风险指引型PSA应用的不确定性分析方法研究

不确定性是这个世界上所有事物与生俱来的特征,本文深入探讨了核电厂PSA不确定性的来源及其分类,研究了参数的不确定性、建模的不确定性以及模型的不完备性的处理方法。结合风险指引型PSA应用相关技术导则、法律法规等文件中对不确定性的相关要求,分别给出了在风险指引PSA应用中此三类不确定性的分析方法和相关可接受准则。风险指引PSA应用的不确定性分析是应用申请中不可或缺的重要组成部分,可为综合决策者提供足够的技术支持。     

风险指引型技术在核电厂维修策略优化中的应用探索

利用概率安全分析技术来支持核电厂的日常风险管理及安全决策,已成为目前国际上的主流分析方法。本文采用概率论和确定论相结合的风险指引综合决策方法对RRI/SEC热交换器维修策略调整的可行性、合理性和安全影响进行论证,结果表明RRI/SEC热交换器的预防性维修项目可以在功率运行期间实施。应用风险指引型技术对维修策略调整,有助于提高维修工作安排的灵活性、优化大修资源配置、增强维修质量和设备可靠性,对机组安全稳定运行具有重要意义。     

内部事件一级PSA在设计中的应用

"华龙一号"(HPR1000)是中国核工业集团开发的第三代核电堆型,在"华龙一号"研发和设计过程中全面应用了概率安全分析(PSA)技术,其中基于内部事件一级PSA开展的重要应用包括:(1)安全系统设计方案评估;(2)设计扩展工况清单选取;(3)事故规程优化;(4)技术规格书开发;(5)防人因设计."华龙一号"通过创新性...     

风险指引型管理方法及在大亚湾核电厂运行安全管理中的应用

安全评价是核电厂运行安全管理中重要的工作内容。本文运用以概率论为基础的概率风险评价方法(PSA)，在分析核电厂安全评价工作特点的基础上．介绍运用PSA方法进行核电厂安全评价的一般过程与方法．最后结合大亚湾核电厂应用PSA进行设备检修的实例，说明其具有可操作性与科学性。     

核电厂风险指引管理研究

风险指引管理是确定论与概率论方法相结合的一种新的安全管理模式.为了促进我国这项工作的开展,有必要对国内外的相关法规、标准和实践进行全面和系统的研究.本文介绍了核电厂风险指引决策的基本原则、方法与风险接受准则,讨论了风险指引决策对概率安全评价(PSA)的要求,并对我国核电厂采用风险指引管理提出了建议.     

非能动核电厂风险指引管理中的风险接受准则研究

由于非能动核电厂的风险水平较低,所以目前风险指引管理中的风险接受准则对其存在适用性问题.本文首先对现有的风险接受准则进行分析,然后针对非能动核电厂可能的风险可接受准则方案和现有接受准则的充分性进行描述,并对非能动核电厂风险准则中存在的大量放射性释放(LRF)问题进行研究,最后提出非能动核电厂风险指引管理中风险接受准则的...     

风险指引型技术在核电厂设备检修策略优化中的应用研究

为了提高核电厂的经济性,核电厂通过对运行管理进行优化以提高其能力因子和运行灵活性,如优化大修期间设备检修策略以缩短大修工期。本文引入风险指引型理念对核电厂设备检修策略优化方法进行研究,并以某核电厂的余热排出系统热交换器检修策略调整为例,即内部表面目视检查由当前的每2年一列交叉检查变更为每4年检查顺序检查。使用该方法进行了详细的论证与计算。通过分析认为通过风险指引型技术方法对该电厂余热排除系统热交换器检查策略调整是合适的,能继续遵守纵深防御原则且不挑战核电厂的安全裕量,变更所引入的风险是可接受的。     

概率截断对PSA中RAW重要度的影响研究

重要度分析是核电站概率安全评价中一个重要组成部分.本文首先简要介绍了RAW重要度;在详细分析了截断对重要度的影响后,引入了安全风险因子,指出影响该因子的相关因素,同时也提出只有保证安全风险因子在一定的变化范围之内即减少可能丢失的割集,才能最终降低截断对重要度的影响.最后给出一种解决概率截断问题的思路即使用自动概率截断和McFarm相结合的方法,较好地解决了因概率截断而对重要度计算误差问题.     

