Development and application of an extensible engineering simulator for NPP DCS closed-loop test

LinAo Nuclear Power Plant (NPP) Phase II is a newly-built CPR1000 reactor in China, and many new technologies including the incorporation of digital control system (DCS) substituting traditional analog control systems have been applied. This is the first time for Chinese engineers to setup and adjust the DCS configurations. Both the lack of the operating experiences and the plant safety requirements from the government make a necessity of the closed-loop DCS test before commercial plant operation. The most practical way is to build a digital plant as the controlled target and this digital plant is used to provide the plant thermal–hydraulic parameters and feedbacks for the DCS. Though the RELAP5 code has been developed for the best-estimate transient simulation of light water reactor coolant systems and is used worldwide, its functionality is too limited to implement a digital plant, such as the simulation of the complicated plant control and protection systems, the 3-dimensional neutron kinetics and the fluid network for the plant auxiliary systems. To overcome these drawbacks, a RELAP5-based extensible simulator has been built to satisfy the new requirements for the implementation of a digital plant. Any simulation code of desired functionality can be integrated into this simulator as a simulation module once it applies a set of well-defined data exchange interfaces. At the present stage, a RELAP5 module, a control system modeling module, a software–hardware data bridge module and some other auxiliary modules have been integrated into the simulator. There are more than 60 systems that need to be tested with the DCS in LinAo Phase II, and the whole testing work is separated into several phases. In this paper, we take the testing of the pressure control system and water level control system of pressurizer as example. A typical transient of 10% load step change from 100%FP (full power) to 90%FP was performed for the closed-loop DCS test. The necessity and capability of this RELAP5-based engineering simulator has been demonstrated.     

Quality assurance for a nuclear power plant simulator by applying standards for safety-critical software

Nuclear power plant simulators are playing a more important role in nuclear power plant lifecycle analysis, and the quality of the simulators should be verified to ensure the safety of nuclear power plants. Currently, there is no systematic quality assurance method for nuclear power plant simulators. In this paper, a systematic quality assurance method for nuclear power plant simulators is proposed basing on experiences with safety-critical software. Key aspects of the method are discussed. In addition, application of this method to a real project is also described as a practical reference.     

核电站仪控系统数字化开发仿真测试技术研究

在核电站应用数字化仪表与控制 (I&C)取代模拟 I&C 系统,已成为必然的发展趋势。本文分析了核电站全范围模拟机的蒸汽发生器数学模型,研制开发独立的核电站蒸汽发生器实时仿真系统,并与控制系统形成能够相互作用的闭环系统,用于数字化仪控系统改造提供仿真对象及进一步控制方案研究。在仿真过程中,除了仿真模型之外,其他的硬件和软件由真实的控制系统构成。不但为核电站仪表与控制 (I&C)系统数字化开发提供理论分析,也为今后现场调试工作创造有利条件。     

RELAP5作为核电站模拟器热工水力系统程序的改造

RELAP5程序由于其非实时计算、无动态输入输出功能以及计算流程难以控制等原因．不适合作为核电站模拟器的热工水力系统程序、RELAPSIM程序在RELAP5基础上经过实时计算功能改造、数据动态交互功能改造、计算流程控制功能改造后,能够完成实时热工水力计算,数据动态交互以及启动、停止、冻结、运行、快照、复位计算流程等功能,满足了作为核电站模拟器的热工水力系统程序的要求。本文主要介绍了RELAP5程序的改造方法和原理以及改造后的RELAPSIM程序测试和结果。     

核电站紧凑型工程模拟器开发

以恰西玛(chashma)核电站为背景，研制开发了核电站工程模拟器。该工程模拟器由仿真计算程序、数据通讯模块及人机界面3部分组成。本文从计算机系统、热工水力模型、程序设计语言，人机界面设计及数据通讯5个方面详细介绍了该工程模拟器的设计方案及具体实现过程。     

实时风险监测系统在田湾核电站的研发与应用

风险监测系统是概率安全分析最重要的应用工具之一,在全球各核电站风险决策领域中得到了广泛应用。本文详细介绍了田湾核电站风险监测模型的研发与管理,系统网络架构与多用户应用、实时监测研发与应用、计划自动导入和计算时效性提升等重要特征,以及在电站的针对性应用方案研究,最后讨论了风险监测系统存在的问题与挑战。     

核动力装置运行过程可靠性研究现状与发展

核动力装置运行过程的可靠性研究是保障装置安全，提高装置效能的重要基础，本文从运行过程可靠性问题的背景出发，研究了可靠性分析方法的适应性，并论述了问题研究的技术途径。     

基于RELAP5的核动力装置故障诊断数据源研究

为解决故障状态下的核动力装置数据源问题,本文建立了核动力装置一、二回路系统的模型,选择秦山一期核电站为对象,利用RELAP5对蒸汽发生器U型管破裂进行计算.通过结果分析可知所建立的模型节点划分是合理的、数据卡编制准确,基于该模型产生的数据可信.将开发的数据与基于神经网络的故障诊断系统联调,诊断测试结果表明数据准确、充分,可以为核动力装置的故障诊断系统的研究提供数据支持.     

