Interpretation of Müller-Klemm's Semi-Empirical Formula by Mason-Monchick Approximation for Thermal Diffusion Factor of Hydrogen Isotope Molecules

Comprehensive analysis on core support barrel (CSB) movements in the ULJIN Nuclear Plant is performed through noise and structural analysis techniques. Noise signals are taken from the lower channel outputs of ex-core instrumentation system during the full power reactor operation period. Then they are converted into auto-power spectral densities (APSDs) and coherence functions in the frequency range of 0–50 Hz to obtain the vibratory information of CSB movements.

From APSDs, the three different vibration peaks of CSB are detected around the frequencies of 8, 15 and 20 Hz, distinctly. These results are also agreed well with those obtained from the structural analysis by ANSYS version 4.3 computer program, which is the finite element method (FEM). Three different vibration mechanisms of CSB at each resonant peaks are identified as two types of the beam mode vibrations (vig., pendulum motion and torsional motion) and the shell mode vibration, respectively.     

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.     

Development of a measurement system of gap between CSB and RV to shorten a nuclear reactor installation period

To decrease the timescale for the installation of a nuclear reactor, a reduced-scale model system is proposed for Korea's third generation nuclear reactor, the APR1400 (advanced power reactor 1400). The construction period of a nuclear power plant is one of the most important factors to make a company competitive in international nuclear energy markets. Our study is related to the modularization of reactor internals to reduce the construction period of nuclear power plants. Generally, reactor internals comprise three components: the core support barrel (CSB), the lower support structure (LSS)/core shroud (CS), and the upper guide structure (UGS). The existing method of assembly is very complicated and requires approximately 8-10 months to complete. The installation of the reactor vessel (RV) is a critical process during the construction period. The proposed method for the modularization of reactor internals can shorten the construction period by a minimum of 2 months compared to the existing method. In order to modularize the reactor internals, gaps between the CSB snubber lug and RV core-stabilizing lug must be measured using a remote method from outside the RV. Therefore, the development of a remote measurement system was necessary. In this paper, we select a suitable sensor and develop a reduced-scale model system for the physical simulation of gap measurement.     

An improved method for reactor coolant pump abnormality monitoring using power line signal analysis

An improved method to detect the reactor coolant pump (RCP) abnormality is suggested in this work. The monitoring parameters that are acquired from power line signal analysis are motor torque, motor speed and characteristic harmonic frequencies. The combination of Wigner–Ville Distribution (WVD) and feature area matrix comparison method is used for abnormality diagnosis. For validation of the proposed method, the test was performed during cool-down phase and heat-up phase in nuclear power plant (NPP) by cross-comparison with RCP vibration monitoring system (VMS). Using pump internal inspection results, the diagnosis prediction is verified.     

Development of a remote measurement system for the gap between RV and CSB for RVI-modularization

This paper describes the development of a measuring system to measure gaps between the reactor vessel (RV) and the core support barrel (CSB) remotely with the aim of reactor vessel internals (RVI)-modularization. A remote measurement system was developed for use at actual construction sites of nuclear power plants using a measurement sensor, a threaded connection jig, and a zero-point adjustment device. With these, a reduced-scale model system was validated. With the remote measurement system, experiments and analyses were performed using mockups for both the RV and the CSB to verify the applicability of the proposed system in a construction project. From the data acquired by the remote measurement system, shims were separately made and adjusted.After installing the shims on RV core-stabilizing lugs, the gaps satisfied requirements within the permissible range of 0.381-0.508 mm. We evaluated the reliability and applicability of the remote measurement method and concluded that the remote measurement system enables RVI-modularization with a significantly reduced construction period.     

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.     

Component vibration of VVER-reactors — diagnostics and modelling

Flow induced vibrations of reactor pressure vessel (RPV) internals (control element and core barrel motions) at VVER-440 reactors have lead to the development of dedicated methods for on-line monitoring. These methods need a certain developed stage of the faults to be detected. To achieve a real sensitive early detection of mechanical faults of RPV internals, a theoretical vibration model was developed based on finite elements. The model comprises the whole primary circuit including the steam generators (SG). By means of that model all eigenfrequencies up to 30 Hz and the corresponding mode shapes were calculated for the normal vibration behaviour. Moreover the shift of eigenfrequencies and of amplitudes due to the degradation or to the failure of internal clamping and spring elements could be investigated, showing that a recognition of such degradations even inside the RPV is possible by pure excore vibration measurements. A true diagnostics, that is the identification of the failed component, might become possible because different faults influence different and well separated eigenfrequencies.     

