scholarly journals Evolved Fuzzy Control System for a Steam Generator

2010 ◽  
Vol 5 (2) ◽  
pp. 179 ◽  
Author(s):  
Daniela Hossu ◽  
Ioana Făgărășan ◽  
Andrei Hossu ◽  
Sergiu St. Iliescu
Keyword(s):  
Control System ◽  
Water Level ◽  
Steam Generator ◽  
Measurement Errors ◽  
Nuclear Power ◽  
Power Plants ◽  
Fuzzy Model ◽  
Operating Conditions ◽  
Steam Generators ◽  
Control Approach

Poor control of steam generator water level is the main cause of unexpected shutdowns in nuclear power plants. Particularly at low powers, it is a difficult task due to shrink and swell phenomena and flow measurement errors. In addition, the steam generator is a highly complex, nonlinear and time-varying system and its parameters vary with operating conditions. Therefore, there is a need to systematically investigate the problem of controlling the water level in the steam generator in order to prevent such costly reactor shutdowns. The objective of this paper is to design, evaluate and implement a water level controller for steam generators based on a fuzzy model predictive control approach. An original concept of modular evolved control system, seamless and with gradual integration into the existent control system is proposed as base of implementation of the presented system.

scholarly journals A hybrid PSO-PID approach for trajectory tracking application of a liquid level control process

2015 ◽  
Vol 5 (2) ◽  
pp. 63-73 ◽  
Author(s):  
Turker Tekin Erguzel
Keyword(s):  
Water Level ◽  
Pid Controller ◽  
Nuclear Power ◽  
Power Plants ◽  
Control Process ◽  
Nonlinear Process ◽  
Reference Trajectory ◽  
Proportional Integral Derivative ◽  
Level Control ◽  
Steam Generators

Water level control is a crucial step for steam generators (SG) which are widely used to control the temperature of nuclear power plants. The control process is therefore a challenging task to improve the performance of water level control system. The performance assessment is another consideration to underline. In this paper, in order to get better control of water level, the nonlinear process was first expressed in terms of a transfer function (TF), a proportional-integral-derivative (PID) controller was then attached to the model. The parameters of the PID controller was finally optimized using particle swarm optimization (PSO). Simulation results indicate that the proposed approach can make an effective tracking of a given level set or reference trajectory.

Age-Related Degradation of Steam Generator Internals Based on Industry Responses to Generic Letter 97-06

2002 ◽  
Author(s):  
M. Subudhi ◽  
E. J. Sullivan
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Pressurized Water Reactor ◽  
Long Term Effects ◽  
Steam Generators ◽  
Pressurized Water ◽  
Age Related ◽  
The U.S

This paper presents the results of an aging assessment of the nuclear power industry’s responses to NRC Generic Letter 97-06 on the degradation of steam generator internals experienced at Electricite de France (EdF) plants in France and at a United States pressurized water reactor (PWR). Westinghouse (W), Combustion Engineering (CE), and Babcock & Wilcox (B & W) steam generator models, currently in service at U.S. nuclear power plants, potentially could experience degradation similar to that found at EdF plants and the U.S. plant. The steam generators in many of the U.S. PWRs have been replaced with steam generators with improved designs and materials. These replacement steam generators have been manufactured in the U.S. and abroad. During this assessment, each of the three owners groups (W, CE, and B&W) identified for its steam generator models all the potential internal components that are vulnerable to degradation while in service. Each owners group developed inspection and monitoring guidance and recommendations for its particular steam generator models. The Nuclear Energy Institute incorporated in NEI 97-06, “Steam Generator Program Guidelines,” a requirement to monitor secondary side steam generator components if their failure could prevent the steam generator from fulfilling its intended safety-related function. Licensees indicated that they implemented or planned to implement, as appropriate for their steam generators, their owners group recommendations to address the long-term effects of the potential degradation mechanisms associated with the steam generator internals.

Wear Scar Progression of Impact-Fretting at Elevated Temperature for Steam Generator Tubes in Nuclear Power Plants

2006 ◽  
Vol 326-328 ◽  
pp. 1251-1254 ◽  
Author(s):  
Chi Yong Park ◽  
Jeong Keun Lee
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Conditions ◽  
Wear Scar ◽  
Fretting Wear ◽  
Test Material ◽  
The Impact ◽  
Steam Generator Tubes

Fretting wear generated by flow induced vibration is one of the important degradation mechanisms of steam generator tubes in the nuclear power plants. Understanding of tube wear characteristics is very important to keep the integrity of the steam generator tubes to secure the safety of the nuclear power plants. Experimental examination has been performed for the purpose of investigating the impact fretting. Test material is alloy 690 tube and 409 stainless steel tube supports. From the results of experiments, wear scar progression is investigated in the case of impact-fretting wear test of steam generator tubes under plant operating conditions such as pressure of 15MPa, high temperature of 290C and low dissolved oxygen. Hammer imprint that is actual damaged wear pattern, has been observed on the worn surface. From investigation of wear scar pattern, wear mechanism was initially the delamination wear due to cracking the hard oxide film and finally transferred to the stable impact-fretting pattern.

