基于CFNN的核蒸汽发生器水位控制

鉴于常规的PID控制存在控制对象参数变化时控制参数无法改变的不足,从而根据一个核蒸汽发生器(NSG)的简化数学模型,将一种补偿模糊神经网络(CFNN)用于NSG水位的控制。该网络由于引入了补偿神经元,使网络的容错性更好,系统更稳定。同时在神经网络学习算法中动态优化补偿模糊运算,使网络更适应,训练速度更快。仿真表明,该方法在装置负荷变化时比常规的PID控制方法超调量小,收敛速度快。该网络能在线调整参数,动态优化模糊规则,适于在线学习控制。该控制方法对NSG水位智能控制研究具有一定意义。     

Auto-tuned PID controller using a model predictive control methodfor the steam generator water level

In this paper, proportional-integral-derivative (PID) control gains are automatically tuned by using a model predictive control (MPC) method. The MPC has received much attention as a powerful tool for the control of industrial process systems. An MPC-based PID controller can be derived from the second-order linear model of a process. The steam generator is usually described by the well-known fourth-order linear model, which consists of the mass capacity, reverse dynamics, and mechanical oscillation terms. However the important terms in this linear model are the mass capacity and reverse dynamics terms, both of which can be described by a second-order linear system. The proposed auto-tuned PID controller was applied to a linear model of steam generators. The parameters of a linear model for steam generators are very different according to the power levels. The PID gains of the proposed controller are tuned automatically. Also, the proposed controller showed fast water level tracking and small shrink and swell performance by changing only the input-weighting factor according to the power level for the water-level deviation and sudden steam flow disturbances supposed to investigate the tracking performance and swell and shrink characteristics     

Design of stability-guaranteed neurofuzzy logic controller for nuclear steam generators

A neurofuzzy logic controller (NFLC), which is implemented by using a multilayer neural network with special types of fuzzifier, inference engine and defuzzifier, is applied to the water level control of a nuclear steam generator (SG). This type of NFLC has the structural advantage that arbitrary two-input, single-output linear controllers can be adequately mapped into a set of specific control rules of the NFLC. In order to design a stability-guaranteed NFLC, the stable sector of the given linear gain is obtained from Lyapunov's stability criteria. Then this sector is mapped into two linear rule tables that are used as the limits of NFLC control rules. The automatic generation of NFLC rule tables is accomplished by using the back-error-propagation (BEP) algorithm. There are two separate paths for the error back propagation in the SG. One considers the level dynamics depending on the tank capacity and the other takes into account the reverse dynamics of the SG. The amounts of error back propagated through these paths show opposite effects in the BEP algorithm from each other for the swell-shrink phenomenon. Through computer simulation it is found that the BEP algorithm adequately generates NFLC rule tables according to given learning parameters.     

基于模糊控制的核电蒸汽发生器水位的串级自抗扰控制

针对核电站蒸汽发生器所存在的非线性,时变性,大时滞等特点,本文提出了基于模糊控制的蒸汽发生器水位的串级自抗扰控制方案。该方案采用双闭环控制,内环采用带前馈的一阶线性自抗扰控制调节阀,并分别前馈补偿蒸汽扰动和给水扰动,外环采用二阶模糊自抗扰,设计了新型的幂次控制率。仿真结果表明,该控制方案对蒸汽发生器水位具有良好的控制效果,与串级ADRC-PID控制系统相比,不仅具有优良的鲁棒性和抗干扰能力,而且具有可行性。     

Design of an intelligent fuzzy logic controller for a nuclear research reactor

The main objective of this paper is to design an intelligent controller system based on the concepts of fuzzy logic. This latter will be used to control the power of a nuclear reactor. The principle of this controller is based on rules established from experiments used with a classical controller and from the knowledge and the expertise of the operators of the reactor. This intelligent controller could be used in parallel with the actual system, which is semiautomatic, as a decision aided system to assist the operators in the control room.     

P controller with partial feed forward compensation and decoupling control for the steam generator water level

In this paper, a P controller with partial feed forward compensation and decoupling control for the steam generator water level is presented. While taking the steam flowrate as a disturbance to water level, the controller design can be completed in three stages. (1) Main circuit controller is designed without regard to disturbance. Since the transfer function of the steam generator model contains integrate element and differential element, the proportional (P) controller can selected as main circuit controller instead of PID controller for steam generator water level. (2) Partial feed forward compensation is introduced to remove the disturbance from the steam flowrate. If disregarding the differential element, the partial feed forward compensation's designing turns to be very simple. Partial feed forward compensation coefficient is set as reciprocal of P controller gain. (3) The coupling effects between the water level regulating and steam flowrate disturbance can be decreased by model reference decoupling control. The proposed methodology shows satisfactory transient responses, disturbance rejection and robustness.     

