Development of diagnosis algorithm for the check valve with spectral estimations and neural network models using acoustic signals

We have developed a method for detecting and diagnosing a disk wear failure and a foreign object failure among the various failure modes of check valves. The method is based on the acoustic emission sensors which can detect the sound wave of the leakage flow and the estimation of the power spectral densities with an auto-regressive model. For validating the method, we implemented a hydraulic test loop with an artificially failed check valve. We have found that the frequency spectrums from the acoustic signals are strongly dependent on the failure modes of the check valve and that they are nearly independent of the failure size and operating pressure through an estimation of the power spectral density with an auto-regressive signal processing model. In addition, the root mean square values of the acoustic signal and the amplitudes of the power spectral density as well as the loop pressure have a strong dependency on the failure size in each failure mode of the check valve. We developed a diagnosis algorithm by using neural network models in order to identify the type and size of the failure in the check valve. The diagnosis algorithm consists of a hierarchical model composed of three back-propagation neural networks. The results of our research and the experiments show that the diagnosis algorithm is proven to be a good solution for identifying the failures of the check valves without any disassembling work.     

Degradation mechanism of check valves in nuclear power plants

A large amount of deposit was detected at a 6-in. check valve installed in a nuclear power plant. To identify the origin of the deposit, the chemical composition of the deposit was analyzed. In this paper, in order to find out the degradation mechanism of the 6-in. check valve, vibration and temperature monitoring of the check valve are carried out during plant heat up and standby conditions. And some degradation mechanisms of the check valve are investigated. Results show that the degradation was caused by valve chattering due to the acoustic resonance.     

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.     

Early results of gate valve flow interruption blowdown tests

This paper presents the preliminary results of the USNRC/INEL** high-energy BWR line break flow interruption testing. Two representative nuclear valve assemblies were cycled under design basis vector water cleanup pipe break conditions to provide input for the technical basis for resolving the Nuclear Regulatory Commission's Generic Issue 87. The effects of the blowdown hydraulic loadings on valve operability, especially valve closure stem force, were studied. The blowdown tests showed that, given enough thrust, typical gate valves will close against the high flow resulting from a line break. The tests also showed that proper operator sizing depends on the correct identification of values for the sizing equation. Evidence exists that values used in the past may not be conservative for all valve applications. The tests showed that improper operator lock ring installation following test or maintenance can invalidate in-situ test results and prevent the valve from performing its design function.     

Safety Relief Valve Performance for Two-Phase Flow

The performance of main steam safety relief valve has been evaluated with respect only to the steam. In the present study, two-phase flow and subcooled water blow-out tests with model valves were performed in order to evaluate the valve's characteristics and performance. From the test results, it was made clear that not only for the steam but also for the two-phase flow the measurement data were hardly affected by scaling and also that the reaction force of the fluid to the valve stem was hardly dependent upon the void fraction. Analytical study was performed using the two-phase flow model in the valve. The results of the analysis showed good agreement with the test data. It was shown from the test and analysis results that the reaction force of the two-phase flow and subcooled water to the valve stem was almost as much as that of the steam flow, and the integrity of the safety relief valve could be maintained.     

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.     

大亚湾核电站闸阀锅炉效应状况及其改进

基于国际上相关的闸阀锅炉效应的试验和研究结果,利用甄别方法及概率安全分析(PSA)方法对大亚湾/岭澳核电站每个机组的闸阀锅炉效应进行了分析.结果表明,大亚湾/岭澳核电站每个机组的闸阀RIS063/064VP及PTR022VB需要改进.为了防止阀腔超压,改进的措施分别为:RIS063/064VP增加带有双向逆止阀的的旁路,PTR022VB则在高压侧的闸板上钻孔.     

大型非能动压水堆核电厂安全注射系统旋启式止回阀阻力变化对长期堆芯冷却影响分析

大型非能动压水堆核电厂在发生失水事故(LOCA)后的长期堆芯冷却阶段依靠重力向堆芯注入应急冷却水,其注射管线上设置的旋启式止回阀的阻力可随流量变化,管线的阻力可能将非预期地增加。根据旋启式止回阀阻力特性,为失水事故最佳估算系统分析程序添加相应的计算功能,对压力容器直接注射(DVI)管线双端断裂事故后长期堆芯冷却工况进行了计算分析。结果表明:安全注射管线上旋启式止回阀阻力变化对大型非能动压水堆核电厂LOCA后长期冷却的影响较小;在安全裕量不足的情况下,旋启式止回阀的阻力特性将影响到非能动注射管线的安全注射功能的执行。     

核电厂一回路压力边界止回阀在线密封性能测试

320 MW压水堆一回路压力边界止回阀为核Ⅰ级关键设备,严密性要求非常高,直接关系到主系统的内泄漏率.焊接式止回阀维修后常采用密封面色印检查的方式,对其密封性能进行判断.如果管道内有存水或者湿热水汽,会影响到色印检查的准确度.针对在线止回阀密封性试验的特殊性,有的核电厂采用水压压降法试验设计过在线检测装置,但存在一些缺点和使用上的限制.文章采用低压气密封试验流量测定法,设计出可靠、便携的试验装置,对压力边界止回阀检修后密封性做出准确、定量的判断.     

