异种材料典型螺纹联接件在预紧和热循环条件下的力学行为

在核电厂堆内构件中,如果螺纹联接结构的联接件和基体采用两种不同的材料,由于螺纹联接件材料的热膨胀系数小于基体材料,在升温过程中将产生较大的附加应力.本文以典型螺纹联接件M12为例,模拟堆内热循环载荷条件,对其进行有限元仿真计算分析.在与试验结果及理论分析、经验公式互相比较的基础上,对预紧力矩、螺纹联接件和基体的变形量、联接件在预紧力、升温、降温等不同载荷条件下的应力等参量变化情况进行了研究,为工程设计提供了可靠的依据.     

防断组件及其支承柱高周疲劳分析

反应堆堆内构件是反应堆冷却剂系统中的重要设备,其设计结构要求在全寿期内保持高度可靠性。在国内外核电厂运行过程中,曾发生堆内构件因流致振动而出现故障和损坏事件,直接影响了反应堆的安全运行和经济效益。本文以堆内构件防断组件及其支承柱(SCSS)为研究对象,研究其在流致振动载荷和泵致振动载荷下的动态响应,并对结构进行谱分析和谐响应分析。最后根据ASME锅炉及压力容器规范对防断组件及其支承柱各部件进行高周疲劳评定,计算结果表明各部件交变应力强度满足规范限值的要求。     

反应堆压力容器上部堆内支承构件的流致振动分析

核电厂运行工况下,反应堆压力容器(RPV)上部堆内支承构件处在一个高速的横向流流场中。在横向流作用下,RPV上部堆内支承构件产生明显的振动。这类振动行为体现在涡旋导致的结构振动和流弹不稳定。研究RPV上部堆内支承构件在横向流作用下的振动行为特性,并根据ASME规范对其进行流致振动分析评定。研究结果表明:特定结构在冷启堆或冷停堆时更容易产生流致振动,而非在满功率运行工况下。     

中国实验快堆堆内构件应力分析与评定

本文是中国实验快堆堆内构件主要部件的应力分析与评定汇总报告.主要构件包括堆内支承结构、堆芯支承结构、堆内热屏蔽等7类设备.堆内各部件采用有限元方法按其特点进行整体分析或部件分析.文章首先建立结构的计算模型,然后,对有限元计算模型进行在自重、流体流动压差、冷却剂流动引起的结构振动和温差载荷条件下的静态分析计算和结构的模态分析以及地震载荷下的动态分析.最后,按规范要求对堆内各结构在承受的各种载荷条件下进行载荷组合与评定.     

用于快堆非能动停堆系统的磁性连接对的设计优化与完整性综合评价

按照ASME规范要求,完成了用于快堆非能动停堆系统的磁性连接对的设计;对工况进行了分析,总结出了磁性连接对在正常运行工况下的载荷,并利用大型有限元分析软件ANSYS对连接对进行计算,得到正常运行工况下的应力值。按照ASME规范,对磁性连接对在正常运行工况下进行了应力强度分析和评定及疲劳评定。结果显示,磁性连接对在正常运行工况下的一次和二次应力强度及疲劳评定均满足ASME规范的要求。     

CPR1000核电厂核反应堆压力容器应力分析

本文所计算的核反应堆压力容器是保证核安全的一道重要屏障,因此,要参照相应的规范和标准对其进行强度方面的分析和校核.通过有限元软件ANSYS建立压力容器的三维模型,计算压力容器在设计工况以及试验工况下,在压力、温度、堆内构件重力和接管载荷等各种载荷作用下的应力强度,并严格参照规范标准RCC-M B篇规定的各种工况下的应力准则,对压力容器进行强度评定.评定的结果表明,压力容器在计算的几类工况下,均符合规范标准RCC-M的强度要求.本工作的计算和分析也为我国核工业未来的设备设计制造走上国产化、标准化奠定了一定的基础.     

核反应堆堆内构件焊接工艺评定

核反应堆堆内构件、零部件及焊缝、焊点较多,存在焊接接头型式、母材及其厚度、焊接工艺方法、焊接位置和方向等的不同,导致焊接工艺评定的复杂性。文章介绍了焊接工艺评定的一般变素、堆内构件焊缝分布及堆内构件的焊接工艺评定。并针对RCC-M规范、ASME规范及国内相关标准对焊接工艺评定要求的差异,结合堆内构件焊接工艺评定过程中尺寸稳定化处理、焊接接头的横向拉伸试样、手工焊与自动焊的定义、破坏性试验的复验要求等方面的争议,提出了个人的理解和认识。     

