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.     

A study on the stem friction coefficient behavior of motor-operated valves

This paper describes how to calculate the stem friction coefficient of safety related motor operated valves (MOVs) that reflects potential degradation with time by using diagnostic signals acquired in static field tests that have been conducted more than two times per valve. Based on the calculated stem friction coefficients, their behaviors with time were analyzed considering various parameters that could cause potential degradation. Most friction coefficients change randomly rather than increasing or decreasing continuously over time. From those trends, a threshold coefficient, which represents the highest expected value of the friction coefficient, was calculated and provided.     

Transient dynamic and inelastic analysis of shells of revolution — a survey of programs

Advances in the limits of structural use in the areospace and nuclear power industries over the past years have increased the requirements upon the applicable analytical computer programs to include accurate capabilities for inelastic and transient dynamic analyses. In many minds, however, this advanced capability is unequivocally linked with the large scale, general purpose, finite element programs. This idea is also combined with the view that such analyses are therefore prohibitively expensive and should be relegated to the “last resort” classification. While this, in the general sense, may indeed be the case, if the user needs only to analyze structures falling into limited categories, however, he may find that a variety of smaller special purpose programs are available which do not put an undue strain upon his resources. One such structural category is shells of revolution.This survey of programs will concentrate upon the analytical tools which have been developed predominantly for shells of revolution. The survey will be subdivided into three parts: (a) consideration of programs for transient dynamic analysis; (b) consideration of programs for inelastic analysis and finally; (c) consideration of programs capable of dynamic plasticity analysis. In each part, programs based upon finite difference, finite element, and numerical integration methods will be considered. The programs will be compared on the basis of analytical capabilities, and ease of idealization and use. In each part of the survey sample problems will be utilized to exemplify the state-of-the-art.     

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.     

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).     

Incipient fault detection of motor-operated valves using wavelet transform analysis

The necessity of improvements in monitoring and diagnosis methods started to be of extreme relevance in the predictive maintenance field, establishing the reliability and readiness of system components as an achievable goal. Taking into account these reasons, this paper presents an approach for incipient fault detection of motor-operated valves (MOVs) using wavelet transforms. The technique applied in this paper is the wavelet transform analysis using wavelet toolbox, where the main goal is to obtain more detailed information contained in the measured data, identifying and characterizing the transient phenomena in the time and frequency domains, correlating them to failure situations in the incipient stage. The wavelet analysis has provided good results establishing a new qualitative methodology for monitoring and diagnostics of motor-operated valves.     

Leak-Before-Break analysis for Pickering ‘A’ Unit 1 and Unit 4 large diameter main steam line pipes

This paper presents a Leak-Before-Break (LBB) analysis of large diameter main steam line pipes (i.e. NPS 28″ and 30″) running from reactor building to main steam balance header in Pickering nuclear plant Unit 1 and Unit 4. Recent development in LBB technology summarized in U.S. Nuclear Regular Commission report NUREG/CR-6765 was adopted. Based on the tiered approach of LBB philosophy, this LBB analysis belongs to level 2 or level 3 LBB analysis. Detailed fracture tolerance analyses and leakage rate calculations were performed. EPFM (elastic plastic fracture mechanics) theory of J-integral, resistance curve versus ductile crack extension was adopted in carrying out all fracture tolerance analyses. Through-wall cracks in axial and circumferential directions on both straight pipes and elbows were postulated and analyzed. The loads applied on the postulated cracked pipes were obtained from detailed piping stress analysis under deadweight load, design pressure, thermal expansion, seismic design based earthquake (DBE) and thrust load due to the opening of relief valves. J-resistance data were derived from the lowest fracture toughness testing data obtained from Ontario Power Generation's PHT (primary heat transport) LBB material testing programs. A margin of 2 on crack size was chosen in establishing maximum allowable crack sizes. Leakage rates were calculated using SQUIRT Windows Version 1.1 program. The fluid inside the main steam line pipes was assumed single phase steam at 100% quality. One tenth of the calculated leakage rates was proposed as the requirement for minimum leakage detection capability. The paper concludes that the absence of through-wall crack larger than 91.16 mm in length should be maintained in order to ensure the structural integrity of large diameter main steam line pipes. In lieu of this crack size requirement, a reliable leakage detection capability which could quantify mass steam leakage rate of 0.01678 kg per second, or volume leakage rate of 1.01 l/min, should be in place. If both of the above two requirements are met, the Leak-Before-Break of these large diameter main steam line pipes is warranted.     

