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Numerical simulations for steam-water CCFL tests using the 1/3 scale rectangular channel simulating a PWR hot leg

机译:使用1/3比例矩形通道模拟压水堆热腿的蒸汽-CCFL试验的数值模拟

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摘要

In reflux condensation, steam generated in the reactor core and water condensed in a steam generator (SG) form countercurrent flow in a hot leg, which consists of a horizontal pipe, an elbow and an inclined pipe. Both countercurrent air-water and steam-water tests were previously carried out at Helmholtz-Zentrum Dresden-Rossendorf (HZDR) using the 1/3 scale rectangular channel simulating a PWR hot leg installed in the pressure chamber of the TOPFLOW facility. In this paper, in order to evaluate the effects of fluid properties, the steam-water CCFL (countercurrent flow limitation) tests at HZDR were simulated using the CFD software, FLUENT 6.3.26. The computational domain included the reactor vessel simulator, hot leg and SG inlet chamber in order to avoid uncertainties of boundary conditions at both ends of the hot leg. The VOF (volume of fluid) method and two-fluid (2F) model were used. In the 2F model, the combination of three correlations on the interfacial friction coefficients, which had been validated for the 1/15 and 1/5 scale tests at Kobe University, was used as a function of local void fractions. The CCFL characteristics predicted by the 2F and VOF agreed relatively well with the steam-water CCFL data at HZDR but overestimated the effects of fluid properties on CCFL. The VOF simulations were better able to reproduce the fluid properties than the 2F simulations.
机译:在回流冷凝中,反应堆堆芯中产生的蒸汽和在蒸汽发生器(SG)中冷凝的水在热管中形成逆流,该热管由水平管,弯头和倾斜管组成。先前在Helmholtz-Zentrum Dresden-Rossendorf(HZDR)上使用1/3比例的矩形通道模拟了安装在TOPFLOW设备压力室中的PWR热压腿,进行了逆流空气和蒸汽测试。在本文中,为了评估流体特性的影响,使用CFD软件FLUENT 6.3.26对HZDR的蒸汽-水CCFL(逆流限制)测试进行了模拟。计算域包括反应堆容器模拟器,热管和SG进气室,以避免热管两端边界条件的不确定性。使用了VOF(流体体积)方法和两流体(2F)模型。在2F模型中,界面摩擦系数的三个相关性的组合(已通过神户大学的1/15和1/5比例测试验证)被用作局部空隙率的函数。 2F和VOF预测的CCFL特性与HZDR的蒸汽水CCFL数据相对吻合,但高估了流体特性对CCFL的影响。与2F模拟相比,VOF模拟能够更好地再现流体特性。

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  • 来源
    《Nuclear Engineering and Design》 |2012年第8期|p.14-23|共10页
  • 作者单位

    Institute of Nuclear Safety System, Inc., 64 Sata, Mihama-cho, Mikata-gun, Fukui 979-1205, Japan;

    Institute of Nuclear Safety System, Inc., 64 Sata, Mihama-cho, Mikata-gun, Fukui 979-1205, Japan;

    Institute of Nuclear Safety System, Inc., 64 Sata, Mihama-cho, Mikata-gun, Fukui 979-1205, Japan;

    Helmholtz-Zentrum Dresden-Rossendorfe. V., P.O. Box 510 119, D-01314 Dresden, Germany;

    Helmholtz-Zentrum Dresden-Rossendorfe. V., P.O. Box 510 119, D-01314 Dresden, Germany;

    Graduate School of Engineering, Kobe University, Nada, Kobe 657-8501, Japan;

  • 收录信息 美国《科学引文索引》(SCI);美国《工程索引》(EI);美国《生物学医学文摘》(MEDLINE);
  • 原文格式 PDF
  • 正文语种 eng
  • 中图分类
  • 关键词

  • 入库时间 2022-08-18 00:44:00

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