核电站DCS仿真的三种实现形式探讨

目前国际上通用的核电站模拟机DCS仿真主要有纯模拟(Simulation)、虚拟实物模拟(Emulation)以及实物模拟(Stimulation)这3种实现途径,在实际项目选型时需要从经济性、精确性、变更易维护以及对工程项目的时间进度影响等方面综合考虑。本文具体考虑了上述3种仿真实现途径的优缺点,并根据不同项目特点给出选型建议。     

A framework for nuclear power plant emergency response system

<正>The purpose of this study is to establish an intelligent expert system for nuclear power plant emergency response.A new framework of environmental risk management methodology by the concept of pattern recognition was introduced in this paper.A knowledge-based decision support system for emergency response and risk management of nuclear power plant was also discussed.The mathematical pattern relationship of accidental release effects on neighboring area and the corresponding response measures were presented in this paper.With this decision system,the decision maker can specify the procedure and minimize their human error in the decision process.The improvement of risk response and the quality of management system could be upgraded by this system.Besides,the methodology can also be served as a basis for the future development of environmental risk response system design.     

核电站模拟器在三里岛事件后的新要求和新发展

本文介绍了美国三里岛事件后核电站控制室设备的改进对核电站模拟器的新要求、模拟器本身在三里岛事件中所暴露的不足及随后模拟软件的新发展,也简略地介绍了目前模拟器在世界范围的发展动向。     

核安全文化的发展与应用

安全文化已对核能企业的安全性产生了重大影响。本文分析了核安全文化产生的背景，介绍了核安全文化在一些国家和组织应用发展的状况，提出了推行安全文化过程中应注意的 上问题，讨论了评价安全文化绩效的原则。     

Fuzzy probability on reliability study of nuclear power plant probabilistic safety assessment: A review

Fault tree analysis (FTA) is a graphical model which has been widely used as a deductive tool for nuclear power plant (NPP) probabilistic safety assessment (PSA). The conventional one assumes that basic events of fault trees always have precise failure probabilities or failure rates. However, in real-world applications, this assumption is still arguable. For example, there is a case where an extremely hazardous accident has never happened or occurs infrequently. Therefore, reasonable historical failure data are unavailable or insufficient to be used for statistically estimating the reliability characteristics of their components. To deal with this problem, fuzzy probability approaches have been proposed and implemented. However, those existing approaches still have limitations, such as lack of fuzzy gate representations and incapability to generate probabilities greater than 1.0E-3. Therefore, a review on the current implementations of fuzzy probabilities in the NPP PSA is necessary. This study has categorized two types of fuzzy probability approaches, i.e. fuzzy based FTA and fuzzy hybrid FTA. This study also confirms that the fuzzy based FTA should be used when the uncertainties are the main focus of the FTA. Meanwhile, the fuzzy hybrid FTA should be used when the reliability of basic events of fault trees can only be expressed by qualitative linguistic terms rather than numerical values.     

核电运行人才的培养

以秦山第二核电厂为例,介绍了系统性培训方法SAT和运行人才培养体系、培训管理组织机构和责任、核电厂运行人才培养的逐级培训体系、在岗培训制度、运行处的岗位授权制度、人才培养的激励机制等,为核电厂运行人才培养提供参考。     

秦山三期重水堆核电站风险监测器研发进展

核电站风险监测系统(Risk Monitor)可对核电站的运行风险进行实时监测和预测,是概率安全评价(PSA)技术的高级应用之一.FDS团队广泛调研了国际现有核电站风险监测系统的研发现状,深入研究了风险监测系统涉及的各种关键算法并探索了相关实现技术,基于前期自主研发的大型集成概率安全分析软件RiskA发展了通用核电站风...     

基于多层流模型的核电站警报分析系统的开发

开发了基于多层流模型(MFM)的核电站警报分析系统。系统通过在复杂的故障状态下自动识别主要根本原因,能够减轻运行人员的工作负荷。另外,由于多层流模型提供了一组蕴涵因果关系的图形符号,操作维护人员可以通过符号分析来验证诊断结果,从而可以提高警报分析过程的可理解性以及维护工作的可靠性。取自RELAP5/MOD2的19组数据用于评价系统性能。仿真实验结果显示了该系统在停堆前具有较好的及时检测和诊断故障的能力。     

核电机组电气设计分析

近年来,我国核电事业得到快速发展,一大批核电项目陆续开工建设,其中大部分为引进技术的二代改进型和三代核电机组。由于技术输出国的标准规范与我国现有的核电设计标准不一致,以及考虑厂址适应性等问题,我国对引进的核电机组存在逐步消化、吸收并改进的过程。本文对核电机组的电气设计进行探讨,希望对今后同类工程具有参考作用。     

Development policy on new generation of nuclear power combined with desalination in China

The potential market for desalination industry is forecasted in China for a long term. A co-generation policy is proposed in power production and desalination. It has been predicted that the desalination would become a huge industry in China provided that the technology of desalination is improved and fresh water cost reduced to a certain Ievel accepted by Chinese residents.     

秦山核电厂的老化及寿期管理

介绍了核电厂老化及寿期管理的相关背景以及国外核电厂在延寿方面采取的两种主要模式,即执照更新模式和PSR模式。结合目前秦山核电厂开展的主要老化管理工作,提出了秦山核电厂延寿的设想,并对核电厂寿期管理中存在的问题进行了讨论。     

Corrosion mitigation in coolant systems in nuclear power plants

Nuclear power plants have suffered various failures through corrosion causing economic losses, increasing the radiation exposure to personnel and increasing the possibility of environmental risk. Many examples of different corrosion mechanisms and their consequences for nuclear power plant (NPP) working conditions are recognized and described. Nevertheless, several issues related to the corrosion of materials used for NPP constructions are still unexplained. This paper gives short, basic information about selected methods of the corrosion reduction and corrosion inhibitors used in coolant systems in nuclear power plants, mainly in pressurized water reactors PWRs and boiling water reactors BWRs. Present data are based in the open scientific and technical literature since 1990.