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.     

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.     

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.     

Progress for on-line acoustic emission monitoring of cracks in reactor systems

This paper reviews accomplishments and planned tasks for the NRC-sponsored research program concerned with “Acoustic Emission/Flaw Relationships for Inservice Monitoring of Nuclear Reactor Pressure Boundaries”. The objective of the acoustic emission (AE) monitoring program is to develop and validate the use of AE methods for continuous surveillance of reactor pressure boundaries to detect flaw growth. Topics discussed include testing AE monitoring on reactors, refinement of an AE signal identification relationship, study of slow crack growth rate effects on AE generation, and activity to produce an ASTM standard for AE monitoring and to gain ASME code acceptance of AE monitoring.     

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.     

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.     

Wear in check valves due to flow induced motion

Experiments showing the frequency and amplitude of the flow induced motion of the gate for a 2- and a 4-in. swing check valve have been performed. The gate motion is due to turbulence in approach flow. We have found the dominant turbulent frequency of the approach flow is about half the natural frequency of the valves. The valves appear to be almost critically damped. Because of this, the valves respond almost as they would to a static force of the magnitude characteristic of the turbulent fluctuation in the flow. Both the dimensionless exciting force and the damping ratio have been found to be independent of valve size so the above statements are true for larger valves also. The recommended valve oscillation amplitudes and frequencies are used to calculate the wear at the shaft and at the stop. For an unpegged check valve, such as one of the 10-in. valves which was used at the San Onofre Nuclear Generation Station, it was found that shaft bearing wear would amount to 0.27 in.³/year and stop wear to 0.03 in.³/year.     

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。      

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.     

Evaluation of nuclear facility decommissioning projects program-status

Major studies have been undertaken in recent years by the US Nuclear Regulatory Commission (NRC) and others on the technology, safety and costs associated with decommissioning nuclear facilities. The program described in this presentation is being undertaken by the NRC to compile and evaluate the activities of ongoing decommissioning projects. Assessment and evaluation of the methods, impacts, and costs will provide a basis for evaluating licensee's decommissioning proposals and for future decommissioning direction and regulation.Program participants include the US Nuclear Regulatory Commission (NRC) through the Office of Regulatory Research, UNC Nuclear Industries (UNC) through the Decommissioning Programs Department, and nuclear facility licensees.     

Analytically predicted versus measured response of a free-standing steel containment vessel subjected to safety-relief valve discharge loads

Following the actuation of safety-relief valves in BWR nuclear power plants, first water then air and steam are cleared from the discharge lines through quencher devices into a suppression pool. This clearing results in water spike, air bubble, and condensation pressure loads applied to structures in the pool, and the surrounding containment vessel.The Leibstadt Nuclear Power Plant has the only free-standing steel Mark III containment vessel in the world. All other steel Mark III containment vessels have concrete backing in the suppression pool region, which dampens clearing load responses. As such, it is of interest to note how this steel vessel responds to discharge pressures, and compare these responses to analytically predicted results.The purpose of this paper is to compare the analytical results used to design the steel containment vessel with the responses measured during in-plant testing. The analytical methods considered the effects of fluid-structure interaction. The test program included initial and consecutive actuations of a single valve, and initial actuation of multiple (four) valves. The conclusion of the comparison is that, in general, there are large conservatisms in the analytical predictions versus measured responses.     

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。     

核环境科学40年发展回顾

概述了核环境科学40年的发展历程。介绍了中国辐射防护研究院与中国原子能研究院等单位在核工业创建初期和二次创业过程中合作开展的一些现场调查、厂矿周围地区的污染物监测、生态转移参数测定、大气和地表水野外实验和环境影响评价等工作，以及在非核领域开展的一些大气和地表水、地下水现场实验和环境影响评价工作。1964年－2001年共完成大型实验83项（其中非核领域项目为24项），环境影响评价329项（其中非核领域项目为218项）。最后，评述了中国辐射防护研究院低水平环境放射性测量实验室以及核工业重点实验室－－核辐射环境模拟技术实验室的建设情况和特点，后者包括环境风洞（大气边界层实验室）、地表水污染物迁移实验室、包气带和地下水污染物迁移实验室、核素迁移研究野外实验场、核素在含水层中迁移实验的地下研究设施（URF）。     

核一级低压差旋启式止回阀研制

为了实现核电站关键阀门的国产化,研制了核一级低压差旋启式止回阀。本文介绍了该阀门的技术参数及其研制过程和型式试验的情况。经过各种测试,各种性能指标均达到设计要求。     

Performance of a piping system and equipment in the HDR simulated seismic experiments

This paper describes a portion of the analysis and results of the United States Nuclear Regulatory Commission/Idaho National Engineering Laboratory (USNRC/INEL) participation in the SHAG (Shakergebaude) Seismic Research Program conducted by Kernforschungszentrum Karlsruhe (KfK) at the Heissdampfreaktor (HDR), a decommissioned nuclear reactor. The program analyzed the responses of a piping system and associated line-mounted equipment when subjected to various seismic and hydraulic loadings. Of interest was to evaluate the influence that piping support system flexibility has on piping system responses. The results of the studies will contribute to the technical basis for assessing the responses of light water reactor (LWR) piping and fine-mounted equipment to earthquakes.     

Dynamic load effects on gate valve operability and snubber performance

The Idaho National Engineering Laboratory (INEL) participated in an internationally sponsored seismic research program conducted at the decommissioned Heissdampfreaktor (HDR) located in the Federal Republic of Germany. An existing piping system was modified by installation of 200-mm, naturally aged, motor-operated gate valve from a U.S. nuclear power plant and a piping support system of U.S. design. Using various combinations of snubbers and other supports, six other piping support systems of varying flexibility from stiff to flexible were also installed and tested. Additional valve loadings included internal hydraulic loads and, during one block of tests, elevated temperature. The operability and integrity of the aged gate valve and the dynamic response of the various piping support systems were measured during 25 representative simulations of seismic events.     

Development of Systematic Models for Aerosol Agglomeration and Spray Removal under Severe Accident Conditions

Radionuclide behavior during various severe accident conditions has been addressed as one of the important issues to discuss environmental safety in nuclear power plants. The present paper deals with the development of analytical models and their validations for the agglomeration of multiple-component aerosol and spray removal that controls source terms to the environment of both aerosols and gaseous radionuclides during recirculation mode operation in a containment system for a light water reactor. As for aerosol agglomeration, the single collision kernel model that can cover all types of two-body collision of aerosol was developed. In addition, the dynamic model that can treat aerosol and vapor transfer leading to the equilibrium condition under the containment spray operation was developed. The validations of the present models for multiple-component aerosol growth by agglomeration were performed by comparisons with Nuclear Safety Pilot Plant (NSPP) experiments at Oak Ridge National Laboratory (ORNL) and AB experiments at Hanford Engineering National Laboratory (HEDL). In addition, the spray removal models were applied to the analysis of containment spray experiment (CSE) at HEDL. The results calculated by the models showed good agreements with experimental results.     

Applicability of AE monitoring technique to evaluate structural integrity of fast breeder reactor piping component

The availability of acoustic emission technique for detecting and monitoring defects of piping components in fast breeder reactors has been studied in the FAET Committee of the Japan Welding Engineering Society sponsored by,the Power Reactor & Nuclear Fuel Development Corporation. The researches which extended over four years have covered a wide range of experiments and evaluations in order to ensure the usefulness of the acoustic emission technique as a monitoring tool of plant operation.