Seismic qualification of motor operated valves—alternate approach

This paper presents a potential alternate method for determining operating capacity of motor-operated valves subjected to seismic and other applicable loadings. As a result of programs at nuclear facilities to ensure the operational capability of MOVs (under NRC GL89-10), extensive analytical focus to develop the structural capability of valves has ensued. In the past, seismic qualification of valves typically addressed the strength of the topwork structure to resist inertial loading from excitation of the large valve actuator mass. These evaluations paid little or no consideration to the loading resulting from valve closing forces. The focus of the recent efforts is to develop the maximum operational capability of the valve, in terms of thrust, with consideration of seismic and other services loading as applicable. The alternate method outlined in this paper presents a series of thrust capacity curves, with reduction factors for seismic loading which can be applied and developed to determine safe thrust loadings without performing extensive analytical effort. A similar approach was put forward by the SQUG GIP approach to MOVs to ensure the safe operation of valves based on past earthquake experience. However, the GIP approach cannot be used to determine safe operational loads and thus has limited use in the necessary analysis required for GL89-10 programs at nuclear facilities.     

Application of signature analysis for determining the operational readiness of motor-operated valves under blowdown test conditions

In support of the NRC-funded Nuclear Plant Aging Research (NPAR) program, Oak Ridge National Laboratory (ORNL) has carried out a comprehensive aging assessment of motor-operated valves (MOVs).As part of this work, ORNL participated in the gate valve flow interruption blowdown (GVFJB) tests carried out in Huntsville, Alabama, The tests provided an excellent opportunity to evaluate signature analysis methods for determining the operability of MOVs under accident conditions.ORNL acquired motor current and torque switch shaft angular position signauresnon two test MOVs during several GVFIB tests. The reduction in operating “margin” of both MOVs due to the presence of additional value running loads imposed by high flow was clearly observed in motor current and troque switch angular signatures. In addition, the effects of differential pressure, fluid temperature, and line voltage on MOV operations were observed and more clearly understood as a result of utilizing the signature analysis techniques.     

A study on the actuator efficiency behavior of safety-related motor operated gate and globe valves

Periodically, the operability of the safety-related motor-operated valves (MOVs) in nuclear power plants must be verified. Because the actuator efficiency is one of the most important factors in the determination of the actuator output, it should be considered in ensuring the operability of MOVs during the verification duration. In particular, special consideration should be paid to its potential degradation, but the design efficiency provided by manufacturers is usually used because the actuator efficiency calculation is difficult and requires considerable time and effort. In this paper, a method is introduced to calculate actuator efficiency by using diagnostic signals acquired in field tests. The actuator efficiency was calculated from the estimated motor torque, the stem thrust measured in field tests, and overall gear ratio provided by manufactures. The motor torque was estimated by using an algorithm, which can calculate electric torque from the three phases of currents and voltages, resistances between phases acquired in field tests. The validation of the design efficiencies was evaluated by comparing those efficiencies with the calculated actuator efficiencies. And, the age-related degradation was analyzed through the behavior analysis over time of the calculated actuator efficiencies. Most of the actuator efficiencies were found not to be degraded over time and kept efficiency greater than the design efficiency. However, two actuator efficiencies with lower motor speed, overall gear ratio, and maximum motor torque rating are susceptible to be lower than the design efficiencies. For the two actuators, threshold efficiencies were calculated and provided to replace their design efficiencies.     

Noise analysis experience of a BWR power plant

This paper summarizes activities and experiences concerning noise measurement and analysis in the BWR power plant SHIMANE I since its beginning of operation, of the Hitachi Atomic Energy Research Laboratory (HAERL) in collaboration with Chugoku Electric Power Company. The possibility of using noise analysis for safety monitoring, detecting abnormalities at an incipient stage and further, diagnosing the abnormal condition is discussed. The discussion on noise sources in reactor power fluctuation during normal operation is also briefly summarized.     

Cavitation noise from butterfly valves

Cavitation in valves can produce levels of intense noise. It is possible to mathematically express a limit for a design level of cavitation noise in terms of the cavitation parameter σ. Using the cavitation parameter or limit, it is then possible to calculate the flow conditions at which a design level of cavitation noise will occur. However, the intensity of cavitation increases with the upstream pressure and valve size at a constant σ. Therefore, it is necessary to derive equations to correct or scale the cavitation limit for the effects of different upstream pressures and valve sizes.The following paper discusses and presents experimental data for the cavitation noise limit as well as the cavitation limits of incipient, critical, incipient damage, and choking cavitation for butterfly valves. The main emphasis is on the design limit of cavitation noise, and a noise level of 85 decibels was selected as the noise limit. Tables of data and scaling exponents are included for applying the design limits for the effects of upstream pressure and valve size.     

