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。      

Prediction of major transient scenarios for severe accidents of nuclear power plants

It is very difficult for nuclear power plant operators to predict and identify the major severe accident scenarios following an initiating event by staring at temporal trends of important parameters. In this regard, a probabilistic neural network (PNN) that has been applied well to the classification problems is used in order to classify accidents into groups of initiating events such as loss of coolant accidents (LOCA), total loss of feedwater (TLOFW), station blackout (SBO), and steam generator tube rupture (SGTR). Also, a fuzzy neural network (FNN) is designed to identify their major severe accident scenarios after the initiating events. The inputs to PNN and FNN are initial time-integrated values obtained by integrating measurement signals during a short time interval after reactor scram. An automatic structure constructor for the fuzzy neural network automatically selects the input variables from the time-integrated values of many measured signals, and optimizes the number of rules and its related parameters. In cases that an initiating event develops into a severe accident, this may happen when plant operators do not follow the appropriate accident management guidance or plant safety systems do not work, the proposed algorithm showed accurate classification of initiating events. Also, it well predicted timings for important occurrences during severe accident progression scenarios, which is very helpful to perform severe accident management.     

Probabilistic safety analysis of an accelerator—Lithium target based experimental facility

The international fusion materials irradiation facility (IFMIF) is aimed to provide an intense neutron source by a high current deuteron linear accelerator and a high-speed lithium flow target, for testing candidate materials for future fusion reactors.An activity aimed at the safety assessment of the IFMIF plant as a whole has been carried out applying the probabilistic risk assessment (PRA) approach to identify and quantify in terms of expected frequencies, the dominant accident sequences related to the plant operation, and define the reference accident scenarios to be further analyzed through deterministic transient analysis, in order to verify the fulfilment of the safety criteria.The accident sequences have been modeled through the event tree technique, which allows identifying all possible combinations of success or failure of the safety systems in responding to a selection of initiating events. The identification of accident initiators, provided by the failure mode and effect analysis (FMEA) procedure, is followed by the systems analysis based on fault tree technique, for the unavailability assessment of the safety systems: finally the accident sequence scenarios are assessed by RISK SPECTRUM software.The study has allowed for the development of all accident sequences resulting from selected initiators relative to IFMIF plant and their grouping within sequence families, denoted as plant damage states, on account of the plant response and expected consequences. The frequency assigned to each family sequence is the sum of the contributors relative to all sequences ending into that particular plant state.The outcome of the analysis shows that IFMIF plant is quite safe and presents no significant hazard to the environment: in fact all the sequences implying potential undesired effects as radioactive release to the outside, show very low frequencies, well below the limit for credible accident (1.0E−6/year). In addition, due to the novelty of the design and the large spreading assigned to the failure parameter probabilistic distributions (data utilized in the probabilistic analysis of this one of a kind plant are largely of a generic nature), an uncertainty analysis has been performed to add credit to the model quantification and to assess if the sequences have been correctly evaluated on the probability standpoint.     

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.     

基于离散动态事件树的风险指引的安全裕度分析方法研究

风险指引的安全裕度是近十年来核工业界提出的新的安全理念。本文阐述了基于离散动态事件树的风险指引的安全裕度分析方法,给出该方法下核燃料包壳失效概率均值和标准差的数学表达式。针对简化压水堆模型下的全厂断电事故,提出了基于离散动态事件树的风险指引的安全裕度计算流程,计算了两种离散动态事件树分支规则下燃料包壳失效的风险指引的安全裕度及其不确定性。计算结果表明,不同的分支规则、模型参数分布、系统程序最大时间步长对核燃料包壳失效概率均值和标准差均有显著影响。提出了一种改进的可变概率阈值的分支方法,以更好地平衡风险指引的安全裕度分析过程中计算精度与计算资源的匹配问题。     

Study on construction of loss of component cooling water system initiating event fault tree for fire event PSA

ABSTRACT

In this study, the construction of the loss of component cooling water system (LOCCWS) initiating event (IE) fault tree (FT) for an actual fire event probabilistic safety assessment (PSA) model of the Korean reference nuclear power plant considering only IE initiators was validated. The quantification results of the LOCCWS accident sequences obtained using an LOCCWS IE FT model with only initiators are similar to that with initiators and enabling events. This confirmed that the LOCCWS IE FT for an actual fire event PSA model could be constructed by considering only IE initiators. In addition, the same LOCCWS accident sequences were quantified assuming that fire triggering only the LOCCWS IE leads to reactor shutdown. Compared with the quantification result obtained based on the assumption that any fire included in the fire event PSA leads to reactor shutdown, the core damage frequency (CDF) can be reduced. Thus, it can be concluded that there is a possibility of underestimation of CDF when the LOCCWS IE FT model with only initiators is used and the assumption that fire triggering only the LOCCWS IE results in reactor shutdown is employed for the quantification of LOCCWS accident sequences.     

