水平孔道“O”形环密封结构有限元接触分析

利用ABAQUS软件对研究堆水平实验孔道中异种材料法兰联接的密封结构进行了弹塑性接触计算.利用有限元分步加载技术,模拟了主螺栓预紧和加压过程,研究了主法兰的应力分布和结构的密封性能. .计算结果表明,在预紧状态和设计压力状态下,水平孔道中的异种材料法兰联接.双道"O"形环密封结构完全可以满足强度和密封要求.     

基于RCC-M标准的法兰设计计算方法

以台山核电厂除盐器和安注箱法兰为例,从法兰的运行工况、瞬态分析和计算模型处理等方面详细介绍了在压力、温度、外力和外弯矩等载荷作用下法兰连接的设计计算方法,特别是考虑了容器温度有较大升高瞬态对螺栓和法兰完整性的影响,以及由于过度预紧法兰倾斜角过大对法兰密封性能的影响,为各工程项目中同类法兰的设计计算提供了参考和借鉴。     

HTR-PM螺栓拉伸机无测量杆螺柱预紧力标定试验研究

为了解决高温气冷堆示范工程(HTR-PM)无测量杆螺柱预紧力的控制问题,保证反应堆一回路压力边界的法兰密封,需要对无测量杆螺柱的预紧力进行标定。以HTR-PM中M56无测量杆螺柱为例,采用液压拉伸机对其进行标定试验,找到螺栓拉伸机拉伸预紧力与螺柱残余预紧力的关系曲线;分析了螺栓拉伸机拉伸前后导致螺柱残余预紧力下降的原因,再通过材料力学本构关系,建立了螺栓拉伸机拉紧力与螺柱回弹后残余预紧力的理论关系式。结果表明,试验获得的螺柱联接体系中的残余预紧力及螺母旋紧前的预紧力关系式都与理论分析比较接近;螺栓拉伸机相同出力下,实际设备管嘴法兰螺柱的残余预紧力会比标定值大,但这更有利于法兰面的密封。     

蒸汽发生器一次侧人孔螺栓连接结构三维有限元分析

采用三维有限元模型,对核电厂蒸汽发生器一次侧人孔螺栓进行运行工况下的瞬态分析。综合考虑螺栓温度滞后、法兰转角及法兰与垫片嵌合面的弹塑性接触等因素,解决螺栓孔及人孔盖与法兰之间的空气传热问题,实现螺栓预紧模拟及螺栓载荷的动态提取,较真实地模拟出系统的变形协调性及传热特性,并根据规范对螺栓密封状态进行验证,对法兰连接结构及螺栓本身进行疲劳评价。     

基于泄漏率的核设备密封技术研究

本文以稳压器双锥密封结构为对象,采用多线性随动强化模型模拟了垫片的压缩回弹力学行为,探究了降温速率和螺栓预紧力变化对密封面接触压力的影响,利用基于泄漏率的密封模型对密封性能进行了科学表征,建立了以最大允许泄漏率为准则的螺栓法兰垫片连接设计方法,解决了依据ASME和GB-150规范密封设计系统在使用过程中多次发生泄漏的问题,为工业领域螺栓法兰密封系统设计提供了新的思路。     

化学和容积控制系统下泄热交换器抗震分析

采用有限元方法对辽宁红沿河核电厂一期工程设备中的化学和容积控制系统(RCV)下泄热交换器进行了抗震计算分析,载荷包括自重、压力、温度、接管载荷和地震.根据RCC-M和ASME规范对计算结果进行评定.结果表明,RCV下泄热交换器的设计满足规范要求.     

高温高压容器封头安装中螺柱应力试验

针对高温高压容器封头法兰连接中,加载后螺柱预紧力分布不均匀问题,提出在螺柱表面粘贴电阻应变片测量螺柱应变变化的测试过程与基本操作方法.采用正弦模型和指数衰减模型预测螺柱应力的变化,以探索适合的预紧方式和制定合理预紧程序.对试验数据的分析结果表明,螺柱的应力沿法兰周向成正弦规律变化;单根螺柱应力在预紧过程中满足指数衰减规...     

蒸汽发生器试验本体快速冷却过程中主蒸汽法兰泄漏有限元分析

采用线弹性瞬态热固耦合有限元方法对高温气冷堆蒸汽发生器试验本体主蒸汽法兰在快速降温试验过程中出现的泄漏现象进行了分析。建立了主蒸汽联箱及法兰螺栓连接模型，模拟了主蒸汽法兰的预紧、加压、升温和瞬态降温过程，分析得出了导致法兰密封结构泄漏的主要因素是快速降温过程中法兰的局部变形及螺栓残余预紧力降低，导致密封面张开量大于金属O型环的可靠密封回弹量。在此基础上模拟了不同降温速率下法兰密封面的张开位移，结果表明，限制蒸汽降温速率可改善压力容器法兰的密封性能。     