Dependent failure analysis of NPP data bases

A technical approach for analyzing plant-specific data bases for vulnerabilities to dependent failures has been developed and applied. Since the focus of this work is to aid in the formulation of defenses to dependent failures, rather than to quantify dependent failure probabilities, the approach of this analysis is critically different. For instance, the determination of component failure dependencies has been based upon identical failure mechanisms related to component piecepart failures, rather than failure modes. Also, component failures involving all types of component function loss (e.g., catastrophic, degraded, incipient) are equally important to the predictive purposes of dependent failure defense development. Consequently, dependent component failures are identified with a different dependent failure definition which uses a component failure mechanism categorization scheme in this study. In this context, clusters of component failures which satisfy the revised dependent failure definition are termed common failure mechanism (CFM) events.Motor-operated valves (MOVs) in two nuclear power plant data bases have been analyzed with this approach. The analysis results include seven different failure mechanism categories; identified potential CFM events; an assessment of the risk-significance of the potential CFM events using existing probabilistic risk assessments (PRAs); and postulated defenses to the identified potential CFM events.     

次声波图数字化系统

介绍一套次声波形图数字化系统。该系统以图像处理技术为基础提供次声波形图扫描,倾斜矫正,图像剪切,去除水平线和垂直线,波形跟踪等功能。其关键技术包括小波分析、图像细化和像素跟踪。该系统使用户可以用计算机处理分析记录在图纸上的次声波形,使次声研究更加快捷方便,为大气核爆炸监测,爆炸源定位和当量计算,地震预测预报等研究提供一种先进的次声数据分析和管理方法。     

Fault diagnosis and classification based on wavelet transform and neural network

In this paper we present a novel method in fault recognition and classification in Nuclear Power Plant (NPP) using wavelet transform based Artificial Neural Network (ANN). We first simulate 10 design basis accidents (DBA) of a VVER-1000 using 15 input parameters with employing a Multilayer Perceptron (MLP) Neural Network with Resilient Backpropagation (RBP) algorithm. Afterwards we present the application of wavelet transform for its temporal shift property and multiresolution analysis characteristics to reduce disturbing perturbations in input training set data. Simulation of Artificial Neural Network and wavelet transform was performed using MATLAB software. The results show an enhanced accuracy and speed in fault recognition and high degree of robustness.     

Evaluation of check valve monitoring methods

Check valves are used extensively in nuclear plant safety systems and balance-of-plant (BOP) systems. Their failures have resulted in significant maintenance efforts and, on occasion, have resulted in water hammer, overpressurization of low-pressure systems and damage to flow system components. Consequently, in recent years check valves have received considerable attention by the Nuclear Regulatory Commission (NRC) and the nuclear power industry. Oak Ridge National Laboratory (ORNL) is carrying out a comprehensive two phase aging assessment of check valves in support of the Nuclear Plant Aging Research (NPAR) program. As part of the second phase, ORNL is evaluating several developmental and/or commercially available check valve diagnostic monitoring methods; in particular, those based on measurements of acoustic emission, ultrasonics, and magnetic flux. These three methods were found to provide different (and complementary) diagnostic information. The combination of acoustic emission with either ultrasonic or magnetic flux monitoring yields a monitoring system that succeeds in providing sensitivity to detect all major check valve operating conditions. The three check valve monitoring methods described in this paper are still under development and are presently being tested as part of a program directed by the Nuclear Industry Check Valve Group (NIC) in conjunction with the Electric Power Research Institute (EPRI). Phase 1 of this program (water testing) is being carried out at the Utah Water Research Laboratory located on the Utah State University campus.     

Improved plasma position detection method in EAST Tokamak using fast CCD camera

To control the steady-state operation of Tokamak plasma, it is crucial to accurately obtain its shape and position. This paper presents a method for use in rapidly detecting plasma configuration during discharge of the Experimental Advanced Superconducting Tokamak device. First, a visible/infrared integrated endoscopy diagnostic system with a large field of view is introduced,and the PCO.edge5.5 camera in this system is used to acquire a plasma discharge image. Based on the analysis of various traditional edge detection algorithms, an improved wavelet edge detection algorithm is then introduced to identify the edge of the plasma. In this method, the local maximum of the modulus of wavelet transform is searched along four gradient directions, and the adaptive threshold is adopted. Finally, the detected boundary is fitted using the least square iterative method to accurately obtain the position of the plasma. Experimental results obtained using the EAST device show that the method presented in this paper can realize expected goals and produce ideal effects;this method thus has significant potential for application in further feedback control of plasma.     

