基于SPSA的蒸汽发生器液位MPC系统性能优化方法研究

核电厂蒸汽发生器（SG）液位变化过程具有强非线性且存在“虚假水位”现象,传统SG液位控制系统多采用固定参数比例-积分-微分（PID）控制器,但传统PID控制方法不具备自优化、自适应、自学习等能力,使得控制系统性能难以达到并保持最佳。为提高机组瞬态响应能力以及核电厂的稳定性、安全性和经济性,提出了一种基于并行摄动随机逼近（SPSA）算法的模型预测控制（MPC）方法。该方法采用MPC系统代替传统PID控制系统,并利用SPSA实现液位控制系统参数的整定优化,从而实现SG液位控制系统的性能优化。通过仿真试验验证了本方法能够有效提高SG液位控制参数的整定效率以及控制系统稳定性。     

中国先进研究堆功率调节系统的仿真研究及优化设计

基于核电厂SimPort仿真平台构建了中国先进研究堆(CARR)功率调节系统仿真模型,利用该模型对CARR功率调节系统进行了瞬态仿真研究;针对CARR功率调节系统驱动机构的特点,研究了控制棒位移精度以及电磁线圈和衔铁位移延迟对系统稳定性的影响.综合考虑CARR系统的工艺要求和功率控制系统的功能及特点,得到了数字控制器的整定参数和控制棒驱动机构的稳定限值:数字控制器的采样周期T=100ms,比例增益KP=300;控制棒位移精度的稳定限值为0.4mm,电磁线圈和衔铁的稳定限值为6.0mm.     

基于数字调节器的压水堆核电厂稳压器压力控制系统优化设计

基于数字调节器的稳压器压力控制系统,同时针对稳压器升、降压的不同动态特性,设计具有2组调节器参数的数字调节器控制系统,并给出含自动切换功能的控制策略SAMA图设计。最后,在SIMULINK仿真软件平台里实现调节器的2组比例积分(PI)参数整定。仿真结果验证此设计可以更好地将压力控制到稳定值,提高核电厂运行的安全性。     

“核能动力学会反应堆控制第四次专业会议”部分论文摘要

本文是在完成《秦山核电厂正常瞬态分析与控制系统整定值研究》课题后写成的,旨在研究核电厂关键物理参数在正常瞬态期间的变化规律与相互影响,并根据各控制系统的设计要求与性能指标合理地确定出各调节器的调节参数,使得核电厂控制     

基于自适应GA自抗扰控制在蒸汽发生器水位控制中的应用研究

核电厂蒸汽发生器是一个非线性、时变的系统.把自抗扰控制方法引入蒸汽发生器水位的串级控制系统中,能够解决传统PID控制中快速性和超调的矛盾,且能够动态补偿对象模型的内扰动和外扰动.由于自抗扰控制器参数众多,无统一调整规范,参数难以整定,为解决这一问题,采用自适应遗传算法来对自抗扰控制器参数进行优化.计算机仿真结果表明,本...     

基于H∞回路成形方法的蒸汽发生器水位控制系统主控制器参数整定

针对U形管蒸汽发生器水位前馈 反馈串级三冲量控制系统的主控制器PID参数整定困难,且鲁棒性要求高的特点,给出了一种基于H_∞回路成形方法的主控制器PID参数整定方案。在负荷变化情况下,对采用传统方案整定主控制器的PID参数和H^∞回路成形方法整定主控制器的PID参数蒸汽发生器水位控制系统进行了对比仿真实验。仿真结果表明,采用该方法整定的控制系统具有良好的鲁棒稳定性和较好的控制性能。     

用于核电厂数字化仪表控制系统优化设计的CATIA 2程序的改进

对压水堆核电厂瞬态仿真程序CATIA 2程序进行了改进,以满足核电厂数字化仪表控制系统优化设计工作的需要.主要的改进是在CATIA 2程序中,对ΔT保护通道引入了数字化采样模块模型及数字化处理时间步长,以及对ΔT保护通道所用的输入变量引入了采样时间步长,并针对数字化技术提出了一种新的超温ΔT定值方程模型.通过分析比较数值仿真的结果,表明所做的改进是合理和可行的.     

超功率ΔT保护整定值验证方法研究

超功率ΔT保护堆芯以防止超线功率密度引起的燃料元件损坏.以M310型核电厂为例,选取满功率下蒸汽系统管道破裂事故,研究初始工况、破口尺寸、反应性反馈系数、控制棒的调节等因素对核电厂超功率ΔT保护整定值有效性的影响,形成超功率ΔT保护整定值有效性的验证方法.     