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。      

Experiences with diagnostic instrumentation in nuclear power plants

Over the past several years, Westinghouse has developed a coordinated system of on-line diagnostic instrumentation for the acquisition and analysis of data for diagnostics and incipient failure detection of critical plant equipment and systems. Primary motivation for this work is to improve NSSS availability and maintainability through the detection of malfunctions at their inception. These systems encompass the following areas: (1) Vibration Monitoring System for detection of changes in vibrational characteristics of the major components of Nuclear Steam Supply System (NSSS) and Balance of Plant (BOP); (2) Acoustic Monitoring System for detection and location of leaks in the primary system pressure boundary and other piping systems in PWRs; (3) Metal Impact Monitoring for detection of loose debris in the reactor vessel and steam generators; (4) Nuclear Noise Monitoring System for monitoring core barrel vibration; (5) Sensor Response Time Measurement System for detecting any degradation of process sensors; and (6) Transit Time Flow Meter for determining primary coolant flow rate. Summarized in this paper are some of the features of the systems and in-plant experience. These experiences demonstrate that diagnostic systems in combination with analytical and laboratory work for data interpretation do improve plant availability.     

Application of an integrated power line signal analysis methodology to the detection of major rotating machine abnormalities in nuclear power plants

An integrated and improved method to detect and identify the abnormality of motor driven rotating machinery in nuclear power plants (NPPs) using power line signal analysis is suggested in this work. The primary goal of this work is to improve the motor current signature analysis (MCSA) method that has been used as an alternative or supplement of the conventional vibration monitoring system (VMS). Through this work, the integrated system using both modulated flux density model (MFDM) and rotating flux model (RFM) is proposed. The MFDM is based on the fact that the major mechanical vibration of rotating machines can be normalized to the motor air-gap eccentricity and the modulation of air-gap flux density. Therefore, if the major defect such as bearing defect or the shaft deformation is present, it is identifiable through the power line signal resulting from the modulated magnetic density. Moreover, the broken rotor bar state or rotor eccentricity due to electrical imbalance can be analyzed using the RFM. The other important feature of this system is an automated abnormality detection and diagnosis algorithm. It is possible to diagnose the abnormality without relying on experts in NPPs. The verification is done through varying load/torque test experiment as well as via computer simulation in this work. The experimental results show that they are in good agreement with the simulated results.     

防城港核电站堆内中子通量测量系统指套管碰磨分析

防城港核电站1号机组主泵惰走试验期间,在对核电站松脱部件和振动监测系统13路加速度通道进行背景噪声例行检查时发现,通过松脱部件和振动监测系统的声音监听设备监测到,安装于反应堆压力容器底部堆内中子通量测量系统导向管上通道有"哒哒哒"的异常信号。为找出异常信号源,利用松脱部件监测系统声监测功能对压力容器底部监测到的异常信号进行分析,该信号不是由松脱部件产生的信号。通过听音棒的辅助监听,最后综合分析得出该信号是由堆内中子通量测量系统指套管在管道路径上碰磨引起。该事件的分析与解决,不仅解决了工程建设需要,对核安全局批准下一步工作开展提供了支持依据,而且对通过松脱部件监测系统来开展由于流致振动引起的中子通量测量系统指套管异常振动诊断有重大的实用价值。     

Testing a system prototype for out-of-reactor monitoring of reactor power and its distribution in the core volume

