考虑接管载荷的反应堆压力容器接管嘴断裂力学分析

压力容器接管嘴位置结构和载荷情况比较复杂,若假想缺陷位于压力容器进口接管内隅角位置,在考虑接管载荷作用时,裂纹为复合型裂纹。建立含裂纹的三维有限元模型,分析在接管载荷单独作用、内压与接管载荷共同作用下裂纹尖端应力强度因子的分布和变化规律。分析结果表明,在仅考虑接管载荷时,进口接管内隅角位置应力强度因子KI、KII和KIII都比较小,应力强度因子近似对称分布;内压对裂纹尖端的应力强度因子KII和KIII基本没有影响。     

LOCA下具有表面裂纹的反应堆压力容器承压热冲击分析

失水事故（LOCA）瞬态下,具有半椭圆形表面裂纹的反应堆压力容器（RPV）承压热冲击（PTS）问题被研究。采用有限元方法计算瞬态过程的热-应力响应;采用影响函数法计算应力强度因子,分别对母材和堆焊层内的应力进行分解,从而解决了由于堆焊层存在造成的应力拟合困难带来的计算偏差。编制了相应的断裂分析程序,对LOCA下RPV的结构完整性进行了分析。结果表明,在研究的LOCA下,整个瞬态过程中RPV应力强度因子均未超过材料断裂韧性,压力容器结构安全。本文研究为RPV在PTS下的结构完整性评估提供理论指导。     

论高强度钢在核动力反应堆压力容器中的应用

本文论述高强度钢在核动力反应堆压力容器中的应用前景.着重指出,在技术工艺水平高度发展的今天,限制高强度钢应用的主要障碍不在于材料和制造工艺,而在于现行的压力容器规范:建议改变现行规范规定的取用设计许用应力的方法,修改材料的极限强度对设计许用应力的限制,而代之以针对不同材料和不同类型的容器规定一个合适的材料屈强比要求.     

核电站压水堆压力容器瞬态应力场分析

在启动、运行、停堆设计瞬态工况下,用轴对称有限元法和三维有限元法分别分析了整体压力容器和接管段由冷却剂温度、压力、内热源、螺栓力和接管载荷等载荷变化而产生的瞬态温度场和应力场。分析表明:瞬态应力分析的关键部位在接管段;由γ射线辐照产生的二维内热源分布所引起的热应力几乎局限于活性区,其最大值比其它应力要小一至二个量级。     

基于Z′参数的高强铝合金稳态蠕变速率可靠性研究

建立了基于速率温度参数模型的新型高强铝合金Z′参数表达式及稳态蠕变速率 可靠度预测方法,Z′参数表征稳态蠕变速率数据偏离速率温度参数模型主曲线的程度。结合实测数据,研究了Z′参数的统计分布规律。基于Z′参数,分别得出新型高强铝合金实验应力 速率温度参数 可靠度曲线和温度 许用应力 可靠度曲线。结果显示,新型高强铝合金Z′参数满足正态分布规律。稳态蠕变速率预测上限随可靠度的增加而升高。外推至稳态蠕变速率上限10-7 h-1时,新型高强铝合金在50、80 ℃下的许用应力（可靠度997%）分别为4700、2945 MPa。与传统安全系数法相比,基于Z′参数的预测结果更合理。     

反应堆压力容器出口接管力学分析

在核电站的运行过程中,反应堆压力容器出口接管需承受自重、内压、热膨胀、地震和管道载荷.作为保证反应堆安全正常运行的重要部件,必须确保反应堆压力容器出口接管的完整性.本工作应用大型有限元程序ANSYS对压力容器出口接管进行应力强度和疲劳分析,得到出口接管的应力分布状况、最大应力及疲劳使用系数,并按照相关规范的应力限值对出口接管的计算结果进行评定.评定结果表明,出口接管满足规范的要求.     

基于Z′参数的高强铝合金稳态蠕变速率可靠性研究

建立了基于速率温度参数模型的新型高强铝合金Z′参数表达式及稳态蠕变速率-可靠度预测方法,Z′参数表征稳态蠕变速率数据偏离速率温度参数模型主曲线的程度。结合实测数据,研究了Z′参数的统计分布规律。基于Z′参数,分别得出新型高强铝合金实验应力-速率温度参数-可靠度曲线和温度-许用应力-可靠度曲线。结果显示,新型高强铝合金Z′参数满足正态分布规律。稳态蠕变速率预测上限随可靠度的增加而升高。外推至稳态蠕变速率上限10~(-7) h~(-1)时,新型高强铝合金在50、80℃下的许用应力(可靠度99.7%)分别为470.0、294.5 MPa。与传统安全系数法相比,基于Z′参数的预测结果更合理。     