Feedback control of a large steam generator

Feedback control of a large counter current heat exchanger employed as a once-through steam generator for a breeder reactor power plant is presented. The mathematical model of the exchanger consists of a set of nonlinear partial differential equations with boundary conditions specified at three space points. The system structure does not allow the control signals to be distributed in space; they are applied only at the boundary points.The objective of the investigation is to choose a practical implementable time-invariant feedback controller which would give satisfactory regulation of the system. Although the emphasis is placed on a specific once-through steam generator system, the techniques used here are equally applicable to other heat exchanger systems.     

基于模糊神经网络的直流蒸汽发生器压力协调控制

针对直流蒸汽发生器水容积小,蓄热能力小,数学模型具有不确定性和非线性,在负荷变化或受到扰动的动态过程中出口蒸汽的压力难于控制的特点,引入了协调控制思想,将蒸汽发生器出口蒸汽流量与给水流量引入到控制系统,将模糊神经网络应用到直流蒸汽发生器出口压力的控制中,在线调整神经网络的权值,在此基础上设计了基于模糊神经网络的直流蒸汽发生器压力协调控制器。仿真结果表明,该控制较PID控制具有更小的超调量和调整时间,对实际控制系统设计具有一定的参考价值。     

Adaptive fuzzy model based predictive control of nuclear steam generators

Poor control of U-tube steam generators (UTSG) in a nuclear power plant can lead to frequent reactor shutdowns or damage of turbine blades. The dynamics of steam generator vary as power level changes. There is, therefore, a need to systematically design a suitable controller for all power levels. In this paper, we employ the concepts of both predictive control and fuzzy set theory to design an appropriate control for UTSG water level. The controller has three main parts: (1) a TSK fuzzy model used for predicting the future behavior of UTSG, (2) a recursive algorithm to estimate parameters of this model and (3) a model predictive controller used to obtain optimal input control sequence. Simulation results show that the proposed controller has a remarkable performance for tracking the step and ramp reference trajectories while at the same time it is robust against steam flowrate changes.     

Droplets deposition in steam piping connecting steam generator and steam turbine in nuclear plant

A numerical investigation is carried out for turbulent droplet-laden flow of saturated steam produced in a steam generator (SG) that feeds steam turbine (ST) through a long and multi-bend steam piping. The main purpose of the study is to analyze deposition of droplets that form a wall film in the piping system. Commercial CFD code StarCD is used for the solution of turbulent flow field of droplet-laden steam. Turbulence is treated using k–ω model of turbulence. Wall film formation is solved by additional conservation equations. Two tasks were performed: parametric study of the deposition in a 90° elbow positioned with different orientation and the deposition in a more complex piping system. This system starts with outlets from steam generator with five mouthpieces leading to a collector pipe and connecting the steam piping leading to a steam turbine. The steam piping consists of three straight segments of pipes and two 90° elbows in the total length 17 m. The diameter of the steam piping is 0.425 m. Results of the simulations show where droplets deposit and where a liquid separator should be placed to drain away the water film and to avoid droplets from entering the steam turbine.     

Application of a neural network based feedwater controller to helical steam generators

In current generation pressurized water reactors (PWRs), the control of steam generator level experiences challenges over the full range of plant operating conditions. These challenges can be particularly troublesome in the low power range where the feedwater is highly subcooled and minor changes in the feed flow may cause oscillations in the SG level, potentially leading to reactor trip. The IRIS reactor concept adds additional challenges to the feedwater control problem as a result of a steam generator design where neither level or steam generator mass inventory can be measured directly.Neural networks have demonstrated capabilities to capture a wide range of dynamic signal transformation and non-linear problems. In this paper a detailed engineering simulation of plant response is used to develop and test neural control methods for the IRIS feedwater control problem. The established neural network mass estimator has demonstrated the capability to predict the steam generator mass under transient conditions, especially at low power levels, which is considered the most challenging region for a full range feed water controller.     

Theoretical study on a multivariate feedback control of a sodium-heated steam generator

This paper applies the connection of a multivariate feedback controller with a state estimator to a 1-MW sodium-heated steam generator for LMFBR theoretically, to obtain a control strategy which emphasizes, from the view point of safety and availability of the FBR plant, that a superheat of 30°C should be required for the evaporator steam. This involves a trial to study the feasibility for the estimation of such an inaccessible variable as the dry-out location of tubes and utilize the state estimate to design a feedback controller of steam generators. The Kalman filter tested was found to generate reasonable estimates of the transient process variables of the steam generator and can provide a major advantage of regulating steam condition of the system even in the presence of contamination by a rather high level of measurement noise in the view point of economic uses of micro- and/or minicomputers.     

A non-linear analysis of the rotational stiffness of a nuclear steam generator foundation

A non-linear analysis of the moment-rotation relationships of a circular foundation for equipment of a nuclear steam supply system is presented. Equations are derived and numerical results given for 1) normal mild steel anchor bolts and circular base plate arrangement and 2) post-tensioned high-strength steel anchor bolts and circular base plate arrangement. An axisymmetric finite element computer program was applied for finding the stiffness of concrete in compression. The inelastic behavior of steel bolts is accounted for by an iterative procedure.     