10MW高温气冷堆压力容器主法兰结构的有限元接触分析

10MW高温气冷堆（HTR－10）压力容器主法兰是HTR－10的关键结构部件,对HTR－10的正常运行起着重要作用。HTR－10压力容器主法兰采用主螺栓进行联接,采用一道金属“O”形环和一道“Ω”环进行密封。该研究课题对此主法结构进行了弹塑性接触计算,利用有限元分步加载技术模拟了主螺预紧以及加压过程中主法兰的应力和位移情况。计算采用二维轴对称结构模型和MSC MARC2000有限元程序。结果表明,无论是预紧状态还是设计压力状态。HTR－10主法兰能满足强度要求,“O”形环和“Ω”环也都有满足密封要求。     

反应堆压力容器主螺栓预紧工艺优化仿真分析

应用商业软件ANSYS模拟了反应堆压力容器主螺栓十字拉伸3级预紧过程,对其预紧工艺进行了仿真优化分析。分析结果表明:3级加载方式中第1级采用对称加载、第2级采用间隔加载、第3级采用顺序加载得到的主螺栓不均匀度和离散度最小;第3级预紧采用变载荷能明显减小主螺栓的不均匀度和离散度,提高反应堆压力容器的密封性能。      

AP1000核电厂主泵法兰螺栓在役检查系统开发

AP1000核电厂反应堆主泵法兰螺栓是在役检查重要监督项目之一,目前国内尚无针对该部件的在役检查系统及应用案例。本文结合AP1000主泵法兰螺栓结构特点、现场高剂量环境及复杂检查条件分析,设计开发了一套从螺栓中心孔内壁实施超声检测、适用于在役检查要求的主泵法兰螺栓在役超声检查系统。主泵模拟体上的调试试验结果表明,该系统可实现周向运行、垂直方向避障、专用超声探头与螺栓孔精确对中调节等功能,进而实现对主泵法兰螺栓的超声扫查。工程应用结果证明本系统满足AP1000核电厂主泵法兰螺栓在役检查现场要求,具有较高的可靠性和良好的适用性。      

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.     

Structural analyses for equatorial and lower ports of the ITER vacuum vessel

To investigate the structural integrity of the ITER vacuum vessel (VV) and ports, the structural analyses of the regular equatorial and the lower remote handling (RH) ports have been performed. The advanced design of the equatorial regular port adopting a pure friction type flange has been recommended as a reference design by the ITER International Organization. The structural integrity of the equatorial port flange, sealing unit, and connecting duct has been reviewed by conducting nonlinear finite element analyses. The advanced design of the regular equatorial port flange with proper pretension is acceptable in the structural design point of view.From the local analyses for a connecting duct and a sealing unit, it has been found that the stresses are less than the allowable values.The structural analyses of the lower RH port have been also performed to verify the capability for supporting the VV. Since high local stress occurs at the gusset and supporting block, the case study for the lower port has been conducted to mitigate the stress concentration and to modify the component design. The strength of the lower RH port structures can be improved by the design modification of poloidal and toroidal gusset.     

Demonstration of structural performance of IP-2 package by simulation and full-scale horizontal free drop test

Packaging systems for the transportation of radioactive wastes have to be designed according to rigorous acceptance criteria and requirements in order to protect people and environment against radiation exposure and contamination risk. The IAEA requirements for type IP2 (Industrial Package Type 2) packages include to carry out free drop tests that represent normal conditions of transport. In such conditions, obviously, the required containment capability of the package has to be ensured.In this study the mechanical performances of a new Italian packaging system for the transportation of low and intermediate level wastes (LILW) undergoing horizontal free drop test are investigated. Especially, deformations caused in the sealing area of the package, which can affect the capability of the containment system, are evaluated.The carried out numerical analyses and experimental tests, at the lab. Scalbatraio of the DICI- University of Pisa, are presented and discussed.Numerical analyses (by qualified MARC^® code) have been performed to investigate the stress histories in the bolts, lid, and package body as well as the deformations in the sealing area and the compression conditions of the gasket.Localised stress appeared at the flange and at the bottom of packaging system. The maximum stresses resulted lower than the stress limits, so the structural integrity of the package was maintained and confirming its tightness. As a consequence of the primary impact a local deformation appeared at the primary lid, no cliff edge or loss of the safety features resulted.     

Study of ITER equatorial port plug handling system and vacuum sealing interface

In the field of the ITER port plug engineering and integration task, CEA has contributed to define proposals concerning the port plugs vacuum sealing interface with the vessel flange and the equatorial plug handling.The 2001 baseline vacuum flange sealing consisted of TIG welding of a 316L strip plate on to U shapes. This arrangement presented some issues like welding access, implementation of tools, lip consumption, complex local leak test, continuous leak checking. Therefore, an alternate sealing solution based on the use of metallic gaskets is proposed. The different technical aspects are discussed to explain how this design can simplify the maintenance and deal with safety and vacuum requirements.The design of the mechanical attachment and vacuum sealing of the plug has constantly evolved, but the associated remote handling equipment was not systematically reviewed. An update of the cask and maintenance procedure was studied in order to design it in accordance with the last generic plug flange design. This includes a concept of a gripping system that uses the plug flange bolting area and, to help the remote handling process, a cantilever assisting system is suggested to increase the reliability of the transfer operation between vacuum vessel and cask.     

