高级检索

留言板

尊敬的读者、作者、审稿人, 关于本刊的投稿、审稿、编辑和出版的任何问题, 您可以本页添加留言。我们将尽快给您答复。谢谢您的支持!

姓名
邮箱
手机号码
标题
留言内容
验证码

单棒垂直方形通道临界热流密度实验研究

刘伟 郭俊良 张丹 桂淼 胡迎 刘扬

刘伟, 郭俊良, 张丹, 桂淼, 胡迎, 刘扬. 单棒垂直方形通道临界热流密度实验研究[J]. 核动力工程, 2022, 43(1): 42-47. doi: 10.13832/j.jnpe.2022.01.0042
引用本文: 刘伟, 郭俊良, 张丹, 桂淼, 胡迎, 刘扬. 单棒垂直方形通道临界热流密度实验研究[J]. 核动力工程, 2022, 43(1): 42-47. doi: 10.13832/j.jnpe.2022.01.0042
Liu Wei, Guo Junliang, Zhang Dan, Gui Miao, Hu Ying, Liu Yang. Experimental Study on Critical Heat Flux of Vertical Square Channel with Single Rod[J]. Nuclear Power Engineering, 2022, 43(1): 42-47. doi: 10.13832/j.jnpe.2022.01.0042
Citation: Liu Wei, Guo Junliang, Zhang Dan, Gui Miao, Hu Ying, Liu Yang. Experimental Study on Critical Heat Flux of Vertical Square Channel with Single Rod[J]. Nuclear Power Engineering, 2022, 43(1): 42-47. doi: 10.13832/j.jnpe.2022.01.0042

单棒垂直方形通道临界热流密度实验研究

doi: 10.13832/j.jnpe.2022.01.0042
详细信息
    作者简介:

    刘 伟(1989—),男,博士,主要从事反应堆热工水力与安全分析研究,E-mail: liuwei0958@126.com

  • 中图分类号: TL334

Experimental Study on Critical Heat Flux of Vertical Square Channel with Single Rod

  • 摘要: 采用R134a作为流体工质,对单棒垂直方形通道临界热流密度(CHF)进行了实验研究。流道横截面为19 mm×19 mm的方形通道,内置外径为9.5 mm的单根加热棒,用来模拟压水堆中典型栅元通道。实验工况通过流体模化方法覆盖了压水堆典型运行工况。实验结果表明,R134a在方形通道内的CHF参数趋势与圆管中水的CHF参数趋势相同,R134a可以替代水作为模化工质;通过对圆管Bowring关系式和Katto & Ohno关系式进行冷壁因子修正,可用于预测带有冷壁的方形通道的CHF;Katto的流体模化方法适用于带有冷壁的方形通道。

     

  • 图  1  R134a CHF实验装置

    Figure  1.  R134a CHF Experimental Apparatus

    图  2  单棒布置示意图

    Figure  2.  Schematic Diagram of Single Rod Arrangement

    图  3  实验段横截面示意图

    Figure  3.  Schematic Diagram of Cross-section of Experimental Section

    图  4  CHF随临界含汽率的变化趋势

    q—CHF,下同

    Figure  4.  Variation Trend of CHF with Critical Steam Content

    图  5  CHF随压力的变化趋势

    Figure  5.  Variation Trend of CHF with Pressure

    图  6  CHF随入口过冷度的变化趋势

    Figure  6.  Variation Trend of CHF with Inlet Undercooling

    图  7  CHF随质量流速的变化趋势

    Figure  7.  Variation Trend of CHF with Mass Flow Rate

    图  8  Katto & Ohno关系式预测值与实验值的对比

    qpre—CHF预测值;qexp—CHF实验值

    Figure  8.  Comparison between Predicted Value and Experimental Value of Katto & Ohno Relation

    图  9  Bowring关系式预测值与实验值的对比

    Figure  9.  Comparison between Predicted Value and Experimental Value of Bowring Relation

