• Nuclear Engineering and Technology
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• 제52권10호
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• pp.2173-2182
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• 2020

Control valves are widely used to regulate fluid flux in nuclear power plants, and there are more than 1500 control valves in the primary circuit of one nuclear power plant. With their help, the flux can be regulated to a specific level of water or steam to guarantee the energy efficiency and safety of the nuclear power plant. The flux characteristics of the control valve mainly depend on the valve core shape. In order to analyze the effects of valve core shapes on flux characteristics of control valves, this paper focuses on the valve core shapes. To begin with, numerical models of different valve core shapes are established, and results are compared with the ideal flux characteristics curve for the purpose of validation. Meanwhile, the flow fields corresponding to different valve core shapes are investigated. Moreover, relationships between the valve core opening and the outlet flux under different valve core shapes are carried out. The flux characteristics curve and equation are proposed to predict the outlet flux under different valve core openings. This work can benefit the further research of the flux control and the optimization of the valve core for control valves in nuclear power plants.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4181-4194
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• 2022

Main steam safety valves are commonly used in nuclear power plants to provide final protections from overpressure events. Blowdown and dynamic stability are two critical characteristics of safety valves. However, due to the parameter sensitivity and multi-parameter features of safety valves, using traditional method to design and/or optimize them is generally difficult and/or inefficient. To overcome these problems, a surrogate model-based valve design optimization is carried out in this study, of particular interest are methods of valve surrogate modeling, valve parameters global sensitivity analysis and valve performance optimization. To construct the surrogate model, Design of Experiments (DoE) and Computational Fluid Dynamics (CFD) simulations of the safety valve were performed successively, thereby an ensemble surrogate model (E-AHF) was built for valve blowdown and stability predictions. With the developed E-AHF model, global sensitivity analysis (GSA) on the valve parameters was performed, thereby five primary parameters that affect valve performance were identified. Finally, the k-sigma method is used to conduct the robust optimization on the valve. After optimization, the valve remains stable, the minimum blowdown of the safety valve is reduced greatly from 13.30% to 2.70%, and the corresponding variance is reduced from 1.04 to 0.65 as well, confirming the feasibility and effectiveness of the optimization method proposed in this paper.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.849-859
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• 2022

Feed-water valves (FWVs) are used to regulate the flow rate of water entering steam generators, which are very important devices in nuclear power plants. Due to the working environment of relatively high pressure and temperature, there is strength failure problem of valve body in some cases. Based on the thermo-fluid-solid coupling model, the valve body stress of the feed-water valve in the opening process is investigated. The flow field characteristics inside the valve and temperature change of the valve body with time are studied. The stress analysis of the valve body is carried out considering mechanical stress and thermal stress comprehensively. The results show that the area with relatively high-velocity area moves gradually from the bottom of the cross section to the top of the cross section with the increase of the opening degree. The whole valve body reaches the same temperature of 250 ℃ at the time of 1894 s. The maximum stress of the valve body meets the design requirements by stress assessment. This work can be referred for the design of FWVs and other similar valves.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2905-2918
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• 2023

The valve assembly used in nuclear power plants is important safety-related equipment. In the new standard, the physical attributes are measured using a valve diagnosis test, which is used in the expansion to other non-tested valves using a quantitative test-basis methodology. With a motor-operated actuator, the state of stem's lubrication is related to physical attributes such as the stem factor and the friction coefficient. This study analyzed the numerical transient of fluid and solid lubrication with a squeeze film effect due to the loading rate on the stem and the stem nut using the experimental data. The differential equation that governs the motion mechanism of the stem and stem nut is established and analyzed. The flow rate, the fluid and the solid contact forces are calculated with the friction coefficient. Finally, we found that a change in the friction coefficient results from a change of the shear force in the solid contact mode during the interchange process between the solid contact mode and the fluid contact mode. The qualitative understanding of the squeeze film effect is expanded quantitatively for forces, thread surface distance, velocity, and acceleration, with consideration of the metal solid contact and fluid contact.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2509-2522
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• 2021

A valve assembly used in nuclear power plants must be qualified and supervised. New technical standards such as ASME QME-1 2007 particularly require detailed qualification using experiment and analysis. Particularly, diagnostic tests and engineering studies are required for qualification of ASME QME-1 2007. Among these studies, the research on the measurement of friction coefficient and packing stress is important. The irregular change of packing stress along the stroke distance occurs because of the abnormal phenomenon, which must be found and studied with quantitative methods. Packing stress should be analyzed conservatively through experimentation and analysis. In this study, various formulas were applied to measure and calculate coefficient of friction and packing stress. This study can be used in relation to qualification and supervision of packing materials. And the calculation using static diagnosis test can be used to find the packing frictional force in dynamic diagnosis test with flow pressure in a pipe. This study has made it possible to reliably consider packing frictional force generated in a valve body. And so, it is believed that more margin can be secured when evaluating the capacity of valve actuator by applying the accurate frictional force generated in the valve assembly.

