储液容器地震易损性参数计算

本文介绍了核电厂设备的易损性分析方法以及易损性模型的参数化计算方法。对核电厂中的典型储液容器应急补水箱（ASG水箱）使用Housner质量-弹簧简化模型进行了分析。对ASG水箱的各项易损性参数进行了计算，绘制出其易损性曲线，并得出高置信度低失效概率（HCLPF）值。结果表明：ASG水箱的HCLPF值低于安全停堆地震（SSE）水平，属于抗震能力较低的设备，需在结构上进行加强。     

法费森海姆核电站应急水箱应进行改进

【法国《核综论》2000年第9～10月刊报道】 2000年8月初,法国电力公司通知核安全主管部门,费森海姆核电站发现影响PTR系统水箱和蒸汽发生器辅助给水系统(ASG)水箱抗地震能力的设计错误。含有水的PTR系统水箱用于: 打开压力容器更换燃料时给反应堆注水; 确保事故情况下反应堆冷却,并向堆芯安全壳注入系统(RIS)的应急系统和安全壳喷淋系统(EAS)供水。 含有水的ASG水箱在运行设施异常情况下向蒸汽发生器供水以确保堆芯冷却。2000年3月在比热伊核电站上发生过与费森海姆一样的不正常现象。为此,核设施安全局要求法国电力公司进行新的…     

换料水箱抗震裕度分析

为评价核电厂应对超过设计基准的外部事件的能力,要求对核电厂进行安全裕度分析。采用EPRI SMA方法进行换料水箱抗震裕度计算,考虑的主要失效形式是螺栓失效。首先进行罐体在地震和自重载荷下的受力计算,接下来进行罐壁屈曲能力分析并计算螺栓压持力,最终通过倾翻力矩评定和滑动性能评定给出设备的抗震裕度值。     

大亚湾核电站换料水箱底板渗漏分析与评估

2006年9月发现大亚湾核电站2号机组换料水箱底板底部有微量水渗出.本工作通过地震分析、应力计算、断裂力学分析,对发现渗漏的大亚湾核电站2号机组换料水箱进行分析与评估.     

核电厂堆芯补水箱抗震分析

核级设备的抗震性能对核电厂运行安全性至关重要,有必要按照抗震设计规范对核级设备进行抗震分析。根据我国核电厂抗震设计规范要求,对堆芯补水箱进行了抗震分析:建立有限元分析模型;给出与抗震分析相关的载荷组合和应力限值;应用有限元软件ANSYS对堆芯补水箱进行了静力分析、模态分析以及响应谱分析;评估了堆芯补水箱在地震条件下的安全性性能。为核级设备抗震分析提供了参考借鉴。     

核电机组辅助给水系统水温超运行技术规范限值的改造研究

针对夏季高温天气下，辅助给水系统（ASG）水温超过运行技术规范限值而导致机组后撤的问题，提出了对辅助给水贮水箱（ASG001BA）加装热交换器的改造方案，从工艺设计、仪控修改和运行控制角度进行了详细分析和论证。机组实践表明，在蒸汽发生器冷却正常停堆模式下，本文提出的改造方案保证了ASG001BA的水位和水温在运行技术规范要求的范围内，保证了机组安全经济的运行。本文的研究对机组大修优化、提升机组核安全水平具有参考价值。      

功率工况异常重要性判定方法在核安全监管中的应用

介绍了国家核安全局开发的功率工况下核电厂异常重要性判定方法（SDP）的基本原理及方法，并使用该SDP对国内某核电厂发生的汽动辅助给水泵（ASG004PO）不可用事件进行了重要度和敏感性分析，结果表明该汽动辅助给水泵的再循环流量试验周期偏长。本文针对此问题给出了优化建议是将ASG004PO再循环流量试验的周期优化为小于34 d。      

PCCS顶部水箱结构隔震参数的优化研究

以AP1000为研究对象,针对非能动安全壳冷却系统(PCCS)结构系统的地震响应,包括屏蔽厂房的结构运动和重力水箱部分的刚体运动、对流运动和冲击运动等特点,基于顶部水箱及其屏蔽厂房的地震响应特点,分别建立隔震结构和普通结构的分析模型。根据核电厂结构特点和抗震规范要求,提出隔震参数优化分析模型,并基于MATLAB工作平台,实现隔震层周期和阻尼比的参数优化。通过优化隔震结构与普通结构、现行抗震规范隔震模型的比较表明,本文优化方法的隔震结构具有减震效果好、抗震性能稳定等优点。     

