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.     

核电站松脱部件监测系统研制

以秦山核电二期扩建工程松脱部件与震动系统(KIR)供货项目为背景,研制出了松脱部件监测系统(LPMS)C1201。该系统由10个监测通道组成,对压力容器和蒸汽发生器内产生的松脱部件进行实时监测。信号调理和报警处理采用现场可编程门阵列(FPGA)程控技术;松脱部件甄别采用了DSP数字化、松脱部件信号长短有效值浮动甄别、时间延迟和通道复核联合甄别等技术,使系统具有强抗误报警能力;系统自检采用I2C控制与多路分配程控技术;系统采用PXI总线技术,使通道间和模块间同步,满足定位分析需要;采用虚拟仪器和数据库管理技术,界面可视化强,接口开放式;系统还具有软硬件故障自检与故障通道切除、磁盘空间检测、软件停电保护等功能。     

核反应堆故障早期检测和在线状态监测方法

即将颁布的核行业标准《核电厂反应堆堆内构件的振动监测》(简称标准)对早期监测反应堆压力容器堆内构件蜕化的方法、故障检测仪表和监测程序提出了要求,适用于以中子波动信号和反应堆压力容器振动信号为基础的堆内构件和一回路部件的动态特性的监测。本文强调了标准所包含的核电站新型仪表控制系统的监测方法。新系统不同于传统的监测系统,它的主要目的是早期故障检测．以便向电厂操纵员和检查维修人员提供有用的状态信息。     

裂变室探测器在高温气冷堆核电站示范工程核测量系统中的应用

高温气冷堆核电站示范工程是我国中长期发展规划中的重大专项之一,也是我国第一座拥有自主知识产权的核电站。文章介绍了高温气冷堆核电站核测量系统的工作原理和系统组成,剖析了裂变室中子探测器的特点及其在核测量系统中的应用。裂变式探测器所具有的γ信号甄别能力强、中子通量测量范围宽等优点,应用于高温气冷堆核测量系统,可以满足大空间、宽范围的堆外中子通量测量需求,有利于简化系统组成、提高经济性。与其他压水堆核电站的核测量系统相比较,应用裂变室中子探测器具有一定的优越性。     

基于虚拟仪器的核电站松脱部件监测系统

采用LabVIEW为开发平台,集虚拟仪器技术、设备组态图形化技术和数据库管理技术于一体,实现了反应堆压力容器和蒸汽发生器松脱部件的在线实时监测.该监测系统在核反应堆松脱部件模拟试验装置上进行了试验验证,实现了松脱部件监测的基本功能.     

Primavera6.0在核电厂设计进度中的应用探索

通过简化假设,分析了中子传输矩阵的物理意义,推导出中子传输矩阵数学模型,并利用以往的数据进行了验证.同时根据矩阵的共轭梯度算法理论,研究利用堆外核探测器系统(RPN)的功率量程通道(PRC)6节电离室信号及堆内中子通量测量系统(RIC)获得的堆内通量分布信号计算中子传输矩阵的方法. 这种算法得到的中子传输矩阵,可以植入冷却剂丧失(LOCA)监测系统(LSS系统).通过LSS系统可以实 时监测堆芯轴向功率分布,进而监测堆芯轴向线功率密度.     

基于错对中分析的M310堆型中子测量管座位置度设计

压力容器中子测量管座位置度是影响压力容器中子测量管座与堆内构件仪表套管错对中最为重要的因素之一。中子测量管座与仪表套管的错对中越大,中子注量测量通道内指套管的磨损越剧烈,中子注量测量探测器也越容易损坏,进而造成较大的安全隐患和经济损失。本文通过对中子测量管座与仪表套管的错对中分析,并结合在运核电厂制造经验反馈,提出中子测量管座相对合理的位置度设计要求。     

通道缺陷模式下的松动部件撞击试验研究

利用在役核电站停堆检修的窗口，对核电站蒸汽发生器进行模拟松动部件撞击的试验，给出了松动部件监测系统（LPMS）在经过长期运行后的主要缺陷模式及处理方法。在指出LPMS故障自检功能中存在的盲区的基础上，分析了存在缺陷的通道对松动部件冲击信号的响应特征。研究表明，通道接触不良、电荷累积和多通道间信号干扰是造成通道信号失真的主要因素；电荷累积会对信号通道造成静电阻塞；多通道间的信号干扰是产生误报警、通道过载断路等现象的重要原因之一。      

田湾核电厂1号机组主泵松脱部件报警事件诊断分析

对田湾核电厂松脱部件监测系统(LPMS)监测到的1号机组的松脱部件报警事件信号进行诊断分析.诊断认为引发这些的报警信号的事件发生在4号主泵附近,是由于4号主泵或其附近有松动件或动静摩擦导致的,不是脱落件.根据分析结果和检修建议,在田湾核电厂1号机组第1次大修(T101)期间,对4号主泵进行检查并对压力容器底部进行目视检...     