    表  1  Katto模化方法

    Table  1.   Modeling Method of Katto

    类型准则
    几何相似 ${\left( {\dfrac{{{L_{\rm{h}}}}}{{{D_{\rm{h}}}}}} \right)_{{\rm{R134a}}}} = {\left( {\dfrac{{{L_{\rm{h}}}}}{{{D_{\rm{h}}}}}} \right)_{{\rm{Water}}}}$
    水力学相似 ${\left( {\dfrac{{{\rho _{\rm{f}}}}}{{{\rho _{\rm{g}}}}}} \right)_{{\rm{R134a}}}} = {\left( {\dfrac{{{\rho _{\rm{f}}}}}{{{\rho _{\rm{g}}}}}} \right)_{{\rm{Water}}}}$
    热力学相似 ${\left( {\dfrac{{\Delta {h_{{\rm{in}}}}}}{{{h_{{\rm{fg}}}}}}} \right)_{{\rm{R134a}}}} = {\left( {\dfrac{{\Delta {h_{{\rm{in}}}}}}{{{h_{{\rm{fg}}}}}}} \right)_{{\rm{Water}}}}$
    韦伯数(We)相同 ${\left( {\dfrac{ {G\sqrt D } }{ {\sqrt { {\rho _{_{\rm{f} } } }\sigma } } } } \right)_{ {\rm{R134a} } } } = {\left( {\dfrac{ {G\sqrt D } }{ {\sqrt { {\rho_{ _{\rm{f} }} }\sigma } } } } \right)_{ {\rm{Water} } } }$
      Lh—加热高度;Dh—热当量直径;Δhin—入口焓升;hfg—汽化潜热;G—质量流速;σ—表面张力;ρfρg—液相和气相密度;D—管道直径
    下载: 导出CSV

    表  2  实验工况

    Table  2.   Experimental Conditions

    参数R134a
    压力/MPa 1.8、2.1、2.7、4 10.9、12.5、14.1、15.6
    质量流速/[kg· (m2·s)−1] 600~2100 800~3000
    入口过冷度(ΔTin)/℃ 10~40 利用Δhin/hfg相等模化
    下载: 导出CSV

    表  3  本实验的不确定度

    Table  3.   Uncertainty of Experiment

    参数不确定度/%
    长度、直径、厚度 ±0.1、±0.5、±3.0
    压力 ±0.7
    温度 ±1.3
    质量流速 ±1.7
    功率 ±2.2
    热流密度 ±5.1
    下载: 导出CSV
  • [1] 徐济鋆, 贾斗南, 沸腾传热和汽液两相流[M]. 北京: 中国原子能出版社, 2001: 300-301.
    [2] KATTO Y, OHNO H. An improved version of the generalized correlation of critical heat flux for the forced convective boiling in uniformly heated vertical tubes[J]. International Journal of Heat and Mass Transfer, 1984(27): 1641-1648.
    [3] BOWRING R W. A simple but accurate round tube, uniform heat flux, dryout correlation over pressure range 0.7-17 MN/m2 (100-2500 psia): AEEW-R-789[R]. Winfrith, England: UK Atomic Energy Authority, 1972.
    [4] HALL D D, MUDAWAR I. Critical heat flux for water flow in tubes-II subcooled CHF correlations[J]. International Journal of Heat and Mass Transfer, 2000(43): 2606-2640.
    [5] ALEKSEEV G V, ZENKEVITCH B A, PESKOV O L, et al. Burn-out heat fluxes under forced water flow[C]. England: Third United Nations International Conference on the Peaceful Uses of Atomic Energy, 1964.
    [6] LEE K L, BANG I C, CHANG S H. The characteristic and visualization of critical flux of R-134a flowing in a vertical annular geometry with spacer grids[J]. International Journal of Heat and Mass Transfer, 2008, 51(1-2): 91-103. doi: 10.1016/j.ijheatmasstransfer.2007.04.024
    [7] LIU Y, LIU W, SHAN J Q, et al. A mechanistic bubble crowding model for predicting critical heat flux in subchannels of a bundle[J]. Annals of Nuclear Energy, 2020(137): 107085.
    [8] CHENG X, ERBACHER F J, MULLER U. Critical heat flux in uniformly heated vertical tubes[J]. International Journal of Heat and Mass Transfer, 1997(40): 2929-2939.
    [9] PIORO I L, GROENEVELD D C, CHENG S C, et al. Comparison of CHF measurements in R-134a cooled tubes and the water CHF look-up table[J]. International Journal of Heat and Mass Transfer, 2001, 44(1): 73-88. doi: 10.1016/S0017-9310(00)00093-4
    [10] TONG L S. An evaluation of the departure from nucleate boiling in bundles of reactor fuel rods[J]. Nuclear Science and Engineering, 1986, 33(1): 7-15.
  • 加载中
图(9) / 表(3)
计量
  • 文章访问数:  520
  • HTML全文浏览量:  44
  • PDF下载量:  40
  • 被引次数: 0
出版历程
  • 收稿日期:  2020-10-17
  • 修回日期:  2020-11-20
  • 刊出日期:  2022-02-01

目录

    /

    返回文章
    返回