热离子空间堆控制系统研究

空间堆与传统地面反应堆的运行特性和控制方式不同，研究空间堆的控制系统十分必要。基于Matlab&Simulink工具箱，建立了热离子空间堆仿真模型。建立的模型包括点堆中子动力学模型、热工水力模型和热电转换模型。基于热离子空间堆仿真模型，建立了核功率、电功率以及冷却剂出口温度控制系统，并进行了控制参数整定和仿真分析。对电功率和冷却剂出口温度控制系统进行了优化，设计了串级控制系统。仿真结果表明：选用核功率作为被控量的控制系统控制效果较好；选用电功率和冷却剂出口温度作为被控量的控制系统控制效果较差；核功率的超调量较大。使用串级控制器优化后，控制性能得到了显著的提升。     

AP1000乏燃料池冷却系统的仿真实现

三门核电站采用AP1000全范围模拟机,所有电站模型按照系统类型和建模工具的不同可以划分为堆芯系统、一回路热工水力系统、气液两相流体系统、气体或者液体单相流体系统、仪控系统以及电气系统。本文主要介绍AP1000电站流体系统的建模原理和仿真过程,以AP1000乏燃料池冷却系统为例,使用GSE公司的图形化建模工具JTOPMERET对建模和仿真过程进行说明。     

自动控制系统数字仿真软件包

本文介绍一种按环节离散化的自动控制系统数字仿真软件包。软件包可以对典型环节仿真,还可以由用户根据实际系统组态后进行仿真。该软件包尤其适用于热工自动控制系统的数字仿真,对锅炉给水控制系统和汽温控制系统可直接进行仿真。本文对自动控制系统的优化设计、工业现场控制参数整定具有一定的指导意义。     

基于微机的200Wm核供热堆仿真系统PCNHR1.0的研究

为了使更多的技术人员形象地了解低温堆供热站的原理 ,在国际原子能机构的支持下 ,清华大学核研院研究开发了基于微机的 2 0 0Mw低温核供热堆仿真系统。它采用两回路、一维漂移流热工水力学模型 ,点中子堆物理以及控制系统模型。能对核供热堆的稳态运行、瞬态过程和事故进行仿真 ,仿真精度接近系统分析结果。在奔腾或以上的微机上 ,WIN DOWS95 /98/NT操作系统下 ,能对过程进行实时仿真 ,而且大多数过程能达到 10倍以上超实时。软件采用VisualC+ + 和FORTRAN混合语言编程 ,运用先进的多线程编程模型、进程通讯和进程控制技术实现了耦合计算和同步控制。     

基于微机的200Wm的核供热堆仿真系统PCNHR1．0的研究

为了使更多的技术人员形象地了解低温堆供热站的原理，在国际原子能机构的支持下，清华大学核研究研究开发了基于微机的200M低温核供堆仿真系统。它采用两回路、一维漂移热工水力学模型，点中子堆物理以及控制系统模型。能对核供热堆的稳态发行、瞬态过程和事故进行仿真，仿真精度接近系统分析结果。在奔腾或以上的微机上，WINDOWS95／98／NT操作系统下，能对过程进行仿真实仿真，而且大多数过程能达到10倍以上超实时，软件采用VisualC 和FORTRAN混合语言编程，运用先进的多线程编程模型，进程通讯和进程控制技术实现了耦合计算和同步控制。     

基于热工水力程序cosFlow的消防系统仿真

AP1000中消防系统是核电站中的一个子系统,其主要作用包括探测火灾及灭火,为失效时CCS(设冷水系统)中的热交换器提供冷却和安全喷淋等.上述系统主要由火灾探测和报警设备、消防水供给设备、自动和手动灭火设备等组成.通过COSINE中的热工水力程序cosFlow对其进行建模仿真,模拟了三种不同稳态运行工况下的消防系统,分...     

紧凑型船舶动力定位控制半物理仿真系统研究

根据紧凑型船舶动力定位控制研究及设计的需求,设计了对应的半物理仿真系统。该系统建立船舶运动及海洋环境计算机模型模拟器、全尺寸控制器和全尺寸操作终端,并通过基于DirectX的三维动画模拟器进行船舶三维运动效果展示。全尺寸控制器与船舶上实际使用的控制器在网络接口、信号输入/输出接口上完全一致,以便提前进行船舶动力定位控制系统半物理仿真及验证,降低了调试成本和风险,提高在水池及海洋进行实物调试的效率。该仿真验证平台考虑了船舶动力定位控制系统实际运行环境,引入了外载荷模型,提高了仿真验证的可信度。文中给出了仿真实例,验证了系统的有效性。     