主螺栓断裂对压力容器密封性能、应力及疲劳的影响分析

在理论分析和数值仿真技术基础上,研究并提出了一种主螺栓断裂对反应堆压力容器（RPV）密封性能、螺栓应力及疲劳的影响分析方法,采用该方法对主螺栓断裂影响进行了评价分析。结果表明,该方法适用于分析1根或多根主螺栓断裂情况对压力容器安全性能的影响,可以用于核电厂在运行中发生类似问题时判断反应堆能否继续运行。      

核压力容器法兰螺栓结构的有限元分析

法兰螺栓结构是核压力容器的重要部件,它对反应堆的正常运行起着非常重要的作用.该研究工作对压力容器上顶盖和蒸汽发生器二次侧人孔法兰螺栓结构均做了等效简化来进行多种轴对称分析与三维分析,并进行了比较,还探讨并比较了实现法兰螺栓的预紧效果的不同方法.研究表明,轴对称模型能较为准确地得到应力强度,且比较保守.     

压力容器开孔接管处表面斜裂纹应力强度因子数值分析

通过在裂纹前沿设置奇异单元,建立了圆筒形压力容器开孔接管处表面斜裂纹断裂力学有限元分析模型;运用可视化编程语言VB编制了压力容器开孔接管处表面斜裂纹应力强度因子参数化分析软件,实现了与有限元分析软件ANSYS的连接与调用;根据内压作用时多种几何参数情况下应力强度因子的计算值,绘制了裂纹前沿应力强度因子随裂纹相对深度、裂纹对应弧度、支管与主管直径比及裂纹倾斜角的变化曲线。     

Three-dimensional transient sealing analysis of the bolted flange connections of reactor pressure vessel

The failure of sealing system of the bolt flange connections is the primary failure mode of the nuclear reactor pressure vessel (RPV). For the safety and integrity of RPV, it is important to predict the sealing behaviour of the bolt flange connections under various loading conditions. Based on the finite element (FE) method for coupled thermal elastoplastic contact problems, a three-dimensional (3D) transient sealing analysis program of nuclear reactor pressure vessels is developed with the consideration of the non-linearity from both surface and material, transient heat transfer and multiple coupled effects. A contact correction approach is proposed to simulate the loading of the bolt connection under the condition of pre-stressing. An automatic pre-processing program is developed for FE modelling of RPVs. Using these programs, a 1:4 scaled model of a 300 MW RPV is analyzed under the loading conditions including pre-stressing, pressurization, heating and cooling. The computational results obtained are in a good agreement with the data of experimental tests. These programs are also successfully used in analyzing the full-scale model of the RPV in a nuclear power plant.     

主螺栓断裂对压力容器密封性能、应力及疲劳的影响分析

Base on the mechanics theory and numerical simulation technique, a method to analyze the effect of the main bolt break on the stress, fatigue and seal is studied in this paper, and is adopted to evaluate and analyze the fracture influence of main bolt. The results show that this method is applicable for the analysis of the RPV safety performance induced by one bolt break or several bolts break accident, and for the determination if the nuclear reactor can be operated when similar problems occur.     

在役核电厂人员闸门传动齿轮优化改进

采用ANSYS有限元分析程序,对国内某在役核电厂人员闸门断齿传动齿轮进行了应力分析,找出了传动齿轮断齿失效的根本原因:传递扭矩过大而引起齿轮过载,应力计算值超过了材料许用应力限值,致使传动齿轮产生根切断齿事故。结合传动齿轮断齿失效根本原因,从材料选择和结构设计2个方面对传动齿轮进行了优化改进。应力分析与评定结果显示,优化改进后的传动齿轮和与之配合的扇形齿轮应力计算值均小于材料许用应力限值,优化改进方案有效降低了传动齿轮的应力水平,有效提高了人员闸门齿轮传动操作的安全性和可靠性。      

国产反应堆压力容器用16MND5钢的蠕变损伤本构模型研究

为了获得反应堆压力容器（RPV）材料在高温下的蠕变行为,保证RPV在严重事故工况下的完整性,本研究对国产RPV用16MND5钢的高温蠕变性能进行了测试,获得了600~900℃下材料的蠕变性能,并基于应变强化的基本蠕变本构模型与基于延性耗竭理论的蠕变损伤模型,建立了适用于16MND5钢的蠕变损伤本构模型,给出了材料的蠕变损伤模型参数。结果表明,本文提出的蠕变损伤本构模型的有限元模拟数据与试验数据符合性较好,验证了此蠕变损伤模型的正确性。该方法可用于严重事故情况下RPV的蠕变损伤分析,为RPV的完整性分析提供支持。      

在役核电厂人员闸门传动齿轮优化改进

ANSYS finite element analysis program was used to analyze the stresses of the transmission gear for an in-service nuclear power plant in China, and the causes for tooth rupture were found. The transmission torque was so large that the calculated stress exceeded the allowable stress limit of the material, which caused a root rupture accident of the transmission gear. Combined with the causes of tooth rupture for transmission gear, optimization was performed based on material selection and structural improving design. The optimized design scheme passed stress analysis and evaluation. The results show that the calculated stresses of the optimized transmission gear and the sector gear were smaller than the allowable stress limits of the material. The optimized scheme effectively reduced the stress level of the transmission gear, and obviously improved the safety and reliability of the transmission gear operation for personnel airlock.     

