Recent Patents on Valve Mechanism Device

2020 ◽  
Vol 13 (3) ◽  
pp. 230-241
Author(s):  
Ye Dai ◽  
Hui-Bing Zhang ◽  
Yun-Shan Qi
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Flow Direction ◽  
Research Progress ◽  
Valve Mechanism ◽  
Advantages And Disadvantages ◽  
Valve Structure ◽  
Valve Leakage ◽  
Safety And Reliability

Background: Valves are an important part of nuclear power plants and are the control equipment used in nuclear power plants. It can change the cross-section of the passage and the flow direction of the medium and has the functions of diversion, cutoff, overflow, and the like. Due to the earthquake, the valve leaks, which will cause a major nuclear accident, endangering people's lives and safety. Objective: The purpose of this study is to synthesize the existing valve devices, summarize and analyze the advantages and disadvantages of various devices from many literatures and patents, and solve some problems of existing valves. Methods: This article summarizes various patents of nuclear-grade valve devices and recent research progress. From the valve structure device, transmission device, a detection device, and finally to the valve test, the advantages and disadvantages of the valve are comprehensively analyzed. Results: By summarizing the characteristics of a large number of valve devices, and analyzing some problems existing in the valves, the outlook for the research and design of nuclear power valves was made, and the planning of the national nuclear power strategic goals and energy security were planned. Conclusion: Valve damage can cause serious safety accidents. The most common is valve leakage. Therefore, the safety and reliability of valves must be taken seriously. By improving the transmission of the valve, the problems of complicated valve structure and high cost are solved.

scholarly journals Experimental Research of Valve Tightness at Different Closure Forces

2018 ◽  
pp. 14-17
Author(s):  
A. V. Koroliov ◽  
P. Y. Pavlyshyn ◽  
I. V. Bandurko
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Valve Closure ◽  
Criterial Equation ◽  
Experimental Data Processing ◽  
Rotational Moment ◽  
Valve Leakage ◽  
Incomplete Closure ◽  
Almost All

Power valves are installed on almost all the pipelines of nuclear power plants performing the functions of regulation and shutting off the flow, so its failure often leads to emergencies. A particularly large number of failures is observed in motor-operated valves. Incorrect setting of the limiting clutch leads either to incomplete closure of the valve or to rod failure. Therefore, the valves are equipment of a nuclear power plant, which often falls into repair shops. Failures leading to an increase of valve leakage are especially dangerous for nuclear power plants. In this case, leakage of high-pressure valves leads to erosion of the sealing surfaces, which only increases the leakage. Thus, it is very important to determine the optimum rotational value when the valve is closed. The lack of conditions for closure force in the standards for valve leakage complicates the issue. A bench that allows working in the air with a pressure up to 3.5 MPa was developed on valve rod to study dependence of valve leakage on the rotational moment. Four independent parameters were measured: air pressure in front of the valve under study, closure force of the valve, volume of air loss through the valve and leakage time. A standard stop valve with a nominal diameter of 15 mm and a nominal pressure of 64 atm was used for the study. The determined dependence of the leakage on torque value allows recommending a gentler mode of valve closure without significantly reducing its tightness. As a result of experimental data processing, a criterial equation is obtained linking a leakage rate, pressure drop on the valve and a rotational moment value. The received criterial equation will allow defining the compromise between valve closure force and permissible leak level according to regulatory requirements. The analysis of the “leakage/rotational moment” diagram showed the possibility to reveal the damaged valves. This possibility may be used during the incoming inspection of the valves supplied to NPP, which should significantly improve the reliability of their operation.

Maintenance and Testing Errors in Nuclear Power Plants: A Preliminary Assessment

1980 ◽  
Vol 24 (1) ◽  
pp. 280-284
Author(s):  
S. Keith Adams ◽  
Zeinab A. Sabri ◽  
Abdo A. Husseiny
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Management Practices ◽  
Individual Responsibility ◽  
Human Factors Engineering ◽  
Hardware System ◽  
Safety And Reliability ◽  
Maintenance Work ◽  
Erroneous Data

Safety and reliability in the operation of nuclear power plants have been topics for intense analysis and debate in recent years. Once crucial factor which has received little attention until very recently, is human reliability in the operation, maintenance and testing of nuclear plants. Maintenance and testing errors are of particular importance in that, despite the fact that they are relatively few in number compared with hardware related events, their occurrence programs the system for future operator errors bv introducing erroneous data in operator-hardware system communication or by eliminating a portion of that communication. Erroneous data or a lack of anticipated feedback following control actions by the operator can tend to cause further inappropriate responses and to escalate the frequency of operator errors. Such effects are most undesirable during emergencies. A number of improvements based upon improved human factors engineering will be necessary to eliminate maintenance errors of the type encountered in this study. These include: standardization of design for given types of subsystems, standardization of maintenance operations, work simplification involving both physical and psychological aspects of maintenance work, the use of training simulators to increase the amount of practice in performing maintenance operations, automanual control to show predicted outcome of results, assignment of individual responsibility and accountability to specific hardware subsystems, measures of job performance against some criterion, programmed instruction, improved divisions of responsibility among maintenance, operating and technical personnel, and improved management practices.

scholarly journals Use of Reliability Block Diagram and Fault Tree Techniques in Reliability Analysis of Emergency Diesel Generators of Nuclear Power Plants