浅谈风险指引型核安全法规体系

简要介绍了我国核安全法规的来历和发展现状,分析了现行核安全法规存在的不足。同时介绍了风险指引型安全管理方法的发展、现状以及所取得的成就,说明风险指引是核安全法规未来改进的方向。结合美国正在进行的构建风险指引型核安全法规体系的发展计划,讨论了新法规体系的原则及其特点。最后根据现实情况,就我国推行风险指引型安全管理和构建风险指引型核安全法规体系,提出了一些建议。     

风险指引的核电厂应急柴油发电机返厂大修分析

为了支持田湾核电站1、2号机组的1台应急柴油发电机进行返厂大修,本文提出了1台应急柴油发电机不可用的恢复时间的优化方案,并采用确定论与概率论相结合的风险指引型方法对优化方案的可行性进行了分析。分析结果表明,1台应急柴油发电机不可用的恢复时间优化满足相关法规导则和传统工程分析的要求,且对电厂带来的风险是很小的、可接受的。因此,可以开展1台应急柴油发电机进行返厂大修工作。      

Implication of PSA uncertainties on risk-informed decision making

Nuclear power plants risk-informed policy is introduced in order to improve safety decision making and regulatory efficiency. The corresponding regulatory guides define the acceptable risk measures and their changes resulting from the modifications in the licensed design of the nuclear power plant. The risk measures used in the acceptance guidelines are the core damage frequency and large early release frequency.The risk measures and their corresponding changes are assessed by the Probabilistic Safety Assessment. The uncertainties of Probabilistic Safety Assessment should be appropriately addressed in the context of the decision making, considering their implication on the obtained results. The Probabilistic Safety Assessment uncertainties include epistemic uncertainties resulting from parameter, model, and completeness uncertainties.The paper presents the obtained results from the uncertainty analysis of the Probabilistic Safety Assessment of the reference nuclear power plant and their implication on risk-informed decision making. The paper focuses particularly on parameter and model uncertainties. The analysed modification is extension of the test interval of the emergency diesel generators. The core damage frequency is the used risk measure in the analysis.The need for the appropriate consideration of the uncertainties in the Probabilistic Safety Assessment in order to adequately support the risk-informed decision making is identified. The deficiency of usage of percentile measures is identified and acknowledged. The need for the adaptation of the risk-informed decision-making principles considering new nuclear power plants is recognized.     

Recent applications of PSA for managing nuclear power plant safety

The safety design and regulation of nuclear power plants has traditionally been based upon deterministic approaches that consider a set of challenges to safety, e.g. design basis accidents, and determine how those challenges should be handled. The approach has been very successful since no plant designed or regulated to United States standards has ever harmed a member of the public. The arbitrary nature of these safety criteria, the potential inconsistencies in the judgments on relative probabilities, and the lack of definition for ‘safety’ became increasingly evident during the 1960s. Probabilistic approaches to reactor safety were proposed ^1,2,3 but did not take off in the United States until publication of the Reactor Safety Study ⁴ in 1975. Even as the methodology matured, there remained a challenge to integrate it into the regulatory process. This article will describe this integration process. A probabilistic approach to regulation enhances and extends the traditional deterministic approach by introducing the concept of safety (risk) significance that allows the designer/operator to focus on important issues. Emphasis was initially placed on relative risk but now regulatory decision-making is employing both relative and absolute risk. Measures of importance will be defined. Risk information can be used to prioritize the allocation of resources and three examples will be described. Equipment configuration control systems are being installed and used at nuclear power plants to enhance safety and to reduce Operating and Maintenance costs; they will be described. Finally, the US Nuclear Regulatory Commission's introduction of risk-informed decision-making into the regulatory process will be discussed.     

