基于acslX软件的AP1000稳压器压力控制系统动态仿真

针对AP1000稳压器内部传热传质过程的特点,结合西屋AP1000的相关参数,在上海核工程研究设计院控制系统模型的基础上,对其中稳压器的二区数学模型进行完善和改进,利用acslX软件建立稳压器三区动态数学模型,并严格按照西屋AP1000稳压器的压力控制逻辑,对建立起来的数学模型进行了相应的控制仿真实现。通过比较改进前和改进后模型试验结果与相关设计文件的差异,验证了改进后模型较改进前具有更好的精确性、可扩展性,同时该模型可为今后CAP1400稳压器的仿真工作打下一定基础。     

稳压器汽腔小破口失水事故的模拟

在研究稳压器汽腔小破口失水事故过程物理现象的基础上,对北京核电厂模拟培训中心模拟器中的反应堆冷却剂系统数学模型进行了改进,将两相流模拟分支加入到该系统模拟程序中去,得到了满意的模拟效果。文中给出了改进前、后模拟效果的对比,以及改进的结果与三里岛事故记录的对比。     

稳压器垂直支承锚固件安装超差的分析评估

压水堆核电厂稳压器垂直支承锚固件为一次预埋件，在安装过程中容易产生超差。由于其结构较复杂，评估超差对后续稳压器安装就位的影响较为困难。本文对稳压器垂直支承结构的可调节裕量进行分析，并基于稳压器垂直支承锚固件各套管的实际超差情况，结合稳压器现场安装要求，通过模拟数值计算确定了稳压器的最佳就位位置。根据稳压器支承结构特点，提出了减小现场安装难度的优化建议如下:增加稳压器垂直支承结构中下环板套管孔径，有效增加稳压器垂直支承结构中下环板套管的可调裕量；采用球面垫圈代替稳压器支承结构中采用的平垫圈，从而放宽稳压器支承结构对地脚螺栓安装的垂直度要求。      

稳态条件下稳压器喷淋降压模型及数值分析

本文基于非稳态球形传热模型,在单液滴与饱和蒸汽之间的传热模型的基础上,根据稳压器喷雾流量、雾化颗粒特性参数及反应堆一回路系统参数,建立了反应堆一回路系统稳压器的喷淋降压动态特性的数学模型。根据某核电机组热态功能试验时的试验条件,使用FORTRAN语言编写程序,选用迭代的方法,对稳压器喷淋降压的动态特性进行了数值模拟分析计算。计算结果表明,通过数值计算得到的喷淋降压速率与实际试验结果符合性较好。     

压水堆稳压器数学模型优化及动态仿真

根据压水堆稳压器的特点及性能要求,对稳压器数学模型进行优化。对一些热工参数与温度的关系进行线性化处理,得到稳压器的优化数学模型。基于MATLAB环境进行稳压器动态仿真研究。通过仿真研究,验证模型优化的准确性和动态仿真的实时性,实现压力安全系统动态仿真。     

核电站培训仿真器稳压器模型的改进

本文的改进模型采用两相非平衡态模型,考虑了稳压器内汽泡上升、液滴下降、喷淋凝结等各种传热传质过程。本文给出的模型与实验数据比较,吻合程度很好,并能满足核电站仿真器实时计算的要求。     

小型堆包壳破损状态下利用稳压器去除放射性研究

利用MELCOR程序对小型船用堆稳压器喷雾除气过程及停堆过程进行建模,进而模拟核动力装置从功率运行至降功率除气,以及除气结束后停堆消除稳压器气腔的全部物理过程。通过对反应堆关键运行参数变化趋势的仿真分析,验证了模拟的物理过程的合理性。结合建立的除气及停堆仿真模型,计算分析了包壳破损状态下,稳压器喷雾除气、停堆过程对稳压器内惰性气体含量的影响,评估了稳压器高点放气和喷雾除气对放射性物质的去除作用。研究结果能为小型堆包壳破损状态下放射性安全管理策略提供指导和帮助。     

