Research on the attribution evaluating methods of dynamic effects of various parameter uncertainties on the in-structure floor response spectra of nuclear power plant

Consideration of the dynamic effects of the site and structural parameter uncertainty is required by the standards for nuclear power plants （NPPs） in most countries. The anti-seismic standards provide two basic methods to analyze parameter uncertainty. Directly manually dealing with the calculated floor response spectra （FRS） values of deterministic approaches is the first method. The second method is to perform probability statistical analysis of the FRS results on the basis of the Monte Carlo method. The two methods can only reflect the overall effects of the uncertain parameters, and the results cannot be screened for a certain parameter＇s influence and contribution. In this study, based on the dynamic analyses of the floor response spectra of NPPs, a comprehensive index of the assessed impact for various uncertain parameters is presented and recommended, including the correlation coefficient, the regression slope coefficient and Tornado swing. To compensate for the lack of guidance in the NPP seismic standards, the proposed method can effectively be used to evaluate the contributions of various parameters from the aspects ＆sensitivity, acuity and statistical swing correlations. Finally, examples are provided to verify the set of indicators from systematic and intuitive perspectives, such as the uncertainty of the impact of the structure parameters and the contribution to the FRS of NPPs. The index is sensitive to different types of parameters, which provides a new technique for evaluating the anti-seismic parameters required for NPPs.

Construction parameters of graded sand-gravel foundation on seismic response law of nuclear safety grade underground corridor

The treatment of soft soil foundation under nuclear safety grade corridors with graded sand and gravel materials has a good development prospect.It is of great engineering value to explore the influence of construction parameters of graded sand and gravel foundation on the seismic response of gallery structures.Taking the safety grade underground corridor of a nuclear power plant as the engineering background,the equivalent linear method is used to consider the nonlinear dynamic characteristics of graded sand and gravel.The energy transfer boundary is applied at the truncation boundary to simulate the dissipation effect of scattered wave fluctuation energy and the ground motion input.The thicknessless contact element is introduced to consider the contact effect between the corridor structure and the graded sand and gravel foundation,so as to establish the calculation model of the dynamic interaction between the graded sand and gravel foundation and the corridor structure.Furthermore,the influence of the relative compactness and the foundation treatment depth on the seismic response of the corridor structure is studied,and the calculation results of the acceleration response spectrum and relative displacement of the corridor structure are analyzed.The calculation results show that the two construction parameters have different degrees of influence on the seismic response of corridor structure.The research results can provide reference for the engineering design and construction of underground corridors,and provide technical support for the application of graded gravel materials in soft soil foundation treatment.

Reliability-Centered Maintenance-Based Maintenance Decision in New Nuclear Power Plants

Taking the project of introducing reliability-centered maintenance( RCM) into maintenance decision in an AP1000 nuclear power plant( NPP) under construction as the research object,an improved RCM methodology was proposed, and the application software and an RCM-based maintenance strategies management system were designed. In the pilot project,the RCMbased maintenance decision methodology had been applied to determining the maintenance strategies for two systems. Both the decision process and the results were described in this paper. The achievements of this project promoted the introduction and routinization of an advanced and effective maintenance decision mode in nuclear power field,which could provide valuable reference for new NPPs in China.

Identification of Parameters in 2D-FEM of Valve Piping System within NPP Utilizing Seismic Response

Nuclear power plants(NPP)contain plenty of valve piping systems(VPS’s)which are categorized into high anti-seismic grades.Tasks such as seismic qualification,health monitoring and damage diagnosis of VPS’s in its design and operation processes all depend on finite element method.However,in engineering practice,there is always deviations between the theoretical and the measured responses due to the inaccurate value of the structural parameters in the model.The structure parameters identification of VPS within NPP is still an unexplored domain to a large extent.In this paper,the initial 2D-finite element model(FEM)for VPS with a DN80 gate valve was updated by utilizing seismic response.The objective function used in the model updating procedure is the vibration control equation error of the VPS.The experimental results show that the updated 2D-FEM can accurately predict the original dynamic characteristic of the VPS.It was also found the Rayleigh damping coefficients corresponding to the VPS vary slightly with the change in seismic excitation amplitude.The research displayed the complete procedure of updating the complex structured initial FEM by utilizing seismic response,and the results show that the parameters can be accurately identified even if the seismic response used for updating merely contained the fundamental frequency information of the structure.

Study of operator reliability in nuclear power plants

The study of the reliability of operator’s diagnosis and decision-making during abnormal events with HCR model has been introduced. A new approach has been proposed by the authors to deal with the classification of HIs. The modification of T0.5 by PSFs coefficients has been discussed. In addition, it is emphasized that T0. 5 and SD(σ) of response times can be used as a scale to predict and evaluate the reliability of crew/operator behavior. These processings provide improvement and refinement to HCR model and its practical application. The experiment for operator response time of Chinese NPP using a lull-scale simulator has proved the above-mentioned processings and presented the quantitative results of operator reliability in Chinese NPP.

