多模块高温气冷堆地震概率安全分析关键技术研究

高温气冷堆核电厂采取多个反应堆模块匹配1个汽轮机的设计方式,即1台高温气冷堆机组会包含多个反应堆模块,这使多个高温气冷堆模块在地震外部事件下存在明显的相关性,因此在利用概率风险分析方法来全面地识别和评价高温气冷堆的地震风险时,需要从机组的角度充分考虑和模化机组内多个反应堆模块间的相关性。高温气冷堆示范电站已完成了较为完整的单模块地震概率安全分析,本文将以该分析结果为基础梳理出高温气冷堆多模块地震概率安全分析的关键技术要素并进行研究,研究内容包括多模块事件序列建模和地震相关性失效评价等关键技术,并针对多模块高温气冷堆提出了应用策略。然后以双模块设计的高温气冷堆示范电站为对象,以地震导致丧失厂外电始发事件为代表,对多模块高温气冷堆地震概率安全分析进行了实例分析获得远低于概率安全目标的释放类频率,且分析得到了高温气冷堆多模块事件序列建模策略与地震相关性失效的评价路线可行这一重要结论。     

国内核电厂丧失厂外电源始发事件频率分析及风险评价

丧失厂外电源（LOOP）是核电厂概率安全分析（PSA）中一个重要的始发事件。本文首先介绍了基于运行经验的始发频率分析方法,然后根据国内核电厂数据进行初步分析并与通用数据比较,最后以一个能动核电厂为例分析LOOP对堆芯损伤频率（CDF）的影响。分析表明,在功率运行工况下国内核电厂LOOP始发频率要明显低于通用数据（约为1/3）,而停堆工况下的频率则接近。LOOP事件是影响核电厂风险的重要因素,本文的分析可为形成国内通用的LOOP始发频率提供参考。     

小型自然循环铅基快堆内部始发事件选取研究

始发事件是铅基反应堆确定论安全分析和概率安全评价的起点和基础,对反应堆优化设计和安全运行具有重要指导作用。本文基于小型自然循环铅基快堆SNCLFR-100当前的设计方案,参考其他先进快堆始发事件选取经验,以广义“堆芯熔化”作为顶层目标事件,采用主逻辑图（MLD）方法推导其内部始发事件,最后得到一组较完整的内部始发事件清单。本文研究可为自然循环铅基快堆安全分析工作的开展提供理论依据。      

核电厂主控室火灾场景分析及风险评价

主控室是核电厂火灾概率安全评价的主要关注对象之一。本文对典型核电厂的主控室火灾场景进行分析并对由其导致的反应堆堆芯损坏频率进行计算评价,主要使用事件树方法演绎火灾场景,通过火灾模拟计算确定火灾场景的危害,最后在电厂内部事件一级PSA模型的基础上建立火灾风险评价模型,完成主控室火灾风险定量化。火灾演绎分析结果获得了4个火灾场景,分别能够导致不同的电厂始发事件,并对相关的操纵员动作产生较大影响。风险定量化结果表明:主控室火灾导致的堆芯损坏频率为1.953×10~(-9)/堆年。     

多堆厂址始发事件分析探讨

多堆厂址一级概率安全评价（PSA）研究中，机组数目的增加使得建模工作量剧增，给整个核电厂的风险评估带来困难。结合已有基础，本文研究了多堆厂址始发事件分析的筛选方法，提出利用堆芯损伤频率（CDF）上下限值评估方法，分析厂址内不同机组数对厂址CDF的影响。结果表明，双机组厂址适合优先进行具体分析。针对双机组核电站，对多堆厂址内各始发事件进行筛选。结果表明，丧失厂外电、丧失热阱等事件适合建模分析，并对其他筛选结果给出后续分析建议，为多堆厂址一级PSA后续事故序列建模工作提供了重要基础。     

