The Research on Effects of the Medium Blocking on Motion Performances of Main Steam and Feed Water Isolation Valves With its Driving Device

2017 ◽  
Author(s):  
Dalei Pan ◽  
Hua Zhong ◽  
Shenjie Gu ◽  
Guangyue Guo ◽  
Hongbo Kuang
Keyword(s):  
Simulation Model ◽  
Nuclear Power ◽  
Power Plants ◽  
Feed Water ◽  
Closing Time ◽  
Blocking Effects ◽  
Simulation Results ◽  
Electromechanical Equipment ◽  
Main Steam ◽  
Resistance Coefficients

This article presents the study of the medium block effects on the driving device performances of main steam and feed water isolation valves, which are typical electromechanical equipment in nuclear power plants. The equipment simulation model is established, which can calculate motion performances under all working situations. The medium blocking force models during the motion of valves are established and the blocking effects on the driving device performances are calculated and analyzed. The simulation results show that the medium block has little effects on the opening performances of the driving device, while it has important effects on the closing performances, especially on the quick closing, which suggests that the equipment should be lubricated to reduce resistance coefficients and the main parameters of electromechanical equipment should be tuned and calibrated to guarantee the quick closing time.

The optimization methodology research on the motion performances of the driving device of main steam and feed water isolation valves

2017 ◽  
Vol 232 (6) ◽  
pp. 1069-1078
Author(s):  
Dalei Pan ◽  
Shenjie Gu ◽  
Guangyue Guo ◽  
Hongbo Kuang ◽  
Hua Zhong ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Maximum Pressure ◽  
Feed Water ◽  
Performance Indexes ◽  
Motion Performance ◽  
Optimization Methodology ◽  
Electromechanical Equipment ◽  
Main Steam

The driving device of the main steam and feed water isolation valves is one of the most important electromechanical equipment in nuclear power plants, and its motion performances are related to the safety of nuclear power plants. This article proposes an optimization methodology to research the motion performances. In the methodology, inputs are the motion performance indexes and the major influencing parameters with no need for complex mathematical models of the electromechanical equipment; the co-simulation model or the prototype are adopted to illustrate the influence rules of major parameters on the motion performance indexes; objective functions for optimization, which combine motion performances with weight factors, can reveal the influence rule curves. Then based on the methodology, motion performances of the equipment are analyzed and the key indexes are selected. Besides, the maximum pressure of the driving device is chosen as the major parameter and a dimensionless objective function synthesizing the motion performance indexes is proposed. Finally, the influence rule curve where the dimensionless function varies with the maximum pressure is obtained by the co-simulation and the experimental study on the prototype verifies the results, which provides references for the further research and development in the engineering application. For other electromechanical equipment, the method is an efficient tool to design, verify, and optimize the performances.

JSME Code on Protection Design Against Postulated Pipe Rupture for Nuclear Power Plants

2004 ◽  
Author(s):  
Keisuke Kitsukawa ◽  
Hiroshi Yokota ◽  
Koichi Murayama ◽  
Hiroshi Ueda ◽  
Yasukazu Takada ◽  
...  
Keyword(s):  
Power Generation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Water System ◽  
Jet Impingement ◽  
Feed Water ◽  
Main Steam ◽  
Leak Before Break

As one of the Codes for Nuclear Power Generation Facilities, “Rules on Protection Design against Postulated Pipe Rupture for Nuclear Power Plants (JSME S ND1-2002)” has been developed by the JSME Committee on Power Generation Facility Codes from October 2001 and published in December 2002. The code covers the design for protection against postulated pipe rupture in nuclear power plants and gives the basic plan of protection design, locations of postulated pipe rupture, methods of determining the rupture type and opening area, and a procedure for evaluating jet impingement phenomena. It is a special feature of the code that the LBB (Leak Before Break) concept is applied to the determination of the piping rupture type, which belongs to RCPB, or to the main steam and feed water system inside the PWR containment vessel. Types of piping material applicable to the LBB concept are austenitic stainless steel, carbon steel and low-alloy steel. Furthermore, the code provides the flow and the method of LBB evaluation. In this paper, we describe the major rules of the code, including an outline of LBB evaluation methods.

