非能动系统可靠性评价方法的研究

本文首先详细解释了非能动系统可靠性概念,分析各种非能动系统可靠性评价方法的特点,对比各种方法之间的区别,并指出这些可靠性评价方法共同存在的不足:没有一种方法可同时兼顾非能动系统设备可靠性与功能可靠性,不能科学地整合两者的可靠性,并且未将非能动系统整体可靠性融合进概率安全评价(PSA)模型;针对各种方法存在的不足,本文在国内外研究基础上提出研究问题与思路,而且展望了非能动系统可靠性评价方法未来的发展方向。     

AP1000非能动余热排出系统可靠性与概率安全评价模型的融合

研究AP1000非能动余热排出系统可靠性,应用重要度和敏感性指标全面详细地分析系统设备可靠性,得到导致系统失效的设备各种失效模式的重要度和敏感性排序结果;整合了系统设备可靠性与物理过程可靠性,将整体可靠性融合进概率安全评价模型。分析结果表明:研究非能动余热排出系统可靠性时不仅需要分析其设备可靠性,还应该重点考虑系统物理过程可靠性,应整合两种可靠性并将系统整体可靠性融合进概率安全评价模型,综合分析非能动余热排出系统可靠性。     

Fuzzy methodology applied to Probabilistic Safety Assessment for digital system in nuclear power plants

A fuzzy inference system (FIS) modeling technique to treat a nuclear reliability engineering problem is presented. Recently, many nuclear power plants (NPPs) have performed a shift in technology to digital systems due to analog obsolescence and digital advantages. The fuzzy inference engine uses these fuzzy IF-THEN rules to determine a mapping of the input universe of discourse over the output universe of discourse based on fuzzy logic principles. The risk priority number (RPN) (typical of a traditional failure mode and effects analysis - FMEA) is calculated and compared to fuzzy risk priority number (FRPN), obtained by the use of the scores from expert opinions. It was adopted the digital feedwater control system as a practical example in the case study. The results demonstrated the potential of the inference system to this class of problem.     

ZBDD algorithm features for an efficient Probabilistic Safety Assessment

This paper explains a Zero-suppressed Binary Decision Diagram (ZBDD) algorithm and introduces advanced ZBDD algorithm-based features that are implemented into a fault tree solver Fault Tree Reliability Evaluation eXpert (FTREX). The ZBDD algorithm and its advanced features have been developed for solving a fault tree in Probabilistic Safety Assessment (PSA) of a nuclear power plant. The ZBDD can be interpreted as a factorized structure of minimal cut sets (MCSs). A ZBDD algorithm was developed in 2004 for performing a Boolean operation of ZBDDs. The ZBDD algorithm is based on a set of new ZBDD operation formulae. The ZBDD algorithm is known as an efficient replacement of a cutset-based algorithm that is based on traditional Boolean algebra.This paper explains how to perform a delete-term operation and a rule-based post-processing of MCSs by the ZBDD algorithm and demonstrates the efficiency of the ZBDD algorithm by performing benchmark tests. By using the ZBDD algorithm in this study, a long run time for (1) solving a fault tree, (2) performing a delete-term operation to handle negates, and (3) performing a rule-based post-processing of MCSs could be significantly reduced. Since the ZBDD algorithm is based on the factorized form of MCSs, it uses much less memory than the cutset-based algorithm.Due to the small memory requirement of the ZBDD algorithm from solving a fault tree to performing a rule-based post-processing, a much smaller truncation limit can be used than that in the cutset-based algorithm. By lowering the truncation limit, accurate PSA results such as a core damage frequency and importance measures could be calculated by the ZBDD algorithm.     

超临界水冷堆研发进展

超临界水冷堆（SCWR）是第四代核能系统国际论坛（GIF）确定的6种堆型中唯一的水冷堆。本文描述了SCWR的技术特点,回顾了我国SCWR的研发历程,简要梳理了国际上加拿大、欧盟、日本等在SCWR方面的最新研发情况。最后,本文总结了SCWR的技术优势、面临的技术挑战和发展机遇。      

概率安全评价软件RiskA中的非逻辑处理方法

非单调关联系统广泛存在于实际工程应用中,传统针对单调系统的处理方法不适合于这类系统的处理。因此,如何处理针对非单调关联系统所建立的模型成为概率安全评价软件研发面临的问题之一。本工作在调研一些国际流行概率安全评价软件非逻辑处理方法的基础上,探讨了非逻辑求解难点,基于RiskA的数据结构,设计并实现了非逻辑处理模块,并通过例题验证了RiskA软件非逻辑处理模块的正确性和可靠性。     

Probabilistic Safety Assessment for the Transport of Radioactive Waste to a UK Repository at Sellafield