Conclusion Prototype tests were performed for a representative set of reactor conditions: the power distribution was performed by a wide range of changes in the position of three groups of control and protection system regulatory devices, the A0 values were varied in the range from –0.40 to 0.16, and the coefficient of non-uniformity k_v in the range from 1.7 to 2.6. The results showed the high accuracy and effectiveness of out-of-reactor monitoring of the power and its distribution throughout the core volume.For an LWR, the out-of-reactor monitoring system assemblies can be placed in radiation shielding channels; for this three assemblies, each having three detectors, is sufficient. Chambers having an energy range of standard neutron flux monitoring equipment can be used as detectors.Determining the thermal power and the coefficients of nonuniformity of its distribution in the core does not begin to exhaust the possibilities for out-of-reactor monitoring. Algorithms already exist or are being developed which would allow increased accuracy in the monitoring of power and its distribution, localization of the region or fuel element with the greatest energy loading, detection of stuck control rods and nonfunctional thermal monitoring sensors, and diagnosis of fluctuations and position shifts of internal reactor vessel structures. Thus a reliable, cheap, rapdily responding core condition diagnostics systems can be constructed on the base of out-of-reactor detectors.Translated from Atomnaya Énergiya, Vol. 64, No. 3, pp. 174–180, March, 1988.     

一种用于旋转设备运行状态无线监测的智能无线振动传感器设计

为提高已投入运行核动力装置旋转设备的运行数据采集和状态监测能力,需要解决安装传感器和敷设配套线缆困难的问题。本文采用现场可编程门阵列（FPGA）作为主控单元,设计了一种基于Zigbee物联网通信技术的智能无线振动传感器,并给出了其电路构成、工作原理,以及嵌入式控制软件的工作流程。通过对此传感器进行性能测试,结果表明该传感器功耗低,实现了对振动信号的连续采集、智能分析与上传。该无线传感器安装简单,无需敷设供电和信号线缆,可应用于构建核动力装置旋转设备的状态监测系统。      

BEACON堆芯在线监测系统概述

历经20多年的发展,堆芯在线监测系统的核心技术在国际上的发展已趋于成熟,但国内仍无可与之相比的综合系统。通过对全球应用最广泛的堆芯监测系统--BEACON进行调研分析,为自主研发此类系统提供技术指引。BEACON利用堆外中子探测器、堆芯出口热电偶、堆内可移动探测器的测量数据,进行堆芯监视、测量数据分析以及预测。在节点均匀的假设下,以有效快中子群（EFG）模型简化扩散方程求解,再用格林函数对全堆芯插值,最后利用样条函数拟合探测器数据,给出较为准确的堆芯功率分布。BEACON的显著技术特点包括能进行非稳态下的堆芯监测,引用节点展开法（NEM）增加堆芯功率重构的准确性,以及使用单点校准技术增加两次全堆校准间的间隔。     

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。     

CPR1000机组100D型主泵振动故障诊断

CPR1000机组各运行模式下100D型主泵的振动现象表明,当处于蒸汽发生器冷却正常停堆或余热排出冷却正常停堆工况时,主泵电动机的瓦振幅值往往存在大范围冲击波动甚至触发振动高报警的现象。根据机械振动原理综合分析电动机瓦振、主泵轴位移信号的频域和时域特征,诊断振幅波动受某7~9Hz的低频随机振动影响;通过分析堆内构件振动噪声监测系统采集的信号判断该低频振动对应一回路主冷却剂流动过程中诱发的堆芯吊篮梁式振动。根据流体诱发振动理论分析了影响主泵电动机振动波动的主要因素,并通过主泵历史运行记录进行了验证。系统性提出优化CPR1000机组运行策略缓解主泵电动机振动波动的建议,为主泵安全稳定运行提供参考。      

基于功率密度谱的压水堆核电厂中子噪声特性研究

分析了压水堆核电厂中子噪声功率密度谱的计算方法,利用该方法以核电厂堆内构件振动监测系统长期的监测数据为基础,计算了中子噪声的功率密度谱,分别分析了百万千万级核电厂、不同功率核电厂和不同燃料周期核电厂中子噪声功率密度谱特性。结果表明,通过分析压水堆核电厂的中子噪声功率密度谱特性,能有效的认识压水堆核电厂堆内构件的振动行为,为压水堆核电厂堆内构件状态分析提供了基础。      

多样化驱动系统保护功能及保护参数信号确定

多样化驱动系统（DAS）为反应堆紧急停堆和驱动专设安全设施提供了与保护和安全监测系统（PMS）不同的多样化的后备。本文结合概率安全评价（PSA）分析工具,以功率运行内部事件PSA模型始发事件导致的堆芯损伤频率为度量,筛选出需要DAS提供保护功能的系统,并通过分析事件进程确定了DAS驱动各系统的保护参数信号。结果表明：通过该方法可合理确定DAS的保护功能和参数信号。