2019 ◽  
Vol 4 (4) ◽  
pp. 814-823 ◽  
Author(s):  
Vanderley de Vasconcelos ◽  
Wellington Antonio Soares ◽  
Antônio Carlos Lopes da Costa ◽  
Amanda Laureano Raso
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Block Diagram ◽  
Electrical Power ◽  
Fault Tree ◽  
Risk Level ◽  
Reliability Block Diagram ◽  
Advantages And Disadvantages ◽  
Diesel Generators

Nuclear power plants (NPPs) are subjected to events such as equipment failures, human errors and common-cause failures, in an environment of complex maintenance, inspection and testing managements. These events will affect the reliability of safety-related systems, as well as the risk level of the plant. Reliability block diagram (RBD) is often used to analyze the effect of item failures on system availability, taking into account their physical arrangement in the system. Fault tree (FT) is a commonly used technique for analyzing risk and reliability in nuclear, aeronautical and chemical industries. It represents graphically the basic events that will cause an undesired top event. Loss of electrical power is one of the main events that influences safe operation of NPPs, as well as accident prevention and mitigation. In case of unavailability of offsite power, emergency diesel generators (EDGs) supply onsite electrical power. This paper carries out reliability analyses of EDGs of NPPs using both RBD and FT techniques. Each technique has its own advantages and disadvantages, allowing a variety of qualitative and quantitative analyses. Outcomes using these two techniques are compared for a typical NPP EDG system.

Innovative Robot Technologies for Nuclear Power Plants Inspection and Maintenance

2014 ◽  
Author(s):  
Seungho Kim ◽  
Kyung Min Jung ◽  
Sung Uk Lee ◽  
Hocheol Shin ◽  
Chang Hoi Kim ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
High Dose ◽  
Nuclear Facilities ◽  
Human Errors ◽  
Safety Issues ◽  
Safety And Reliability ◽  
Inspection And Maintenance ◽  
Remote Inspection

Nuclear energy has become the main energy source in Korea, but the safety issues are being debated since the Fukusima accident. In order to guarantee the safety and reliability of nuclear power plants, the uncertainty of human errors are being minimized by utilizing innovative technologies for inspection and maintenance. KAERI has developed robotic systems to upgrade and ensure the safety of nuclear facilities, to detect unusual conditions of facilities through remote inspection and to prevent human workers from high dose radiation with efficient plant maintenance.

Development of Rules on Concrete Containment Vessels for NPP

2016 ◽  
Author(s):  
Tomohiro Nishizawa ◽  
Takao Nishikawa ◽  
Katsuki Takiguchi ◽  
Hiroyuki Sugita ◽  
Yasuo Hijioka ◽  
...  
Keyword(s):  
Reinforced Concrete ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Containment Vessel ◽  
Safety And Reliability ◽  
Near Future ◽  
Design And Construction

“Rules on Concrete Containment Vessels for Nuclear Power Plants” was first published in 2003. It was revised in 2011 and the latest Edition was published in 2015. Endorsement of this Code is expected in the very near future. There are two types of reactor containments: steel containment vessels (hereinafter referred to as “SCV”) and concrete containment vessels (hereinafter referred to as “CCV”). The former is addressed by Rules on Design and Construction for Nuclear Power Plants, and the latter by Rules on Concrete Containment Vessels for Nuclear Power Plants, including Reinforced Concrete Containment Vessel (hereinafter referred to as “RCCV”), Pre-stressed Concrete Containment Vessel (hereinafter referred to as “PCCV”), and hybrid containment vessels which are required further safety and reliability.

An Accurate Probabilistic Model for Estimating the Life Cycle Cost of Degrading Components in Nuclear Power Plants

2010 ◽  
Author(s):  
T. Cheng ◽  
M. D. Pandey ◽  
W. C. Xie
Keyword(s):  
Life Cycle ◽  
Life Cycle Cost ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Nuclear Plants ◽  
Proposed Model ◽  
Safety And Reliability ◽  
Finite Time Horizon ◽  
Functional Components

Degradation of systems and components operating in harsh environment has an adverse effect on safety and reliability of nuclear power plants. Condition-based maintenance (CBM) programs are used to preventively maintain degrading components, which minimize the risk of failure. However, maintenance programs can be costly due to frequent inspection, increased outage time, and redundant maintenance of functional components. The optimization of maintenance programs over the life cycle of systems is an important issue for the plant managers. The paper presents an advanced model for the evaluation of life cycle cost of degrading components in the nuclear plants, which can be used for the maintenance optimization. The proposed model is based on the more precise finite time horizon formulation, instead of using asymptotic formulae that may lead to inaccurate results in practical settings.

scholarly journals Research Progress of Reactor Thermal-Hydraulic Characteristics Under Ocean Conditions in China

2020 ◽  
Vol 8 ◽  
Author(s):  
Xin Jie
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Flow Instability ◽  
Reactor Core ◽  
Research Progress ◽  
Future Research ◽  
Hydraulic Characteristics ◽  
Bubble Behavior ◽  
Ocean Conditions

Floating nuclear power plants are affected by sea wind and waves, which will produce various forms of movement, and cause changes in the thermal and hydraulic characteristics of the reactor core and threaten the safety of reactor operation. In response to the R&D and design requirements of floating nuclear power plants, the research progress on the thermal-hydraulic characteristics of reactors under ocean conditions in China are reviewed in this paper. The emphasis is put on flow heat transfer, bubble behavior, flow instability, and critical heat flux under ocean conditions. Research progresses as well as the issues that need to be focused on in the future research are discussed in detail.

Sign in / Sign up

Export Citation Format

Share Document