Impact assessment of nuclear power plant discharge on zooplankton abundance and distribution in coastal waters of Kalpakkam, India

Background:Kalpakkam coastal area serves as fishing ground,influenced by anthropogenic effluents from its surrounding populations and backwaters.Methods:The monthly abundance and distribution of zooplankton and some physico-chemical parameters were investigated in coastal waters near a nuclear power plant,Kalpakkam 2011.Zooplankton and surface water samples collected from three different locations(stations P1,P2 and P3)in the heated and non-heated parts of the coastal waters were analyzed using standard methods.Results:Among all plankton genera in the cooling water,zooplankton was observed most sensitive group to entrainment-induced stress.Thirty genera were identified over the period of the investigation.Zooplankton population followed a trend as P1>P3>P2.A clear monthly variation and slight differences among stations were observed in terms of the zooplankton genera.The investigations revealed that changes in zooplankton genera across water bodies could be associated with differing temperature.Conclusion:Our results clearly demonstrate that heated water discharge from the Madras Atomic Power Station has a negligible effect on zooplankton population at engineered canal near Plutonium Recycle Project(station P2).There is no impact on zooplankton population due to thermal effluent discharge into the coastal area(stations P1 and P3).Hence,further investigation is required to estimate the impact zone at either side during different seasons.

模拟事故工况下非能动核电厂安全相关涂层的可靠性测试及评估方法研究

安全相关涂层在非能动核电厂中起着重要的作用,涂层的失效会影响核电厂安全系统的功能执行,影响核安全。国内外核监管机构对于在设计基准事故(DBA)工况下涂层系统的可靠性及评估方法非常重视。文章结合非能动核电厂涂层系统的工程应用,针对其在DBA下的可靠性及评估方法进行了研究。研究表明:在DBA下非能动核电厂安全相关涂层的可靠性要综合考虑涂层的模拟DBA性能、干膜密度、导热性能等。而非能动核电厂安全相关涂层工程应用,则需从涂层的模拟DBA性能、干膜密度、导热性能、涂层碎片(数量、大小、位置和性能等)以及包络涂层碎片后的碎片裕量等角度进行综合评估,以确定在事故工况下涂层的可靠性,不对系统安全产生影响,保证核电厂更安全、高效和经济性运行。

2023年冬季宁德核电冷源海域棘头梅童鱼(Collichthys lucidus)资源声学评估

滨海核电厂冷却水源的取水安全正受到海洋生物暴发引起的堵塞威胁,鱼类堵塞核电冷源的事件时有发生。声学技术在海洋生物探测方面具有高效、实时的优势,因此,被广泛应用于海洋生态系统研究。为了解冬季宁德核电厂冷却水源取水海域棘头梅童鱼(Collichthys lucidus)的密度和运动规律,利用声学技术对相关海域冷却水源生物量进行了测量和评估。研究结果显示,棘头梅童鱼为研究海域的优势种,拖网调查显示平均尾数密度为1964尾·km^(-2),平均质量密度为0.483 kg·km^(-2)。棘头梅童鱼丰度在研究海域向核电厂方向减少,质量密度以核电厂温排水区域站点最高。通过对鱼类回波特征和个体轨迹的追踪统计,分析了该海域冬季内单体目标强度检测及其空间结构分布、资源现存量和资源丰度密度的时空分布变化。声学走航海域鱼类平均密度为2.78×10^(5)尾·n mile^(-2),生物量密度均值为1.41×10^(4) kg·n mile^(-2),棘头梅童鱼资源总量为7.47×10^(5) kg。研究表明,声学技术在宁德核电厂取水海域冬季冷却水源生物量测量中具有明显优势。通过不断改进和优化方法,声学技术将能更好地支持核电厂取水海域的生态环境监测与资源管理,为冷却水源取水安全提供科学依据。

核电厂动力机器模型基础动力参数测试的优化

核电厂汽轮机厂房多位于坚硬的岩石地基上,模型基础动力参数测试共振频率达到100 Hz以上,土质地基上常用的测试方法由于其机械式激振器扰频上限低的缺陷,对核电厂岩石地基适用性很差。在激振器、垫层、激振器与模型基础的连接方式、测试方式等方面进行研究,优化、改进了测试设备和测试方法,开发了全自动常扰力测试控制系统,使之更适合核电厂岩石地基,测试结果更准确地代表岩石地基的动力特性。