核电厂继发丧失厂外电源事件风险研究

继发丧失厂外电源事件是一类特殊的核电厂始发事件，核电机组如位于较小规模的电网中，其自身的事故紧急停堆将有可能引起所在电网失稳，从而带来事故叠加外电网丧失的风险。因此当核电机组处于典型的“大机小网”系统时，核电厂继发丧失厂外电源风险不可忽视。本文采用概率安全分析方法，探讨了继发丧失厂外电源的风险，对比了不同反应堆型、不同联网模式下继发丧失厂外电源导致的功率工况内部事件堆芯损坏频率，并给出了核电厂在运行和管理方面的建议。     

固态钍基熔盐堆概率安全评价始发事件分析研究

始发事件分析是反应堆概率安全评价的起点。本文以10 MW固态钍基熔盐堆(Thorium Molten Salt Reactor,TMSR-SF1)为研究对象,采用主逻辑图分析方法,基于TMSR-SF1的最新概念设计,在参考已有氟盐冷却高温堆、高温气冷堆和钠冷快堆的始发事件清单和始发事件分析理论的基础上,针对TMSR-SF1始发事件分析进行初步探索研究,初步确定了TMSR-SF1的始发事件清单,共得到了TMSR-SF1的37个始发事件(功率运行情况下),并按照故障类型分类的方法对这些始发事件进行分组,共分为6组。为TMSR-SF1下一步的深入分析研究始发事件及其概率安全评价(Probabilistic safety assessment,PSA)中事故序列分析奠定了重要基础,也为安全分析的完整性提供了支持。     

高温气冷堆地震丧失厂外电的风险评价

地震导致丧失厂外电是核电厂地震情况下的典型始发事件。本研究使用地震概率安全分析方法,以高温气冷堆为研究对象,得到其在地震丧失厂外电事故下的风险水平。研究范围包括分析地震导致丧失厂外电的事故发展情景分析,筛选地震设备清单并结合现场巡访进行调整,建立地震导致丧失厂外电的风险评价模型,并对超过高温气冷堆风险接受准则剂量（概率安全目标）的放射性释放的频率结果进行了间隔分析、割集分析和重要度分析。本文工作可为高温气冷堆的地震概率安全分析在方法实施、建模假设、过程分析等方面提供有益的参考。     

应用Jeffreys方法计算可靠性参数和始发事件频率的无信息先验

在核电厂概率安全评价（PSA）分析中,有些始发事件频率或设备失效记录在工业界几乎无历史数据。为了计算这些无信息先验的可靠性参数和始发事件频率,可采用Bayesian统计学中的Jeffreys方法。本文阐述了Jeffreys先验和简化的受限无信息先验分布（SCNID）的数学原理,分别导出了Gamma-Poisson模型和Beta-Binomial模型的Jeffreys无信息先验公式和不确定性区间。结合反应堆冷却剂小破口失水事故(SLOCA)实例介绍了如何应用Jeffreys先验计算始发事件频率。结果表明,Jeffreys方法是一种计算无信息先验的有效方法。     

HFETR一级PSA始发事件分析

在高通量工程试验堆（HFETR）一级概率安全分析（PSA）中,始发事件分析是首要任务。首先综合应用了工程评价、参考以往的始发事件清单、演绎分析和运行经验总结等方法,确定了HFETR运行阶段一级PSA始发事件清单,然后对始发事件进行适当的归并分组,最后结合故障树分析、HFETR运行事件统计及参照国内外相同类型研究堆等方法,给出了各始发事件组的频率,为后续开展HFETR一级PSA奠定了基础。      

Analysis of human-induced initiating events in a multi-unit loss of offsite power scenario

ABSTRACT

Human-induced initiators (category-B actions) are the initiators that are caused by human errors and are rarely explicitly identified and modeled in probabilistic safety assessments (PSAs). The current concern over the safety of multi-unit nuclear power sites is also a motivation for this research. This study proposes a novel process for identifying and quantifying category-B actions and ultimately, how to derive a human-induced initiating event frequency in a multi-unit scenario. Hence, this study fundamentally applies a scenario–system–action search scheme using maintenance and testing procedures, quantifies the human error probability by using the cause-based decision tree and technique for human error rate prediction method, models category-B human actions in the developed fault trees, and derives the human-induced initiating event frequency. The procedure, which is used in this approach, essentially involves system analysis, fault tree development, human error identification, screening, and quantification. The multi-unit loss of offsite power is used as an example accident situation which illustrates the application of the suggested method. Hence, the human-induced initiating event frequency for the loss of off-site power scenario for two units is derived. The application of this method would advance the efforts concerning multi-unit nuclear power plant (NPP) site risk analysis.     