• Nuclear Engineering and Technology
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• 제54권6호
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• pp.2107-2119
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• 2022

Accurate remaining useful life (RUL) prediction for critical components of nuclear power equipment is an important way to realize aging management of nuclear power equipment. The electric gate valve is one of the most safety-critical and widely distributed mechanical equipment in nuclear power installations. However, the electric gate valve's extended service in nuclear installations causes aging and degradation induced by crack propagation and leakages. Hence, it is necessary to develop a robust RUL prediction method to evaluate its operating state. Although the particle filter(PF) algorithm and its variants can deal with this nonlinear problem effectively, they suffer from severe particle degeneracy and depletion, which leads to its sub-optimal performance. In this study, we combined the whale algorithm with regularized particle filtering(RPF) to rationalize the particle distribution before resampling, so as to solve the problem of particle degradation, and for valve RUL prediction. The valve's crack propagation is studied using the RPF approach, which takes the Paris Law as a condition function. The crack growth is observed and updated using the root-mean-square (RMS) signal collected from the acoustic emission sensor. At the same time, the proposed method is compared with other optimization algorithms, such as particle swarm optimization algorithm, and verified by the realistic valve aging experimental data. The conclusion shows that the proposed method can effectively predict and analyze the typical valve degradation patterns.

• 한국압력기기공학회 논문집
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• 제17권2호
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• pp.75-82
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• 2021

The 3D printing technology is one of the fourth industrial revolution technology that drives innovation in the manufacturing process, and should be applied to nuclear industry for various purposes according to the manufacturing trend change. In nuclear industry, it can be applied to manufacture obsolete items and new designed parts in advanced reactors or small modular reactors (SMRs), replacing the traditional manufacturing technologies. A gate valve body was manufactured, which was obsolete in nuclear power plant, using DED(Directed Energy Deposition) metal 3D printing technology after restoring design characteristics including 3D design drawing by reverse engineering. The 3D printed valve body was assembled with commercial parts such as seat-ring, disk, stem, and actuator for performance test. For the valve assembly, including 3D printed valve body, several tests were performed, including pressure test, end-loading test, and seismic test according to KEPIC MGG and KEPIC MFC. In the pressure test, hydraulic pressure of 391kgf/cm² was applied to 3D printed valve body, and no leak was detected. Also the 3D printed valve assembly was performed well in end-loading and seismic tests.

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2669-2681
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• 2021

Valve is widely used in the various industry areas to adjust and control the flow. Cavitation frequently takes place and sometimes is inevitable in various types of valve to cause the erosion damage. Therefore, how to control and minimize the effect of cavitation is still an important topic. This study numerically investigates the cavitating flow in a control valve with a perforated cage. The effects of some parameters on the cavitation are discussed. It also discusses to use the throttling steps to govern the cavitating flow. The results show that the opening degree of valve and the length of downstream divergent connection both influence the cavitation. The increase of the divergent length reinforces the cavitation. And the larger the opening of valve is, the intenser the cavitation is and the more vapor is present. The more throttling steps are helpful to decrease the cavitation.

• Nuclear Engineering and Technology
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• 제54권5호
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• pp.1674-1685
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• 2022

The ball valve is an important device in the pipeline transportation system of nuclear power plants. Its operational stability and safety directly affect the normal working of nuclear power plants. In this study, the transient numerical simulation of the opening and closing process of a ball valve was conducted on the basis of the flow interruption capability experiment of the ball valve by using the moving mesh method and inlet and outlet variable boundary conditions. The flow rate and pressure difference with time of the opening and closing process of the ball valve were studied. The internal flow characteristics of the ball valve under different relative openings were analyzed in conjunction with the typical back-step flow structure. Results show that the transient numerical results agree well with the experimental results. The internal flow characteristics of the ball valve are similar at the same opening during opening and closing process. At small opening, the spool and outlet channels easily form a back-step flow structure. The disappearance and generation of backflow vortices during opening and closing occur at 85% opening and 75% opening, respectively. With the decrease in opening degree, the difference in vortex core area in the flow channel of the ball valve spool in the opening and closing process gradually appears. The research results provide some reference value for the design and optimization of ball valves.

• International Journal of Fluid Machinery and Systems
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• 제3권1호
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• pp.58-66
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• 2010

A steam control valve is used to control the flow from the steam generator to the steam turbine in thermal and nuclear power plants. During startup and shutdown of the plant, the steam control valve is operated under a partial flow conditions. In such conditions, the valve opening is small and the pressure deference across the valve is large. As a result, the flow downstream of the valve is composed of separated unsteady transonic jets. Such flow patterns often cause undesirable large unsteady fluid force on the valve head and downstream pipe system. In the present study, various flow patterns are investigated in order to understand the characteristics of the unsteady flow around the valve. Experiments are carried out with simplified two-dimensional valve models. Two-dimensional unsteady flow simulations are conducted in order to understand the experimental results in detail. Scale effects on the flow characteristics are also examined. Results show three types of oscillating flow pattern and three types of static flow patterns.