基于CANDU型核电站的高精度线性电源分析设计

40 V仪用直流电源国产化势在必行。在对电源应用场景和性能指标深入分析的基础上,设计出一款具有高可靠性、高性能、符合人机交互工学界面等特点的电源,第三方核级抗震试验、型式试验、老化试验、EMC试验结果表明该款40 V仪用直流电源的各项指标可完全比拟原用国外产品,并在电源安全防护和人机交互方面具有更好的性能。

福清核电厂5^(#)、6^(#)机组反应堆保护系统的设计

反应堆保护系统是核电厂应对预期运行事件或设计基准事故而设置的重要安全级仪控系统。为满足“华龙一号”三代核电高安全性目标的设计要求,需要对反应堆保护系统的功能性能进行提升。以“华龙一号”示范工程福清核电厂5#、6#机组为研究对象,通过充分考虑保护系统设计准则要求并基于事故分析结论,提出了更具高可靠性和全面保护功能特征的反应堆保护系统设计方案。详细介绍了系统自动保护功能的设计、系统结构设计以及系统健康管理等内容,并对系统拒动率进行了定量分析。分析结果表明,系统总体拒动率不高于5.0×10^(-7)/指令,达到了国际主流三代核电的可靠性水平。该研究成果可为后续核电厂的反应堆保护系统设计或优化提供借鉴。

核电厂仪控系统本体可靠性分析研究

核电厂仪控系统本体可靠性是影响核电厂安全、稳定运行的重要因素。研究了核电厂仪控系统设计过程中影响仪控系统本体可靠性的影响因素。提出了包含4个主要影响因子和13个子影响因子的人因失误模型。应用灰色关联分析方法对人因失误模型中各影响因子进行定量关联度计算,以识别影响因子的重要程度。对某实际核电机组仪控系统设计活动进行实例分析,成功地识别出了核电厂仪控系统设计过程中影响仪控系统本体可靠性的关键因素。该研究结果有助于更好地识别和管理潜在的人因失误风险,并提高系统的可靠性和安全性。

基于风洞实验的丘陵下垫面气溶胶大气扩散特性研究

核电厂正常运行或发生核事故时,会向大气释放一定量的放射性气溶胶,研究气溶胶在复杂下垫面的大气扩散特性对实现气载流出物辐射影响的准确评价具有重要意义。高斯烟羽或烟团模型被广泛用于大气扩散评价中,模型中的扩散参数描述了气溶胶受湍流作用扩散的强弱,需要针对不同的地形如平坦、丘陵下垫面进行校正。本文针对典型下垫面开展了气溶胶大气扩散风洞实验,采用激光粒子图像扩散参数估计方法,针对释放源高度不同的多组工况,统计分析了竖直截面内高分辨率的气溶胶浓度,根据高斯分布估计了下风向烟羽的上下边界及垂直方向扩散参数。结果表明:烟羽经过山体会出现整体抬升现象,这将导致山体后近地表浓度有所下降;烟羽经过山体前后出现浓度中心线两侧浓度分布不均现象,远离山体烟羽轨迹回归类高斯分布;释放源高度小于山体1/2高度的烟羽经过山体后轨迹趋于一致。高斯模型在丘陵下垫面适用度有限。

核电仪控实时操作系统软件需求研究

实时操作系统作为核电厂数字化仪控系统的核心组成部分,在核电厂数字仪控系统中发挥重要作用。但我国核电厂应用的实时操作系统严重依赖进口。为打破国外技术壁垒、实现核电厂的实时操作系统完全自主化,对适用于核电厂的实时操作系统软件需求进行研究。深入学习了实时操作系统的经典书籍,研究了核电软件法规和实时操作系统相关标准,调查了实时操作系统在核电数字仪控系统中的应用需求,分析了在核电厂成熟应用的友商产品的软件需求。提出了适用于核电厂数字仪控系统的可靠性和安全性的实时操作系统软件需求,包含外部接口需求、功能需求、性能需求和软件系统属性等。该研究为后续实现独立自主的核电仪控实时操作系统软件的设计及研制奠定了基础。

核电厂隔震结构附加侧向阻尼系统分析模型及减震效果

提出对半埋置/全埋置小堆三维隔震结构附加侧向黏滞阻尼的混合控制系统,达到同时控制水平加速度、水平位移以及摇摆反应且不影响竖向隔震效果的目的。基于隔震层及侧向阻尼系统变形分析,建立考虑隔震层平动及摇摆的耦合效应和侧向阻尼器协调转动变形的刚体动力学模型。基于RG1.60谱选取30条地震输入信号开展参数分析,探究阻尼布置参数、附加阻尼力参数及阻尼滞回形状参数对减震效应的影响。研究发现合理的附加阻尼力与隔震层出力比值区间为10%~20%,最优参数下水平加速度位移可同时分别减小20%,40%,最大摇摆反应减小70%。选取合理阻尼参数对某真实核电厂模型进行案例分析,摇摆角、加速度、隔震层位移等地震响应指标均减小,边支座受拉现象消失,案例数值模拟结果与参数分析规律一致,也与理论分析吻合。