多机组核电厂总体风险的一级PSA方法研究

福岛核事故发生后,多机组核电厂的总体风险受到越来越多的关注,但国内外缺乏评价多机组核电厂总体风险的方法或导则。本文结合有关法规对核电厂的总体安全要求,探索将单机组的一级概率安全评价（PSA）方法拓展为多机组的风险评价方法。以双机组核电厂为例,讨论了多机组厂址PSA定量化的一些问题,提出了机组间相关性的一些见解,并阐明了数学原理。本文讨论的方法对研究多机组厂址PSA方法具有重要价值。     

Multiple Failure Accident Sequences for SEAFP Reactor

This paper summarizes several investigations on the identification of possible multiple failure accidents relevant in terms of consequences for the SEAFP reactor. Particularly, on those sequences of events that could induce a risk of radioactive materials bypass through the SEAFP confinement barriers. The analyses here reported are related to the Heat Transfer Systems of both reactor models 1 and 2. The work is carried out within the Safety and Environmental Assessment of Fusion Power—Long Term Programme (SEAL) 95/96. A set of specific initiating events (IEs) has been individuated, on the basis of the previous studies performed in the frame of the first SEAFP program. Basing on pre-existing analyses, each accident initiator has been discussed and several sequences have been described depending on the additional failures which could follow the initiator.     

Risk assessment in setting inservice inspection priorities at light water reactors

As part of the Nondestructive Evaluation Reliability Program, sponsored by the U.S. Nuclear Regulatory Commission, Pacific Northwest Laboratory is developing a method that uses risk-based approaches to establish inservice inspection plans for nuclear power plant components. The method first uses probabilistic risk assessment (PRA) results and Failure Modes and Effects Analysis (FEMA) techniques to identify and prioritize the most risk-important systems and components for inspection. The acceptable level of risk from structural failure for important systems and components is then apportioned as a small fraction of the total PRA estimated risk for core damage. This process determines the target (acceptable) risk and failure probability values for individual components. The Surry Unit 1 Nuclear Power Station was selected for pilot applications of the method. The specific systems addressed are the reactor pressure vessel, the reactor coolant, the low-pressure injection, and the auxiliary feedwater. The results provide a risk-based ranking of components within these systems and relate the target risk to target failure probability values for individual components. These results will be used to guide the development of improved inspection plans for nuclear power plants.     

10 MW固态燃料熔盐实验堆安全分析关键技术初步研究

钍基熔盐反应堆(Thorium Molten Salt Reactor,TMSR)项目是中国科学院科技先导项目之一。基于10 MW热功率熔盐反应堆-固体燃料(Thorium Molten Salt Reactor-Solid Fuel,TMSR-SF)的设计,对TMSR的关键技术安全分析进行了初步研究。TMSR-SF与现有反应堆之间的差异对核安全审查提出挑战,TMSR-SF审查方法的研究将准备其安全审查的技术和要求。固态燃料熔盐实验堆安全分析关键技术初步研究包含4个方面:堆芯核设计关键安全限值、事故序列及验收准则、源项及其审评方法和验收准则、概率安全评价方法和始发事件。首先对其它类型反应堆的安全审查方法进行了研究,对其关键参数和重要规定做了概述,并借鉴了高温气体冷堆和钠冷却快堆的审评要求和方法;然后使用蒙特卡罗和其他方法、模型来计算TMSR-SF的关键参数。应用逻辑图方法讨论概率风险评价(Probabilistic Risk Assessment,PRA)方法和始发事件清单。在本研究中,计算了核心核设计安全限值,研究和讨论事故列表和分类,讨论了TMSR-SF的PRA框架和始发事件清单,该研究将支持TMSR-SF的安全审查和安全设计。     