热电联产压水堆核电机组的控制方案研究

核能热电联产是实现“双碳”目标的重要方法。核能热电联产改造后，供需侧能量平衡关系发生了变化，需要针对热电联产改造后压水堆核电机组开展堆机协调控制方案的研究。首先，以大亚湾900 MW核电机组为研究对象，基于最佳估算程序RELAP5和仿真平台3KeyMaster建立了热电联产模型。热电联产形式包括主蒸汽集管抽汽、中间抽汽和背压机排汽。接着，以主蒸汽集管抽汽为例，通过典型扰动的动态特性分析并验证了所建立的模型。然后，提出了将总蒸汽流量折算为等效负荷的方法，对反应堆功率控制中冷却剂平均温度控制系统进行改造。最后，对设计的控制系统改造方案进行了验证。验证结果表明，该方案能够满足反应堆核功率、电功率和供能热负荷协调控制的要求。所设计的控制改造方案为热电联产压水堆核电机组的反应堆功率控制提供了参考。     

基于粒子群算法的1000MW火电机组模型辨识

对于火电厂热工控制系统,建立精确的热工过程模型是保证控制质量的基础.将传统的阶跃响应法用于热工过程模型辨识,由于现场数据的不规则性使得阶跃响应法的经验结果精度不高.针对火电厂热工过程对象的特点及传统模型辨识的缺陷,将粒子群算法用于火电厂热工过程模型的辨识,以1000MW超超临界机组的风煤比作为控制对象,对空预器进口氧量变化系统进行系统辨识,辨识曲线能够很好的反应实际输出曲线,证明了该方法的有效性和可靠性,相比传统的辨识方法,将粒子群算法用于模型辨识提高了辨识的精确性与快速性.     

基于组态软件和仿真软件的层流冷却实验平台

针对热轧层流冷却过程的自动化系统及优化控制算法难以在实际运行环境中进行充分的调试和验证、基于仿真软件发的实验平台则脱离工业实际的监控与运行环境、购买硬件控制系统进行实物/半实物仿真价格昂贵等问题,本文提出了采用组态软件、工业控制器模拟软件、仿真软件混合编程开发层流冷却虚拟实验平台的方法。该平台采用C#语言开发参数优化子系统、组态软件开发管理监控子系统、工业控制器模拟软件运行回路控制程序、MATLAB软件开发对象模型子系统,通过OPC软总线等技术实现了各个子系统的无缝连接。本文最后采用实际工业数据验证了平台的有效性,表明其广阔的应用前景。     

Nonlinear observer based control for travelling wave nuclear reactors based on the Lyapunov approach during load following operation

Power control of the nuclear reactor is one of the most important subjects in each nuclear power plant. In this paper, a nonlinear controller using sliding mode method which is a robust nonlinear controller is designed to control a Traveling Wave Nuclear Reactor (TWR) power. The reactor core is simulated based on the point kinetics equations and six delayed neutron groups. Considering the limitations of the delayed neutron precursors densities measurement, a sliding mode observer is designed to estimate their values and finally a sliding mode control based on the sliding mode observer is presented to control the reactor core power. The stability analysis is given by means Lyapunov approach, thus the control system is guaranteed to be stable within a large range. Sliding Mode Control (SMC) is one of the robust and nonlinear methods which have several advantages such as robustness against matched external disturbances and parameter uncertainties. Since it has systematic design procedure, it is one of the most powerful solutions to design many practical control systems. The designed control system is evaluated in the presence of disturbances and uncertainties. The results show the robustness and performance of the used control system.     

A case study and troubleshooting experience of safety power channel of instrumentation and control system of the BAEC TRIGA Research Reactor

Microsystem Technologies - Bangladesh Atomic Energy Commission TRIGA Research Reactor (BTRR) is a MK II type nuclear research reactor with a maximum thermal output power of 3 MW. Nuclear...     

核电站数字化反应堆保护系统中央处理器负荷率分析与测试

核电站对数字化反应堆保护系统的一书央处理器的负荷率有严格要求。本文首先对核电站数字化反应堆保护系统中央处理器的负荷率进行了理论分析,得出了负荷率计算公式;然后设计了相应的负荷率测试方法0j测试装置,完成了实际的测试工作;对测试所得实验数据进行处理,得出测试结果,结果表明数字化反应堆保护系统的中央处理器负衍率符合技术嘤求,且主控CPU的负荷率比备用CPU负荷率要高。     

大亚湾核电站堆芯参数测量（RIC）系统的国产化改造

RIC（堆芯测量）系统是核电站的重要测控系统之一,其部分功能是在KIT（集中数据采集处理）系统中实现的。大亚湾核电站近期对其原有KIT系统进行了全面的国产化升级改造,与之关联的RIC系统也同时进行了改造。该文介绍RIC系统的技术改造方案、构成和配置、工作原理和主要功能,着重描述了核反应堆中子注量率的采集和数据处理过程。使用效果表明,新RIC系统的技术指标完全满足工艺要求。     