Crack Initiation Factor by Impact Fretting Wear at INCONEL Alloy for Steam Generator Tube in Nuclear Power Plants

2007 ◽  
Vol 26-28 ◽  
pp. 1269-1272
Author(s):  
Chi Yong Park ◽  
Jeong Kun Kim ◽  
Tae Ryong Kim ◽  
Sun Young Cho ◽  
Hyun Ik Jeon
Keyword(s):  
Steam Generator ◽  
Wear Mechanism ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Operating Conditions ◽  
Fretting Wear ◽  
Steam Generator Tube ◽  
Inconel Alloy ◽  
The Impact

Inconel alloy such as alloy 600 and alloy 690 is widely used as the steam generator tube materials in the nuclear power plants. The impact fretting wear tests were performed to investigate wear mechanism between tube alloy and 409 stainless steel tube support plates in the simulated steam generator operating conditions, pressure of 15MPa, high temperature water of 290°C and low dissolved oxygen(<10 ppb). From investigation of wear test specimens by the SEM and EDS analysis, hammer imprint, which is known to be an actual damaged wear pattern, has been observed on the worn surface, and fretting wear mechanism was investigated. Wear progression of impact-fretting wear also has been examined. It was observed that titanium rich phase contributes to the formation of voids and cracks in sub-layer of fretting wear damage by impact fretting wear.

Structural Integrity Role in Plant Life Extension - A Case Study

2015 ◽  
Vol 750 ◽  
pp. 295-306
Author(s):  
Jin Hua Shi
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Structural Integrity ◽  
Operating Conditions ◽  
Life Extension ◽  
Guide Tube ◽  
Gas Temperature ◽  
Steam Generators ◽  
Steam Generator Tube ◽  
The Uk

The steam generators at Advanced Gas-Cooled Reactor (AGR) nuclear power stations in the UK are potentially life-limiting components. Enhancing the capability to monitor the steam generators has been identified as having the potential to provide key evidence in justifying the extension of the generating lifetime of the stations. It has been proposed to install new temperature measuring instrumentation to monitor reactor gas temperature and to provide additional data regarding steam generator operating conditions. The modification will be to introduce thermocouples to the bore of an intact steam generator tube to facilitate temperature measurement at or near to the locations of interest. The modified steam generator tube will be sealed at the feed header upstand. Between the upper surface of the superheater header tubeplate and the wall of the superheater header, the thermocouple bundle and sheath will be contained within a rigid stainless steel guide tube. The guide tube will be attached at both ends by welds, each forming a pressure boundary. At the tubeplate a weld will separate the bore of the sealed guide tube from the steam space within the superheater header; a weld between the guide tube and the superheater header will separate the steam space within the superheater header from atmosphere outside the header. In order to obtain a better design, three 3-dimentional finite element models have been created using ABAQUS. A series of cyclic pressure, and start-up and shutdown thermal transient stress analyses have been carried out to provide stress values for structural integrity assessments to be conducted using ASME III, Subsection NH and R5.

The Effect of Transient Operation on Gasket Contact Pressure in Steam Generator Primary Side Obround Manway

2009 ◽  
Author(s):  
Nick Idvorian ◽  
Nansheng Sun
Keyword(s):  
Contact Pressure ◽  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Complex Interaction ◽  
Time Varying ◽  
Transient Operation ◽  
Steam Generators ◽  
Start Up ◽  
Leak Tightness

The Steam Generators used in CANDU Nuclear Power Plants employ a primary-side obround manway design with two cover plates on each side of the opening. During transient operation of the units (heat up, start up, operation, shut down and cool down) a complex interaction between the manway components occurs, which not only causes time-varying gasket contact pressure but also induces a non-uniform gasket contact pressure along the gasket width. As a result, the leak tightness of the joint may be affected. The effect of different stud preloads on the gasket contact pressure during transient operation is investigated by the 3D FE Analysis of a primary-side obround manway of CANDU’s steam generator. The results are used to justify the seal performance of the manway structure under conditions of different stud pretension loads.