核电站实时风险管理系统部件重要度计算方法研究

在PSA中,重要度分析都是基于基本事件进行计算的,而在核电站实时风险管理系统Risk Moni- tor中,往往计算的是部件重要度,通常1个部件一般包括1个或多个基本事件,因而不能简单按照求解基本事件重要度的方法来进行。目前虽然有一些关于部件重要度的计算方法,但是均未得到广泛的认可。本文在分析了常用部件重要度计算方法基础上,引入了一种新的部件重要度即部件概率微分重要度CPDIM,该重要度主要基于微分重要度DIM,利用DIM的1个明显特性即可加性,来求解部件的概率微分重要度。通过实例证明该重要度可作为核电站关键风险部件选择的重要度依据。     

Application of risk-informed design methods to select the PSACS ultimate heat sink

In the early phases of advanced system design, information is scarce. The technologies, components and processes to be used have not been specified adequately or are not well understood and uncertainties are very large. Yet, it is during these early phases that design teams and other stakeholders are required to make critical decisions to guide the development of the system. To aid in this decision making, a formal process is proposed based on the Analytic-Deliberative Decision-Making Process (ADP) that allows stakeholders to synthesize rationally their knowledge and experience and facilitate learning and sharing of best practices. The ADP identifies and prioritizes attributes relevant to a decision problem and supports the formulation of metrics to measure the performance of different design options. This paper reports on an application of the ADP to the selection of an ultimate heat sink for the Flexible Conversion Ratio (FCR) reactor's Passive Secondary Auxiliary Cooling System (PSACS). Two ultimate heat sink options are identified and evaluated, air and water.     

二级PSA中人员可靠性分析方法研究

在概率安全分析(PSA)中,人员可靠性分析(HRA)是必不可少的组成部分。国内在一级PSA中的HRA做了大量的研究工作,已有良好的基础和工程实践,但由于核电厂严重事故下人员响应的复杂性,有关二级PSA的HRA还处于摸索阶段。通过研究二级PSA中人员响应特点,调研国内外在二级PSA中采用的HRA方法,最后以我国某三代压水堆核电厂严重事故下一回路快速卸压为例,采用THERP、HCR+THERP以及SPAR-H三种方法,分别进行了HRA,并给出相应的结论和建议。     

不确定性分析在概率安全评价中的应用

分析了概论安全评价（PSA）中存在的完整性，模型假设条件及输入数据的不确定性和它们的来源。针对输入参数的不确定性，阐述了Risk Spectrum软件关于不确定性分析的原理，方法和误差因子的选取。对输入参数的不确定性进行定量计算后，得到13个初因和各工况的堆芯损坏频率的均值。介绍了表征不确定性的概率密度函数和累计密度函数曲线。     

A risk-informed approach to assessment of DHC initiation in pressure tubes of CANDU reactors

The delayed hydride cracking (DHC) of flaws and cracks in pressure tubes is a serious form of degradation in the reactor core. CSA standard N285.8 (2005) recommends deterministic and probabilistic procedures for assessing the potential of DHC initiation from flaws that are generated by fretting or any other mechanism. Although the deterministic method is simple, it lacks a quantitative risk-informed basis for the assessment. On the other hand, a full probabilistic method based on simulation is tedious to implement. This paper presents an efficient, reliability-based approach that bridges the gap between a simple deterministic analysis and full Monte Carlo simulations. In the proposed method, a deterministic DHC initiation criterion is calibrated to specified target probability levels. The main advantage of the proposed approach is that it provides a practical, risk-informed basis for DHC initiation assessment while retaining the simplicity of the deterministic method.     

PSA吸附罐吸附剂床层高度的不停车检测

本文介绍了利用γ射线对扬子石化公司芳烃厂的１００＃ＰＳＡ吸附罐内吸附剂高度的不停车检测，并给出了检测原理和方法。