辅助给水系统超流量分析及改进

采用k-ε湍流模型模拟辅助给水系统(ASG)孔板的三维流动状态,获得孔板流速分布、压降分布及流量与压降关系等特性。建立一维的系统仿真模型并验证了模型的有效性,结合数值模拟得到的孔板特性参数,对ASG役前调试期间除氧器超流量报警问题进行仿真验证和分析,提出报警信号延迟的改进方案,有效地解决了除氧器超流量报警的问题。     

堆芯补水箱地震易损性分析

本文采用有限元软件ANSYS建立AP1000核电站堆芯补水箱(CMT)三维有限元模型,通过模态分析获得其结构特征,采用时程分析法较为真实地模拟CMT地震下响应。通过地震易损性数学模型,对CMT的各项易损性参数进行分析,获得了其抗震能力中值A_m、随机性标准差β_R以及不确定性标准差β_U,计算出其高置信度低失效概率(HCLPF)值。结果表明:CMT的HCLPF值明显高于设计安全停堆地震强度0.3g,说明其具有较高的抗震能力,且HCLPF值略高于采用确定论方法得到的值。对易损性参量误差敏感性分析发现β_R取值变化对CMT的条件失效概率和HCLPF值影响较小,可简化部分随机性误差的考虑,使得易损性分析更简洁。     

安全壳外挂水箱结构强度分析

我国自主设计的第3代核电站安全壳外挂水箱用于超设计基准事故下内层安全壳的长期排热,这是确保安全壳完整、核电厂安全的重要设施。因此,有必要对外挂水箱在极限安全地震动与温度异常工况组合作用下的结构强度进行分析。建立带有外挂水箱的外层安全壳有限元模型,开展网格敏感性分析,并通过模态分析研究结构的振动特性。采用时程分析法,对结构同时施加温度和地震动载荷,基于流固耦合方法分析水体与结构的相互作用,研究外挂水箱结构的地震动响应以及水箱内水体振荡特性。研究表明,水体在水箱凹沉处水面振荡幅度较大,在EL Centro地震动、人工合成地震动和长周期地震动工况下外挂水箱的最大拉应变和最大压应变绝对值均小于C60混凝土许用应变值。     

Simplified method for evaluating the seismic buckling capacity of unstiffened steel containment structures

A simplified method is presented for evaluating the seismic buckling capacity of unstiffened, free-standing steel containment structures. The method is consistent with current US Nuclear Regulatory Commission seismic design standards and with containment buckling interaction equations given in the American Society of Mechanical Engineers Boiler and Pressure Vessel Code which includes the influence of geometrical imperfections of the shell on buckling. Stresses to be considered in the interaction equations are determined from beam theory using standard response spectrum analysis. An empirical correction factor is developed to account for hoop stresses that are not explicitly represented in the beam theory. As the results of these analyses are very sensitive to the damping that is assumed, the extensive three-dimensional finite element analyses that were performed to develop the hoop stress reduction factor were also used to study the sensitivity of containment buckling to the assumed damping. Experiments on model containment structures were then performed to further investigate the damping properties exhibited by these structures. The study in concluded by showing that the simplified method reasonably predicts seismic buckling capacities when compared with independently determined predictions from detailed finite element analyses.     

Dynamic buckling experiments of fluid-structure-coupled co-axial thin cylinder

The purpose of this paper is to clarify dynamic buckling behaviours such as buckling mode and buckling pressure for thin cylindrical shells immersed in fluid subjected to seismic excitations. For this purpose, dynamic buckling experiments of thin cylindrical shells placed inside a rigid liquid container are carried out using a shaking table. These shells and the container are intended to represent thermal baffles and a main vessel of a fast breeder reactor, respectively. The fluid pressure caused by horizontal excitation induces buckling deformation which involves flower-shaped deformation, which is a type of external pressure buckling. The buckling pressure is measured with various types of the test cylinders under seismic excitations and this pressure is confirmed to agree with static buckling pressure predicted by static buckling analysis. It is also found that sub-harmonic vibration occurs under a certain sinusoidal excitation inducing a sudden increase in response displacement at a lower pressure level than the buckling pressure under seismic excitations. Based on these experiments, it is pointed out that, in seismic design, to prevent the buckling of thermal baffles, static buckling analyses can be used as long as sub-harmonic vibration does not occur.     