Parameter Optimization for AP1000 Steam Generator Feedwater Control System Using Particle Swarm Optimization Algorithm

2016 ◽  
Author(s):  
Shifa Wu ◽  
Pengfei Wang ◽  
Jiashuang Wan ◽  
Xinyu Wei ◽  
Fuyu Zhao
Keyword(s):  
Particle Swarm Optimization ◽  
Control System ◽  
Water Level ◽  
Steam Generator ◽  
Parameter Optimization ◽  
Particle Swarm ◽  
Operating Conditions ◽  
Control Performance ◽  
Level Control ◽  
Swarm Optimization

The U-tube Steam Generator (UTSG) of AP1000 Nuclear Power Plant (NPP) is the crucial component transferring heat from the primary loop to the secondary loop to make steam. The UTSG of AP1000 NPP is a highly complex, nonlinear and time-varying system and its parameters vary with operating conditions. Therefore, it is difficult and challenging to well control the water level of AP1000 UTSG by tuning the PID controller parameter in a traditional way, especially when the system is undergoing a sharp transient. To achieve better control performance, the Particle Swarm Optimization (PSO) algorithm was applied for the parameter optimization of the AP1000 UTSG feedwater control system in this study. First, the mathematical model of AP1000 UTSG was established and the objective function was developed with the system constraints considered. Second, the simulation platform was built and then the simulation was conducted in MATLAB/Simulink environment. Finally, the optimized parameters were obtained and the feedwater control system with optimized parameters was simulated against that without optimized. The simulation results demonstrate that optimized parameters of AP1000 UTSG feedwater control system can significantly improve the water level control performance with smaller overshoot and faster response. Therefore, the PSO based optimization method can be applied to optimizing AP1000 UTSG feedwater control system parameters to provide much better control capabilities.

Startup and Shutdown Steam Generator Separator Piping Forces for a Coal-Fired Supercritical Power Plant

2012 ◽  
Author(s):  
Rajgopal Vijaykumar ◽  
Julie M. Jarvis ◽  
Allen T. Vieira ◽  
James Humphrey ◽  
Dong Zheng
Keyword(s):  
Control System ◽  
Steam Generator ◽  
Power Plants ◽  
Cold Water ◽  
Control Valve ◽  
Operating Conditions ◽  
Operating Characteristics ◽  
Two Phase ◽  
Wide Range ◽  
Valve Opening

Coal-fired supercritical power plants have steam generator liquid/vapor separator systems used during transition to/from “drum” mode and “once-through” mode, which undergo flow transients, involving control systems and valve openings, during startup and shutdown. These transients result in fluid acceleration which can produce significant reaction loads on piping systems (20 kips or higher). The evaluation of these loads is used to design piping supports and to assess possible control system and valving modifications. The computation of these transient loadings is challenging because the conditions in steam generator separator systems range from supercritical to subcritical, two phase, cold water or steam conditions occurring over a wide range of pressures and valve operating characteristics. A transient analysis of a typical separator-condensate line is performed using computer codes RELAP5/MOD3.2 and R5FORCE for the hydrodynamic forcetime history. A range of hydraulic loads associated with a range of operating conditions is provided in this paper using different boundary conditions for separator tank pressure, initial temperature of water in pipe lines, and control valve opening/closing times. These sensitivity runs show the benefit of plant control system changes to prevent the control valves opening above 1400 psia, increasing the control valve opening time to over one second, and the effects of keeping the separator-condensate line hot.

scholarly journals Steam Generator Water-Level Control

1966 ◽  
Vol 88 (2) ◽  
pp. 343-354 ◽  
Author(s):  
Amir N. Nahavandi ◽  
Abram Batenburg
Keyword(s):  
Mathematical Model ◽  
Control System ◽  
Water Level ◽  
Steam Generator ◽  
Natural Circulation ◽  
Level Control ◽  
Steam Generators ◽  
Flow Rates ◽  
Mass Flow Rates ◽  
The Difference

A combined digital-analog mathematical model for the dynamic analysis of vertical U-tube natural-circulation steam generators is presented. The application of this model to the optimal design of a water-level controller for a steam generating unit is demonstrated. It is shown that a control system consisting of standard proportional and reset controls on water-level deviation from a desired set point and the difference between the steam and feedwater mass flow rates can be successfully employed for the control of water level in such a plant. The optimum values, as well as the range of the controller parameter sellings for which the steam generator exhibits a desired stable response, are determined.

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