An Experimental Measurement of Toughness Locus for a Ductile Material

2005 ◽  
Vol 297-300 ◽  
pp. 2410-2415 ◽  
Author(s):  
Dong Hak Kim ◽  
Jeong Hyun Lee ◽  
Ho Dong Kim ◽  
Ki Ju Kang
Keyword(s):  
Crack Length ◽  
Nuclear Power ◽  
Power Plants ◽  
Test Procedure ◽  
Stress Triaxiality ◽  
Crack Tip ◽  
Potential Drop ◽  
Image Correlation ◽  
Ductile Material ◽  
Main Steam

A toughness locus Jc-Q for a ductile steel, SA106 Grade C used in the main steam piping of nuclear power plants, has been experimentally evaluated. Along with the standard fracture test procedure for J-R curve, Q as the second parameter governing stress triaxiality nearby the crack tip is measured from the displacements nearby the side necking which occurs near the crack tip on the lateral surface of a fracture specimen. The displacements nearby the side necking are measured from the digital images taken during the fracture experiment based on Stereoscopic Digital Photography (SDP) and high resolution Digital Image Correlation (DIC) software. The crack length is monitored by Direct Current Potential Drop (DCPD) method and the J-R curve is determined according to ASTM standard E1737-96. The effects of crack length, specimen geometry and thickness of specimen are studied, which are included in the toughness locus Jc-Q.

Simulation Analysis and Optimization of Lubricating Oil System

2021 ◽  
Author(s):  
Qiongxiao Wu ◽  
Jianjun Wang ◽  
Jingming Chen ◽  
Pengzheng Li
Keyword(s):  
Simulation Model ◽  
Simulation Analysis ◽  
Sudden Change ◽  
Pressure Change ◽  
Lubricating Oil ◽  
Maximum Pressure ◽  
Closing Time ◽  
Original Design ◽  
Relief Valve ◽  
Simulation Results

Abstract Based on the one-dimensional simulation model of lubricating oil system is established and analyzed by using FLOWMASTER software, this paper proposes a new method of optimizing lubricating oil system by PID technology. Ensure that the configuration requirements and control strategies of the relevant accessories of the simulation model are satisfied with the design requirements. Firstly, by simulating lubricating oil pressure fluctuation and lubricating oil flow distribution under Open/Close Valve in different opening and closing time, the optimal opening/closing time of Open/Close Valve is determined to be 0.2 s and 0.5 s respectively. Secondly, by writing the controller script file combined with a controller to realize automatic unloading relief valve simulation, determine the relief valve pressure regulating range of 0∼0.38 MPa, For precision of constant pressure valve of oil spill, the simulation results show that the average 10 m3/h flow caused by pressure changes of about 0.06 MPa. Under the flow sudden change signal of about 40 m3/h, the maximum pressure change is less than 0.1 MPa. Through the simulation results, it is found that most of the lubrication parts in the original design have the phenomenon of flow redundancy, which causes unnecessary pump power loss. The system is optimized by PID technology. By comparing the simulation results before and after optimization, it is found that the speed of constant displacement pump could be changed in time by PID controller, and the flow redundancy could be improved significantly, so the lubricating oil system could be lower consumption and achieve the purpose of optimization.

Suppression Methods of Acoustic Noise Generated in Main Steam Stop Valve

2018 ◽  
Author(s):  
Shiro Takahashi ◽  
Eiji Ozaki ◽  
Atsuyuki Minenaga
Keyword(s):  
Pressure Loss ◽  
High Speed ◽  
Nuclear Power ◽  
Power Plants ◽  
Acoustic Noise ◽  
Acoustic Resonance ◽  
Stop Valve ◽  
Cfd Analyses ◽  
Main Steam Line ◽  
Main Steam

The main steam stop valve (MSSV) is installed in the main steam line in thermal and nuclear power plants. The MSSV is a safety valve that instantaneously shuts off the steam flowing into the steam turbine in an emergency. However, as high-speed steam flow goes through the MSSV during even the rated operation, acoustic sound or noise is generated in the MSSV. Moreover, there is a possibility that flow-induced acoustic resonance occurs in the MSSV. Flow-induced acoustic resonance must be suppressed to decrease the sound noise. Reducing the pressure loss of the MSSV is also an important issue that cannot be neglected with respect to the plant thermal efficiency. Therefore, we have developed the MSSV which can suppress the flow-induced acoustic resonance and its pressure loss. To develop this MSSV, we conducted scale air tests and computational fluid dynamics (CFD) analyses that are described in this paper. Mach and Strouhal number of the test conditions were the same as those of an actual plant. Reynolds number was sufficiently large to obtain the developed turbulent flow. An unsteady compressible CFD analysis was also conducted using large eddy simulation as a turbulence model. We developed new tilted triangular tabs and installed them in the MSSV to suppress the intense vortex generation and pressure loss. As a result, the sound noise due to the flow-induced acoustic resonance was completely attenuated and pressure loss was reduced compared to the case using the current tilted tabs. CFD results also showed that the tilted triangular tabs could suppress the generation of intense vortexes and the flow-induced acoustic resonance.