Flow-induced vibration of nuclear steam generator U-tubes in two-phase flow

Flow-induced vibration characteristics of a U-tube bundle were experimentally investigated in air-water two-phase flow. The test section was equipped with 39 U-tubes, simulating the innermost region of an actual steam generator. The U-tubes were made of Inconel 690 with a diameter of 19.05 mm. The horizontal region of the U-tubes had a rotated square array with a pitch of 31.11 mm and a p/d of 1.633. The U-tubes and supporting structures have almost the same prototypical geometries. Vibration responses of six U-tubes were measured with ten 3-axis accelerometers. Two sets of experiments were performed to investigate an onset of fluid-elastic instability, damping ratio, and hydrodynamic mass of the U-tubes. The experiments were performed for a void fraction of 70-95%. The instability constant (K) of the Connors’ equation for the present U-tube bundle was evaluated to be in the range of 6.5-10.5.     

Data-driven controller of nuclear steam generators by set membership approximation

Poor control of U-tube steam generators (UTSG) in a nuclear power plant can lead to frequent reactor shutdowns or damage of turbine blades. The steam generator is a highly complex, non-linear and time-varying system and its parameters vary with operating conditions. Therefore, it seems that design of a suitable controller is a necessary step to enhance plant availability factor. In this paper, a data-driven controller approximated by set membership approach is presented for the water-level control of U-tube steam generators in nuclear power plants. This controller is capable of learning the control action principles from the data obtained using other methods of automatic or manual control. Simulation results of the approximated controller demonstrate its capability in regulating the water level under random disturbances and reference level changes.     

Assessment of nuclear energy sustainability index using fuzzy logic

Nuclear energy is increasingly perceived as an attractive mature energy generation technology that can deliver an answer to the worldwide increasing energy demand while respecting environmental concerns as well as contributing to a reduced dependence on fossil fuel. Advancing nuclear energy deployment demands an assessment of nuclear energy with respect to all sustainability dimensions.In this paper, the nuclear energy, whose sustainability will be assessed, is governed by the dynamics of three subsystems: environmental, economic, and sociopolitical. The overall sustainability is then a non-linear function of the individual sustainabilities. Each subsystem is evaluated by means of many components (pressure, status, and response). The combination of each group of indicators by means of fuzzy logic provides a measurement of sustainability for each subsystem.     

压水堆蒸发器水位调节失控跳堆分析及在线诊断调整

分析了核电站引发蒸发器给水控制系统(ARE)水位跳堆的主要原因和薄弱环节,建立了在功率运行状态下进行ARE主阀调节偏差在线调整和旁路阀调节偏差在线调整技术,提出并实施处理方案,解决了压水堆机组在启停过程中水位容易发生失控的技术难题.     

Prediction of crack coalescence of steam generator tubes in nuclear power plants

Prediction of failure pressures of cracked steam generator tubes of nuclear power plants is an important ingredient in scheduling inspection and repair of tubes. Prediction is usually based on nondestructive evaluation (NDE) of cracks. NDE often reveals two neighboring cracks. If the cracks interact, the tube pressure under which the ligament between the two cracks fails could be much lower than the critical burst pressure of an individual equivalent crack. The ability to accurately predict the ligament failure pressure, called “coalescence pressure,” is important. The failure criterion was established by nonlinear finite element model (FEM) analyses of coalescence of two 100% through-wall collinear cracks. The ligament failure is precipitated by local instability of the ligament under plane strain conditions. As a result of this local instability, the ligament thickness in the radial direction decreases abruptly with pressure. Good correlation of FEM analysis results with experimental data obtained at Argonne National Laboratory’s Energy Technology Division demonstrated that nonlinear FEM analyses are capable of predicting the coalescence pressure accurately for 100% through-wall cracks. This failure criterion and FEA work have been extended to axial cracks of varying ligament width, crack length, and cases where cracks are offset by axial or circumferential ligaments.     

Leak behavior of SCC degraded steam generator tubings of nuclear power plant

A forced outage due to a steam generator tube leak in a Korean nuclear power plant has been reported [Kim, J.S., Hwang, S.S., et al., 1999. KAERI Internal Report (Korean). Destructive analysis on pulled tubes from Ulchin unit 1. Korea Atomic Energy Research Institute]. Primary water stress corrosion cracking has occurred in many tubes in the plant, and they were repaired using sleeves or plugs. In order to develop proper repair criteria, it is necessary to understand the leak behavior of the tubes containing stress corrosion cracks. Cracked specimens were prepared using a room temperature cracking technique, and the leak rates and burst pressures of the degraded tubes were determined both at room temperature and at a high temperature. Some tubes with 100% through wall cracks did not show a leakage at 10.8 MPa, which is the typical pressure difference of the pressurized water reactors (PWRs) during a normal operation. In some tests, the leak rates of the tubes increased with time at a constant pressure. In a high temperature pressure test at 282 °C one specimen showed a very small leakage at 18.6 MPa, which stopped after a small increase in the test pressure. Because stress corrosion cracks can develop at relatively low stresses, even 100% through wall cracks can be so tight that they will not leak at a normal operating pressure.