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An integral 3D full-scale steady-state thermohydraulic calculation of the high temperature pebble bed gas-cooled reactor HTR-10

机译:高温卵石床气冷抗液型HTR-10的整体3D全尺度稳态热液压计算

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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.
机译:根据IAEA的说法,非常高温反应器(VHTR)是核反应堆的下一代候选者之一。预测高温反应器的热液压性能是技术开发的重要贡献。对HTR-10气冷卵石床高温试验反应器的稳态热液行为的评价是IAEA的国际科学界提出的挑战。本文采用了使用实际规模的三维计算热液模型来提出了一种用于高温气体床核反应堆稳态的热液压研究的方法。本文的重点是讨论改进的方法。通过使用其他计算代码获得的实验数据和数据进行分析。开发了CFD方法以分析主要的热液压参数。 ANSYS CFX的全多孔介质方法用于模拟卵石床反应器芯。在卵石床模拟中实施了可变孔隙度模型,以考虑墙壁的近距离。反应器芯冷却系统的效果从反应器压力容器表面的可变边界条件建模。在鹅卵石,冷却剂和结构元件中获得的温度值被确认为正常的操作限制。也获得了基准的冷却剂速度和压降曲线,这是了解反应器的热液态行为的重要信息。另外,显然存在热液三维效应,显示3D热液型建模的重要作用。用本文载有,使用CFD应用于HTR-10的核热液型建模,增加了高温气体冷却卵石床核反应堆中发生的现象的理解。

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