Design development of the HTGR core and its support structure seismic considerations

This paper discusses the development of the core support structure design from that employed on Fort St Vrain to recently announced contracts by Philadelphia Electric, Delmarva Power and Light and Southern California Edison for the large HTGR. Particular emphasis is given to the seismic considerations in the design of the structure for the large HTGR. The overall configuration of each reactor type is critically compared. Although similar components are employed, the basic difference in layout configuration results in significant conceptual differences in the structural and mechanical requirements of the core support components. The configuration and major components for the large reactor are described in some detail. The essential features and function of components are discussed. The graphite components in the core support floor and permanent reflector are designed to form a tight array during reactor normal operating conditions. This composite structure resists compressive loading due to differential gas forces and concrete pressure vessel movement. This tight array concept has important advantageous effects on primary coolant flow distribution and seismic capability.The paper discusses the inherent requirements and methodology in developing a standard plant design for high seismic sites. A design suitable for 0.15 g operating basis earthquake and 0.25 g safe shutdown earthquake has been developed which is applicable for over 80% of the expected sites in the USA. The HTGR core and support structure consists of many thousand graphite elements. It behaves as an inelastic body having random response when subjected to seismic excitation. The paper describes simplified analytical models which have been developed to investigate this phenomenon. An overview of a test program to substantiate and correlate with the analytical models is provided. The program addresses the interelement collision forces and frequencies of elements within the core and the load/deflection at the boundary. Various one, two and three-dimensional scale models have been tested. A summary of the objectives of the program is provided.     

Conceptual evaluation of fluid–structure interaction effects coupled to a seismic event in an innovative liquid metal nuclear reactor

The seismic response analysis of such liquid storage systems, especially liquid metal reactors, as for example the eXperimental Accelerator Driven System (XADS), was examined taking into account mainly the coupling effects of the fluid–structure interaction and their influence on its relevant internal systems and components.Therefore this paper deals with the structural analyses of the seismically induced hydrodynamic responses, in the event of a safe shutdown earthquake (SSE), and the free oscillation (known as sloshing waves) of a metal liquid coolant as well as the dynamic buckling effects on involved structures.To the mentioned purpose the interaction and coupling effects among the main reactor vessel structures and the primary coolant response were investigated by means of a numerical evaluation (with a qualified finite element code) because of the lack of analytical linear theories that in any case are not adequate to describe all the complex phenomena related to the seismic loading.For the numerical modelling procedure, 3D finite element models were set up to analyse the propagation of seismic waves as well as its derived structural effects, such as the fluid steep waves motion, the local buckling bulges, etc., taking into account the geometrical and material nonlinearities of the RPV and the considered simplified internals.The obtained numerical results in terms of stress intensity and of the capability of the structures to resist relevant seismic loads are, thus, presented and discussed. Moreover the performed analyses allowed to highlight the structures mostly affected by the assumed loading conditions in order to achieve data useful for an upgrading of the design geometry, if any, for the considered reactor.     

PRTHRUST-J1 code for calculation of blowdown thrust force and its experimental verification

This paper presents an outline of the PRTHRUST-J1 code for calculating blowdown thrust force and gives two numerical examples to show the effectiveness of this code. One numerical example is the problem of saturated steam blowdown. The blowdown thrust forces obtained from the PRTHRUST-J1 code were compared with those of the simplified method of Moody. Fairly good agreement was found between these two results. The other numerical example is the problem of jet discharging tests with stop valve performed in Japan Atomic Energy Research Institute. Analysis was carried out by varying the discharge coefficient. The analytical blowdown thrust force and pressure in the discharging nozzle were compared with experimental results. Qualitative agreement was found between the analytical and experimental results of the blowdown thrust force. Generally speaking, the blowdown thrust forces obtained from the experiment were between the analytical results for discharge coefficients of 1.0 and 0.6.     

Analysis of flow induced valve operation and pressure wave propagation for single and two-phase flow conditions

The flow induced valve operation is calculated for single and two-phase flow conditions by the fluiddynamic computer code DYVRO and results are compared to experimental data. The analysis show that the operational behaviour of the valves is not only dependent on the condition of the induced flow, but also the pipe flow can cause a feedback as a result of the induced pressure waves. For the calculation of pressure wave propagation in pipes of which the operation of flow induced valves has a considerable influence it is therefore necessary to have a coupled analysis of the pressure wave propagation and the operational behaviour of the valves.The analyses of the fast transient transfer from steam to two-phase flow show a good agreement with experimental data. Hence even these very high loads on pipes resulting from such fluiddynamic transients can be calculated realistically.     

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.     

Scram and nonlinear reactor system seismic analysis for a liquid metal fast reactor

