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Investigation of cross flow mixing on rod bundle safety margins using subchannel analysis framework

机译:使用子通道分析框架研究棒材束安全裕度上的错流混合

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摘要

An investigation on effect of cross flow mixing on rod bundle thermal hydraulic safety margins is carried out using a sub-channel analysis framework. In the present study, general purpose automated sub-channel analysis framework is developed to carry out the thermal hydraulic analysis of nuclear reactor core. The framework involves the preprocessor to generate the sub-channel layout of the fuel assemblies of different sizes and shapes such as square and triangular pitched pin bundle. The analyses are carried out for different simulated rod bundles of square pitched array of typical nuclear fuel assembly. The parameters considered in this study are the size of the bundle, p/d ratio, intensity of turbulent mixing, mass flux, uniform and non-uniform axial power distribution. The coolant local conditions are significantly affected due to the inter channel mixing between sub-channels. The sub-channel layouts for fuel bundles of different sizes varying from 2 x 2 to 10 x 10 and 17 x 17 for same p/d and w/d ratios of 1.05 to 1.3 are generated. The analyses are carried out by varying the intensity of turbulent mixing parameter (c) from 0 (no mixing) to 0.1 (high mixing) using the developed framework. It is found that hot channel temperature decreased by 26% in smaller bundle due to strong interaction between wall and corner channel. In case of large sized assembly above 5 x 5, hot channel temperature is less affected. The estimation of coolant temperature, fuel pin surface temperature and local quality are also carried out without considering the inter-channel mixing. The results are compared with a change in the local conditions for different degree of mixing among the fuel bundle sub-channels. The developed Framework is used for performing the sensitivity-studies during the design and analysis phase of nuclear reactor core for estimation of thermal hydraulic safety margins in an efficient way with minimum human intervention.
机译:使用子通道分析框架对错流混合对棒束热工水力安全裕度的影响进行了研究。在本研究中,通用通用自动子通道分析框架被开发来进行核反应堆堆芯的热力水力分析。该框架包括预处理器,以生成不同尺寸和形状的燃料组件(例如方形和三角形变桨销束)的子通道布局。对典型核燃料组件的方形节距阵列的不同模拟棒束进行了分析。在这项研究中考虑的参数是束的大小,p / d比,湍流混合的强度,质量通量,均匀和不均匀的轴向功率分布。由于子通道之间的通道间混合,冷却剂的局部条件受到显着影响。对于相同的p / d和w / d比为1.05至1.3的不同尺寸的燃料束,其子通道布局从2 x 2到10 x 10和17 x 17不等。通过使用开发的框架将湍流混合参数(c)的强度从0(不混合)更改为0.1(高混合)来进行分析。发现较小的管束中的热通道温度下降了26%,这是由于壁和角通道之间强烈的相互作用。对于大于5 x 5的大型组件,热通道温度的影响较小。冷却剂温度,燃料销表面温度和局部质量的估计也可以在不考虑通道间混合的情况下进行。将结果与燃料束子通道之间不同混合程度的局部条件变化进行比较。所开发的框架用于在核反应堆堆芯的设计和分析阶段进行敏感性研究,从而以最少的人工干预以有效的方式估算热工水力安全裕度。

著录项

  • 来源
    《Nuclear Engineering and Design》 |2018年第8期|30-43|共14页
  • 作者

    Moorthi A.; Sharma Anil Kumar;

  • 作者单位

    Bhabha Atom Res Ctr Facil, Kalpakkam, Tamil Nadu, India;

    Indira Gandhi Ctr Atom Res, Fast Reactor Technol Grp, Safety Engn Div, Kalpakkam, Tamil Nadu, India;

  • 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

    Sub-channel analysis; Rod bundle; Cross flow; Mixing; Framework;

    机译:子通道分析杆束错流混合框架;
  • 入库时间 2022-08-18 00:40:47

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