【作者】 徐铭宇；

【导师】 李润方；

【作者基本信息】 重庆大学 ， 机械设计及理论， 2004， 硕士

【摘要】 本论文来源于“十五”国防预研关键技术项目中关于反应堆程序开发项目(编号:1010104020102)的子课题“反应堆压力容器三维瞬态密封分析程序开发”。本课题研究目的是开发先进的核容器密封分析软件系统,以利于提高设计质量和设计水平,确保核容器密封系统设计的可靠性,具有重要的理论意义和实用价值。核反应堆压力容器包容高温、高压且具有放射性的工作介质,工作条件十分苛刻,法兰密封失效是压力容器的最基本失效形式。密封分析不仅涉及法兰密封面间弹塑性接触这类双重非线性、热性能的非线性和传热的瞬态性,而且涉及各种非线性的耦合作用。目前国内外的商用软件均没有这种全面功能。本文的研究是在已开发的反应堆压力容器三维法兰密封分析程序系统的基础上进行的,主要内容是对该容器进一步的改进和扩充,以拓展分析功能,增大解题规模,提高求解速度。论文包括以下各项主要工作:① 本文改进了反应堆压力容器模拟体有限元自动建模程序。应用映射网格法生成了模拟体不同分析区域的有限元网格,实现了反应堆压力容器模拟体参数化有限元分析的计算模型和分析数据的自动生成。② 新开发了高效的线性方程组求解器,包括具有分块消元功能的LDLT分解求解程序和与之结合使用的带宽优化程序,使解题规模显著增大,求解效率大大提高。③ 对三维瞬态密封分析程序进行了功能扩展和算法改进,新增了程序重启动功能,方便地实现了程序的断点控制;改进了螺栓预紧工况下平面螺栓和球面螺栓接触状态修正方法,使其能更好的模拟实际情况。④ 在程序单元库中新增了三维可变节点二次等参单元,包括四面体10节点二次单元、五面体15节点二次单元以及六面体20节点二次单元,推导了各种等参单元的形函数,修改了密封分析程序中的相应模块。⑤ 应用改进后的反应堆压力容器瞬态密封分析程序对压力容器模拟体1/4模型进行了冷态和热态密封分析。该计算模型超过10万个自由度,涉及预紧、加压、升温、恒温、降温等复杂运行工况。与中国核动力运行研究所在压力容器模拟体上测试所得的应力、温度、螺栓力、分离量等大量数据对比,在冷态和热态两种工况下,密封分析结果的定性规律和定量数据都与实验结果吻合良好。更多还原

【Abstract】 This paper comes from the project of “the 3-D Program Development for Transient Sealing Analysis in Reactor Pressure Vessel”. This project is the subproject of the national defense key technology pre-research project on the program development of reactor design in the tenth Five-year Plan. The aim of this research project is to develop advanced program system of sealing analysis for the nuclear pressure vessel, to improve the design quality and level, to ensure the reliability of the sealing system in the advanced pressurized water reactor vessel, to increase the design quality and ensure the design reliability of the nuclear vessel sealing system.The sealing failure of flange is the primary failure mode of the nuclear reactor pressure vessel because it works in formidable environment with high temperature, high pressure and radioactive working substance. The sealing analysis involves not only the elasto-plastic nonlinear between sealing surfaces, thermal nonlinear and transient characteristics of thermal transfer, but also couple interaction among them. Till now, no matter the domestic or the foreign commercial software doesn’t has the comprehensive function. The main task of this paper is to improve and extend the sealing analysis program based on the developed 3-D program system of reactor pressure vessel, to increase analysis functions as well as solving scale and speed.The main work of this paper includes:①The automatic modeling program of finite element analysis for the reactor pressure vessel model is developed and improved. The finite element mesh for arbitrary region of the model can be generated based on the mapping mesh method. Using this program, the parametric FE model and calculation data of the RPV model can be obtained.②An efficient linear equations solver is developed. It includes a block elimination LDLT decomposition and solution program as well as a corresponding bandwidth optimization program. It can enlarge the solving scale and improve the solving efficiency. ③The following function expansion and algorithm improvement has been done to the 3-D Transient Sealing Analysis program: restart function is added to realize the breakpoint control, the contact state correction method of the plane bolt and the spherical bolt at the pre-tension work condition is improved to simulate the real situation better.④Several 3-D node-variable quadratic isoperimetric elements are added to the element library of the program, such as the 10-node tetrahedron quadratic element, the 15-node pentahedron <WP=7>quadratic element and the 20-node hexahedron quadratic element. The shape functions of the isoperimetric elements are deduced and added to the corresponding module of the program. ⑤The improved transient sealing analysis program for reactor pressure vessel is applied to analyze the 1/4 structure of the model under the cool and the thermal work conditions. The scale of the calculation module is more than 100 thousand degrees of freedom. The load cases include bolt pre-tensioning, compression, heating, thermostated heating and cooling process. The computational results are compared with the experimental data including stress, temperature, bolt force and separation, which are obtained from the pressure vessel module by the China Nuclear Power Operation Research Institute. Either under cooling or heating cycle, the computational results are in good concordance with the experimental data in both the qualitative regularity and the quantitative data.更多还原