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.     

安全壳及其内部结构安全的可靠性分析方法研究

安全壳及其内部结构作为核电站专设安全设施,其安全及可靠性问题关系到复杂工况情境下核电站的安全保障。本文结合模糊FMECA方法与改进的FTA定量计算方法,探讨研究针对安全壳及内部结构安全及可靠性的分析方法。通过模糊FMECA研判系统所含高危害度的失效故障模式,进一步构造Bow-tie模型,利用模糊数学和灰色关联度理论改进的FTA定量计算方法计算得出最小割集底事件组对应故障致因触发顶事件的可能性。该分析方法解决了系统结构复杂、底事件定量信息贫乏的问题,为安全壳及内部结构的设计与建设提供理论决策支持。     

FMECA在N36锆合金管材研制工艺设计中的应用

文章以N36锆合金管材试制工艺流程为对象,利用FMECA进行工艺设计优化,对工艺故障模式进行系统地识别、分析了故障模式成因及对后续工序和产品使用的影响,分析了风险优先数(RPN),建立了FMECA分析表,识别出了高RPN数值的故障模式及薄弱工艺环节。以降低RPN为目标,采取了改进措施。结果表明:(1)N36锆合金管材研制过程应用FMECA能够系统识别薄弱工艺环节,对工艺设计优化起到有效的作用;(2)FMECA分析表明,铸锭熔炼和锻造、管坯轧制等过程集中了高RPN值的故障模式,将其作为关键工艺实施工艺确认和工艺优化可降低RPN;(3)工艺设计FMECA是一个迭代过程,随试制进展应不断采取有效的改进措施。     

压水堆堆内构件老化评估方法及其应用

本文介绍了压水堆堆内构件老化分析评定方法。该方法通过建立老化机理筛选准则、初步分类、堆内构件的故障模式、影响及危害分析(FMECA)以及识别主要潜在老化部件等步骤,最终完成堆内构件老化程度的评估。老化评估结果为堆内构件的老化管理奠定基础。该老化评估方法已首次成功应用于秦山CNP320机组和CNP650机组堆内构件的老化评估。     

ITER实验包层吹洗气前处理系统的FMECA研究

实验包层(TBM)输出吹洗气前处理系统将安装在国际热核聚变实验堆(ITER)装卸TBM的通道内(Port Cell),它的功能是将TBM输出的含氚吹洗气进行过滤、除HTO、冷却、调流量等处理,处理后输出到氚提取系统。介绍了该系统的工艺过程和系统组件,以氚释放危险为判据,对该系统进行FMECA(故障模式、影响及危害性分析),并作出分析表。找出了几种故障模式或薄弱环节,进行了尝试性的风险优先数和故障模式危害度计算,提出了设计改进措施和使用补偿措施;最后确定了需要重点关注的4种需导致释非正常过量释放的潜在故障模式。这些故障分析为降低系统氚过量释放危险设计提供了依据,也为TBM其他附属氚系统的安全分析奠定了基础。     

HWRR堆本体退役危害分析

根据重水研究堆(HWRR)堆本体结构层次、拆除次序和退役工艺,分别采用主逻辑图法和故障模式影响及危害度分析(FMECA)对HWRR堆本体退役活动中潜在的危害开展分析,确定其主要的潜在危害,并对其进行分类。     

中国实验快堆通风控制系统的可靠性分析

介绍中国实验快堆通风控制系统的功能和组成结构,并本文以故障模式影响分析(FMEA)和FTA为基础进行了可靠性分析,采用故障树进行定量分析和计算,得到该故障树的失效概率和最小割集,从而为中国实验快堆通风系统风险管理提供数据支持。     

新产品开发中3F一体化技术的应用

3F一体化技术是由以FMECA与FTA综合分析方法为基础构成的潜在故障预测和改善功能模块及以FRACAS为原型的已知故障闭环控制功能模块组成的可靠性分析技术.描述了将3F一体化技术应用于新产品开发全过程的研究思路,重点阐述3F一体化技术在新产品开发过程中的运行模型及实施流程,并以案例演示其具体实施过程.最后,对3F一体化技术及其在新产品开发中的应用,提出了有待发展研究的课题.     

多方法融合的反应堆紧急停堆子系统安全性分析

针对反应堆紧急停堆子系统,将故障模式影响分析（FMEA）、故障树分析（FTA）、系统理论的过程分析（STPA）3种独立的基本分析方法进行组合,形成仪表控制系统设计阶段的失效和故障基本项覆盖统计表格。STPA方法能够很好地弥补了FMEA和FTA方法的不足。同时,在仪控系统的设计阶段,STPA方法非常适合发现反应堆紧急停堆子系统涉及的软件类、系统交互以及通信类的故障和安全问题。      

泰山核电厂安全壳喷淋系统可靠性分析

本文用失效模式和效应分析(FMEA)及故障树(FTA)方法,对秦山核电厂安全壳喷淋系统(CSIS)可靠性进行分析.     

