Advance Search
Volume 44 Issue 5
Oct.  2023
Turn off MathJax
Article Contents
Zhang Yunshan, Zhou Xiafeng. Study on JFNK Global Solution Method of of Full-core Thermal Sub-channel Model[J]. Nuclear Power Engineering, 2023, 44(5): 39-46. doi: 10.13832/j.jnpe.2023.05.0039
Citation: Zhang Yunshan, Zhou Xiafeng. Study on JFNK Global Solution Method of of Full-core Thermal Sub-channel Model[J]. Nuclear Power Engineering, 2023, 44(5): 39-46. doi: 10.13832/j.jnpe.2023.05.0039

Study on JFNK Global Solution Method of of Full-core Thermal Sub-channel Model

doi: 10.13832/j.jnpe.2023.05.0039
  • Received Date: 2022-11-14
  • Rev Recd Date: 2022-12-23
  • Publish Date: 2023-10-13
  • The thermal sub-channel model of reactor takes into account many coupling factors such as axial flow, lateral mixing and turbulent mixing in detail, and it is the key model of thermal hydraulic analysis of the core. However, these factors bring difficulties and challenges to the numerical simulation of sub-channel. In order to improve the computational efficiency and convergence of the thermal sub-channel model, this paper develops a global solution framework for the thermal sub-channel model of the whole core based on the Jacobian-Free Newton-Krylov (JFNK) global solution method (hereinafter referred to as JFNK method), and establishes a residual system based on physical pretreatment according to the existing code model and framework to enhance the convergence rate of JFNK method. Numerical solutions show that the speedup of JFNK methods is about five compared with the fixed-point iterative method used in the original sub-channel models and there are higher efficiency for JFNK methods as convergence precision improves, which indicates the potential and efficiency advantages of JFNK methods for complicated thermal sub-channel model.

     

  • loading
  • [1]
    ROWE D S. COBRA-III: a digital computer program for steady state and transient thermal--hydraulic analysis of rod bundle nuclear fuel elements: BNWL-B-82[R]. Richland: Battelle Pacific Northwest Labs. , 1971.
    [2]
    ROWE D S. COBRA IIIC: digital computer program for steady state and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements: BNWL-1695[R]. Richland: Battelle Pacific Northwest Labs. , 1973.
    [3]
    STEWART C W, WHEELER C L, CENA R J, et al. COBRA-IV: the model and the method: BNWL-2214[R]. Richland: Pacific Northwest Laboratories, 1977.
    [4]
    BASILE D, BEHGI M, CHIERICI R, et al. COBRA-EN: Code System for Thermal-Hydraulic Transient Analysis of Light Water Reactor Fuel Assemblies and Cores: RSICC Code Package PSR-507[R]. U.S.: Oak-Ridge, 2001.
    [5]
    PLAS R. FLICA III M-reactors or test loops thermohydraulic computer code: CEA-N--2418[R]. Gif-sur-Yvette: CEA Centre d'Etudes Nucleaires de Saclay, 1984.
    [6]
    DENG J, LU Q, LIU Y, et al. Review of sub-channel code development for pressurized water reactor and introduction of CORTH-V2.0 sub-channel code[J]. Progress in Nuclear Energy, 2020, 125: 103373. doi: 10.1016/j.pnucene.2020.103373
    [7]
    HU J W, SALKO JR R K, WYSOCKI A J. CTF 4.0 theory manual[R]. Oak Ridge: Oak Ridge National Lab. , 2019.
    [8]
    TOUMI I, BERGERON A, GALLO D, et al. FLICA-4: a three-dimensional two-phase flow computer code with advanced numerical methods for nuclear applications[J]. Nuclear Engineering and Design, 2000, 200(1-2): 139-155. doi: 10.1016/S0029-5493(99)00332-5
    [9]
    ESMAILI H, KAZEMINEJAD H, KHALAFI H, et al. Subchannel analysis of annular fuel assembly using the preconditioned Jacobian-free Newton Krylov methods[J]. Annals of Nuclear Energy, 2020, 146: 107616. doi: 10.1016/j.anucene.2020.107616
    [10]
    PORTER N W. Development of a novel residual formulation of CTF and application of parameter estimation techniques[D]. Raleigh: North Carolina State University, 2018.
    [11]
    PORTER N W, MOUSSEAU V A, AVRAMOVA M N. CTF-R: a novel residual-based thermal hydraulic solver[J]. Nuclear Engineering and Design, 2019, 348: 37-45. doi: 10.1016/j.nucengdes.2019.04.006
    [12]
    PLAS R. FLICA III M-reactors or test loops thermohydraulic computer code[R]. Gif-sur-Yvette: CEA Centre d'Etudes Nucleaires de Saclay, 1984.
    [13]
    MOORTHI A, SHARMA A K, VELUSAMY K. A review of sub-channel thermal hydraulic codes for nuclear reactor core and future directions[J]. Nuclear Engineering and Design, 2018, 332: 329-344. doi: 10.1016/j.nucengdes.2018.03.012
    [14]
    KNOLL D A, KEYES D E. Jacobian-free Newton–Krylov methods: a survey of approaches and applications[J]. Journal of Computational Physics, 2004, 193(2): 357-397. doi: 10.1016/j.jcp.2003.08.010
    [15]
    CHAN T F, VAN DER VORST H A. Approximate and incomplete factorizations[M]//KEYES D E, SAMEH A, VENKATAKRISHNAN V. Parallel Numerical Algorithms. Dordrecht: Springer, 1997: 167-202.
    [16]
    ZHOU X F, ZHONG C M, ZHANG Y Y. Jacobian-free Newton Krylov two-node coarse mesh finite difference based on nodal expansion method for multiphysics coupled models[J]. Annals of Nuclear Energy, 2022, 168: 108915. doi: 10.1016/j.anucene.2021.108915
    [17]
    EISENSTAT S C, WALKER H F. Choosing the forcing terms in an inexact Newton method[J]. SIAM Journal on Scientific Computing, 1996, 17(1): 16-32. doi: 10.1137/0917003
    [18]
    PERNICE M, WALKER H F. NITSOL: a Newton iterative solver for nonlinear systems[J]. SIAM Journal on Scientific Computing, 1998, 19(1): 302-318. doi: 10.1137/S1064827596303843
    [19]
    广东核电培训中心. 900MW压水堆核电站系统与设备[M]. 北京: 原子能出版社, 2005: 611.
    [20]
    梁志滔. 压水堆核电站堆芯子通道分析[D]. 广州: 华南理工大学, 2011.
  • 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(11)  / Tables(3)

    Article Metrics

    Article views (1483) PDF downloads(366) Cited by()
    Proportional views
    Related

    /

    DownLoad:  Full-Size Img  PowerPoint
    Return
    Return