基于动态故障树的核反应堆稳压器数字压力控制装置可靠性研究

以配置四取中逻辑输入模块的核电厂稳压器数字压力控制装置为研究对象，建立其故障树模型，包括四取中逻辑的动态部分和其他设备的静态部分，采用马尔科夫方法分析动态部分，再根据逻辑关系分析整体故障树，最后，围绕可靠度和重要度评价四取中逻辑的可靠性及其对整个装置可靠性的提升效果，结果表明:四取中逻辑在可靠性方面优化程度相对较高。      

基于BDMP的AP1000设备冷却水系统动态可靠性分析

设备冷却水系统（CCS）是一类具有双重冗余的可修复系统，冗余设备的交替运行以及故障设备的修复等因素对可靠性分析结果影响较大。传统故障树分析因缺乏对时间因素的描述，对于此类动态时序问题，分析假设过于保守。针对故障树分析的上述局限性，采用布尔逻辑驱动的马尔可夫过程（BDMP）方法对先进压水堆（AP1000）CCS进行了动态可靠性分析。然后基于BDMP模型导出了最简故障组合，计算了系统失效概率，并分析了系统失效的主要贡献因素。结果表明，CCS失效概率对设备冷却水泵的共因失效因子敏感，降低泵的共因失效可提高CCS可靠性。      

核电厂数字化仪控系统动态流程图方法分析研究

传统静态故障树方法对数字化仪控系统(DCS)的可靠性评价存在较大局限性,尤其是对包含动态特性和反馈信号的控制系统。作为新型概率安全分析方法的动态流程图法(DFM)具有动态特性,可表征系统变量和时间的关系、模拟逻辑信号控制,能够综合分析评价软件失效、硬件失效以及外部环境对系统可靠性的影响。本文选取CNP1000稳压器压力控制系统作为分析对象,假设发生稳压器高压信号缓解失效事件;分别应用DFM机制和故障树机制对其进行PSA建模分析,计算得到各自顶事件的发生概率、质蕴含与最小割集。最后通过两种模型结果的对比,总结分析出DFM方法较传统故障树方法的优势之处。     

秦山核电厂厂用供电系统动态分析和电源故障时主泵惰走的研究

利用数学模型对秦山核电厂厂用供电系统动态和在电源故障时主泵惰走进行了分析研究。为了简化分析和计算,在建立各类设备的数学模型时,只需正确地模拟那些对运行状态影响较大的因素,忽略一些次要的因素。通过秦山核电厂在电源事故情况下,厂用母线电压,频率的动态过程研究和主泵惰走的分析研究,阐述了这一问题分析研究的方法,给出计算结果并通过动态模拟试验进行验证。本文分析计算的结果可作为核电厂安全分析的支持性材料。     

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。     

秦山核电厂厂用供电系统动态分析和电源故障时主泵惰走的研究

利用数学模型对秦山核电厂厂用供电系统动态和在电源故障时主泵惰走进行了分析研究。为了简化分析和计算,在建立各类设备的数学模型时,只需正确地模拟那些对运行状态影响较大的因素,忽略一些次要的因素。通过秦山核电厂在电源事故情况下,厂用母线电压、频率的动态过程研究和主泵惰走的分析研究,阐述了这一问题分析研究的方法,给出计算结果并通过动态模拟试验进行验证。本文分析计算的结果可作为核电厂安全分析的支持性材料。     

海洋条件下冷却剂系统自然循环仿真模型

在对核反应堆堆芯冷却剂系统分析的基础上,建立了海洋条件下冷却剂系统自然循环运行状态仿真模拟的数学模型。模型分为稳态计算和动态计算两部分,包括修正的单通道模型、强迫外干扰力模型、两相沸腾计算和动态计算模型等。对于海洋条件,主要是针对左右摇摆的情况进行了计算,并对结果作了简单的分析。计算结果表明模型是可行的。     

考虑多失效行为的核电厂可修系统可靠性数值仿真

为分析含有顺序、冗余及功能相关等多失效行为的核电厂安全系统的可靠性,提出了基于动态故障树模型的可靠性数值仿真方法。通过对部件多失效行为的随机模拟及动态逻辑门成功准则的判定,实现了对含有多失效行为的核电厂安全系统的可靠性数值模拟。案例分析结果表明,该方法能对多失效行为的复杂系统进行可靠性分析,具有较强的适应性。     