堆内构件设计与研制

堆内构件是核电站反应堆中的主要设备。为了保证核电站反应堆安全可靠地运行,对 堆内构件的结构设计、应力分析、热工水力、以及工艺试制等方面,进行了大量的试验研究和计算分析,予以不断完善和改进。其中设计了对中定位、反应堆驱动线、高温高压密封等独特结构;建立了完整的分析法设计的工作流程和方法,研究中涉及到流-固耦合、辐照-热-应力耦合、冲击-振动-弹塑性动力学、相似理论-模型试验等交叉学科;并且攻克了薄板加工、精密镗孔、精密焊接、驱动线对中等一系列制造技术难关。为新的核电厂的设计     

辐照监督管支承定位结构的载荷分析与应力评定

某电厂发现辐照监督管组件与保护套管的连接支承处发生严重磨损,且支承弹簧失效.为此,改造了辐照监督管支承定位结构,为论证其满足在核电厂中的使用寿命要求,需对改造后的结构进行载荷分析和应力评定.首先,分析了改造后的结构在运行工况下所承受的载荷;然后,对其进行了不同载荷组合下的应力分析和疲劳分析;最后,按照ASME规范的要求,进行了各种使用限制下的应力评定.     

堆内构件设计与研制

堆内构件是核电站反应堆中的主要设备。为了保证核电站反应堆安全可靠地运行,对堆内构件的结构设计、应力分析、热工水力、以及工艺试制等方面,进行了大量的试验研究和计算分析,予以不断完善和改进。其中设计了对中定位、反应堆驱动线、高温高压密封等独特结构;建立了完整的分析法设计的工作流程和方法,研究中涉及到流-固耦合、辐照-热-应力耦合、冲击-振动-弹塑性动力学、相似理论-模型试验等交叉学科;并且攻克了薄板加工、精密镗孔、精密焊接、驱动线对中等一系列制造技术难关。为新的核电厂的设计和制造,提供了借鉴与参考。     

CPR1000堆内构件螺栓紧固件的转化设计

以中国改进型三环路压水堆(CPR1000)堆内构件的螺栓联接拧紧力矩作为问题研究的出发点,探讨堆内构件的螺栓联接件翻版设计中,以国标米制替代统一英制的具体步骤和方法,列出了在转化设计中必须考虑的影响螺栓联接拧紧力矩大小的螺栓结构要素,以确保CPR1000堆内构件螺栓联接结构的可靠性,避免在反应堆运行过程中因螺栓联接结构的松动或紧固件脱落而威胁到反应堆的安全运行.     

压水堆核电厂堆内构件安装加工件控制

对压水堆核电厂反应堆关键设备的堆内构件安装过程的加工件控制所存在的问题进行深入分析,提出堆内构件安装加工件的测量、加工、检查等控制环节的关键要求,归纳总结了堆内构件安装加工件的控制要素,结合笔者的施工经验,分析堆内构件安装加工件的控制关键点,提出了堆内构件安装加工件质量控制和风险防范的措施,对堆内构件安装加工件控制具有指导意义。     

堆本体抗震试验动力相似关系推导与验证

为确保堆本体抗震试验中流体对流效应、脉冲效应和堆本体结构响应的准确性,需保证重力、流体与固体惯性力、结构弹性力和结构应变的相似性。本文从固体结构的振动方程、不可压牛顿流体的动力学方程、流固交界面的边界条件和环形柱体域内液体线性晃动的动力学公式出发,基于控制方程的量纲分析法,推导了考虑液体晃动效应的堆本体地震响应动力相似关系。基于上述相似关系建立了堆容器堆内构件和堆容器内自由液面流体域的缩尺模型,通过有限体积法分析堆容器堆内构件原型和缩尺模型中液体的晃动固有频率、晃动波高、压力以及液体晃动对堆容器支承裙的倾覆力矩。结果表明本文动力相似关系具有合理性和准确性,可用于堆本体缩尺模型的抗震试验研究。     

Age-related degradation of boiling water reactor vessel internals

Researchers at the Idaho National Engineering Laboratory performed an assessment of the aging of the reactor internals in boiling water reactors (BWRs), and identified the unresolved technical issues related to the degradation of these components. The overall life-limiting mechanism is intergranular stress corrosion cracking (IGSCC). Irradiation-assisted stress corrosion cracking, fatigue, and thermal aging embrittlement are other potential degradation mechanisms. Several failures in BWR internals have been caused by a combination of factors such as environment, high residual or preload stresses, and flow-induced vibration. The ASME Code Section XI in-service inspection requirements are insufficient for detecting aging-related degradation at many locations in reactor internals. Many of the potential locations for IGSCC or fatigue are not accessible for inspection.     