TRANTH软件基于稳压器安全阀组流量试验的确认与评估

热工水力瞬态分析软件TRANTH用于分析核电厂安全性,其中,考虑了两区质量守恒和能量守恒的关键模型之一稳压器模型可对稳压器安全阀、释放阀、电加热器、喷淋和相关系统进行模拟。在软件开发完成后需进行相关软件验证,故结合方家山核电厂1号机组稳压器安全阀流量试验数据和软件模拟结果,验证稳压器模型。结果表明,模拟计算结果与现场试验数据符合度高,模型精度满足工程设计要求。      

核电厂稳压器水封结构数值模拟研究

建立了含不凝性气体的气-液两相流三维计算流体力学(CFD)模型,运用Fluent软件对核电厂稳压器新型水封结构流场进行模拟,通过水密封建立过程中稳压器压力和不凝性气体含量的影响分析,研究了水密封建立过程的热工特性。结果表明,对于新型水封结构,水密封建立时间随稳压器压力的增大而缩短,随不凝性气体含量的增大而增长。     

核电站稳压器压力非自衡系统的预测控制

针对稳压器压力具有大惯性、多干扰、复杂非线性、难以获得精确数学模型的特点，以及稳压器压力开环不稳定的动态特性，本文采用非自衡系统的模型自衡化进行动态矩阵预测控制器设计，获得控制信号输出以解决该非自衡系统预测控制的工程实现问题，并以MATLAB/Simulink为仿真平台搭建控制系统。仿真对比和扰动测试表明，该控制系统具有良好的控制性能和抗扰能力，同时为动态矩阵控制算法适用于非自衡系统提供了一种可行方案。      

Application of artificial intelligence techniques in modeling and control of a nuclear power plant pressurizer system

In pressurized water reactor (PWR) nuclear power plants (NPPs) pressure control in the primary loops is fundamental for keeping the reactor in a safety condition and improve the generation process efficiency. The main component responsible for this task is the pressurizer. The pressurizer pressure control system (PPCS) utilizes heaters and spray valves to maintain the pressure within an operating band during steady state conditions, and limits the pressure changes during transient conditions. Relief and safety valves provide overpressure protection for the reactor coolant system (RCS) to ensure system integrity. Various protective reactor trips are generated if the system parameters exceed safe bounds. Historically, a proportional-integral-derivative (PID) controller is used in PWRs to keep the pressure in the set point, during those operation conditions. The purpose of this study is two-fold: first, to develop a pressurizer model based on artificial neural networks (ANNs); secondly, to develop fuzzy controllers for the PWR pressurizer modeled by the ANN and compare their performance with conventional ones. Data from a 2785 MWth Westinghouse 3-loop PWR simulator was used to test both the pressurizer ANN model and the fuzzy controllers. The simulation results show that the pressurizer ANN model responses agree reasonably well with those of the simulated power plant pressurizer, and that the fuzzy controllers have better performance compared with conventional ones.     

A linearized non steady state model for the pressurizer of the safe integral reactor concept

A linearized model for the pressurizer of the Safe Integral Reactor (SIR) is developed. The mathematical model derived from the general conservation equations includes all the important thermal-hydraulic processes occurring in the pressurizer. These processes are spray condensation, interfacial condensation, vapour rise, condensate fall, and heat tranfer from heaters. Also, the model takes into account some special functional features of the passive spray and the fluidic diodes. The insurge and outsurge events are modelled separately and the pressurizer response to insurge and outsurge flows are compared for the SIR, the H. Robinson, and the Shippingport reactors. It is concluded that the SIR passive spray concept is not as favourable as valved spray in inverting the pressure rise during an insurge flow.     

