基于RELAP5-HD的先进压水堆仿真研究

发布时间：2018-06-17 11:57

本文选题：AP1000 + RELAP5-HD　；参考：《哈尔滨工程大学》2014年硕士论文

【摘要】：AP1000是西屋公司在继承传统压水堆成熟技术,并吸取其长期积累的运行经验的基础上开发出来的三代+压水堆,它是一个革新性的设计,符合美国核管会安全评审要求,并满足先进轻水堆用户要求文件。AP1000是一个单堆布置的两环路核电厂,其净电输出功率为1117MWe。与传统压水堆核电厂相比,其最主要的特点就是使用了非能动安全系统,利用非能动特性,如压缩气体储能,重力势能,自然循环等代替能动设备如泵,交流电源等进行驱动,从而使得电厂的安全性和可靠性得到大幅提升。AP1000的非能动堆芯冷却系统包括非能动余热排出系统和非能动安全注射系统,以及用于有效衔接高、中、低压安注的自动卸压系统。AP1000核电站的整体系统结构,运行模式和特点以及其非能动设计理念与我国目前大量运营的反应堆相比有较大不同。为了熟悉先进压水堆的系统结构,全面掌握其运行特点,充分理解其非能动设计理念,并且对堆芯下降腔等具有典型多维流动的部件进行模拟,有必要使用具有多维组分模拟功能的RELAP-HD程序,对AP1000进行建模仿真研究。RELAP5-3D是RELAP5系列程序的最新版本,与之前的RELAP5版本相比,RELAP5-3D最重要的改进在于多维水力学部件和多维中子动力学模型的引入。GSE公司将RELAP5-3D嵌入其实时仿真平台SimExec上,形成了 RELAP5-HD,它可以在不损害RELAP5-3D最佳估算程序的完整性的前提下实时地进行电厂运行状态的热工水力求解。本文首先利用RELPA5-HD程序建立了 AP1000核电厂的模型,主要包括其压力容器、蒸汽发生器、主管道、稳压器、非能动堆芯冷却系统,以及控制系统等。压力容器内的下降通道和堆芯用多维组分进行模拟。对该模型进行了稳态调试,并将最终的稳态结果与AP1000电厂额定值进行比较,以验证稳态模型的适用性。同时,还在稳态情况下对压力容器下降通道和堆芯内的多维流动进行了分析。利用AP1000核电厂对冷段10-in小破口失水事故的响应,对非能动堆芯冷却系统模型及控制系统进行了验证。最后,使用经过验证的模型,对压水堆核电厂内的典型事故,如主给水丧失事故,主蒸汽管道破裂事故等进行了模拟,分析了 AP1000非能动堆芯冷却系统对非LOCA事故的响应,并对主蒸汽管道破裂事故中压力容器内的多维流动和不对称现象进行了分析。仿真结果表明,事故中非能动堆芯冷却系统都能自动投入,有效导出堆芯余热,确保反应堆安全。在主蒸汽管道破裂事故中,由于环路以及非能动系统响应的不对称性,压力容器的下降通道和堆芯出口处也会有明显的不对称现象。
[Abstract]:AP1000 is a third-generation PWR developed by Westinghouse on the basis of inheriting the mature technology of traditional PWR and absorbing its long accumulated operating experience. It is an innovative design that meets the requirements of the safety assessment of the American Nuclear Regulatory Commission. AP1000 is a two-loop nuclear power plant with a single reactor arrangement and its net output power is 1117MWe. Compared with the traditional PWR nuclear power plant, its main feature is the use of passive safety system, the use of inactive characteristics, such as compressed gas energy storage, gravity potential energy, natural circulation instead of active equipment such as pumps, AC power, etc. This greatly improves the safety and reliability of power plants. AP1000's inactive core cooling system includes inactive residual heat removal systems and inactive safety injection systems, and is used to effectively connect high, medium, The whole system structure, operation mode and characteristics of the automatic pressure relief system .AP1000, and its inactive design concept are quite different from those of a large number of reactors in our country at present. In order to be familiar with the system structure of the advanced PWR, fully grasp its operating characteristics, fully understand its inactive design concept, and simulate the typical multi-dimensional flow components such as the falling chamber of the reactor core, It is necessary to use the RELAP-HD program, which has the function of multi-component simulation, to model and simulate AP1000. RELAP5-3D is the latest version of the RELAP5 series. The most important improvement over previous RELAP5 versions is the introduction of multidimensional hydraulics components and multidimensional neutron dynamics models. GSE has embedded RELAP5-3D on its real-time simulation platform, SimExec. The RELAP5-HD is formed, which can be used to solve the thermohydraulic problems of power plant operation in real time without compromising the integrity of the RELAP5-3D optimal estimation program. In this paper, the model of AP1000 nuclear power plant is established by using RELPA5-HD program, including its pressure vessel, steam generator, main pipeline, voltage regulator, inactive core cooling system and control system. The drop channel and core in the pressure vessel are simulated with multi-dimensional components. The steady-state model is debugged and compared with AP1000 power plant rating to verify the applicability of the steady-state model. At the same time, the multi-dimensional flow in the drop channel and core of the pressure vessel is analyzed under steady state. The model and control system of inactive core cooling system are verified by the response of AP1000 nuclear power plant to the small break water loss accident of 10-in cold section. Finally, the typical accidents in PWR nuclear power plant, such as main water supply loss accident, main steam pipeline rupture accident and so on, are simulated by using the verified model, and the response of AP1000 inactive core cooling system to non-LOCA accident is analyzed. The multi-dimensional flow and asymmetry in the pressure vessel in the main steam pipeline rupture accident are analyzed. The simulation results show that the non-active core cooling system of the accident can be automatically put into operation, and the residual heat of the reactor core can be effectively derived to ensure the safety of the reactor. In the main steam pipeline failure due to the asymmetry of the loop and the response of the inactive system there will also be obvious asymmetry in the descending channel and the core outlet of the pressure vessel.
【学位授予单位】：哈尔滨工程大学
【学位级别】：硕士
【学位授予年份】：2014
【分类号】：TM623;TL421.1

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