高频地震激励对核级阀门危害性分析

利用试验和修正后的集中质量有限元模型预测安装在管道中阀门在不同频率成分地震激励下的响应,研究高频地震激励对管道中质量较大核级阀门的危害性。研究结果表明：高频地震激励对核级阀门的危害在于使阀门以其自身固有振型发生共振,此时阀门顶部取代阀门与管道连接位置成为阀门中响应最大的位置,这会导致安装于阀门顶端的驱动机构遭受苛刻的地震工况。增加管道阻尼和阀门刚度能有效降低高频激励对阀门的危害,但增加阀门刚度会导致管道响应增大。利用等效静力法对阀门进行抗震鉴定时,分析结果对阀体水平部位内力估计不足,对阀体垂直部分、阀盖等阀门上部构件的内力估计结果具有较大裕度。     

Hazard Analysis of Nuclear Safety Class Valve Induced by High Frequency Seismic Excitation

The test and the updated lumped mass finite element model were used to predict the response of the valve installed in the pipeline under the seismic excitation of different frequency components, and the hazard of high frequency seismic excitation to large-mass nuclear safety class valves in the pipeline was studied. The results show that the high frequency seismic excitation causes the nuclear safety class valve to resonate with its own mode of vibration. At this moment, the top of the valve replaces the position where the valve is connected to the pipe to become the position with the largest response amplitude in the valve, which causes the drive mechanism installed on the top of the valve to suffer severe seismic conditions. Increasing the pipe damping and valve stiffness can effectively reduce the hazard of high frequency excitation to the valves, but increasing the valve stiffness will lead to the increase of the pipe response amplitude. When the equivalent static method is used for seismic identification of the valve, the analysis result is insufficient to estimate the internal force of the horizontal part of the valve body, and has a large margin to estimate the internal force of the vertical part of the valve body, the valve cover and other upper parts of the valve.     

船用核动力装置止回阀的流固热耦合研究

针对船用核动力装置中止回阀的泄漏问题，利用流固热耦合仿真方法研究了温度快速变化对止回阀的影响，结果表明:止回阀的等效应力和变形量随温度的降低而降低；密封压垫和四合环最大等效应力位于阀门管道两端，最大变形量位于阀门前后部位；密封压垫的最大变形量和收缩率都比四合环大；由于高温高压的作用，密封压垫与阀盖之间产生了明显的间隙，易发生泄漏，且该间隙随温度的降低而扩大，可能加剧泄漏。      

Analysis of potential waterhammer at the Ignalina NPP using thermal–hydraulic and structural analysis codes

The RBMK (Russian acronym for ‘channeled large power reactor’)-1500 reactors at the Ignalina nuclear power plant (NPP) have a series of check valves in the main circulation circuit (MCC) that serve the coolant distribution in the fuel channels. In the case of a hypothetical guillotine break of pipelines upstream of the group distribution headers (GDH), the check valves and adjusted piping integrity is a key issue for the reactor safety during the rapid closure of check valve. An analysis of the waterhammer effect (i.e. the pressure pulse generated by the valves slamming closed) is needed. The thermal–hydraulic and structural analysis of waterhammer effects following the guillotine break of pipelines at the Ignalina NPP with RBMK-1500 reactors was conducted by employing the RELAP5 and PipePlus codes. Results of the analysis demonstrated that the maximum values of the pressure pulses generated by the check valve closure following the hypothetical accidents remain far below the value of pressure of the hydraulic tests, which are performed at the NPP and the risk of failure of the check valves or associated pipelines is low. Sensitivity analysis of pressure pulse dependencies on calculation time step and check valve closure time was performed. Results of RELAP5 calculations are benchmarked against waterhammer transient data obtained by employing structural mechanics code BOS fluids.     

A study of the characteristics of the acoustic emission signals for condition monitoring of check valves in nuclear power plants

The objective of this study is to demonstrate that a condition-monitoring system based on acoustic emission (AE) detection can provide timely detection of check valve degradation and service aging so that maintenance or replacement can be preformed prior to the loss of safety function. This research is focused on the investigation and understanding of the capability of the acoustic emission technique to provide diagnostic information on check valve failures.AE testing for a check valve under controlled flow loop conditions was performed to detect and valve degradation such as wear and leakage due to foreign object interference. It is clearly demonstrated that the distinction of different types of failure were successful by systematically analyzing the characteristics of various AE parameters.     

