒模态分析的结果,根据压水堆燃料棒的流场分布特征,采用功率谱密度表征湍流激励,结合相关功率谱密度试验参数,求解了各阶模态的振动位移均方值,基于ARCHARD磨损公式计算了燃料棒刚凸位置的磨损深度。由于制造工艺、运输、辐照的影响,格架对燃料棒的夹持作用可能松弛。依次假设格架单个刚凸及弹簧松弛,研究了松弛对燃料棒模态、流致振动以及磨损的影响。结果表明：格架弹簧的松弛对固有频率的影响可忽略;原振幅较大的位置附近刚凸松弛对固有频率影响明显;堆芯入口及出口的横向流速较大,燃料棒底部和顶部的湍流激励振幅较大,这些位置的刚凸支承松弛使湍流激励振幅明显增大,中间位置的刚凸支承松弛对振幅影响较小;刚凸支承松弛对磨损深度的影响与对湍流激励最大振幅的影响趋势基本一致。磨损除了与湍流激励振幅相关,还与固有频率相关,顶部振型和频率乘积的影响大于底部格架位置,顶部格架刚凸松弛对磨损影响最大。     

某核电厂吊篮梁型振幅偏大问题分析与诊断

通过对方家山、宁德核电机组多个燃料循环周期的堆外中子噪声信号进行分析,得出了吊篮梁型振动频率和幅度的特性。将此特性应用到其他核电机组,发现某核电机组吊篮梁型振幅偏大。对该机组吊篮梁型振动频率和幅度的变化趋势、频率的漂移量、幅度的增长率进行分析与诊断,判断吊篮并没有发生明显的支撑劣化情况,并给出了该机组可继续运行的条件。      

运行核电厂堆芯吊篮壳型振动特性研究

反应堆堆芯吊篮的振动反映了吊篮及相关堆内构件的振动情况与设备稳定性，是评估反应堆安全运行的重要依据。本文采用中子噪声技术和信号时频域分析方法，重点研究了堆芯吊篮壳型振动特性，通过分析某核电厂特定机组近几年的监测数据，获得堆芯吊篮壳型振动模态参数的变化趋势。结果表明，在每个燃料周期内，吊篮壳型振动频率有逐渐变小趋势，每经历一次大修后，振动中心频率基本恢复至上一个燃料循环的初始振动频率处。研究结果有助于了解堆芯吊篮在多个燃料循环周期内壳型振动的特性和成因，为堆芯吊篮早期故障诊断奠定基础。      

Flow-induced vibrations in closed side branch pipes and their attenuation methods

Some problems due to flow-induced vibrations related to closed side branch pipes have been observed in thermal and nuclear power plants. Fluctuating pressure generated in the main pipes was unusually, acoustically excited in closed side branch pipes, and intense vibrations were caused at pipes and components. For example, flow-excited acoustic resonance in closed side branches of stub pipes of safety relief valves caused the failure of steam dryers in the United States Quad City Unit 2 nuclear power plant. Furthermore, there was a possibility that residual air or gas in a closed side branch pipe unexpectedly caused severe vibrations of low frequency in the feed water piping system. We have investigated the root cause and influence of air on severe vibrations. Intense fluctuating pressure was often caused by water hammer due to valve closure and it became larger in the closed side branch pipes. We showed that an additional side branch with an orifice was very effective to suppress the flow-induced acoustic resonance. Design methods of the orifice to attenuate fluctuating pressure generated by water hammer were presented considering Mach number, the pressure loss coefficient of orifice and the intensity of particle velocity. Moreover, suitability of the characteristic curve method was confirmed for evaluation of the attenuation effect of an orifice on fluctuating pressure generated by water hammer. Finally, we considered some flow-induced vibration problems related to closed side branch pipes and their attenuation methods.     

基于中子噪声分析技术的核电站振动监测系统研制

以秦山核电二期扩建工程松脱部件与振动监测系统(KIR)供货项目为背景,研制出了VMS C1201堆内构件振动监测系统.该系统由4个加速度通道和8个中子噪声通道组成,采用PXI总线技术以及虚拟仪器、数据库管理和监测报告自动生成技术,信号调理采用现场可编程门阵列(FPGA)程控技术,各通道信号采用同步处理技术;监测软件采用原始数据存储,并提供开放式接口.该系统具有时程分析、自谱与互谱分析以及压力容器、吊篮和燃料组件振动监测功能.     

核承压热交换器两相流流致振动现象研究

为保障核承压热交换器的安全运行,采用数值模拟以及软件计算相结合的方法,对核承压热交换器两相流流致振动现象及减振措施进行了探究。研究结果表明:基于流致振动发生机理,热交换器横流速度、固有频率、卡门旋涡脱落频率以及紊流抖振频率为重点分析因素;由公式得出流量、换热管直径、换热管壁厚、管束排列等对流致振动有直接影响,无支撑跨距是影响管束流致振动较大因素;最易发生流致振动的部位包括入口区域、出口区域、折流板缺口区域以及无支撑跨距大管束;设计中,应在流量、换热管直径、壁厚、无支撑跨距、管束排列及入口防冲挡板设置等方面优化,以减小流致振动危害。     

Air/steam flow and steam wetness dependence on acoustic resonance in safety relief valves

Many experimental studies related to the flow-induced acoustic resonance closed side branches have been reported. However, few studies have reported on the effects of air/steam flow and steam wetness dependence on fluctuating pressure amplitude. Therefore, we investigated the effect of air/steam flow and steam wetness dependence on fluctuating pressure amplitude by conducting a high temperature and high pressure tests at the Hitachi Utility Steam Test Leading Facility (HUSTLE). The test section consisted of a main pipe and a side branch. The side branch was mounted on the long straight main pipe. Fluctuating pressures at the end face of the side branches were measured. The following two results were obtained; the first is that the air/steam flow had little effect on the fluctuating pressure amplitude normalized by dynamic pressure and frequency normalized by the resonance frequency; the second is that under the acoustic resonance (St = 0.41) and non-resonance (St = 0.55) conditions, fluctuating pressure and frequency changed little with steam wetness. The steam wetness during the boiling water reactor operation was less than 0.1%; thus, there was no effect of steam wetness on the acoustic pressure amplitude and the frequency under this operating condition.