Advance Search
Volume 43 Issue 3
Jun.  2022
Turn off MathJax
Article Contents
Li Xiang, Sun Wan, Ding Shuhua, Huang Tao, Li Zhongchun, Pan Liangming. Study on CCFL Characteristics in Downcomer during Discharge Phase of LOCA with RELAP5 Code[J]. Nuclear Power Engineering, 2022, 43(3): 58-65. doi: 10.13832/j.jnpe.2022.03.0058
Citation: Li Xiang, Sun Wan, Ding Shuhua, Huang Tao, Li Zhongchun, Pan Liangming. Study on CCFL Characteristics in Downcomer during Discharge Phase of LOCA with RELAP5 Code[J]. Nuclear Power Engineering, 2022, 43(3): 58-65. doi: 10.13832/j.jnpe.2022.03.0058

Study on CCFL Characteristics in Downcomer during Discharge Phase of LOCA with RELAP5 Code

doi: 10.13832/j.jnpe.2022.03.0058
  • Received Date: 2021-04-21
  • Accepted Date: 2021-12-07
  • Rev Recd Date: 2021-09-02
  • Publish Date: 2022-06-07
  • Under the loss-of-coolant accident (LOCA), the two-phase countercurrent in downcomer is extremely likely to cause the vapor-liquid counter-current flow limitation (CCFL), which is not conducive to the smooth entry of emergency coolant into the core, which greatly affects the safety performance of the nuclear reactor system. Based on RELAP5 code, the Wallis overflow relation is used to model the UPFT experimental device and calculate the water injection behavior in the downcomer during discharge phase of LOCA; The validity of the model is verified by comparing the water storage capacity of lower chamber, the pressure in the downcomer and the transient changes of steam flow at the break, and the distribution characteristics of vapor phase velocity field and liquid phase volume fraction in the downcomer are analyzed. The results show that the flow irregularity caused by the three-dimensional characteristics of the channel structure in the downcomer affects the characteristics of the vapor-liquid CCFL. With the increase of steam flow, the greater the pressure gradient and upward flow velocity gradient in the connection area between the break loop and the downcomer, the division method with fewer nodes is difficult to truly reflect the vapor-liquid overflow relationship in the local area of the downcomer channel; The cooling water injected in the circuit close to the break is more difficult to reach the lower chamber, while the cooling water in the circuit far from the break can easily enter the lower chamber; The superheated steam is cooled by the cooling water during the flow process, resulting in condensation, resulting in the steam flow at the outlet being less than the inlet steam flow, and the condensation effect decreases with the increase of the inlet steam flow. The model and method established in this study can be applied to the prediction of vapor-liquid CCFL in the downcomer channel during discharge phase of LOCA.

     

  • loading
  • [1]
    AL ISSA S, MACIAN R. A review of CCFL phenomenon[J]. Annals of Nuclear Energy, 2011, 38(9): 1795-1819. doi: 10.1016/j.anucene.2011.04.021
    [2]
    AL ISSA S, MACIAN-JUAN R. Experimental investigation and CFD validation of countercurrent flow limitation (CCFL) in a large-diameter PWR hot-leg geometry[J]. Journal of Nuclear Science and Technology, 2016, 53(5): 647-655. doi: 10.1080/00223131.2015.1125312
    [3]
    GLAESER H, ROHATGI U S. Scaling ability of the counter-current flow limitation (CCFL) correlations for application to reactor thermal hydraulics[J]. Nuclear Engineering and Design, 2019, 354: 110226. doi: 10.1016/j.nucengdes.2019.110226
    [4]
    CHEN C Y, SHIH C, WANG J R, et al. Sensitivity study on the counter-current flow limitation in the DEG LBLOCA with the TRACE code[J]. Annals of Nuclear Energy, 2013, 57: 121-129. doi: 10.1016/j.anucene.2013.01.025
    [5]
    NIKITIN K, MUELLER P, MARTIN J, et al. BWR loss of coolant accident simulation by means of RELAP5[J]. Nuclear Engineering and Design, 2016, 309: 113-121. doi: 10.1016/j.nucengdes.2016.09.008
    [6]
    TAKEDA T, OHTSU I. RELAP5 uncertainty evaluation using ROSA/LSTF test data on PWR 17% cold leg intermediate-break LOCA with single-failure ECCS[J]. Annals of Nuclear Energy, 2017, 109: 9-21. doi: 10.1016/j.anucene.2017.05.007
    [7]
    江灼威,蔡杰进. LOCA事故时一回路冷却剂管肘部回流流动极限研究[J]. 核科学与工程,2019, 39(3): 414-422. doi: 10.3969/j.issn.0258-0918.2019.03.011
    [8]
    SIDDIQUI H, BANERJEE S, ARDRON K H. Flooding in an elbow between a vertical and a horizontal or near-horizontal pipe: Part I: Experiments[J]. International Journal of Multiphase Flow, 1986, 12(4): 531-541. doi: 10.1016/0301-9322(86)90058-3
    [9]
    DAMERELL P S, SIMONS J W. 2D/3D program work summary report: NUREG/IA-0126[R]. Washington, DC: Nuclear Regulatory Commission, 1993
    [10]
    DAMERELL P S, SIMONS J W. Reactor safety issues resolved by the 2D/3D Program: NUREG/IA-0127[R]. Washington, DC: Nuclear Regulatory Commission, 1993.
    [11]
    U. S. NRC. RELAP5/MOD3.3 code manual Vol. 1: code structure, system models, and solution methods. Nuclear safety analysis division Office of Nuclear Regulatory Research U. S. Nuclear Regulatory Commission: RELAP5/MOD3.3 Code Manual: NUREG/CR-5535[R]. Rockville: Information Systems Laboratories, Inc. , 2001
    [12]
    WANG M J, ZHAO H, ZHANG H P, et al. Research on the designed emergency passive residual heat removal system during the station blackout scenario for CPR1000[J]. Annals of Nuclear Energy, 2012, 45: 86-93. doi: 10.1016/j.anucene.2012.03.004
    [13]
    WALLIS G B. One-dimensional two-phase flow[M]. New York: McGraw-Hill, 1969.
  • 加载中

Catalog

    通讯作者: 陈斌, bchen63@163.com
    • 1. 

      沈阳化工大学材料科学与工程学院 沈阳 110142

    1. 本站搜索
    2. 百度学术搜索
    3. 万方数据库搜索
    4. CNKI搜索

    Figures(8)  / Tables(1)

    Article Metrics

    Article views (216) PDF downloads(37) Cited by()
    Proportional views
    Related

    /

    DownLoad:  Full-Size Img  PowerPoint
    Return
    Return