核电站数字化保护系统控制层通信协议的设计

核电站数字化保护系统中的控制层通信承担着重要的数据和信息传输任务,为控制单元提供现场数据和保护输出,是核电站数字化保护系统中的重要组成部分。现有的控制层通信协议无法满足核电站数字化保护系统确定性、可靠性、独立性和实时性的要求。采用基于状态、通信周期固定、通信链路冗余、只传输固定的数据集、帧长度和格式固定等方式,设计了控制层通信协议。根据核电标准要求,针对通信协议中可能存在的故障类型进行了分析,同时给出了错误检测方法和相应的处理措施,以满足核电站数字化保护系统确定性、可靠性、独立性和实时性的要求。该通信协议已经通过权威机构认证,所涉及的硬件产品已经通过核安全级设备的鉴定,后续可应用于军工、航天等高可靠性行业。

核电厂严重事故人员可靠性分析研究现状与展望

福岛事故缓解过程中暴露出的较多人因问题,使得核电厂严重事故场景下的人员可靠性逐渐成为研究热点。为分析国内外核电厂严重事故人员可靠性的研究现状,探究后续的研究方向,本文基于文献调研、核电厂走访、专家访谈等方法,阐述了开展核电厂严重事故下人员可靠性分析研究的必要性,从国内外核电厂严重事故管理、HRA方法、严重事故人员可靠性研究等3个方面对核电厂严重事故人员可靠性分析的研究现状进行评述,指出国内外针对核电厂严重事故场景的HRA方法研究还处于初步阶段,并题出未来的研究工作应从应急人员行为模型、人员失误因果模型、人因数据的收集和规范化等方面开展。

多机组核电厂机组间协作行为对组织可靠性的影响分析

“一址多堆”核电厂设计方案为核电的安全运营维护带来了新的挑战,组织人因失误是导致多机组核电厂事故的最常见的致因因素。目前学界针对多机组运行模式下的组织人因失误机理尚未形成统一的理论,相关分析方法也是在单机组背景下开发的,针对多机组核电厂机组间协作行为对组织可靠性的影响机理方面的研究不足。本文以Furuta提出的组织可靠性模型为基础,结合多机组场景下的组织行为特征,建立多机组核电厂组织可靠性数学网络模型;以双机组核电厂为例,模拟分析了机组间协作行为对多机组核电厂组织可靠性的影响。研究表明:机组间协作可以有效提升资源不足情况下的组织可靠性。相较于单机组独立运行,机组间协作对于组织内个别成员能力不足的容忍度更高。

基于简化模型马尔可夫法的核电厂仪控系统可靠性评估方法

核电厂数字化仪控系统作为核电厂“中枢神经”,其可靠性决定了核电厂是否能安全稳定运行。在使用马尔可夫法针对核电厂数字化仪控系统进行可靠性建模分析时,往往由于系统过于复杂、状态交互关系较多,出现组合爆炸的情况,使得模型难以构建。本文提出了一种应用于复杂系统的马尔可夫模型简化方法,通过对马尔可夫模型中的状态进行检测,筛选出具有相同转移率的状态后组成简化状态组,将简化状态组中的成员进行合并简化。进一步地,通过选取1oo2(1 out of 2)架构作为实施案例,对提出的简化方法进行验证,利用本研究提出的简化方法,将原有的1oo2架构8阶模型简化到6阶,有效降低了马尔可夫模型状态数量,提升建模及计算分析效率,一定程度上解决了组合爆炸问题。本研究提出的马尔可夫简化方法可应用于复杂系统建模分析,为核电厂数字化仪控系统可靠性评价提供参考。

核电厂复杂系统的维修规则可靠性指标的制定技术改进

本文概述了核电厂维修规则性能指标的制定方法,结合CPR1000核电机组低压安注系统的维修规则的可靠性指标的制定过程,特别针对工艺流程复杂的系统的可靠性指标制定提出技术改进。本文提出通过“按功能”或者“按设备”的方法,对系统中的若干个相关设备进行分组以设置“设备级”指标,以解决对设备组合设定指标的困难;针对复杂系统可靠性指标计算,运用泊松分布统一描述需求失效和运行失效,并使用失效率期望值来简化计算过程,以解决复杂系统中大量设备的失效率计算困难。本文对其中运用的数学工具均进行了合理的解释和证明。