上海EBIT动态控制电路设计

通过对上海EBIT装置离子注入引出系统的分析,采用FPGA为控制逻辑单元来设计EBIT装置的动态控制电路,实现了对EBIT装置中离子“注入-引出-电离”的动态控制。上海EBIT动态控制系统包括了基于FPGA的控制电路的设计和用户控制界面的设计。采用FPGA来设计EBIT动态控制系统,提高系统的控制速度和精度。     

蒸汽排放系统蒸汽冷凝器动态特性仿真研究

根据核动力商船高压定压蒸汽冷凝器的结构及工作特性,建立了适合蒸汽冷凝器实时仿真分析计算的两相流仿真模型。利用该仿真模型对蒸汽冷凝器进行了稳态计算和动态仿真分析,并将其稳态计算结果与蒸汽冷凝器试验结果进行对比验证。结果表明：该模型能准确模拟蒸汽冷凝器的动态特性,满足核动力装置蒸汽排放系统实时仿真分析要求。研究结果对二回路系统启动及停运的运行和控制系统设计具有指导意义。     

大亚湾核电站外电源系统可靠性分析中GO法的应用

应用GO法进行了大亚湾核电站外电源系统的可靠性分析，导出了具有共有信号时系统稳态可靠性指标的直接定量公式和两状态单元进行状态概率动态计算的公式，提出了求解系统是小故障事件集合的方法，求出外电源系统的所有最小故障事件集合及其故障概率，并进行了电网失稳后外源系统供电恢复概率的动态计算。结果表明，在可修系统的稳态和动态可靠性分析中GO法是一种十分简便而有效的方法。     

核电厂丧失厂外电源事件下柴油发电机系统动态响应分析

将系统可靠性分析方法GO法与Markov法相结合,对核电厂概率安全分析(PSA)中厂外电源丧失(LOOP)后柴油发电机应急响应系统在24h内缓解全厂断电(SBO)事件中的动态过程进行分析,解决了维修相关存在下可修系统可靠性精确计算问题,并通过创建GO法“备用门”操作符真实地模拟应急响应系统工作的逻辑关系.通过将2种可靠性分析方法相结合使用的尝试,使之与柴油发电机应急响应系统存在维修相关的实际情况相适应,拓展了2种方法的分析领域,同时能够更为精确地得出SBO对系统安全运行的影响.     

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.     

A new assessment method for demonstrating the sufficiency of the safety assessment and the safety margins of the geological disposal system

A new method for demonstrating the sufficiency of the safety assessment and safety margins of the geological disposal system has been developed. The method is based on an existing comprehensive sensitivity analysis method and can systematically identify the successful conditions, under which the dose rate does not exceed specified safety criteria, using analytical solutions for nuclide migration and the results of a statistical analysis. The successful conditions were identified using three major variables. Furthermore, the successful conditions at the level of factors or parameters were obtained using relational equations between the variables and the factors or parameters making up these variables. In this study, the method was applied to the safety assessment of the geological disposal of transuranic waste in Japan. Based on the system response characteristics obtained from analytical solutions and on the successful conditions, the classification of the analytical conditions, the sufficiency of the safety assessment and the safety margins of the disposal system were then demonstrated. A new assessment procedure incorporating this method into the existing safety assessment approach is proposed in this study. Using this procedure, it is possible to conduct a series of safety assessment activities in a logical manner.     

Discrete event dynamic system (DES)-based modeling for dynamic material flow in the pyroprocess

A modeling and simulation methodology was proposed in order to implement the dynamic material flow of the pyroprocess. Since the static mass balance provides the limited information on the material flow, it is hard to predict dynamic behavior according to event. Therefore, a discrete event system (DES)-based model named, PyroFlow, was developed at the Korea Atomic Energy Research Institute (KAERI). PyroFlow is able to calculate dynamic mass balance and also show various dynamic operational results in real time. By using PyroFlow, it is easy to rapidly predict unforeseeable results, such as throughput in unit process, accumulated product in buffer and operation status. As preliminary simulations, bottleneck analyses in the pyroprocess were carried out and consequently it was presented that operation strategy had influence on the productivity of the pyroprocess.     