Safety management in NPPs using an evolutionary algorithm technique

The general goal of safety management in Nuclear Power Plants (NPPs) is to make requirements and activities more risk effective and less costly. The technical specification and maintenance (TS&M) activities in a plant are associated with controlling risk or with satisfying requirements, and are candidates to be evaluated for their resource effectiveness in risk-informed applications. Accordingly, the risk-based analysis of technical specification (RBTS) is being considered in evaluating current TS. The multi-objective optimization of the TS&M requirements of a NPP based on risk and cost, gives the pareto-optimal solutions, from which the utility can pick its decision variables suiting its interest. In this paper, a multi-objective evolutionary algorithm technique has been used to make a trade-off between risk and cost both at the system level and at the plant level for loss of coolant accident (LOCA) and main steam line break (MSLB) as initiating events.     

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。     

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。      

基于蒙特卡罗抽样和最大-最小法的地震易损度算法研究

本文探索并研究了一种新的地震易损度算法,基于蒙特卡罗（MC）抽样和最大-最小法计算了单个设备和多个设备组合的最小割集的易损度。最小割集包括3种类型：纯地震失效最小割集、包含非事件的最小割集、地震失效和随机失效混合割集。对于仅包含地震失效的事故序列,可直接采用基于蒙特卡罗抽样和最大 最小法的易损度算法进行计算。涉及地震失效和随机失效混合的事故序列,可采用极限近似方法（MCUB）或其他割集定量化算法进行计算。经对比,基于蒙特卡罗抽样和最大 最小法的地震易损度算法计算结果与理论值一致,为工程应用中的地震易损度计算提供了另一种可行的算法。     

Fundamental safety strategy against severe accidents on prototype sodium-cooled fast reactor

The accident categories of severe accidents (SAs) for prototype sodium-cooled fast reactor (SFR) which need proper measures were investigated through the internal event probabilistic risk assessment (PRA) and event tree analysis for the external event and six accident categories, unprotected loss of flow (ULOF), unprotected transient over power (UTOP), unprotected loss of heat sink (ULOHS), loss of reactor sodium level (LORL), protected loss of heat sink (PLOHS) and station blackout (SBO), were identified. Fundamental safety strategy against these accidents is studied and clearly stated considering the characteristics and existing accident measures of prototype SFR, and concrete measures based on this safety strategy are investigated and organized. The sufficiency of these SA measures is confirmed by comparing the evaluated core damage frequency (CDF) and containment failure frequency (CFF) to the target value, 1×10^?5 and 1×10^?6 per plant operating year, respectively, which were selected based on the IAEA's safety target. However, the target value of CDF and CFF should be satisfied considering all the SAs caused by both internal and external events. External event PRA for prototype SFR is now under evaluation and we set out to satisfy the target value of CDF and CFF considering both internal and external events.     

Dynamic scenario quantification for level 2 PRA of sodium-cooled fast reactor based on continuous Markov chain and Monte Carlo method coupled with meta-model of thermal–hydraulic analysis

In probabilistic risk assessment (PRA), an event tree (ET) methodology is widely used to quantify accident scenarios which result in core damage and fission products release. However, the current approach using the ET methodology is not applicable to evaluate dynamic characteristics of accident progression, when the accident progression is time-dependent and headings in the ET have inter-dependency between events. Thus, a dynamic approach of accident scenario quantification is necessary to evaluate more realistic PRA.

This research addressed this need by developing a dynamic scenario quantification method for the level 2 PRA by coupling of Continuous Markov chain and Monte Carlo (CMMC) method and a plant thermal–hydraulic analysis code for a sodium-cooled fast reactor (SFR).

The CMMC method is applied to protected loss of heat sink (PLOHS) accident of the SFR to analyze dynamic scenario quantifications. The coupling method requires heavy computational cost and it makes difficult to quantify the whole accident scenarios by comparing the results from existing plant state analysis codes. Thus, a meta-analysis coupling method is proposed to obtain dynamic scenario quantifications with reasonable computational cost. Also, a categorizing method is used to depict analytical results in a transparent manner.     