Design and Development of Virtual DCS Debugging and Research Platform Based on NPP Simulation Model

2018 ◽  
Author(s):  
Caike Zhang ◽  
Jingwen Qi ◽  
Chun Liu ◽  
Chenglong Xie ◽  
Peibang Liu ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Simulation Model ◽  
Nuclear Power ◽  
Power Plants ◽  
Professional Training ◽  
Nuclear Power Plants ◽  
Testing Procedures ◽  
Communication Program ◽  
Research Platform

At present, DCS is widely used as the control system for nuclear power plants both at home and abroad, which prompting many companies to research the technology of DCS debugging. In this paper, taking a certain nuclear power plant within China for reference, the virtual DCS debugging and research platform which based on the full-scope nuclear power plant simulation model is developed. It was developed by first establishing a simulation model on the RINSIM Simulation Platform and ordering a customized set of virtual DCS system, then developing a communication program between the simulation model and the virtual DCS system. Users can observe the actual effects and results if following the pre-designed testing procedures after the configuration of control logics, HMI (Human Machine Interface) and I/O communication interfaces. The virtual DCS platform is aimed at assisting with technology research of DCS project for similar nuclear power plants and also can provide professional training for associated personnel of nuclear power plant.

scholarly journals Logistical Simulation Modeling for Planning a Soil Remediation Process

2019 ◽  
Vol 2019 ◽  
pp. 1-13
Author(s):  
David Kessel ◽  
Jihan Jeon ◽  
Jaeyeon Jung ◽  
Eutteum Oh ◽  
Chang-Lak Kim
Keyword(s):  
Simulation Model ◽  
Soil Remediation ◽  
Nuclear Power ◽  
Power Plants ◽  
Discrete Event ◽  
Soil Washing ◽  
Batch Process ◽  
Washing Process ◽  
Plant Operations ◽  
Remediation Process

This paper describes the development of a discrete event simulation model using the FlexSim software to support planning for soil remediation at Korean nuclear power plants that are undergoing decommissioning. Soil remediation may be required if site characterization shows that there has been radioactive contamination of soil from plant operations or the decommissioning process. The simulation model was developed using a dry soil separation and soil washing process. Preliminary soil data from the Kori 1 nuclear power plant was used in the model. It was shown that a batch process such as soil washing can be effectively modeled as a discrete event process. Efficient allocation of resources and efficient waste management including volume and classification reduction can be achieved by use of the model for planning the soil remediation process. Cost will be an important criterion in the choice of suitable technologies for soil remediation but is not included in this conceptual model.

Investigation of Blast Wave Effects on Containment Wall and Steam Generator

2018 ◽  
Author(s):  
Tae Jin Kim ◽  
Yoon-Suk Chang
Keyword(s):  
Blast Wave ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Failure Criteria ◽  
High Energy ◽  
Element Analysis ◽  
Secondary Damage ◽  
Main Steam Line ◽  
Main Steam

When a sudden rupture occurs in high energy lines such as MSL (Main Steam Line) and safety injection line of nuclear power plants, ejection of inner fluid with high temperature and pressure causes blast wave, and may lead to secondary damage of adjacent major components and/or structures. The objective of this study is to assess integrity of containment wall and steam generator due to the blast wave under a postulated high energy line break condition at the MSL piping. In this context, a preliminary analysis was conducted to examine the blast wave simulation using coupled Eulerian-Lagrangian technique. Subsequently, a finite element analysis was carried out to assess integrity of the structures. As typical results, strain and stress values were calculated at the containment wall and steam generator, which did not exceed their failure criteria.

Sign in / Sign up

Export Citation Format

Share Document