Formation and growth of cracking in feed water pipes and RPV nozzles

The observation of numerous small and large cracks in ferritic feed water pipes of boiling (BWR) and pressurized water reactors (PWR) in the last few years has led to basic research into the causes of cracking and the crack growth mechanisms.In horizontal feed water pipe sections connected to nozzles of reactor pressure vessels (RPV) of BWR's as well as of steam generators (SG) of PWR's, circumferential macro and micro cracks were detected. These cracking phenomena could be observed in base material of pipes as well as in weld seam regions. The examination of the stress state displayed that the cracked pipe regions have been exposed to a number of cyclic thermal transients (thermal shock, flow stratification) during start-up (hot stand-by) and shut-down periods of the plants. During thermal transient periods, local and global cyclic stresses in the referred pipe cross sections have been induced which in interaction with the influence from environment (in operation as well as in shut-down periods) and local geometrical imperfections led to the initiation and formation of macro and micro cracks.In the reactor water clean-up system of BWR through which reactor water is fed from the RPV to the main feed water line, two longitudinally welded elbows have been detected to be severely cracked. Both elbows have been subjected to an internal pressure corresponding to RPV and additionally to a relevant “in-plane” bending moment. These longitudinal cracks were found to be started from the inner elbow surface. In one case the longitudinal crack was situated in the base material and was enlarged to leakage. In the second elbow the longitudinal crack was located in the heat affected zone (HAZ) of a longitudinal weld. In both cases the macro cracks started either from corrosion pits located in defective areas of the magnetic protection layer or from geometrical notches (weld root). The semi-elliptic small cracks got linked to more extended shallow cracks.Formation and growth mechanism of these cracks have been studied at the MPA Stuttgart in laboratory under simulated operation conditions which were held as realistic as possible compared with those in nuclear power plants.The results of experimental studies in laboratory as well as conclusions based on the above mentioned cracking phenomena in piping have been used as basic information for a realistic design of large scale (RPV) thermal shock experiments under operation conditions. The formation and growth mechanism of these cracks and their detection by means of NDE during thermal transients at the inner surface of RPV nozzle and at the adjacent cylindrical areas of RPV shell will be described.     

The HTTR reactor pressure vessel and its integrity against a PTS event

Before manufacturing the real steel to be used in the reactor pressure vessel (RPV) of the high temperature engineering test reactor (HTTR) the vessel manufacturer and materials supplier made a sample steel by the same procedure as for the real steel (2.25Cr-1 Mo) and conducted many tests to obtain material strength data for its base and weld metals. The test results showed that the sample steel satisfied the HTTR design requirements. Vessel cooling panels are set on the inner surface of the biological shielding concrete around the RPV, and are circulated with cooling water at 0.5 MPa and 40°C to cool the shielding concrete during normal operation of the reactor. By supposing that the cooling panel breakes and the water discharges to the RPV outer surface heated at 400°C, the stress distribution generated in the vessel wall by a pressurized thermal shock (PTS) event can be calculated using a finite element method code. This paper describes some of the results obtained from the material testing of the sample steel and the estimated result using the scheme developed for a light water reactor pressure vessel, to clarify the integrity of the HTTR-RPV under a PTS event.     

Structural integrity assessment of the reactor pressure vessel under the external reactor vessel cooling condition

Since the suggestion of external reactor vessel cooling (ERVC), the effects of melting and cooling on the response of structural integrity of the reactor pressure vessel (RPV) under core melting accident conditions have been investigated. To investigate the initial behavior of RPV lower head and the effects of analysis conditions on the structural integrity of RPV, the transient analysis is utilized considering the transient state. To obtain an analogy with real phenomena, the material properties were determined by combining and modifying the existing results considering phase transformation and temperature dependency. The temperature and stress analyses are performed for core melting accident by using ABAQUS. Finally, the potential for vessel damage is discussed using the Larson-Miller curve and damage rule. In addition, the results by transient analysis are compared with those by steady state analysis and the effects of analysis conditions on structural integrity are reviewed.