Development of Telerobotic Systems for Reactor Decommissioning, (III)

This paper describes the telerobotic system for reactor decommissioning in the scope of engineering demonstration of dismantling radioactive reactor internals of an experimental boiling water power reactor JPDR. The total system consists of a telerobotic manipulator system equipped with a multi-functional amphibious slave manipulator with a load capacity of 25 daN, a chain-driven transport system, and a computer-assisted monitoring and control system. Preceding to the application of the telerobotic system to actual dismantling operation, a mockup test was performed of dismantling the simulated reactor internals of actual-size by the method of underwater plasma arc cutting in order to study the performance of the telerobotic system in a realistic environment. The system was then successfully applied to dismantling the actual reactor internals according to the JPDR decommissioning program.     

压水堆核电厂堆内构件制造的质量监督

分析了堆内构件制造工艺中的重点和难点,就如何进行堆内构件制造的质量控制与监督进行了探讨.特别对堆内构件中重要部件的制造过程和堆内构件装配过程中质最控制的重点、难点进行了详细的阐述,给出了堆内构件制造驻厂监造中的主要关注点,同时也给出制造过程质量监督中其他还需要注意的要点,如文件控制、人员控制.     

压水堆堆内构件寿命管理中的关键问题

介绍了对堆内构件中潜在失效部件应开展的研究内容，包括预防措施、潜在的失效部件、失效模式，失效过程和基本接受条件；指出了堆内构件的主要失效模式和危险区域．并对这些失效模式和危险区域进行了分析对需修复或更换的部件．提出了需进一步研究的内容。     

压水堆结构设计中应注意的问题

根据工程经验,在压水堆结构设计中,应对反应堆的结构布置、功能要求、接口控制、热工水力特性、驱动线性能、结构完整性以及堆内构件的功能准则和堆结构对燃料组件的影响等方面进行综合评价,以确保反应堆的顺利装配,实现反应堆的功能并确保反应堆安全运行.     

Operational feedback on internal structure vibration in 54 French PWRs during 300 fuel cycles

EDF has acquired extensive feedback on vibration of reactor vessel internals by analysing ex-core neutron noise on its 54 pressurized water reactors during the course of over 300 fuel cycles.

This feedback has been built up by processing more than 3,000 vibratory signatures acquired since the startup of its reactors. These signatures are now centralized for the whole of France in the “SINBAD” data base.

Signature processing has enabled:

1. • distinguishing between mechanical phenomena and signature variation linked to unit operation: in particular, the impact on signature level of unit operating parameters such as initial fuel enrichment and burn-up rate was assessed;

2. • among the purely mechanical phenomena, pointing up slight changes in position of vessel internals and the first signs of structural wear: relaxation (in the hold-down spring and fuel rod assemblies) and wear on surfaces of contact between internals and reactor vessel were detected;

3. • lastly and most importantly, automatic recognition of the various types of vibratory behavior of internals.

It was consequently possible to draw up user requirement specifications for automated monitoring of internals, which should soon be integrated in PSAD, a system which groups several reactor monitoring functions.     

The numerical simulation of the WWER440/V213 reactor pressure vessel internals response to maximum hypothetical Large-break Loss of Coolant Accident

The reactor internals are designed to ensure cooling of the fuel, to ensure the movement of emergency control assemblies under all operating conditions including accidents and facilitate removal of the fuel and of the internals following an accident.This paper presents preliminary results of the numerical simulation of the WWER440/V213 reactor vessel internals (RVI) dynamic response to maximum hypothetical Large-break Loss of Coolant Accident (LOCA). The purpose of this analysis is to determine the reactor vessel internals response due to rapid depressurization and to prove no such permanent (plastic) deformations occur in the RVI which would prevent timely and proper activation of the emergency control assemblies.In the case of the LOCA accident it is assumed rapid “guillotine” break of one of the main coolant pipes and rapid depressurization of the primary circuit. The pressure wave spreads at the speed of sound, enters the reactor pressure vessel and causes deformation and stress in reactor vessel internals.The finite element model was created by MSC.Patran (Patran, 2010) and dynamic response was solved using MSC.Dytran (Dytran, 2008) finite element code. The model consists of reactor vessel internals (Lagrangian solid elements) and water coolant (Euler elements) inside the reactor. Arbitrary Lagrangian Eulerian (Belytschko et al., 2003) coupling was used for simulation of the fluid-structure interaction. The calculation assumes no phase change in the water. No comparison with the experiment was performed up to now, because the required experimental data are not accessible for this type of the reactor.The most important acceptance criteria for the reactor internals demands that the movement of the emergency control assemblies under all operating conditions including accident is ensured (BNS, 2008). The numerical simulation of the WWER440/V213 reactor internals response to a LOCA accident showed that the acceptance criteria for RVI is fulfilled and required NPP safety standards are satisfied.