Aseismic safety analysis of a prestressed concrete containment vessel for CPR1000 nuclear power plant

Abstract： The containment vessel of a nuclear power plant is the last barrier to prevent nuclear reactor radiation. Aseismic safety analysis is the key to appropriate containment vessel design. A prestressed concrete containment vessel （PCCV） model with a semi-infinite elastic foundation and practical arrangement of tendons has been established to analyze the aseismic ability of the CPR1000 PCCV structure under seismic loads and internal pressure. A method to model the prestressing tendon and its interaction with concrete was proposed and the axial force of the prestressing tendons showed that the simulation was reasonable and accurate. The numerical results show that for the concrete structure, the location of the cylinder wall bottom around the equipment hatch and near the ring beam are critical locations with large principal stress. The concrete cracks occurred at the bottom of the PCCV cylinder wall under the peak earthquake motion of 0.50 g, however the PCCV was still basically in an elastic state. Furthermore, the concrete cracks occurred around the equipment hatch under the design internal pressure of 0.4MPa, but the steel liner was still in the elastic stage and its leak-proof function soundness was verified. The results provide the basis for analysis and design of containment vessels.

Impact Analysis of the 2011 Fukushima Nuclear Power Plant Accidents by Running Spectrum Analysis on Newspaper

Huge amount of digital data of the Great East Japan Earthquake is provided by the highly-developed digital data technology. But the method and technique for analysis of these huge digital data are not developed sufficiently. This paper proposes a running spectrum technique for text data and analyzing changes of disaster phase during the disaster management cycle. Impact analysis of the nuclear power plant accidents have been performed by using Fukushima Minpo newspaper for its verification. The result shows the dynamic characteristics of the nuclear power plant accidents. As the time interval B becomes longer, the analysis data is used from wide range period along with the smoothing effect. When observing different time intervals B, fewer keywords have been ranked in the longer time intervals of B. The proposed technique is a powerful tool to effective and efficient disaster response and management. analyze effectively the huge amount of digital data for the

Development of nuclear power plant real-time engineering simulator

A nuclear power plant real-time engineering simulator was developed based on general-purpose ther- mal-hydraulic system simulation code RELAP5. It mainly consists of three parts:improved thermal-hydraulic system simulation code RELAP5,control and protection system and human-machine interface. A normal transient of CHASHMA nuclear power plant turbine step load change from 100% to 90% of full power,was simulated by the en- gineering simulator as an application example. This paper presents structure and main features of the engineering simulator,and application results are shown and discussed.

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.

Upgrade to Nuclear Power Plant Krsko Internal Flooding Probabilistic Safety Analysis

The original internal flooding probabilistic safety analysis （PSA） study of Krsko Nuclear Power Plant （two-loop Pressurized Water Reactor （PWR） plant of Westinghouse design） was performed in mid nineties and limited to reactor core damage risk （Level 1 PSA）. In 2003, it was, together with other safety and hazard analyses, subject to the Periodic Safety Review （PSR）. In the PSR, it was stated that methodological PSA approaches and guidelines have evoluted during the past decade and several observations were provided, concerning the area screening process, residual risk and treatment of plant damage states and risk from radioactivity releases （i.e., Level 2 PSA）. In order to address the PSR observations, upgrade ofKrsko NPP internal flooding PSA was undertaken. The area screening process was revisited in order to cover the areas without automatic reactor trip equipment. The model was extended to Level 2. Residual risk was estimated at both Level 1 and Level 2, in terms of core damage frequency （CDF） and large early release frequency （LERF）, respectively.

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.

Theoretical analysis and experimental study on the dynamic behavior of a valve pipeline system during an earthquake

To study the dynamic behavior of pipeline systems installed with large-mass valves within nuclear power plants during earthquakes,seismic simulation tests are carried out on a pipeline system equipped with a DN80 gate valve,and the FEM updating technique is used to identify the stiffness distribution of the valve.By conducting tests and a numerical analysis,the following conclusions are obtained:After a large-mass valve is installed in the pipeline,the system shows higher sensitivity to intermediate and high frequency components in the earthquake than low frequency components.It is possible for the intermediate frequency components to be amplified by the valve in the horizontal direction,while the pipes tend to amplify the high frequency components in horizontal and vertical directions.Changes in the high-order modes of the system depend on valve stiffness distribution.Since the existence of a valve makes pipeline system damping distribute with an obvious non-proportional feature,when the response spectrum method is used to calculate the response of the pipeline system,it could result in an underestimation of low-damping positions and overestimation of high-damping positions.