核动力堆热功率跟踪系统的动态矩阵控制研究

动力堆控制系统是一个带不确定参数及干扰的复杂非线性系统 ,采用常规的古典控制很难保证其热功率精确跟踪负荷的变化。本研究工作利用系统单位阶跃响应序列建立了一种非参数模型 ,并应用动态矩阵控制原理提出了一种动力堆热功率跟踪数字控制新方法。该方法跟踪调节性能好、鲁棒性强、能消除不可测干扰。通过仿真检验和调试证明了该方法的正确性和有效性 ,并实现了热功率—负荷的高精度匹配     

Visual simulation of thermal fluid dynamics in a pressurized water reactor

We present a simulation and visualization system for a critical application—analysis of the thermal fluid dynamics inside a pressurized water reactor of a nuclear power plant when cold water is injected into the reactor vessel. We employ a hybrid thermal lattice Boltzmann method (HTLBM), which has the advantages of ease of parallelization and ease of handling complex simulation boundaries. For efficient computation and storage of the irregular-shaped simulation domain, we classify the domain into nonempty and empty cells and apply a novel packing technique to organize the nonempty cells. This method is implemented on a GPU cluster for acceleration. We demonstrate the formation of cold-water plumes in the reactor vessel. A set of interactive visualization tools, such as side-view slices, 3D volume rendering, thermal layers rendering, and panorama rendering, are provided to collectively visualize the structure and dynamics of the temperature field in the vessel. To the best of our knowledge, this is the first system that combines 3D simulation and visualization for analyzing thermal shock risk in a pressurized water reactor.     

核电厂模拟机电气系统建模与仿真

核电厂全范围模拟机电气系统数量众多导致仿真计算量巨大,设计参数的缺失影响仿真精度,系统数据接口之间的数值传输需要实时准确.为提高仿真计算效率、精度和保证实时数据传递,阐述了系统仿真策略和参数分类存取方法;基于电动机、静态负荷及低压开关模型和仿真,重点讨论了模块参数的反算功能和接口数据实时传递原理.以某1000 MW核电厂电动机及电气系统主接线为例,对仿真模型和方法进行了分析和验证.仿真结果证明:采用电气系统仿真策略和仿真方法,有效减少了全范围模拟的电气系统数量,对设备参数进行分类存取,提高了仿真计算效率;关键模型参数的自动反算和系统设备接口数据的及时准确传输,保证了计算的简洁性和数据传输的实时性;同时模型的仿真精度满足设计要求.仿真模型方法及开发工具在核电厂模拟机电气系统建模与仿真中具有广泛的适用性.     

Self-tuning control application to a nuclear power plant

The aim of this paper is to investigate the application possibilities of a self-tuning control method to a pressurized-water reactor (PWR) nuclear power plant. A self-tuning control algorithm which incorporates pole assignment into the generalized minimum variance strategy with a particular form of cost function is employed. This algorithm enables closed-loop system stability characteristics to be readily specified and facilitates reference following. The control system design is based on a second-order linear model with unknown, time-varying parameters. To ensure that this low-order model describes the complex real dynamics well enough for control purposes, control parameters are updated on-line with a recursive estimation sequences of the extended least-squares method. Weighting polynomials are also adjusted on-line to keep closed-loop poles at the desired location and to satisfy the zero steady-state condition and disturbance rejection. The purpose of this control system is to hold the average coolant temperature in the reactor as near as possible to a desired but changing reference value in the load-following mode of the nuclear power plant. The position of the control rods is an appropriate control variable. Simulation results are very successful and show the possibilities of the adaptive control application to actual plants.     

A fuzzy rule base for the improved control of a pressurized waternuclear reactor

In France, nuclear energy provides about 80% of the whole electricity production. A modulation of the nuclear power plants must be able to respond to the demand on the network. The pressurized water nuclear reactor has to yield correctly a load set point. Fundamentally, two parameters are concerned in leading this task to a successful conclusion: the power axial-offset and the control rods position. The focus of this study is the automation of the control of the power axial-offset by adding soluble boron and by minimizing the volume flows through the water pump. It is also important to take into consideration the liquid waste volume. Water or boron is injected into the reactor primary circuit. At the present time this task is still performed manually by an operator, for all previous attempts to automate it failed. A nonfuzzy device, earlier developed by Electricite de France, was intensively tested at Cruas, France, power plant and allowed us to prove the feasibility of automating the boration-dilution function. But it could not be definitely adopted because it was too difficult to be tuned for industrial purposes. That device, sketchily described in the paper, gave rise to the development of a real-time fuzzy controller for the power axial-offset and the control rods insertion in a pressurized water reactor (PWR). The fuzzy controller, which is the main subject of the paper, expresses more naturally the human expertise, thus avoiding the previous issue of empirical tunings. It was implemented in simulation using Matlab-Simulink on a Sun workstation. Two realistic tests discussed show that the fuzzy controller runs as efficiently as an expert operator does