基于多区非平衡模型的稳压器动态仿真与验证

采用FORTRAN程序设计语言,基于稳压器多区非平衡模型和吉尔数值求解方法,开发了稳压器动态仿真程序NEP MR。计算结果与美国希平港核电站稳压器74MW和105MW甩负荷的试验结果进行对比,总体趋势符合良好。NEP MR程序可较好地反映瞬态甩负荷过程中稳压器内部控制体的温度分层现象。在稳压器喷淋器开启和关闭瞬间,汽相和液相控制体之间的界面流量出现突跳现象。计算结果证明了多区非平衡模型建模和NEP MR程序仿真的合理性。     

Benchmark calculations of pressurizer model for Maanshan nuclear power plant using TRACE code

The pressurizer plays an important role in controlling the pressure of the primary coolant system in pressurized water reactor (PWR) nuclear power plants. An accurate modeling of the pressurizer is needed to determine the pressure response of the primary coolant system, and thus to successfully simulate overall PWR nuclear power plant behavior during transients. The purpose of this study is to develop a pressurizer model, and to assess its pressure transients using the TRACE code version 5.0. The benchmark of the pressurizer model was performed by comparing the simulation results with those from the tests at the Maanshan nuclear power plant. Four start-up tests of the Maanshan nuclear power plant are collected and simulated: (1) turbine trip test from 100% power (Test PAT-50); (2) large-load reduction at 100% power (Test PAT-49); (3) net-load trip at 100% power (Test PAT-51); and (4) net-load trip at 50% power (Test PAT-21). The simulation results show that the predictions of the pressure response are in reasonable agreement with the power plant's start-up tests, and thus the pressurizer model built in this study is successfully verified and validated.     

基于LSSVM的核电厂稳压器压力控制系统隐蔽攻击方法研究

隐蔽攻击对大型核电厂稳压器压力控制系统的安全、稳定、高效运行构成了严重威胁,实现隐蔽攻击的关键是建立高精度的对象估计模型。本文提出了一种基于最小二乘支持向量机（LSSVM）的隐蔽攻击方法,通过LSSVM算法进行系统辨识,获得稳压器受攻击区域高精度的估计模型,随后利用该估计模型结合隐蔽控制器实施攻击,实现稳压器压力控制系统在无噪声、有噪声及包含非线性环节情况下的隐蔽攻击。仿真结果表明,该攻击方式对稳压器压力控制系统造成一定破坏的同时具有高度的隐蔽性。      

核动力稳压器多区非平衡模型

提出了核动力稳压器多区非平衡模型，运用该模型对希平港核电站稳压器的动态特性进行了仿真。仿真结果与试验数据符合良好，证明了本文所给出的稳压器模型的正确性。     

Development of nonequilibrium pressurizer model with noncondensable gas

To investigate thermal–hydraulic characteristics of a steam–gas pressurizer in the integral type reactor, the steam–gas pressurizer model based on the two-region nonequilibrium concept was developed and introduced into RETRAN-3D/INT code. The model includes an explicit solution method for the one-dimensional governing equations and the equation of the state solution method to determine the thermal–hydraulic state of the steam–gas pressurizer volume. In addition, the wall condensation model based on the diffusion layer modeling was included to consider the effect of the noncondensable gas. The developed model was verified with the results from the pressurizer insurge experiment conducted at Massachusetts Institute of Technology. From the verification results, it was concluded that the developed steam–gas pressurizer model can sufficiently predict the pressurizer transient and it can be used as a component model of the one-dimensional system code based on the homogeneous equilibrium model.     

非凝性气体对汽-气稳压器稳压的影响研究

通过分析相间的传热传质过程以及非凝性气体存在时壁面蒸汽冷凝过程,建立了汽 气稳压器模型,研究了非凝性气体对稳压过程的影响,描述了稳压器的稳压特性,并将模型计算结果与MIT稳压器实验数据进行了对比。结果表明：当不含非凝性气体时,计算精度高,相对偏差在0.8%内,压力峰值为0.647 MPa;当非凝性气体含量从0增至20%时,计算精度相对减小,最高相对偏差为15.4%;压力峰值从0.647 MPa增至1.02 MPa。研究表明非凝性气体对稳压器稳压过程具有重要影响作用,随着非凝性气体的种类和含量的变化,稳压器内稳压过程发生显著变化。