红沿河核电厂地震仪表系统震后数据分析

2023-08-23红沿河核电厂地震仪表系统成功记录到大连市普兰店区发生的4.6级地震,这是我国首次完整记录到核电厂附近中强震数据。对该系统记录到的地震数据进行分析,时程信号波形回放表明,本次地震记录数据清晰完整,自由场数据峰值与地震衰减经验公式计算结果相符。系统7个监测点加速度峰值对比分析表明,核电厂厂房对地震加速度信号具有放大效应,放大系数与建筑物标高正相关。加速度峰值数据频谱分析结果显示,核电厂厂房地震响应数据的卓越频率主要集中在10~20 Hz范围内,该范围内的地震加速度信号对厂房仍具有较大破坏性,这一结论与核电厂地震仪表准则NB/T 20076-2012中规定的地震触发频带(1~10 Hz)不符。鉴于我国核电厂地震仪表系统的地震触发滤波通频带设定范围为1~10 Hz,这一缺陷会降低地震仪表系统纵深安全防御的性能。

地震载荷对核电阀门可运行性影响机理研究

核电阀门在地震期间可否运行主要通过试验进行鉴定,为了降低核电阀门设计成本,以某闸阀为例,总结基于分析法的核电阀门可运行性抗震鉴定流程,并阐明地震对阀门可运行性的影响机理。运用有限元法研究阀门启闭过程中遭受地震载荷时其各部位的应力与接触压力的变化,分别基于可运行性和结构完整性对阀门进行抗震校核。结论如下:地震对阀门可运行性的影响在于放大了阀门滑动摩擦部位的接触压力,导致阀门闭合所需要克服的摩擦力增加,延后阀门闭合动作;与结构完整性分析相比,基于可运行性分析的抗震鉴定过程更加全面、苛刻。

基于中国场地相关谱的核电厂结构地震易损性与风险研究

随着概率地震危险性分析与分解理论与应用的发展,场地相关谱生成理论也得到了不断发展,基于中国场地相关谱的核工程等重要基础设施地震易损性与风险分析研究还较为匮乏。该文总结了标量型中国概率地震危险性分析与分解理论方法,提出了向量型中国概率地震危险性分析与分解、条件型中国概率地震危险性分析基本原理,给出了基于中国概率地震危险性分析与分解的我国场地一致危险谱、条件均值谱、广义条件均值谱和条件一致危险谱生成理论和方法,总结了基于中国场地相关谱的核电厂结构地震易损性与风险分析理论基础,以我国某核电厂厂址及核电厂安全壳结构为算例,生成算例厂址场地相关谱,计算不同场地相关谱条件下核电厂安全壳结构地震易损性与风险。分析结果表明:不同场地相关谱条件下,我国核电厂安全壳结构安全裕量都较大;基于条件均值谱计算得到的风险结果偏于不保守。

非核级蓄电池柜安装方式对抗震性能影响研究

电气设备安装方式一般分为螺栓连接和焊接。为验证2种安装方式对电器设备抗震性能的影响,将2台完全一样的蓄电池柜用不同安装方式固定在振动台上进行抗震试验。试验结果表明:焊接机柜的加速度响应大于螺栓连接机柜的,共振频率较高,机柜位移较小,整体刚度较大,焊接部位出现较大的应力;螺栓连接机柜因力矩作用易出现底梁紧固失效,电池及机柜变形大、应力大的问题。为提高机柜的抗震性能,需采取必要的措施来减少机柜变形及蓄电池与机柜之间的碰撞。大质量、高重心的电池柜宜采用焊接方式安装,以减少几何变形对蓄电池柜抗震性能的影响。

基于有限元法的核电厂防异物堵板分析与优化

为探究不同材料蜂窝结构堵板对核电厂作业工况的减重效果与相应的力学性能,借助有限元分析软件(ANSYS)分别对各个材料以实心堵板、蜂窝堵板和空心结构堵板进行静态变形计算与评定和动态仿真分析。首先进行静态载荷分析,对堵板的关键部位施加法向垂直荷载模拟工作中持续荷载;对堵板采用动态冲击仿真分析与实验,比较了五种材料的三种不同结构堵板主要受载荷部位的变形影响;通过冲击仿真,对比动力学冲击形变量与塑性变形量,判断不同材料的力学特性以找到选取最优解。结果表明:在施加静态载荷与动态冲击时,蜂窝结构堵板的减重效果最佳,在不同的加载条件下性能最好;7010铝合金重量轻,承载能力强,能很好地吸收冲击,更适合应用于核电厂堵板。采用7010蜂窝铝板,综合加工工艺优化与样件实验,堵板重量由19.3 kg降低为13.2 kg,极大地提高了产品的实用性。

IEC 63096核电厂仪控系统网络安全管控标准与国内等级保护相关标准的协调分析

伴随着全球工业数字化、网络化和智能化的发展,传统基于物理隔离的保护方式已无法确保核电厂仪控系统的网络安全。核电厂仪控系统遭受网络攻击不仅可能导致生产过程中断,还可能引起核安全事件,因此核电厂仪控系统网络安全已引起世界各国及相关国际组织的高度关注。国际电工委员会于2020年发布了关于核电厂仪控系统网络安全防范管控的国际标准IEC 63096,为核电厂仪控系统各相关方提供了基于网络安全防范等级和生命周期阶段的具体指引,用于指导核电厂实施网络安全管控措施,以预防、检测和处置网络攻击。同时,等级保护制度作为我国网络安全的基础制度,是国内各核电厂必须开展的规定工作。为此,本文分别对IEC 63096以及等级保护系列标准进行了介绍,重点对二者在安全等级及管控措施方面的协调性进行了分析,从而帮助核电厂在进行网络安全管控措施的部署时有效降低时间成本和投资成本。