核电厂运行规程整定值研究

核电厂运行规程与核电厂的安全运行密切相关,整定值内容作为运行规程的重要组成部分,其确定过程是运行规程开发技术的难点。针对示范工程非能动核电厂运行规程,对规程文件体系和整定值内容进行分析,结合核安全法规要求对规程整定值信息的功能、整定值结构进行研究,提出了确定运行规程整定值的系统性方法和总体流程。     

CAP1400核电厂设计分析器系统调试研究

本文系统研究了CAP1400设计分析器系统调试的难点及解决方案。根据分析器平台要求对各系统单机版程序及模型数据进行了适应性改善,成功地将CAP1400核电厂RELAP5工艺模型、SCADE电厂控制模型及人机显示画面等模型集成到了设计分析器平台,并分别进行了单系统调试及系统联合调试。在此基础上演示了线性升降负荷运行瞬态的调试成果。本文研究的主要工程价值在于为CAP1400核电厂控制系统验证、整定值分析等设计验证工作提供了一个综合性的仿真平台,并为相应的设计验证工作提供了很好的反馈。     

稳压器压力控制系统动态仿真

本文旨在通过核电厂控制系统的全数字仿真来验证稳压器的压力控制系统的设计，并根据瞬态分析的结果来确定各控制环节的参数，分析结果为秦山核电厂调试和最终安全分析报告提供了依据，并与实际高度结果比较验证了分析模型与方法的合理性。     

Load following control and global stability analysis for PWR core based on multi-model,LQG, IAGA and flexibility idea

The work is to design a nonlinear Pressurized Water Reactor (PWR) core load following control system and analyze the global stability of this system. On the basis of modeling a nonlinear PWR core, linearized models of the core at five power levels are chosen as local models of the core to substitute the nonlinear core model in the global range of power level. The combination control strategy of the Linear Quadratic Gaussian (LQG) control and the Proportional Integral Derivative (PID) control with an optimization process of Improved Adaptive Genetic Algorithm (IAGA) proposed is used to contrive a combined controller with the robustness of a core local model as a local controller of the nonlinear core. Based on the local models and local controllers, the flexibility idea of modeling and control is presented to design a decent controller of the nonlinear core at a random power level. A nonlinear core model and a flexibility controller at a random power level compose a core load following control subsystem. The combination of core load following control subsystems at all power levels is the core load following control system. The global stability theorem is deduced to define that the core load following control system is globally asymptotically stable within the whole range of power level. Finally, the core load following control system is simulated and the simulation results show that the control system is effective.     

Application of a self-organizing fuzzy logic controller to nuclear steam generator level control

In this paper, the self-organizing fuzzy logic controller is investigated for the water level control of a steam generator. In comparison with conventional fuzzy logic controllers, this controller performs the control task with no initial control rules; instead, it creates control rules and tunes input membership functions based on the performance criteria as the control behavior develops, and also modifies its control structure when uncertain disturbance is suspected. Selected tuning parameters of the self-organizing fuzzy logic controller are updated on-line in the learning algorithm, by a gradient descent method. This control algorithm is applied to the water level control of a steam generator model developed by Irving et al. The computer simulation results confirm the good performance of this control algorithm for all power ranges. This control algorithm can be expected to be used for the automatic control of a feedwater control system in a nuclear power plant with digital instrumentation and control systems.     

压水堆核电站稳压器压力和水位的解耦控制研究

针对核电站稳压器压力和水位的耦合现象对控制性能带来的影响,本文通过系统辨识得到加热器和上充阀门对稳压器压力、水位的被控特性数学模型,根据实验现象和耦合原理搭建出压力和水位的耦合数学模型,采用对角矩阵法得出解耦器并简化,在MATLAB/simulink仿真平台上验证基于此解耦器的稳压器压力和水位的解耦控制系统,取得了较好的解耦效果,提高了稳压器的控制性能,有助于核电站运行的稳定性和安全性。     

多用途重水研究堆数字化功率调节系统仿真分析

基于MATLAB/Simulink平台构建了功率调节系统闭环内所有环节的数学模型,采用离散控制理论对简化的系统模型进行理论计算,获得了保持系统稳定的控制器参数范围。采用完整的系统模型在Simulink平台环境下进行仿真分析,确定了合适的采样周期及其最佳的PID控制器参数。仿真结果表明,数字化功率调节系统能有效克服反应性的扰动,并具备良好的随动特性。     