基于故障树分析的AP1000非能动余热排出系统的设计与优化

非能动余热排出(PRHR)系统是AP1000核电站重要安全系统之一。采用故障树方法对该系统可靠性进行评价,得到系统可能失效机理,并运用Risk Spectrum软件进行定量分析,得出系统失效概率约为9.215×10^-5。结果表明,余热排出热交换器入口管线上电动阀失效关闭是导致系统失效的最主要因素。基于计算结果提出两种方法对系统进行优化：1） 在另一回路增加同样1套PRHR系统;2） 在原有系统基础上增加一PRHR热交换器。通过故障树分析计算,并分别从系统可靠性、复杂性、经济性等方面对两种方法进行比较发现,方法2更具可行性,建议工程上采用此方法对系统进行优化。     

Development of diagnosis algorithm for the check valve with spectral estimations and neural network models using acoustic signals

We have developed a method for detecting and diagnosing a disk wear failure and a foreign object failure among the various failure modes of check valves. The method is based on the acoustic emission sensors which can detect the sound wave of the leakage flow and the estimation of the power spectral densities with an auto-regressive model. For validating the method, we implemented a hydraulic test loop with an artificially failed check valve. We have found that the frequency spectrums from the acoustic signals are strongly dependent on the failure modes of the check valve and that they are nearly independent of the failure size and operating pressure through an estimation of the power spectral density with an auto-regressive signal processing model. In addition, the root mean square values of the acoustic signal and the amplitudes of the power spectral density as well as the loop pressure have a strong dependency on the failure size in each failure mode of the check valve. We developed a diagnosis algorithm by using neural network models in order to identify the type and size of the failure in the check valve. The diagnosis algorithm consists of a hierarchical model composed of three back-propagation neural networks. The results of our research and the experiments show that the diagnosis algorithm is proven to be a good solution for identifying the failures of the check valves without any disassembling work.     

RAMI analysis of the ITER CIS

ITER is the first worldwide international project aiming to design a device that proves the physics and technological basis for fusion power plants to produce nuclear fusion energy. In the project, the RAMI approach (reliability, availability, maintainability and inspectability) has been adopted for technical risk control to guide the design of components in preparation for operation and maintenance. RAMI analysis of the ITER central interlock system (CIS), which shall provide investment protection for the ITER systems was performed on the conceptual design. A functional breakdown was prepared in a bottom-up approach, resulting in the system being divided into 5 main functions and 7 sub-functions which are described using the IDEFØ method. Reliability block diagrams (RBDs) were prepared to estimate the reliability and availability of each function under stipulated operating conditions. Initial and expected scenarios were analyzed to define risk-mitigation actions. The inherent availability of the ITER CIS expected after implementation of mitigating actions was calculated to be 99.86% over 2 years, which is the typical interval of the scheduled maintenance cycles. A failure modes, effects and criticality analysis (FMECA) was performed to initiate risk mitigation action. Criticality matrices highlight the risks of the different failure modes with regard to the probability of their occurrence and impact on operations. It was assessed that the availability of the ITER CIS, with appropriate mitigating actions applied, meets the project availability requirement for the system.     

CARR保护系统可靠性分析

保护系统是保证中国先进研究堆（CARR）安全的重要系统之一,它的高可靠性是保证其完成预期功能的重要条件。本工作以保护系统为研究对象,利用故障模式及影响分析和故障树等可靠性分析方法,建立相应模型,对保护系统进行了定性和定量的分析,得到了保护系统发生故障的概率和最小割集,其可靠性水平达到了CARR工程的设计要求,验证了设计,为CARR其他系统分析和验证奠定了基础。     

核电厂主蒸汽隔离阀控制系统故障分析及可靠性提升

主蒸汽隔离阀是核电厂核岛与常规岛间主蒸汽管线上最重要的隔离设备,主要介绍了秦山第二核电厂1/2号机组主蒸汽隔离阀控制系统功能及控制原理、故障模式分析。结合历史故障统计得出系统的薄弱点—限位开关,并对限位开关故障失效机理和根本原因分析进行了重点阐述,最后针对系统的可靠性提升从人、机、料、法、环5方面提出了改进措施,对国内其他在役核电厂主蒸汽隔离控制系统维护和改进有一定的借鉴和参考意义。     

CARR应急供电系统可靠性分析

应急供电系统是保证中国先进研究堆（CARR）安全的重要系统之一,它的高可靠性是保证其完成预期功能的重要条件。本文以应急供电系统为研究对象,利用故障模式及影响分析和故障树等可靠性分析方法,建立相应模型,对应急供电系统进行定性和定量的分析,得到了该系统发生故障的概率和最小割集,找出了薄弱环节,提出了有效提高该系统可靠性的建议和措施,为保证CARR安全可靠地运行提供支撑。     

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.