Abstract

A study has been undertaken to provide a detailed understanding of the radiological and non-radiological risks associated with the transpott of radioactive waste from the sites at which waste is produced in the UK to a proposed deep repository at Sellafield, and to ensure that these risks meet the design targets specified by Nirex. The routine transport collective dose to members of the public was assessed to be 0.2 man.Sv per year, which is only about 0.004% of the natural background dose. Accident frequencies were calculated using event tree methodology. The radiological consequences of accidents were assessed using the probablistic computer code CONDOR. The risk expectation value was calculated to be 1.5 × 10^?5 ? 8.6 × 10^?6 latent cancer fatalities per year (depending on the transport mode scenario). These values are significantly lower than the corresponding prediciions for non-radiological accident fatality rates, 0.05 ? 0.035 fatalities per year. The radiological accident risk for the most exposed individual member of the public was assessed to be 5 × 10^?11 ? 1.7 × 10^?11 per year, very much less than the Nirex target of 5 × 10^?7 per year. Plots of societal risk were shown to lie in the region of ‘negligible risk’, as defined by the UK Health and Safety Commission for non-radioactive dangerous goods transport.     

概率安全评价在核能安全分析领域的应用和发展

概率安全评价(PSA)是核能安全分析领域的两大分析方法之一。本文从PSA概念入手,首先从理论基础、分析视角等多个方面比较了确定论和概率论2种分析方法的差异;其次,梳理PSA在核能安全分析领域的历史进程,通过回顾PSA在技术和法规上的变化,展示了PSA与核能安全在提升过程中相互促进的关系;再次,阐释PSA技术在风险量化预测、平衡安全设计、安全决策、安全监管方面的应用,并通过华龙一号(HPR1000)的实例展示了PSA在核能安全分析中的具体应用方式。最后,对PSA技术未来的发展方向进行了预测,指出确定论和概率论2种分析方法将深入融合,PSA分析从安全目标向任务目标转移、从静态向动态转换、从认知向感知转换的发展方向。     

适用于动态概率安全评价的故障树逻辑简化方法

对故障树进行逻辑简化将有效提高分析计算的速度。根据故障树结构特点,提出了基于贪心算法的故障树逻辑简化方法。该方法已编程实现,并采用实际系统的故障树进行了测试。实践证明,该方法可大幅度提高分析求解速度,同时,该方法所采取的贪心策略又可运用在故障树分析的其他方面。     

AP-1000概率安全评价及其若干相关问题

本文概述了AP-1000的概率安全评价(PSA)及其若干相关问题,如:AP-1000设计平衡、非安全级能动系统的管理"待遇"等问题.     

地下核电厂概率安全评价初步分析

充分考虑地下核电厂的岩土包容性、卸压洞室、隔离门、过滤排放系统设计等,对比地面核电厂,用概率安全评价方法(PSA)研究地下核电厂的大量放射性释放频率(LRF)。分析结果表明,地下核电厂的LRF比同样设计的地面核电厂大约低2个量级,可以实现从设计上实际消除大量放射性释放的安全目标。     

How to effectively compute the reliability of a thermal-hydraulic nuclear passive system

The computation of the reliability of a thermal-hydraulic (T-H) passive system of a nuclear power plant can be obtained by (i) Monte Carlo (MC) sampling the uncertainties of the system model and parameters, (ii) computing, for each sample, the system response by a mechanistic T-H code and (iii) comparing the system response with pre-established safety thresholds, which define the success or failure of the safety function. The computational effort involved can be prohibitive because of the large number of (typically long) T-H code simulations that must be performed (one for each sample) for the statistical estimation of the probability of success or failure. The objective of this work is to provide operative guidelines to effectively handle the computation of the reliability of a nuclear passive system. Two directions of computation efficiency are considered: from one side, efficient Monte Carlo Simulation (MCS) techniques are indicated as a means to performing robust estimations with a limited number of samples: in particular, the Subset Simulation (SS) and Line Sampling (LS) methods are identified as most valuable; from the other side, fast-running, surrogate regression models (also called response surfaces or meta-models) are indicated as a valid replacement of the long-running T-H model codes: in particular, the use of bootstrapped Artificial Neural Networks (ANNs) is shown to have interesting potentials, including for uncertainty propagation. The recommendations drawn are supported by the results obtained in an illustrative application of literature.     

Calculation and Updating of Reliability Parameters in Probabilistic Safety Assessment

The internal events of nuclear power plant are complex and include equipment maintenance, equipment damage etc. These events will affect the probability of the current risk level of the system as well as the reliability of the equipment parameter values so such kind of events will serve as an important basis for systematic analysis and calculation. This paper presents a method for reliability parameters calculation and their updating. The method is based on binomial likelihood function and its conjugate beta distribution. For update parameters Bayes’ theorem has been selected. To implement proposed method a computer base program is designed which provide help to estimate reliability parameters.     

