CFD evaluation on the thermohydraulic characteristics of tube support plates in steam generator

被引:0
|
作者
Zhang, B. [1 ]
Zhang, H. [1 ]
Han, B. [1 ]
Yang, B. W. [1 ]
Mo, S. J. [2 ]
Ren, H. B. [2 ]
Qin, J. M. [2 ]
Zuo, C. P. [2 ]
机构
[1] Xi An Jiao Tong Univ, Sch Nucl Sci & Technol, Xian 710049, Peoples R China
[2] China Nucl Power Design Co Ltd Shenzhen, Shenzhen 518124, Peoples R China
关键词
THERMAL-HYDRAULIC CHARACTERISTICS; SECONDARY-SIDE;
D O I
10.3139/124.110749
中图分类号
TL [原子能技术]; O571 [原子核物理学];
学科分类号
0827 ; 082701 ;
摘要
The integrity and thermal hydraulic characteristics of steam generator are of great concern in the nuclear industry. The tube support plates (TSP), one of the most important components of the steam generator, not only support the heat transfer tubes, but also affect the flow dynamic and thermal hydraulic characteristics of the secondary-side flow inside the steam generator. Different working conditions, ranging from single-phase adiabatic condition to two-phase high-void boiling condition, are simulated and analyzed. Calculated void fraction, under simple geometry, agrees well with the experiment data whilst the simulated heat transfer coefficient is tremendously close to the empirical correlation. Temperature, void fraction, and velocity distributions in different locations show reasonable distribution. The simulation results indicate that TSP can enhance the heat transfer in the secondary side of the steam generator. On the top of TSP, with the increase in cross-section flow area, the back-flow phenomenon occurs, which might lead to the contamination of precipitation.
引用
收藏
页码:299 / 307
页数:9
相关论文
共 50 条
  • [41] CFD analysis of a CiADS fuel assembly during the steam generator tube rupture accident based on the LBEsteamEulerFoam
    Yun-Xiang Li
    Lu Meng
    Song Li
    Zi-Nan Huang
    Di-Si Wang
    Bo Liu
    You-Peng Zhang
    Tian-Ji Peng
    Lu Zhang
    Xing-Kang Su
    Wei Jiang
    Nuclear Science and Techniques, 2023, 34 (10) : 138 - 149
  • [42] CFD analysis of a CiADS fuel assembly during the steam generator tube rupture accident based on the LBEsteamEulerFoam
    Yun-Xiang Li
    Lu Meng
    Song Li
    Zi-Nan Huang
    Di-Si Wang
    Bo Liu
    You-Peng Zhang
    Tian-Ji Peng
    Lu Zhang
    Xing-Kang Su
    Wei Jiang
    Nuclear Science and Techniques, 2023, 34
  • [43] CFD analysis of a CiADS fuel assembly during the steam generator tube rupture accident based on the LBEsteamEulerFoam
    Li, Yun-Xiang
    Meng, Lu
    Li, Song
    Huang, Zi-Nan
    Wang, Di-Si
    Liu, Bo
    Zhang, You-Peng
    Peng, Tian-Ji
    Zhang, Lu
    Su, Xing-Kang
    Jiang, Wei
    NUCLEAR SCIENCE AND TECHNIQUES, 2023, 34 (10)
  • [44] Comparison of Absolute and Differential ECT Signals around Tube Support Plate in Steam Generator
    Shin, Young-Kil
    Lee, Yun-Tai
    Song, Myung-Ho
    JOURNAL OF THE KOREAN SOCIETY FOR NONDESTRUCTIVE TESTING, 2005, 25 (03) : 201 - 208
  • [45] A Study for the Effect of the ECT Signal influenced by Steam Generator Tube Support Plate Blockage
    Jee, DongHyun
    Cho, ChanHee
    ELECTROMAGNETIC NONDESTRUCTIVE EVALUATION (XII), 2009, 32 : 71 - 78
  • [46] Mathematical simulation of thermohydraulic processes in a PGV-1000 horizontal steam generator
    Urban, T.V.
    Melikhov, V.I.
    Melikhov, O.I.
    2002, IAPC Nauka/Interperiodica (49)
  • [47] RADIOGRAPHIC SYSTEM FOR EVALUATION OF STEAM GENERATOR SUPPORT PLATE INTEGRITY.
    McDearman, W.R.
    Radcliff, F.T.
    Jamison, T.D.
    Electric Power Research Institute (Report) EPRI NP, 1981,
  • [48] STEAM-GENERATOR SECONDARY PH DURING A STEAM-GENERATOR TUBE RUPTURE
    ADAMS, JP
    PETERSON, ES
    NUCLEAR TECHNOLOGY, 1993, 102 (03) : 304 - 312
  • [49] Steam generator tube flow rate regulation of HTGR steam generator test loop
    Li, Xiao-Wei
    Wu, Xin-Xin
    Ju, Huai-Ming
    Yuanzineng Kexue Jishu/Atomic Energy Science and Technology, 2012, 46 (SUPPL.2): : 859 - 862
  • [50] Structural integrity evaluation method for overheating rupture of FBR steam generator tube
    Machida, H
    Yoshioka, N
    Ogo, H
    NUCLEAR ENGINEERING AND DESIGN, 2002, 212 (1-3) : 183 - 192