地震作用下桩-土-核电结构相互作用的异步分析方法

为拓展核电厂的选址范围,有必要对非基岩场地桩基情形的核电结构进行地震安全性评估。在目前的桩-土-结构相互作用分析方法中,Winkler地基梁模型以及p-y法都将桩-土-结构相互作用问题进行了简化,难以反映复杂地基情形。整体有限元法可考虑复杂地基情形,但计算量较大,效率较低。本文基于高效的三维时域土-结构相互作用分区分析(Partitioned Analysis of Soil-Structure Interaction,PASSI)方法,实现桩基与土体分别采用不同时间步距的计算方法,避免土体采用桩基相对较小的时间步距而增加不必要的计算量。本文以AP1000核岛结构作为研究对象,建立了桩-土-核电结构相互作用的三维有限元模型并对其进行分析。通过输入脉冲波验证了该异步算法的有效性,并结合运动相互作用和惯性相互作用,分析了桩身最大剪力和最大弯矩的特点。分析了桩-土-核电结构在地震波输入下的响应。由于桩的自由度数相对于土体的自由度数可以忽略不计,采用桩-土异步算法时,桩附加的计算量可以忽略,这种高效方法有望用于大型核电结构的桩-土-结构动力相互作用分析中。

核电厂离线啜吸检测装置抗震性能分析与评定

啜吸检测装置是核电厂燃料操作与贮存系统中的检查设备,分析其在地震载荷下的安全性能具有重要的意义。为了验证设备在地震工况下的安全性和可靠性,校核设备是否满足强度要求,应用有限元软件ABAQUS对啜吸装置进行了建模和抗震计算分析。介绍了设备结构、载荷组合和使用限制、地震反应谱法的分析过程及模型边界条件设定,采用反应谱法计算得到了设备在地震载荷下的响应,并依据相关规范对设备主要部件(筒体、结构件、定位销和锚固螺栓)在静态载荷、地震等多种工况载荷组合作用下的应力进行了强度校核和评定。结果表明,啜吸装置的结构强度满足规范要求,在地震工况载荷作用下能够保证设备结构的完整性并可靠运行。通过抗震计算分析,为啜吸装置优化设计提供了参考依据,对保证核电厂设备的正常可靠运行具有重要意义。

基于故障树分析识别核动力厂电气系统火灾共模故障

本研究在工程实践的基础上,建立了一种基于故障树分析识别电气系统火灾共模故障的方法。通过建立故障树,并采用逻辑代数进行计算分析,识别出电气系统的火灾共模故障。本研究能够系统性地识别出电气系统的火灾共模故障,减少主观判断,提高识别结果的准确性,可以广泛地应用于核动力厂的火灾安全分析和防火安全设计。

基于故障树分析的蒸汽隔离阀健康评估方法

蒸汽隔离阀作为核电站二回路的关键安全设备,在维护核电站安全稳定运行方面具有重要意义,因此对于蒸汽隔离阀系统的健康评估至关重要。本文基于故障树分析法,提出了一种针对复杂蒸汽隔离阀系统的健康评估方法。通过故障模式及影响分析,构建蒸汽隔离阀系统故障树。结合实际工程经验,使用风险等级矩阵评估故障风险,对系统进行定性分析,据此制定专门运维检修策略。引入可靠度函数、故障率函数以及系统故障概率密度函数,建立蒸汽隔离阀系统的健康指标,实现实时监测系统健康状况。该方法为后续蒸汽隔离阀系统的精确智能运维提供了重要参考。

核电厂反应堆冷却剂系统抗震阻尼比研究

基于2023年国内新建核电厂安全审评中核安全监管部门对抗震物项阻尼比取值提出关注的背景。介绍核电工程中抗震分析阻尼比取值依据,指出标准与工程实践之间存在的差异,以及由此产生的困惑;阐述阻尼比在动态分析中的作用原理;对比核电领域不同标准与导则文件对于机械设备阻尼比的要求,指出当前标准的相关要求对于由多种部件组成的组合设备或系统过于保守;重点对压水堆核电厂反应堆冷却剂系统与设备阻尼比进行研究,给出国内外核电工程实践中该系统与设备的阻尼比取值依据,并针对核电工程实践中组合设备或系统阻尼比取值依据不足的问题提出建议。

三维激光投影技术在核电厂施工放样中的应用

施工放样作为核电厂建造的关键基础工作,是其他施工活动的依据,其对工程质量及进度有着直接影响。目前核电建安领域主要采用全站仪进行逐点式放样的施工放样方法,该方法具有放样精度差、施工效率低和人工成本投入大等弊端。本文主要提出了应用三维激光投影技术实现核电厂快速施工放样,该方法具有批量放样和投影精度高的特点,通过核电厂施工放样可行性分析和某核电项目现场应用,并使用全站仪进行对比分析,验证了三维激光投影技术在核电厂施工放样中具有一定的应用前景。