Design options for safety depressurization system

The Korean Next Generation Reactor (KNGR) adopted an advanced design feature, a safety depressurization system (SDS) to rapidly depressurize the primary system in case of events beyond the design basis. Two design approaches are considered for the SDS design. The use of bleed valves similar to the ABB-CE System 80+ is design option 1, while in design option 2, the French Sebim valve is considered to provide the combined function of overpressure protection and rapid depressurization. In this paper, thermal hydraulic analysis using a best-estimate version of CEFLASH-4AS/REM is performed for a total loss of feedwater (TLOFW) event to investigate the feasibility of those two design options. For each design option, various feed and bleed (F and B) procedures are investigated for a TLOFW event. For design option 1, the required bleed capacity is determined from the CEFLASH-4AS/REM simulation according to the EPRI Advanced Light Water Reactor (ALWR) requirements. The analysis results demonstrate that the TLOFW event can be mitigated in a proper manner with a sufficient margin using design option 1. For design option 2, the operator action times for initiating the F and B are investigated by varying the number of Sebim valves and high pressure safety injection (HPSI) pumps. If the operator opens two out of the three Sebim valves in conjunction with the four HPSI pumps before a hot leg saturation condition, the decay heat removal and core inventory make-up function can be successfully accomplished. The results of the present investigation demonstrate that the two design options are both feasible.     

Measuring and analysis methods applied to on-line vibration and noise monitoring in PWR power plants

Measurements and analyses performed in a PWR plant are discussed in the light of their relevance for application to an on-line monitoring system. This is achieved by extensive investigation during the preoperational tests. The results gained are transferred to the situation existing during full power operation by means of spectral analysis and correlation techniques. Theoretical models provide further proof of the results. The final conclusions show the consequences for establishing an on-line monitoring system.     

Neutron noise measurements on pressurized water reactors

Noise analysis contributes to increase significantly understanding of safety and monitoring of PWR. The difficulties of a correct interpretation of noise signal in a power reactor encourage a deeper insight into the theoretical model. Following, this paper is dealing with 4 topics:

• - theoretical model

• - measurements made in PWR's, and evolution of power spectral densities,

• - experimental test of models,

• - future study in this range : methodology for early detection of failure and to indicate incipient failure.

Using the expressions for linear system and for feedback loop, we obtain a simplified diagram for PWR with reactivity, temperature, velocity of fluid inputs and movement of internal structures.

Neutron noise measurements are performed periodically on the “Centrale Nucléaire des Ardennes”. An investigation is also performed, in order to detect core barrel movements.

The change of neutron noise at several power levels is shown, and used to check the model.

The development of signal analysis and models for PWR is investigated in the last chapter.     

Vibration- and pressure signals as sources of information for an on-line vibration monitoring system in PWR power plants

Higher demands with regard to the safety and reliability of reactor primary components require methods to get an idea of the mechanical state of the plant at any time during operation and to recognize failures already in their developing phase. Reactor vibration monitoring systems are being developed which are based on the analysis of vibration signals, neutron noise and pressure fluctuation signals. The special role vibration and pressure signals can play in such a system is investigated by the analysis of extensive preoperational tests at different PWRs. The theoretical foundation for the application of these signals to vibration monitoring are developed in the special case of the Stade nuclear power plant. The pressure vessel of this reactor performs pendular and verical vibrations. They are excited mainly by pressure fluctuations generated by the coolant flow, by standing waves, or by the revolution of the coolant pumps.

For interpreting the spectra measured during the preoperational test and during power operation and for clearing up changes of these spectra, which will signalise incipient failures, model investigations are of predominant importance. Two mechanical models, a pendular and a vertical one, simulate the two kinds of vibration sufficiently which can be seen in comparing the calculated frequency response with the measured vibrations.     

基于平稳小波变换的毛刺检测算法

核爆炸地震监测技术研究中,数据质量检测是地震数据自动处理的基本内容,毛刺是影响数据质量的主要问题数据。基于平稳小波变换和非线性能量检测算法,给出一种毛刺自动检测算法。平稳小波变换弥补了正交小波变换存在的不足,可以使尺度分解结果的长度和原始数据保持一致,具备时移不变性。非线性能量检测算法可以增强记录中的高频信号,对平稳小波变换的结果应用非线性能量检测算法,提高了记录中毛刺检测的准确性,非常适合连续地震监测数据自动处理的需要。实验结果表明,给出的这种算法特别有利于记录中小毛刺的检测,从而能够减小信号检测的误检率。     

Analysis of accidents leading to local flow stagnation in parallel fuel channels of RBMK-type reactors

