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。     

核电站稳压器水位控制系统数字化的实现及优化方案

根据模拟稳压器水位控制系统结构、功能以及运行原理的分析,展开对稳压器水位控制系统数字化的研究。在选择成熟商用单板机的基础上引入了数/模,模/数转换以及控制算法等实现了稳压器水位控制系统,并和稳压器水位仿真模型构成闭环运行,通过虚拟仪器监测此控制系统的运行情况,证实了本系统达到并超过了原有模拟控制系统的功能和性能。     

基于可编程控制器的稳压器压力水位控制研究

对采用可编程控制器构建稳压器压力水位控制系统进行了研究.利用可编程控制器的功能模块、模块化结构、配置的灵活性和软件控制技术,实现了稳压器压力水位控制系统的功能.通过对实验结果的分析,证明可编程控制器应用于稳压器压力、水位控制是有效的、成功的.     

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。     

基于遗传算法的压水堆核电一回路稳压器机理建模与仿真

针对压水堆核电站一回路稳压器实际运行特性,根据能量守恒、质量守恒和动量守恒方程,考虑了喷淋流量、电加热器功率及安全释放阀的影响,建立了一个两相动态非平衡的稳压器机理模型.为提高模型精度,采用遗传算法对该模型的参数进行优化,用以得到一组模型的最优参数.将参数优化算法应用于某900 MW核电站稳压器仿真实例,通过模块封装组建成稳压器压力仿真模型,并与核电厂提供的对应数据做了比较验证了建模方法的正确性及优化方法的有效性.     

核电站稳压器水位控制系统模件偶发故障原因分析及改进建议

大亚湾核电站稳压器水位控制系统在运行中曾多次出现水位定值组件输出不可忽略的脉冲干扰,而且干扰只存在于模拟量存储操作组件上,严重时直接影响到稳压器水位控制,引起水位波动.本文从系统控制和模件本身硬件设计两个方面对核电站稳压器水位控制系统模件偶发干扰故障原因进行分析,采用电源回路测试及系统仿真模拟试验,找出故障原因.建议再...     

稳压器的模型化与仿真

本文提出了完善的核电站稳压器三区不平衡模型及相应的PZR-1动态仿真程序。程序分别采用了亚当斯(Adams)方法、哈明(Hamming)方法、龙格-库塔(Rang-Kutta)法和外推法等不同算法,仿真精度高、速度快。以希平港核电站减负荷试验时稳压器的动态特性对模型和程序进行验证,仿真结果与试验数据符合良好,结果同时表明用亚当斯方法进行仿真更有效。     

岭澳核电站蒸汽发生器水位控制系统改进方案仿真研究

利用核电厂瞬态分析和控制系统优化设计程序CATIA2,依据各典型瞬态试验验收的不同准则,通过核电厂典型瞬态下的数值仿真试验,为岭澳核电站解决主给水流量系统管路及调节阀振动问题拟采用的蒸汽发生器给水控制系统改进方案进行了仿真验证。结果表明:该改进方案可行,并且对反应堆控制系统所带来的影响是可接受的。     

核动力装置稳压器安全阀可靠性建模

基于对核动力装置稳压器安全阀系统功能及任务剖面的分析,建立了基本可靠性框图和任务可靠性框图,并采用普通慨率法得到了相应的可靠性数学模型.该可靠性模型为下一步进行故障模式、影响及危害性分析(FMECA)和可靠性分配提供依据.     

稳压器模糊控制系统初步研究

通过合理地确定模糊集、隶属度函数和模糊控制规则等。初步研究和设计了核电站稳压器的模糊控制器。仿真结果表明,与传统的比例－积分－微分控制器相比,模糊控制器在提高稳压器压力、水位控制系统的动态和稳态性能方面具有明显的优点。     

Modeling and simulation of cogeneration nuclear power plant for seawater desalination

Nuclear desalination of seawater remains a very viable option to solving the perennial fresh water shortage problem along the coast of Ghana especially as Ghana prepares to install the first nuclear power plant. There is, therefore, the need for research to be conducted into nuclear seawater desalination technology as part of the nuclear power program of Ghana so as to develop the needed human resources in Ghana. In this research, cycle analysis of the cogeneration nuclear power plant was conducted to determine its efficiency and desalination steam requirements. An analytical model of the thermo vapour compression (TVC) desalination process was also developed to investigate the effect of design and operating parameters on parameters controlling the cost of producing fresh water from TVC process. Steady state mass and energy balances as well as empirical correlations derived from experiments were used to model the TVC, which was coupled to a cogeneration nuclear power plant to supply the needed steam for the desalination. The model was developed on a computer code, using FORTRAN language. The results showed that the thermal performance of the TVC desalination process improves with the efficiency of the cogeneration nuclear power plant but decreases with increasing steam consumption rates.     