Earthquake vibrations cause large forces and stresses that can significantly increase the scram time required for safe shutdown of a nuclear reactor. The horizontal deflections of the reactor system components cause impact between the control rods and their guide tubes and ducts. The resulting frictional forces, in addition to other operational forces, delay the travel time of the control rods. To obtain seismic responses of the various reactor system components (for which a linear response spectrum analysis is considered inadequate) and to predict the control rod drop time, a non-linear seismic time history analysis is required. Nonlinearities occur due to the clearances or gaps between various components. When the relative motion of adjacent components is large enough to close the gaps, impact takes place with large impact accelerations and forces.This paper presents the analysis and results for a liquid metal fast rector system which was analyzed for both scram times and seismic responses such as bending moments, accelerations and forces. The reactor system was represented with a one-dimensional nonlinear mathematical model with two degrees of freedom per node (translational and rotational). The model was developed to incorporate as many reactor components as possible without exceeding computer limitations. It consists of 12 reactor components with a total of 71 nodes, 69 beam and pin-jointed elements and 27 gap elements. The gap elements were defined by their clearances, impact spring constants and impact damping constants based on a 50% co-efficient of restitution.The horizontal excitation input to the model was the response of the containment building at the location of the reactor vessel supports. It consists of a 10 sec safe shutdown eathquake (SSE) acceleration-time history at 0.005 sec intervals and with a maximum acceleration of 0.408 g. The analysis was performed with two Westinghouse special purpose computer programs. The first program calculated the reactor system seismic responses and stored the impact forces on tape. The impact forces on the control rod driveline were converted into vertical frictional forces by multiplying them by a coefficient of friction, and then these were used by the second program for the scram time determination.The results give time history plots of various seismic responses, and plots of scram times as a function of control rod travel distance for the most critical scram initiation times. The total scram time considering the effects of the earthquake was still acceptable but about four times longer than that calculated without the eathquake. The bending moment and shear force responses were used as input for the structural analysis (stresses, deflections, fatigue) of the various components, in combination with the other applicable loading conditions.     

Seismic behaviour of piping systems with and without defects: experimental and numerical evaluations

Based on important experimental and analytical programs, CEA has developed a wide expertise in the domain of the seismic studies of piping. The specific behaviour of piping systems, with and without flaw, under seismic loading, has been analysed. CEA has evaluated the margins coming from present design procedures, that are shown to be greatly conservative, and has developed analytical methods devoted to a better evaluation of the global behaviour (internal moments in the piping system or reactions in supports, displacements, rotations). Non-linear time–history procedures have been built that allow for accurate modelisation, with and without defect. For industrial purposes or for sensitivity analyses, simplified methods have been proposed that are much less time consuming than non-linear time–history calculations, but much more accurate than linear methods.     

Seismic reevaluation and upgrading of nuclear power facilities outside the US using US developed methodologies

Several seismic licensing and safety issues have emerged over the past fifteen years for commercial US Nuclear Power Plants and US Government research reactors, production reactors and process facilities. The methodologies for the resolution of these issues have been developed in numerous government and utility sponsored research programs. The resolution criteria have included conservative deterministic design criteria, deterministic seismic margins assessments criteria (SMA) and seismic probabilistic risk assessment criteria (SPRA). The criteria for SMAs and SPRAs have been based realistically on considering the inelastic energy absorption capability of ductile structures, equipment and piping and have incorporated the use of earthquake and testing experience to evaluate the operability of complex mechanical and electrical equipment. Most of the applications to date have been confined to the US, however there have been several applications to Asian, Western and Eastern Europe reactors. This paper summarizes the major issues addressed, the development of reevaluation criteria and selected applications to non US reactors including VVER reactors of Soviet origin.     

Development and application of high volume remote handling systems in support of JET and ITER

The Joint European Torus (JET) Remote Handling System has evolved from a small scale maintenance capability to one of high efficiency large volume installations. The Enhanced Performance 2 shutdown 2010–2011 for example, required the replacement of many thousands of components ranging from about 100 g to 130 kg in weight. The scale of this type of operation and the necessity to maximise operational availability intensified the demands for high productivity whilst maintaining the necessary high standards for precision, reliability, cleanliness, and operational security.This paper discusses the developments in design, control, maintenance, preparation and operation of the current state of the art remote handling facilities at JET. It explores how the experience of over 20,000 h of operations has developed the applied methodology and how this could be appropriate to ITER and other facilities requiring complex remote maintenance, where extensive, high productivity remote handling operations will be essential. It also discusses the advances that have been made in management and presentation of operational data within the command, control and human machine interfaces (HMI) systems, along with the supporting operational databases.     

Assessing the defect-free structure

There has been a strong incentive within Nuclear Electric to develop a comprehensive assessment procedure to evaluate the high temperature response of structures; the so-called R5 procedure is extensive and Vols. 2 and 3 deal with the analytical approach and the assessment of creep fatigue damage for defect-free structures. In this paper we describe the approach which has been adopted and identify the relationship between the analytical requirement and material damage assessment under cyclic loading.     

Evaluation of nuclear system requirements for accommodating seismic effects

An integrated approach is presented for the design of nuclear facilities to accommodate seismic effects. Site evaluation and soil-structure interaction are discussed briefly since they determine the magnitude and characteristics of the input forcing function to be used in the analysis. The evaluation of the requirements of the nuclear system to accommodate the effects of specified seismic input forces involves the classification of components of the system, development of a mathematical model which adequately defines and relates the components of the system, consideration of coupling and resonance effects on the interactions among the components, selection of the methods of analysis and choice of the methods of solving the resulting equations. Specific design considerations and criteria for judging the acceptability of the design are discussed.