摇摆条件下典型通道间湍流的流动传热特性

利用Fluent软件分析了摇摆条件对典型四棒束间的湍流流体流动和传热特性的影响机理。摇摆运动会对棒束间流体的流动传热特性产生一定影响,但不会对绝热通道与加热通道内流体流动相似性产生影响。而当摇摆幅度较大时,径向附加力会使通道横截面上的参数分布发生显著的变化,进而影响流体的流动与传热特性。在摇摆条件下,随着P/D（棒间距/棒直径）的逐渐减小,尤其是小于1.1时,典型棒束间流体的流动传热特性发生明显变化。     

An analysis of check valve performance characteristics based on valve design

Although it had been theorized by nuclear industry valve experts that the two most significant factors in assessing check valve performance were valve type (or design) and operating conditions, until recently, no data was available to support their assumptions. In co-operation with the Nuclear Industry Check Valve Group (NIC), Oak Ridge National Laboratory (ORNL) undertook a review and analysis of check valve failures recorded in the Institute of Nuclear Power Operations’ (INPO) Nuclear Plant Reliability Data System (NPRDS). This study involved the characterization of failures according to several parameters, including valve design (e.g. swing check, lift check). Since the valve design is not inherently included within the NPRDS engineering record for each component in the database, ORNL relied on input from NIC, valve manufacturers and catalogs to supply the missing information. As a result, nearly 60% of the 21 000 check valves listed in the NPRDS component database and 85% of the 838 failures occurring during 1991–1992 were identified according to valve design. This data provided the basis to perform previously unavailable cross-correlations between parameters such as valve design versus failure mode, valve design versus failure discovery method, population/failure distributions by valve design, etc. Performance assessments and predictions based on more specific sets of parameters (as opposed to generic check valve failure rates obtained from standard reference sources that generally ignore the valve design) should result in a significant impact on future nuclear plant operations, including inservice testing (IST) practices, maintenance, and probabilistic risk assessments (PRAs) by providing a means to calculate more appropriate relative (and ultimately absolute) failure rates for check valves.     

Engineering analysis of mass flow rate for turbine system control and design

The mass flow rate is determined in the steam turbine system by the area formed between the stem disk and the seat of the control valve. For precise control the steam mass flow rate should be known given the stem lift. However, since the thermal hydraulic characteristics of steam coming from the generator or boiler are changed going through each device, it is hard to accurately predict the steam mass flow rate. Thus, to precisely determine the steam mass flow rate, a methodology and theory are developed in designing the turbine system manufactured for the nuclear and fossil power plants. From the steam generator or boiler to the first bunch of turbine blades, the steam passes by a stop valve, a control valve and the first nozzle, each of which is connected with piping. The corresponding steam mass flow rate can ultimately be computed if the thermal and hydraulic conditions are defined at the stop valve, control valve and pipes. The steam properties at the inlet of each device are changed at its outlet due to geometry. The Compressed Adiabatic Massflow Analysis (CAMA) computer code is written to predict the steam mass flow rate through valves. The Valve Engineered Layout Operation (VELO) test device is built to experimentally study the flow characteristics of steam flowing inside the control valve with the CAMA input data. The Widows’ Creek type control valve was selected as reference. CAMA is expected to be commercially utilized to accurately design and operate the turbine system for fossil as well as nuclear power plants.     

Large eddy simulation of turbulent flow surrounding two simulated CANDU fuel bundles

Three-dimensional turbulent flow in one and a half simulated 43-element CANDU^®1 fuel bundle at the inlet of a fuel channel is solved using large eddy simulation. Wake generated after the endplates and the flow development in the inlet bundle are investigated based on the simulation results. Spatial distribution and frequency spectra of fluid force components are also examined. The simulation results are compared to experimental data available in the literature as well as the measurements conducted by the authors. The current investigation provides a basic understanding of the fluid excitation in a simulated 43-element fuel bundle. The results may be used in a flow-induced vibration analysis for fuel bundles.     

A study on a characteristic of stem friction coefficient for motor operated flexible wedge gate valve

Stem friction coefficient is a coefficient that represents friction between thread leads of the stem and stem nut. It is an important factor to determine output thrust delivered from the actuator to the valve stem in assessing performance of motor operated valves. This study analyzes the effects of changes in differential pressure on stem friction coefficient, and determines the bounding value of stem friction coefficient. A dynamic test was conducted on multiple flexible wedge gate valves in various differential pressure conditions, and the test data was statistically analyzed to determine the bounding value. The results show that stem friction coefficient in middle and high differential pressure is influenced by fluid pressure, while stem friction coefficient in low differential pressure is almost not affected by fluid pressure. In addition, it is found that the bounding value of stem friction coefficient is higher in a closing stroke than in an opening stroke.     

Mitigating check valve slamming and subsequentwater hammer events for PPFS using MOC

The method of characteristic (MOC) was adopted to analyze the check valve-induced water hammer behaviors for a Parallel Pumps Feedwater System (PPFS) during the alternate startup process. The motion of check valve disc was simulated using inertial valve model. Transient parameters including the pressure oscillation, local flow velocity and slamming of the check valve disc etc. have been obtained. The results showed that severe slamming between the valve disc and valve seat occurred during the alternate startup of parallel pumps. The induced maximum pressure vibration amplitude is up to 5.0 MPa. The scheme of appending a damping torque to slow down the check valve closing speed was also performed to mitigate of water hammer. It has been numerically approved to be an effective approach.