Application of model predictive control strategy based on fuzzy identification to an SP-100 space reactor

In this work, a model predictive control method combined with fuzzy identification, is applied to the design of the thermoelectric (TE) power control in the SP-100 space reactor. The future TE power is predicted by using the fuzzy model identified by a subtractive clustering method of a fast and robust algorithm. The objectives of the proposed fuzzy model predictive controller are to minimize both the difference between the predicted TE power and the desired power, and the variation of control drum angle that adjusts the control reactivity. Also, the objectives are subject to maximum and minimum control drum angle and maximum drum angle variation speed. The genetic algorithm that is effective in accomplishing multiple objectives is used to optimize the fuzzy model predictive controller. A lumped parameter simulation model of the SP-100 nuclear space reactor is used to verify the proposed controller. The results of numerical simulations to check the performance of the proposed controller show that the TE generator power level controlled by the proposed controller could track the target power level effectively, satisfying all control constraints.     

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。     

A PWR hybrid computer model for assessing the safety implications of control systems

As part of the ORNL study of safety-related aspects of control systems, a hybrid computer model was developed to trace the dynamic impact of single and multiple component failures on the control system and remainder of the plant. Since the thrust of this program is to investigate control system behavior, the controls are modeled in detail to accurately reproduce characteristic response under normal and off-normal transients. The balance of the model, including neutronics, thermohydraulics and component submodels, is developed in sufficient detail to provide a suitable support for the control system.The model is being used primarily to address mild to moderate transients that can occur at least partially under action of the non-safety control system. Attention initially focused on overfill events that assumed single or multiple failures of feed valves or the generator low and high level setpoints and trips that regulate the valves. In general, these calculations showed that for single-generator overfeed, water inventory in the affected generator increased to a sufficiently high level to saturate the generator fluid, quench superheat, and inject water into the steam line. Overcooling of the primary side was modest.Other events studied with the model include: insufficient main feedwater cooling induced by a steam generator high level setpoint failing low, potentially drying out the generator; secondary side depressurization induced by turbine bypass valves failing open in loop A or in combination with loop B, at low and high power levels; and steam generator tube ruptures in combination with overcooling incidents.     

蒸汽发生器水位改进分数阶控制器研究

蒸汽发生器在瞬态扰动时存在严重的虚假水位现象,增加了低功率水位控制的难度。为研究蒸汽发生器低功率水位控制问题,利用线性参数变化理论,建立了时变的多胞线性参数变化模型。在此模型基础上,提出了分数阶控制器。依据分数阶微积分理论,设计了串级分数阶PI^λD^μ控制器。根据Oustaloup间接离散化方法实现了分数阶PI^λD^μ控制并对Oustaloup方法进行了改进。研究了在负荷变化时,内环和外环4个阶次参数以及改进算法后2个参数变化对系统控制性能的影响。在不同功率区间,相同负荷变化的情况下,对改进后的串级分数阶PI^λD^μ控制器进行了仿真实验。结果表明,所设计的改进串级分数阶PI^λD^μ控制器能有效抑制干扰,分数阶微积分算子的阶次以及改进的Oustaloup方法引入的系数对控制效果均有一定影响,合理调节参数能明显改善系统的控制性能。     

Development and diversity and defense-in-depth application of ABWR feedwater pump and controller model

This work developed an advanced boiling water reactor (ABWR) feedwater pump and controller model, which was incorporated into Personal Computer Transient Analyzer (PCTran)-ABWR, a nuclear power plant simulation code. The feedwater pump model includes three turbine-driven feedwater pumps and one motor-driven feedwater pump. The feedwater controller includes a one-element/three-element water level controller and a specific feedwater speed controller for each feedwater pump. The performance tests, including step change of dome pressure, feedwater pumps transfer, inadvertent closure of all turbine control valves, and one feedwater pump trip at 100% power, demonstrate the feasibility of dynamic response of stand-alone model and incorporated model. Furthermore, a diversity and defense-in-depth analysis is performed to demonstrate the feasibility for motor-driven feedwater pump as an emergency core cooling system (ECCS) automatic diverse back-up. In Lungmen nuclear power plant (NPP), a diverse manual initiation means for the high pressure core flooder (HPCF) loop C is designed as the back-up of digitalized engineered safety features actuation system (ESFAS). If the motor-driven feedwater pump (MDFWP) can be an automatic digital diverse back-up for ESFAS, Lungmen NPP would be more robust to defend against software common-cause failure (CCF).