Investigation on buckling of FBR vessels under seismic loadings with fluid structure interactions

The critical thin walled shell structures in the reactor assembly of a pool type fast breeder reactor (FBR) are the main vessel, inner vessel and thermal baffles. On these structures, the seismic events impose major forces by developing high dynamic pressures, thereby causing a concern on structural integrity due to buckling. An integrated analysis for determining realistic forces and critical buckling loads at any instant during the seismic event has been carried out for the vessels of a typical 500 MWe pool type fast breeder reactor. The dynamic forces including pressure distributions generated on the vessel surfaces are extracted from the seismic analysis carried out for the reactor assembly. The seismic forces thus computed from axisymmetric analysis are transmitted appropriately to 3D shell geometries for the buckling analysis. In view of high computational time needed for carrying out buckling analysis at every time increment, the elastoplastic analysis is carried out only at a few critical time steps which are identified based on strain energies that are associated with the shear and compressive stresses developed at the portions of the vessels prone to buckle. The shear buckling of main vessel straight portion and buckling of toroidal portion of inner vessel and thermal baffles are found to be important. The possible randomness of support excitation time histories is accounted for by compressing as well as expanding the nominal time histories by 10%. Buckling strength reduction factors due to the initial geometrical imperfections are adopted from the literature. The inner vessel is found to be the most critical component which buckles under seismic forces induced by a safe shutdown earthquake with a load multiplier of 1.52, which is higher than the minimum factor of safety of 1.3 required as per the design code RCC-MR [RCC-MR: edition, 2002. Design and construction rules for mechanical components for FBR nuclear islands, vol. 1, section I, subsection B. AFCEN, Paris, in press] for service level D conditions.     

Pseudo-dynamic buckling experiments on thin cylindrical shells under biaxial seismic loads

A buckling design research program has been carried out to establish seismic design guidelines for a fast breeder reactor. In doing so, the buckling strength of the cylindrical part of the reactor vessel of a fast breeder reactor under horizontal and vertical seismic loads has been clarified. The effects of axial loads on the horizontal seismic responses in pre- and post-buckling states of thin cylindrical shells are investigated. Pseudo-dynamic buckling experiments are performed to study the dynamic buckling characteristics of thin cylindrical structures when subjected to seismic loads. The buckling tests use model cylinders made of an aluminum plate and a biaxial loading test apparatus. The axial seismic loads reduce the lateral load-carrying capacity of the shells in the pre- and post-buckling regions so that they amplify the horizontal response displacement. An amplification factor that accounts for the effects of the vertical loads is presented and its validity is verified experimentally.     

Outline of the seismic buckling design guideline of an FBR — a tentative draft

Central Research Institute of Electric Power Industry (Japan), commissioned by the Ministry of International Trade and Industry, is carrying out the Demonstration Test and Research Program of Buckling of FBR (FY 1987-FY 1993). The first half of the research program was finished after establishing a seismic buckling design guideline (a tentative draft). The purpose of this paper is to describe the dynamic buckling characteristics of FBR main vessels and the outline of the rationalized buckling design guideline for seismic loadings.     

线圈终端盒的改进设计与分析

为满足美国GA公司中心螺线管线圈模型CSM低温电性能测试的需要,基于ITER馈线系统的设计,对线圈终端盒壳体进行了修改设计。采用直立圆筒结构代替横卧立方体结构,优化了壳体安装工艺,提高了空间利用率。在此基础上,对线圈终端盒内部其他部件进行了相应的改进设计,最终实现了线圈终端盒的功能。利用大型有限元分析软件ANSYS对线圈终端盒壳体作弹性应力分析、屈曲分析及地震分析,并将屈曲分析结果与理论计算结果进行了对比。计算分析结果表明,直立圆筒结构形式的线圈终端盒设计合理可靠。     

容积控制箱力学分析及评定

采用有限元软件ANSYS对容积控制箱进行力学分析,遵照规范RCC-M2000和ASME相关规定进行相应评定,计算了多种载荷下系统的抗震性能,同时分析了裙座的屈曲性.结果表明,该容积控制箱的设计满足规范相关条款要求.     

Resolution of flat bottom tank outlier

This paper contains the results of an outlier resolution evaluation for a large flat bottom tank, 40 ft 6 in. in diameter and 32 ft 8 in. in height. The tank was an outlier in both the USI A-46 and IPEEE programs due to insufficient strength of the bolt chair to transfer the bolt load to the side of the tank. The bolt chair evaluation resulting in the outlier was linear elastic. A more sophisticated non-linear analysis was performed of the bolt chair using the program ANSYS. The evaluation resulted in the conclusion that the bolt chair was able to transfer almost the entire yield strength of the bolt without excessive deformation that could ultimately cause overall tank failure. This evaluation tremendously increased the seismic capacity of the tank and resolved the outlier for both programs. The tank outlier evaluation also included a evaluation of soil–structure interaction (SSI) effects on the seismic demand on the tank. However, the formal consideration of SSI had a small effect on the overall seismic demand.