The Very High Temperature Reactor (VHTR) is one of the next generation candidates for nuclear reactors, according to the IAEA. Predicting the thermohydraulic performance of high temperature reactors is an important contribution to technology development. The evaluation of the thermohydraulic behavior of the steady-state of the HTR-10 gas-cooled pebble bed high temperature test reactor was a challenge proposed to the international scientific community by the IAEA. This paper proposes a methodology for the thermohydraulic study of steady-states of high temperature gas-cooled pebble bed nuclear reactors, using real scale three-dimensional computational thermohydraulic modeling. The focus of this paper is to discuss the methodology that was improved. Analyses were carried out from comparative studies with experimental data and data obtained by other computational codes. A CFD method was developed to analyzes the main thermohydraulic parameters. The ANSYS CFX's full porous media approach was used to model the pebble bed reactor core. A variable porosity model was implemented in the pebble bed simulation to consider the closeness of the walls. The effect of the Reactor Core Cooling System was modeled from variable boundary conditions in the reactor pressure vessel surfaces. The temperature values obtained in the pebbles, the coolant, and the structural elements were confirmed to be under the normal operating limits. Also were obtained the threedimensional coolant velocities and pressures drop profiles which are important information to understand the thermohydraulic behavior of the reactor. Additionally, was evidenced the presence of thermohydraulic three-dimensional effects showing the important role of the 3D thermohydraulic modeling. With this paper was bear out that use of CFD applied to nuclear thermohydraulic modeling of HTR-10 increase the understanding of the phenomena occurring in high temperature gas-cooled pebble bed nuclear reactors.

据国际原子能机构称，超高温反应堆（VHTR）是下一代核反应堆的候选者之一。预测高温反应器的热工水力性能是技术发展的重要贡献。对HTR-10气冷卵石床高温试验反应器的稳态热工液压性能的评估是国际原子能机构向国际科学界提出的挑战。本文提出了一种利用实数三维计算热工水力模型对高温气冷卵石床核反应堆稳态进行热工水力研究的方法。本文的重点是讨论改进的方法。通过比较研究对实验数据和其他计算代码获得的数据进行分析。开发了一种CFD方法来分析主要的热工液压参数。ANSYS CFX的全多孔介质方法用于对卵石床反应堆堆芯进行建模。在卵石床模拟中实现了可变孔隙度模型，以考虑墙的紧密度。反应堆堆芯冷却系统的效果是根据反应堆压力容器表面的可变边界条件进行建模的。确认在卵石，冷却剂和结构元件中获得的温度值在正常操作范围内。还获得了三维冷却液速度和压降曲线，这对于了解反应堆的热工行为是重要的信息。此外，有证据表明存在热工三维效应，显示了3D热工建模的重要作用。本文证明了将CFD用于HTR-10核热工水力模型可以增加对高温气冷卵石床核反应堆中发生现象的理解。