压水堆核电站堆芯集中参数模型的微机仿真

阐述了PWR核电站堆芯的模型化问题,提适用于微机仿真的核电站堆芯的物理数学模型,将核电站堆芯分为三大块分别建立模型,中子动力学模块,反应性反馈模块,堆芯热力学模块,建立系统传递函数,运用MATLA仿真,得到良好结果。     

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.     

大型压水堆核电站换料机的强度分析

压水堆核电站换料机对保障核电站安全运行具有重要的作用,对其主要结构的动力计算和强度评定具有重要的意义。本文应用有限元分析软件ANSYS 12对1 000 MW核电站大型换料机进行了有限元建模,并分别在正常工况(启动、制动)、异常工况(OBE)和事故工况(SSE)下进行了动力计算;采用SRSS方法对3个不同方向地震反应谱下的结构响应(内力、应力)进行了工况组合,并进一步考虑了自重条件的不利影响。根据RCCM规范对换料机主要结构、螺栓、焊缝的强度和辅吊支腿的稳定性进行了评定,并在此基础上对抓取燃料组件的指形钩进行了局部强度分析。评定结果表明换料机的强度在不同工况下均满足规范要求。     

大型先进压水堆核电站堆芯装载方案设想

核电站堆芯装载方案是反应堆堆芯设计的重要基础,它首先必须满足核安全的要求,同时还要尽可能地提高经济性。通过分析国内、外百万千瓦级核电站的堆芯装载,对反应堆输出功率、燃料组件数、堆芯平均线功率密度进行比较,给出我国大型先进压水堆核电站示范工程反应堆堆芯装载方案的设想,为技术决策提供参考。     

压水堆核电站燃料传输系统抗震性能有限元分析

压水堆核电站燃料传输系统的主要功能是将核燃料组件在反应堆厂房和燃料厂房之间进行相互传送.为保障其结构在不同工况下能够安全运行,需对其在异常工况(OBE)和事故工况(SSE)下进行抗震性能分析.为此,采用有限元分析软件ANSYS 12.0对某百万千瓦级核电站的燃料传输系统进行了结构建模;对传输系统在运行过程中不同构型下的转运通道、承载器、传输小车、反应室倾翻架、燃料室倾翻架、燃料室轨道结构和反应室轨道结构等主要构件在不同工况下进行了应力分析;采用SRSS方法进行了应力组合,并依据RCCM规范对各构件进行不同工况下的强度评定.结果表明,在异常和事故工况下,燃料传输系统在不同构型下各结构均满足强度要求,具有良好的抗震性能.     

Modeling and optimization of a nuclear power plant secondary loop

Increasing the thermodynamic efficiency of fossil fuel or nuclear power plants can lead to significant economic gains. Consequently, the continuous quest of searching higher efficiency in power plants has resulted in the development of innovative tools to comply with these needs. Although a large inventory of simulation tools is available for industrial applications, sometimes it is more appropriate to develop in-house models that are best suitable for treating specific energy systems. In the present work, a combined simulation-optimization tool was developed and used to optimize the secondary loop of a CANDU-6 nuclear power plant (i.e., Gentilly-2 nuclear station). Based on previous studies an optimizer module has been coupled with a thermodynamic model, written in Matlab, used as a plant simulation tool. It includes models that take into account the responses of major thermal units, i.e., condenser, moisture separator reheater and feedwater heaters. The simulation package is used to estimate the behavior of the power station according to the variation of a given number of plant operation parameters. The methodology permits a set of best trade-off operating conditions of the secondary loop to be determined, and thus providing a better and more realistic support to plant operators. The results also clearly show that there is plenty of potential to improve the overall performance of the power station.