Development of subchannel thermal-hydraulic analysis code for dual cooled annular fuel

被引:0
|
作者
Saffari, A. H. [1 ]
Esmaili, H. [2 ]
机构
[1] Sharif Univ Technol, Dept Energy Engn, Tehran, Iran
[2] Nucl Sci & Technol Res Inst NSTRI, Reactor & Nucl Safety Res Sch, Tehran, Iran
关键词
Annular fuel assemblies; Subchannel analysis; Local thermal-hydraulic parameters;
D O I
暂无
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
Regarding the geometric structural characteristics of innovative dual cooled annular fuel and the possibility of heat split and flow distribution among the internal and external channels, the development of new computational tools is essential for estimating safety margins and accurate assessment of its thermal-hydraulic performance. The SADAF code (Subchannel Analysis Dual cooled Annular Fuel) by COBRA-EN code is developed for this purpose. In the SADAF code, using COBRA-EN code for subchannel analysis in internal and external subchannels, a program has been developed to compute new variables that need to be considered in the thermal-hydraulic assessment. Also, fuel heat transfer calculations were performed by the finite difference method. Benchmark calculations have been performed for Westinghouse 13 x 13 annular fuel rod design, the results were compared with the reference values, and a good agreement was found. Evaluation of the results shows that the total pressure drop predicted by the SADAF code differs from the benchmark results by less than 1%. This difference is about 0.4%, 1.0 K, and 1.0% for fluid temperatures, rod surface temperatures, and flow rates, respectively. It is also shown that these error levels are within acceptable margins. The results analysis exhibited the robustness of the SADAF code in sub-channel analysis of annular fuel assembly.
引用
收藏
页数:12
相关论文
共 50 条
  • [1] Development of subchannel thermal-hydraulic analysis code for dual cooled annular fuel
    Saffari, A. H.
    Esmaili, H.
    PROGRESS IN NUCLEAR ENERGY, 2022, 150
  • [2] Development of a thermal-hydraulic analysis code for annular fuel assemblies
    Han, KH
    Chang, SH
    NUCLEAR ENGINEERING AND DESIGN, 2003, 226 (03) : 267 - 275
  • [3] Development of a thermal-hydraulic analysis code for annular fuel assemblies
    Vishnoi, A. K.
    Chandraker, D. K.
    Vijayan, P. K.
    KERNTECHNIK, 2012, 77 (01) : 12 - 17
  • [4] Development of a thermal-hydraulic subchannel analysis code for motion conditions
    Cai, Rong
    Yue, Nina
    Chen, Ronghua
    Tian, W. X.
    Su, G. H.
    Qiu, S. Z.
    PROGRESS IN NUCLEAR ENERGY, 2016, 93 : 165 - 176
  • [5] ANTEO plus : A subchannel code for thermal-hydraulic analysis of liquid metal cooled systems
    Lodi, F.
    Grasso, G.
    Mattioli, D.
    Sumini, M.
    NUCLEAR ENGINEERING AND DESIGN, 2016, 301 : 128 - 152
  • [6] Subchannel thermal-hydraulic analysis of the fuel assembly for liquid sodium cooled fast reactor
    Wu, Y. W.
    Li, Xin
    Yu, Xiaolei
    Qiu, S. Z.
    Su, G. H.
    Tian, W. X.
    PROGRESS IN NUCLEAR ENERGY, 2013, 68 : 65 - 78
  • [7] Development of Thermal-hydraulic Analysis Code for Spent Fuel Assembly of Sodium Cooled Fast Reactor
    Ma X.
    Lin C.
    Li S.
    Zhou Z.
    Feng Y.
    Zhang D.
    Yuanzineng Kexue Jishu/Atomic Energy Science and Technology, 2020, 54 (04): : 606 - 614
  • [8] Development and validation of a thermal-hydraulic analysis code for dual cooled assemblies in molten salt reactors
    Hu, Siqin
    Zhou, Chong
    Zhu, Guifeng
    Zou, Yang
    Yu, Xiaohan
    Xue, Shuaiyu
    He Jishu/Nuclear Techniques, 2024, 47 (09):
  • [9] DEVELOPMENT OF THERMAL-HYDRAULIC AND SAFETY ANALYSIS CODE FOR A HEAT PIPE COOLED REACTOR
    Jiao, Guanghui
    Xia, Genglei
    Zhou, Tao
    Wang, Jianjun
    PROCEEDINGS OF 2024 31ST INTERNATIONAL CONFERENCE ON NUCLEAR ENGINEERING, VOL 4, ICONE31 2024, 2024,
  • [10] SACOS-PLATE: A new thermal-hydraulic subchannel analysis code for plate type fuel assemblies
    Sun, Ruiyu
    Gui, Minyang
    Wang, Jinshun
    Chen, Ronghua
    Zhang, Kui
    Tian, Wenxi
    Qiu, Suizheng
    Su, G. H.
    ANNALS OF NUCLEAR ENERGY, 2024, 204