“华龙一号”堆芯中子通量测量系统设计

传统的二代核电厂普遍采用定期从压力容器底部插入中子探测器的方式来获得堆芯中子通量密度等信息,这种设计降低了反应堆的固有安全性。根据三代核电设计标准以及"华龙一号"堆芯中子通量测量系统的需求,本文设计了一套可用于实际工程的三代核电堆芯中子通量测量系统。该系统由自给能中子探测器组件、信号处理柜和控制柜组成,在通过系统研发、工厂测试和K3级设备鉴定后已成功应用于"华龙一号"全球首堆核电机组。该系统与国外同类型设备相比,其整体性能指标达到了国际先进水平。     

数字化核测量系统反应堆倍周期计算的敏感性分析

反应堆倍周期是核反应堆工程中的一个重要参数。在反应堆启动和功率提升过程中,操纵员可通过反应堆倍周期来了解反应堆的运行状态,并据此控制反应性。数字化核测量系统通过对与反应堆功率成正比的电压信号进行采样和处理,计算得到反应堆倍周期。在实际的应用中,电压信号往往包含测量噪声,对计算结果带来较大的不确定性。针对数字化核测量系统的倍周期计算问题,对其敏感性进行了分析,并给出相应的算例。     

Experience with identification of loose parts by acoustic monitoring of primary system

Substantial progress has been achieved in the identification of loose parts which had been detected by acoustic monitoring of reactor primary system. Several years of practical experience and the use of the offline digital analysis system MEDEA proved that acoustic monitoring is very successful for detecting component failures at an early stage. ISTec is involved in loose parts monitoring in several nuclear power plants in Germany. Advanced powerful tools for classification and evaluation of burst signals have been realised.

Loose parts monitoring systems, which are installed in all German nuclear power plants (NPPs), indicated specific impact conditions at lower plenum of two BWR's. Flow tests were carried out with various coolant flow rates of internal axial pumps and use of model nuts in one case. More than 2000 different bursts have been analysed to provide information in detail about impact occurrences, their spectral characteristics and impact sequences. Burst shape parameters could be determined and signal amplitudes have been trended. Determination of the sound origin — fixed origin in one case, flow-induced moving origin in the other case - and mass estimation of the loose parts could be performed by application of advanced burst analysis methods. Characteristics of the impact signals are presented in the paper.     

The design of the J-type neutron flux distribution measurement system

This paper introduces the design of a J-type aeroball system that the tube penetrates the lateral wall of reactor pressurized vessel (RPV), then immediately goes down to the vessel bottom and then goes up through the lower core support plate into the reactor core. Some experimental results related to gas flowing within a thin tube are presented in the paper, such as the gas friction drag coefficient on the ball’s way and etc. From theoretical and experimental viewpoints, the feasibility of the system is proved in pneumatic holding-up and β measurement aspects. In order to ensure an enough ratio of signal to noise, the maximum distance between two measuring points in water reactors is given. The paper gives out the measuring number per two-assemblies width N_i=Int(n_i/2+0.51), which is the accuracy relation between the number of fuel assembly and the minimum one of measuring point to only reconstruct 2D neutron flux distribution completely by the measured data.     

核电厂指套管磨损涡流检查深度定量的影响因素研究

反应堆中子注量率测量指套管的涡流检查结果是核电厂维修行动的参考依据。基于现用的涡流检查方法，分析了指套管磨损缺陷的参数变化对涡流检查深度定量的影响，结果表明:周向磨损宽度在210°以内时对涡流测量深度定量的影响较大；轴向磨损长度小于20 mm时，对深度定量的影响明显；月牙型磨损的涡流测量值偏小；在2次磨损的周向位置不变的情况下，磨损叠加对测量无影响，当2次磨损的周向位置变化时，磨损缺陷叠加对测量结果影响较大。      

核电厂指套管磨损涡流检查深度定量的影响因素研究

The result of the eddy current inspection for the thimble tube of a nuclear reactor neutron flux measurement system is the reference for the nuclear power plant to take maintenance actions. Based on the currently-used eddy current inspection method, the effect of the parameter variation of the tube defect on the depth quantification is analyzed. The results show that when the angle of the defect circumferential is less than 210°, it has a greater effect on the eddy current measurement depth quantification and when the defect length is less than 20mm, the effect of the defect depth quantification is obviously. The measurement depth in the eddy current of the crescent-type defect is smaller than the design depth. When the circumferential position of two defects is constant, the overlapped defect has no effect on the measurement, but the overlapped defect will have serious influence for the measurement result if the circumferential position of the two defects changes.     

A real-time monitoring system of core support barrel vibration using FEM data and self-organizing neural networks

In this work a core internal vibration monitoring system which is particularly concerned with the core support barrel (CSB) in ULJIN nuclear power plant unit 1 in Korea is developed. Flow induced vibration and aging processes in the reactor internals cause unsoundness of the internal structure. In particular, the loose-joined flange between the top of the CSB and the head of the vessel may result in core or fuel damage accidents. In order to improve the accuracy of the conventional CSB monitoring system, signals from the piezoelectric accelerometers are used in this work instead of those from the ex-core neutron detectors. This work consists of three parts: one is the development of a suitable tool for detecting the hold down spring broken accident or wearing out of the CSB using the Fuzzy (self-organizing neural network) technique; another is the generation of vibration signals to represent the abnormal states of the CSB by finite element method (FEM) analysis and mock-up experiments; the third is the development of a graphical man-machine interface for the practical use of the monitoring system.     

固体径迹探测器测量反应堆功率研究

在零功率反应堆上利用固体径迹探测器直接测量燃料元件内的裂变率，可得到反应堆的功率。同时测量反应堆某位置的热中子通量密度，继而可得到单位功率的热中子通量密度。因此，通过测量该点的任何热中子通量密度即可得到反应堆的运行功率。该方法可以减少与能谱测量有关的修正工作。由于辐照所需的中子通量密度低、时间短，因此与活化法等相比具有明显的优点。     

反应堆压力容器中子辐照损伤的分析和评价

对于材料已经确定的反应堆压力容器,其辐照脆化效应的主要因素是快中子积分通量。本文应用中子输运格林函数法验算了秦山核电站压力容器1/4厚度处最大快中子通量。分析和评价结果表明,该压力容器的设计对中子辐照是安全的。