基于经验-解析数据的某核电安全壳地震易损性分析

本文基于混合数据的地震易损性分析方法,对我国已运行核电厂地震易损性分析进行研究。首先基于地震危险性分析和分解结果,生成了我国华南地区某核电厂厂址条件谱;然后采用贪心优化算法,选取符合厂址危险性的地震动记录;基于增量动力分析方法,生成我国某核电厂安全壳地震易损性安全系数FS和FSA的解析数据;地震易损性其他参数采用经验数据,基于经验-解析数据,生成了我国某核电厂安全壳地震易损性曲线。建议将基于经验-解析数据的地震易损性分析方法应用于我国核电厂安全壳初步地震易损性分析中。     

Software failure events derivation and analysis by frame-based technique

A frame-based technique, including physical frame, logical frame, and cognitive frame, was adopted to perform digital I&C failure events derivation and analysis for generic ABWR. The physical frame was structured with a modified PCTran-ABWR plant simulation code, which was extended and enhanced on the feedwater system, recirculation system, and steam line system. The logical model is structured with MATLAB, which was incorporated into PCTran-ABWR to improve the pressure control system, feedwater control system, recirculation control system, and automated power regulation control system. As a result, the software failure of these digital control systems can be properly simulated and analyzed. The cognitive frame was simulated by the operator awareness status in the scenarios. Moreover, via an internal characteristics tuning technique, the modified PCTran-ABWR can precisely reflect the characteristics of the power-core flow. Hence, in addition to the transient plots, the analysis results can then be demonstrated on the power-core flow map. A number of postulated I&C system software failure events were derived to achieve the dynamic analyses. The basis for event derivation includes the published classification for software anomalies, the digital I&C design data for ABWR, chapter 15 accident analysis of generic SAR, and the reported NPP I&C software failure events. The case study of this research includes: (1) the software CMF analysis for the major digital control systems; and (2) postulated ABWR digital I&C software failure events derivation from the actual happening of non-ABWR digital I&C software failure events, which were reported to LER of USNRC or IRS of IAEA. These events were analyzed by PCTran-ABWR. Conflicts among plant status, computer status, and human cognitive status are successfully identified. The operator might not easily recognize the abnormal condition, because the computer status seems to progress normally. However, a well trained operator can become aware of the abnormal condition with the inconsistent physical parameters; and then can take early corrective actions to avoid the system hazard. This paper also discusses the advantage of simulation-based method, which can investigate more in-depth dynamic behavior of digital I&C system than other approaches. Some unanticipated interactions can be observed by this method.     

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。     

An Overview of R&D Work Performed in France Concerning the Risk of Brittle Fracture of Transport Casks

Abstract

An overview is given of the methodology followed in France to evaluate the risk of brittle fracture in transport casks. Various aspects are considered in this evaluation. First: the material characterisation: effect of strain rate and temperature on tensile properties and toughness. Secondly: evaluation of applied stress under dynamic loading and criteria to define worst cases (time, distribution). Thirdly: the global validation is presented to demonstrate that the whole methodology is conservative; consideration of representative analytical cases where initiation is obtained under controlled conditions. Results are presented concerning cast ductile iron and carbon steel materials.     

Inhibition of Radioactive Contamination of Stainless Steel in Boiling Water Reactor

In order to investigate the later phase of a sodium-water reaction (SWR) event, the code SELPSTA (Sodium-water reaction Event Later Phase System Transient Analyzer) has been developed and the analysis for the long-term system dynamic responses of a SWR event in KALIMER (Korea Advanced Liquid MEtal Reactor) has been made. The SELPSTA code uses the very simple analysis model applied only to the reaction period characterized by a bulk motion, and makes the very quick and concise computation possible. The code reasonably predicts the quasi-steady system transients and has the superiority in the aspect that the various design parameters or operational characteristics are flexibly applicable. In the long-term period of a SWR event, the system dynamic responses analyzed by the code totally depend on the system design parameters such as the breaking pressure of the rupture disk, the variation of the steam injection rate and the sodium drain tank pressure,etc. Based on these analyses results, it is expected that the numerical quantification method of the SELPSTA code is practicable for the long-term system transient analysis and also makes the design of a pressure relief system against a SWR event in a liquid metal reactor (LMR) possible.