高温气冷堆是具有第四代核能系统安全特征的先进堆型，其安全性能优异，在极端事故条件下能够依靠负反应性温度系数和较大的温升裕度停堆，堆芯余热也能通过自然机制导出。失冷不失压（PLOFC）是高温气冷堆的典型事故场景，该工况下堆内氦气的自然对流以及进一步形成的自然循环直接影响到堆芯的余热载出和堆内温度再分布，是一种重要的自然传热机制。研究停堆后堆内氦气的自然对流特性可以为全面评估反应堆安全性和优化反应堆设计提供参考，具有重要意义。基于CFD软件Fluent的多孔介质模型和UDS功能，本研究开发了能够应用于球床高温气冷堆稳态、瞬态热工水力计算的二维模型。然后将模型应用于HTR-10的满功率稳态运行试验和9 MW功率水平下停堆试验的模拟，获得了堆内的温度分布和流场分布，并与试验数据、其他程序的计算结果进行了对比，验证了本文模型的可靠性，并解释了模型偏差的原因。以9 MW功率水平下停堆试验为例，分析了PLOFC工况下堆内氦气的自然对流特性，包括堆芯及其周围流道氦气自然循环的循环流向和形成原因。为定量评估自然对流强度，定义了自然循环流量的概念，并在此基础上分析了停堆过程中，自然循环流量随时间的变化特点及其对顶反射层散热量的影响。基于本文模型，进一步探究了水冷壁温度、堆芯等效导热系数、燃料球热容和堆芯初始温度四个参数对停堆后氦气自然对流特性的影响。研究发现，堆芯的余热载出和压力壳的散热之间存在解耦现象，即压力壳相关热工参数几乎不影响堆芯的热工行为，所以水冷壁温度对堆内氦气自然对流的影响很小，其他三个参数通过影响氦气温度梯度和粘性进而影响氦气的自然对流强度。堆芯等效导热系数越大，自然对流强度越小；燃料球热容越大，停堆工况前期自然对流强度越小，循环流量越晚达到峰值但衰减也更慢；堆芯初始温度越高，停堆工况前期自然对流强度越小。本研究开发了球床高温气冷堆的热工水力CFD计算模型并对模型进行了验证，然后将模型应用到堆内氦气自然对流特性的研究中，相关工作加深了对氦气自然对流特性的认识，同时也为高温气冷堆的热工水力模拟提供了一套可靠的工具。

The high temperature gas-cooled reactor (HTGR) is an advanced reactor type with safety characteristics of Generation IV nuclear energy systems. It has excellent safety performance, characterized by its self-shutdown ability via negative temperature coefficient of reactivity and large temperature marginas as well as residual heat removal capability through natural mechanisms even under extreme accident conditions. Pressurized loss of forced cooling (PLOFC) is a typical accident scenario of HTGR. Under this condition, the natural convection of helium in the reactor core and the further natural circulation, which is an important natural heat transfer mechanism, directly affects the core residual heat removal and the temperature redistribution in the reactor. It is of great importance for the comprehensive reactor safety evaluation and design optimization to investigate the natural convection characteristics of helium in the reactor after shutdown.Based on the porous media model and user-defined scalar (UDS) of the CFD software Fluent, a two-dimensional model that can be applied to steady-state and transient thermal-hydraulic calculations of the pebble bed HTGR is developed. The model is applied to the simulation of two tests of the HTR-10, i.e., the steady-state full-power operation test and the scram test at 9 MW power level. The temperature field and flow field in the reactor are obtained and compared with the test results and the calculation results from other codes. The reliability of the model developed by this study is verified. Besides, the model deviations are analyzed and explained.Based on the test conditions of the scram test at 9 MW power level, the natural convection characteristics of helium under the PLOFC condition in the reactor are analyzed, including the cause and the flow direction of natural circulation in the core and its surrounding flow channels. To quantitatively evaluate the intensity of natural convection, the concept of natural circulation flow rate is defined, based on which the flow rate over time after shutdown and its impact on the heat dissipation of the top reflector are analyzed.Using the model developed by this study, the influences of the temperature of the water-cooling panel, the effective heat conductivity of the pebble bed, the specific heat capacity of the fuel elements and the initial temperature of the core on the post-shutdown helium natural convection characteristics are further investigated. There is a decoupling phenomenon between residual heat removal from the core and heat dissipation from the pressure vessel, which means that the thermal parameters related to the pressure vessel hardly affect the thermal behavior of the core, so the temperature of the water-cooling panel hardly affect the natural convection of helium in the reactor. The other three parameters affect the natural convection intensity of helium by affecting the temperature gradient and viscosity of helium. The greater the effective heat conductivity of the pebble bed is, the smaller the natural convection intensity is. For larger the specific heat capacity of the fuel elements, the natural convection intensity is smaller at the initial stage after shutdown, and the circulating flow rate reaches its peak later but also attenuates more slowly. And a higher temperature of the initial core provides a smaller natural convection intensity at the initial stage after shutdown.In this study, a thermal-hydraulic calculation model based on CFD method is developed and validated for the pebble bed HTGR, and then is applied to the investigation of natural convection characteristics of helium in the reactor. The work leads better understanding of natural convection characteristics of helium and provides a reliable tool for the thermal-hydraulic simulation of pebble bed HTGRs.