The RBMK-type nuclear power reactors, still operating in Russia, are graphite-moderated with vertical fuel channels, using low-enriched nuclear fuel. The main challenge, which leads to the overheating of the fuel assemblies, fuel channels and other core components in channel type nuclear reactors, is a misbalance between heat generation in core structures and heat sink, which can appear due to the loss of coolant accident. In this accidental case, the emergency core cooling system ensures the core cooling. In RBMK-type reactors this system consists of hydro-accumulators and a number of pumps, taking water from large water reservoirs. This equipment injects water into the fuel channels through the group distribution headers at high pressure. However, the direct supply of cold water from emergency core cooling system into fuel channels is possible only if check valves on group distribution headers are closed properly. If these check valves failed, the part of water would be lost through the break, the flow stagnation in channels could occur, which might lead to overheating of fuel assemblies in the fuel channels. This paper presents the results of deterministic safety analysis, performed using system thermal hydraulic code RELAP5. Using this code the reactor cooling system of RBMK-1500 was modelled and analyses of loss of coolant accidents with failure of few check valves in group distribution headers were performed. The results of the calculations are used for the development of symptom-based emergency operating procedures for RBMK-1500. The basic principles that describe the complex distribution of water flows in vertical forced circulation circuit with parallel fuel channels can be adjusted for the RBMK-1000 reactors, still operating in Russia.     

基于Hilbert-Huang变换和BP神经网络的核级电动阀门退化趋势预测

核级电动阀门服役环境恶劣,极易发生退化失效。为准确预测核级电动阀门性能退化趋势,采用Hilbert-Huang变换（HHT）和反向传播神经网络（BPNN）相结合的方法（HHT-BPNN）对核级电动阀门的退化状态进行预测。本文采用某次核级电动阀门可靠性试验的振动信号对电动阀门退化趋势进行预测,结果显示该方法能准确预测出核级电动阀门的3种退化状态,且其相对误差在可接受范围内。研究表明HHT能够有效提取信号的退化信息,BPNN能够准确预测核级电动阀门的退化趋势,HHT-BPNN预测方法能有效解决核级电动阀门性能退化预测困难的问题。      

A study on the method of impact mass estimation of loose parts

It is a very difficult problem to realize the mass estimation of loose parts in the mechanical equipment. The result of mass estimation will influence the fault diagnosing of equipment, especially in the loose part monitoring system of nuclear power station which can provide important guidance for the type classification of loose parts. This paper is based on experiments, by wavelet energy spectrum method to make estimation for different impact mass, and by using linear interpolation method to establish the scale peak function. The results show that the method has characteristics of small estimation errors and good consistency, strong anti-interference capacity, and it has better actual application value.     

A study on the phenomenon of rate of loading in motor operated gate valves

This study analyzed the rate of loading (ROL) phenomenon, which is generated during the operation of a motor operated valve (MOV) under fluid pressure conditions. ROL is one of the most important parameters for an MOV performance evaluation. This paper includes the analysis results for the characteristics of ROL and the effect of fluid pressure on the ROL. Dynamic and static test were performed to analyze the ROL effect for flexible wedge gate valve. The result of this analysis confirmed that the ROL is generated under fluid pressure condition and that the ROL value under high differential pressure condition appeared to be higher than under low differential pressure condition. According to the test results of multiple valves, the ROL appeared to become higher, as the differential pressure increased, and under the high differential pressure condition, it accounted for approximately 17.6% of the thrust loss. In addition, the ROL effect was negligible in valves with a low differential pressure (below 1100 kPa).     

Evaluation of intentional depressurization strategy in Chinese 600 MWe PWR NPP

Intentional depressurization is one of the effective strategies in preventing high-pressure melt ejection (HPME) and direct containment heating (DCH), which is most feasible for the operating nuclear power plants (NPPs) in China. In order to evaluate this strategy of a Chinese 600 MWe PWR NPP, the plant model is built using SCDAP/RELAP5 code. ATWS, SBO, SGTR and SLOCA are selected as the base cases for analysis of intentional depressurization. The results show that opening safety valves of pressurizer manually when the core exit temperature exceeds 922 K can reduce the RCS pressure effectively and prevent the occurrence of HPME and DCH. Several uncertainties such as the operability of safety valves, ex-vessel failure and the transitory rise of RCS pressure are also analyzed subsequently. The results show that the opening of the safety valves can be initiated normally and that opening three safety valves is a more favorable strategy in the event of possible failure of one or more of the safety valves; the probability of ex-vessel failure is reduced after intentional depressurization is implemented; the transitory rising of reactor coolant system (RCS) pressure when the molten core materials relocate to the lower head of reactor pressure vessel (RPV) will not influence the effect of depressurization.