Fatigue life estimation for internal threads in class 1 components

Heat exchangers, steam generators and other pressure vessels in nuclear power plants are equipped with bolted closures for the purpose of in service inspection and maintenance. The ASME Boiler and Pressure Vessel Code specifies that all Class 1 components meet the fatigue life requirements for Level A and B Service Conditions. In the case of bolted closures, it is often found that the bolt/stud is the most critical part. In many situations, the bolts fail to meet the fatigue requirements for the design life of the equipment. In such cases, the bolts can be replaced after certain duration based upon their fatigue life. However, the mating threads in the flange (which is an integral part of the vessel) are still a concern. While the replacement of the bolts is relatively easy and inexpensive, the corrective action (e.g. replacement or repair) for the flange is usually difficult and expensive, or impossible. Hence, it is important to have a reasonable estimate of the fatigue life of internal threads to alleviate or minimize the concern. In this paper, a simplified approach is presented for this purpose. Considering various bolt sizes, commonly used thread series and typical Class 1 component materials, it is shown that the fatigue life of the internal threads is about three times the fatigue life of the bolt threads. This conclusion greatly reduces or eliminates the concern for in service replacement or repair of the components with internal threads.     

中国先进研究堆一次冷却剂回路密封垫破损事件分析与处理

本文介绍了中国先进研究堆（CARR）一次冷却剂回路密封垫破损事件以及运行人员对事件的处理过程,剖析了事件发生的根本原因即管路法兰原安装的缺陷,分析了事件后果影响,即造成冷却剂泄漏,指出该事件未对工作人员辐射安全造成影响,未对反应堆及关键系统设备造成影响,泄漏水均在厂房内,对环境无影响,反应堆各项参数均未超出技术规格书中规定的运行限值与条件。本文对事件的过程数据与安全分析报告的内容进行了分析比较,针对事件发生的原因以及处理过程中遇到的一些问题,按照纵深防御的原则,分别从密封垫选型更换、增设法兰防护套、增设渗漏信号以及加强控制技术等方面阐述了系统的改进及优化。     

Study of ultimate seismic response and fragility evaluation of nuclear power building using nonlinear three-dimensional finite element model

The probabilistic safety assessment (PSA) is important for nuclear power buildings in Japan because the risk of the occurrence of seismic ground motions beyond the design assumption cannot be denied. In this paper, the building fragility of the seismic PSA was evaluated using a high accuracy analysis model (three-dimensional nonlinear FEM building model considering soil-structure interaction and basemat uplift behavior). First, the response analyses were conducted increasing the input acceleration up to 3500 Gal, until the damage of the building reached the ultimate condition. The damage of the building was estimated from the shear strain, the axial stress, and the consumed strain energy of the shear walls. Then, the influence on the response given by the vertical ground motion and the basemat uplift was evaluated. In addition, considering the shear destruction of the web wall and compressive crash of the flange wall as the fracture modes, the building fragility was evaluated. As a result, it was shown that the investigated method is efficient for more accurate seismic PSA estimation.     

Fabrication of a full-size mock-up for inboard 10° section of ITER vacuum vessel thermal shield

A full-scale mock-up of VVTS inboard section was made in order to validate its manufacturing processes before manufacturing the vacuum vessel thermal shield (VVTS) for ITER tokamak. VVTS inboard 10° section consists of 20 mm shells on which cooling tubes are welded and flange joints that connect adjacent thermal shield sectors. The whole VVTS inboard is divided into two by bisectional flange joint located at the center. All the manufacturing processes except silver coating were tested and verified in the fabrication of mock-up. For the forming and the welding, pre-qualification tests were conducted to find proper process conditions. Shell thickness change was measured after bending, forming and buffing processes. Shell distortion was adjusted after the welding. Welding was validated by non-destructive examination. Bisectional flange joint was successfully assembled by inserting pins and tightening with bolt/nut. Bolt hole margin of 2 mm for sector flange was revealed to be sufficient by successful sector assembly of upper and lower parts of mock-up. Handling jig was found to be essential because the inboard section was flexible. Dimensional inspection of the fabricated mock-up was performed with a 3D laser scanner.     

铝合金局部真空电子束焊接工艺研究

阐述了铝合金法兰环缝典型结构样件局部真空电子束焊接工艺试验的情况。分析了铝合金焊接气孔的产生原因及采取相应的工艺措施。研究出铝合金局部真空电子束焊接工艺。