Seismic individual plant examination of external events of US nuclear power plants: insights and implications

As part of the implementation of the severe accident policy, nuclear power plants in the US are conducting the individual plant examination of external events (IPEEE). Seismic events are treated in these IPEEEs by either a seismic probabilistic risk assessment (PRA) or a seismic margin assessment. The major elements of a seismic PRA are the seismic hazard analysis, seismic fragility evaluation of structures and equipment and systems analysis using event tree and fault tree analysis techniques to develop accident sequences and calculate their frequencies of occurrence. The seismic margin assessment is a deterministic evaluation of the seismic margin of the plant beyond the design basis earthquake. A review level earthquake is selected and some of the components that are on the success paths are screened out as exceeding the review level earthquake; the remaining ones are evaluated for their seismic capacity using information from the original plant design criteria, test data and plant walkdown. The IPEEEs of over 100 operating nuclear power plants are nearing completion. This paper summarizes the lessons learned in conducting the IPEEEs and their applicability to nuclear power plants outside of the United States.     

The NUREG-1150 probabilistic risk assessment for the Sequoyah Nuclear Plant

This paper summarizes the findings of the probabilistic risk assessment (PRA) for Unit 1 of the Sequoyah Nuclear Plant performed in support of NUREG-1150. The emphasis is on the “back-end” analyses, that is, the accident progression, source term, and consequence analyses, and the risk results obtained when the results of these analyses are combined with the accident frequency analysis. The results of this PRA indicate that the offsite risk from internal initiating events at Sequoyah are quite low with respect to the safety goals. The containment appears likely to withstand the loads that might be placed upon it if the reactor vessel fails. A good portion of the risk, in this analysis, comes from initiating events which bypass the containment. These events are estimated to have a relatively low frequency of occurrence, but their consequences are quite large. Other events that contribute to offsite risk involve early containment failures that occur during degradation of the core or near the time of vessel breach. Considerable uncertainty is associated with the risk estimates produced in this analysis. Offsite risk from external initiating events was not included in this analysis.     

Risk managed technical specifications

Application of probabilistic risk assessment (PRA) technology has become an essential component in the decision-making processes associated with the operation and regulation of commercial nuclear power plants (NPPs). As PRA technology has matured, it increasingly has been utilized to provide risk insights in the support of both operational and regulatory decision-making. This paper describes the next significant application of PRA technology to risk inform NPP operation. This Risk Managed Technical Specification (RMTS) application utilizes the results of the plant PRA to determine risk-informed technical specification (TS) allowed out of service times (AOTs). The RMTS process utilizes the PRA results to specify appropriate configuration specific TS AOTs and ensures the risk of events that could result in core damage or large early release are maintained below acceptable levels. In addition, RMTS requires development of integrated risk management actions to actively mitigate risks associated with the inoperability of TS structures, systems and components (SSCs). RMTS has been approved for implementation at commercial NPPs in the United States with the South Texas Project Electric Generating Station (STPEGS) serving as the initial application. In this paper we describe the programmatic requirements necessary to implement RMTS and provide several examples illustrating its application; thus demonstrating the applicability of RMTS to manage nuclear safety risk while simultaneously enhancing operational flexibility.     

Some thoughts on allocation of reliability

The technical feasibility of allocating reliability to reactor systems, subsystems, components, and structures is discussed in this paper. The basic premise for this analysis is that a set of objective functions or safety variables has been defined on a global basis for a class of nuclear power plants. The decision variables, which represent the system, subsystem, component, and structural reliabilities are related to the global objective functions by a risk model obtained from an existing plant-specific probabilistic risk assessment (PRA). A multiobjective optimization technique is employed to obtain the set of decision variables which optimize (minimize) all of the objective functions. A cost function is introduced (and incorporated in the optimization scheme) which measures the cost of increasing reliability. Illustrative calculations were performed for a boiling water reactor with an existing PRA.