基于PSA的核电厂核事故应急演习情景设计

本文以昌江核电厂概率安全分析(PSA)的结论为基础,通过研究相关运行事件/事故,梳理建设过程中不符合项或发生过的事件,筛选厂址区域自然灾害以及外部环境事件,收集国内、外其他核电厂已发生的事件/事故,初步选择了代表性与针对性强的事件/事故,选择合适的事故序列进行组合,形成若干个包络性较强的主事故序列,同时编制了部分可随时插入(在不影响主事故进程的前提下)的情景片段,可组合形成若干个随时间演变的完整事故情景。本文给出了事故情景个例分析,并就如何提高事故情景设计的合理性和针对性提出了建议。     

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。     

Identification of initiating events for pressurized water reactor accidents

In order to aid operators in identifying the different initiating events as defined in the Final Safety Analysis Report (FSAR), we develop a novel identification procedure. The procedure is based on the monitoring of three key system parameters in a pressurized water reactor (PWR), i.e., the pressure, the average temperature, and the temperature difference of the hot-leg and cold-leg of the reactor coolant system. By monitoring the system thermal state diagram in a pressure–temperature space, an operator can easily identify what initiating event is taking place while a static point in the diagram starts to move. The event data pool is first established by storing the transient analysis results for events of different types using the optimal estimated RELAP5 model. Since the variation ranges of system key parameters at a specific time represent the specific character for each initiating event, the identification procedure can easily determine which cases in which the event data pool can be fitted to on-line data using only variation range comparison without complex calculations. This identification method is believed to be able to help the plant operator to identify the different events and then execute the Emergency Operating Procedure more effectively.     

中国实验快堆一回路冷阱工艺间钠火概率安全评价

本文运用事件树方法对中国实验快堆一回路冷阱工艺间发生钠火后的事故场景进行演绎分析,运用故障树方法对钠火相关系统进行可靠性建模。在此基础上计算得到各钠火事故序列的条件发生概率。结果表明:在获得的25个典型钠火事故序列中,19个序列的条件发生概率较低;在发生概率相对较高的6个序列中,4个序列的后果轻微,其余两个序列代表的钠火场景存在一定不确定性,需要在今后的钠火危险性评价中进一步具体研究。     

Human error probability estimation by coupling simulator data and deterministic analysis

Operator error in diagnosis and execution of task have significant impact on Nuclear Power Plant (NPP) safety. These human errors are classified as mistakes (rule base and knowledge based errors), slip (skill based) and lapses (skill based). Depending on the time of occurrence, human errors have been categorized as i) Category ‘A’ (Pre-Initiators): actions during routine maintenance and testing wherein errors can cause equipment malfunction ii) Category ‘B’ (Initiators): actions contributing to initiating events or plant transients iii) Category ‘C’ (Post-Initiators): actions involved in operator response to an accident. There have been accidents in NPPs because of human error in an operator's diagnosis and execution of an event. These underline the need to appropriately estimate HEP in risk analysis. There are several methods that are being practiced in Probabilistic Safety Assessment (PSA) studies for quantification of human error probability. However, there is no consensus on a single method that should be used. In this paper a method for estimating HEP is proposed which is based on simulator data for a particular accident scenario. For accident scenarios, the data from real NPP control room is very sparsely available. In the absence of real data, simulator based data can be used. Simulator data is expected to provide a glimpse of probable human behavior in real accident situation even though simulator data is not a substitute for real data. The proposed methodology considers the variation in crew performance time in simulator exercise and in available time from deterministic analysis, and couples them through their respective probability distributions to obtain HEP. The emphasis is on suitability of the methodology rather than particulars of the cited example.     

西安脉冲堆概率安全分析技术要点及分析框架研究

针对西安脉冲堆（XAPR）自身设计特点及安全特性,研究了XAPR概率安全分析（PSA）的技术特殊要点,提出了XAPR PSA分析框架及技术要素具体实施方法。最后以XAPR堆水池中破口失水事故为始发事件,验证了XAPR PSA研究思路。分析表明:以始发事件为起点、事件序列为主干、放射性释放类为终点的一体化事件树结构分析框架适合于XAPR PSA。      

Assessment methodology for confidence in safety margin for large break loss of coolant accident sequences

Deterministic Safety Analysis and Probabilistic Safety Assessment (PSA) analyses are used to assess the Nuclear Power Plant (NPP) safety. The conventional deterministic analysis is conservative. The best estimate plus uncertainty analysis (BEPU) is increasingly being used for deterministic calculation in NPPs. The PSA methodology integrates information about the postulated accident, plant design, operating practices, component reliability and human behavior. The deterministic and probabilistic methodologies are combined by analyzing the accident sequences within design basis in the event trees of a postulated initiating event (PIE) by BEPU. The peak clad temperature (PCT) distribution provides an insight into the confidence in safety margin for an initiating event.