钠冷快堆钠火概率安全评价方法研究

钠火事故是钠冷快堆的典型和特有事故,且很可能是反应堆总风险的主要贡献因素之一。本文在介绍钠火事故特点的基础上,研究使用概率安全分析评价钠冷快堆钠火风险的方法。以中国实验快堆反应堆大厅钠火事故为实例,计算得到反应堆大厅钠火导致的堆芯损坏频率为1.19×10^-8/(堆•年)。在此基础上进一步讨论目前钠火概率安全评价中尚需研究的关键问题。     

Probabilistic Service-Life Assessment for Safety Analysis for Preparing a Power-Generating Unit of a Nuclear Power Plant for Service-Life Extension

Information about the actual history of cyclic loading in the actual state of the metal in equipment components must be used for analyzing the safety of power-generating units in operating nuclear power plants, in preparation for service-life extension, and to determine the moment of onset of rapid aging of the metal and the end of the period of stable operation.The properties of the exponential distribution function and probability functions which are constructed for various loading scenarios using the hypothesis of probabilistic summation of fatigue damage for estimating the -percentage residual service life of equipment components are examined.Probabilistic estimates of the service life for operating nuclear power plants make it possible to control effectively the residual service life of the components of a nuclear power plant on the basis of information provided by the diagnostics systems and by the systems monitoring the state of the metal and the data on loading parameters from the control systems.     

球床模块式高温气冷堆核电站的概率安全分析框架

通过对国际上相似堆型概率安全分析(PSA)框架的调研,结合球床模块式高温气冷堆(HTR-PM)自身设计特点,提出以始发事件为起点,以事件序列为主干,以释放类为终点的HTR-PM的PSA一体化事件树框架.分析表明,HTR-PM在PSA框架上的特点主要由其设计特点决定.     

核电厂概率安全分析中动态可靠性方法综述

用动态可靠性方法弥补传统事件树/故障树方法的不足,补充和完善现有核电厂的可靠性与安全性评估,已成为核电厂概率安全研究的一新发展点。近30年来,动态可靠性已具有相对成熟的理论基础——概率动力学,并形成了蒙特卡罗（MC）模拟和离散动态事件树（DDET）两类主要方法。本文简要介绍动态可靠性理论和方法的研究现状与技术特点,并对未来趋势进行分析。     

Safety evaluation of the inherent and passive safety features of the smart design

SMART (system-integrated modular advanced reactor) is a 330 MWt advanced integral PWR, which is under development at KAERI for seawater desalination and electricity generation. The conceptual design of the SMART desalination plant produces 40,000 m³/day of potable water and generates about 90 MW of electricity, which are assessed as sufficient for a population of about 100,000. The SMART enhances safety by adopting the inherent safety design features such as the elimination of large break loss of coolant accidents, substantially large negative moderator temperature coefficients, etc. In addition, the safety goals of the SMART are achieved through the adoption of passive engineered safety systems such as an emergency core cooling system, passive residual heat removal system, safeguard vessel, and reactor and containment overpressure protection systems. This paper describes the design concept of the major safety systems of the SMART and presents the results of the safety analyses using a MARS/SMR code for the major limiting accidents including transient behaviors due to desalination system disturbances. The analysis results employing conservative initial/boundary conditions and assumptions show that the safety systems of the SMART conceptual design adequately remove the core decay heat and mitigate the consequences of the limiting accidents, and thus secure the plant to a safe condition.     

Probabilistic model of the reliability of a Du-500 pipeline

A methematical model of the operation of a VVéR-440 Du-500 pipeline is considered. The model is constructed on the basis of accumulation processes. Relations are obtained for the reliability and longevity indicators. The reliability and longevity indicators are estimated using real operating data for pipelines. 5 figures, 6 references. IATé. Translated from Atomnaya énergiya, Vol. 87, No. 6, pp. 469–475, December, 1999.     

Incorporating reliability analysis into the design of passive cooling systems with an application to a gas-cooled reactor

A time-dependent reliability evaluation of a two-loop passive decay heat removal (DHR) system was performed as part of the iterative design process for a helium-cooled fast reactor. The system was modeled using RELAP5-3D. The uncertainties in input parameters were assessed and were propagated through the model using Latin hypercube sampling. An important finding was the discovery that the smaller pressure loss through the DHR heat exchanger than through the core would make the flow to bypass the core through one DHR loop, if two loops operated in parallel. This finding is a warning against modeling only one lumped DHR loop and assuming that n of them will remove n times the decay power. Sensitivity analyses revealed that there are values of some input parameters for which failures are very unlikely. The calculated conditional (i.e., given the LOCA) failure probability was deemed to be too high leading to the identification of several design changes to improve system reliability. This study is an example of the kinds of insights that can be obtained by including a reliability assessment in the design process. It is different from the usual use of PSA in design, which compares different system configurations, because it focuses on the thermal–hydraulic performance of a safety function.