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两流体双压力热工水力系统分析软件LOCUST 2.0架构设计

徐财红 袁红胜 柳焕楠 琚忠云 李昌莹 王婷 厉井钢

徐财红, 袁红胜, 柳焕楠, 琚忠云, 李昌莹, 王婷, 厉井钢. 两流体双压力热工水力系统分析软件LOCUST 2.0架构设计[J]. 核动力工程, 2023, 44(6): 86-94. doi: 10.13832/j.jnpe.2023.06.0086
引用本文: 徐财红, 袁红胜, 柳焕楠, 琚忠云, 李昌莹, 王婷, 厉井钢. 两流体双压力热工水力系统分析软件LOCUST 2.0架构设计[J]. 核动力工程, 2023, 44(6): 86-94. doi: 10.13832/j.jnpe.2023.06.0086
Xu Caihong, Yuan Hongsheng, Liu Huannan, Ju Zhongyun, Li Changying, Wang Ting, Li Jinggang. Architecture Design of Two-Fluid Two-Pressure Thermal-Hydraulic System Analysis Code LOCUST 2.0[J]. Nuclear Power Engineering, 2023, 44(6): 86-94. doi: 10.13832/j.jnpe.2023.06.0086
Citation: Xu Caihong, Yuan Hongsheng, Liu Huannan, Ju Zhongyun, Li Changying, Wang Ting, Li Jinggang. Architecture Design of Two-Fluid Two-Pressure Thermal-Hydraulic System Analysis Code LOCUST 2.0[J]. Nuclear Power Engineering, 2023, 44(6): 86-94. doi: 10.13832/j.jnpe.2023.06.0086

两流体双压力热工水力系统分析软件LOCUST 2.0架构设计

doi: 10.13832/j.jnpe.2023.06.0086
详细信息
    作者简介:

    徐财红(1988—),男,硕士研究生,现主要从事热工水力分析软件开发方面的研究,E-mail: xch_sjtu@126.com

  • 中图分类号: TL334

Architecture Design of Two-Fluid Two-Pressure Thermal-Hydraulic System Analysis Code LOCUST 2.0

  • 摘要: 中国广核集团(CGN)研发了一款新的热工水力系统分析软件LOCUST 2.0,它采用两流体两场双压力七方程模型,从理论上保证了守恒方程的严格适定性。LOCUST 2.0的应用对象为华龙一号失水事故类的安全分析,目前已初步完成代码开发。本文简要描述了该软件的守恒方程、数值算法、软件功能、源代码架构设计;给出了6个典型案例的分析计算,结果表明该软件能够合理预测热工水力进程,性能良好。

     

  • 图  1  节点图举例

    Figure  1.  Example of Node Graph

    图  2  LOCUST 2.0源代码框架

    Figure  2.  Source Code Frame of LOCUST 2.0

    图  3  LOCUST 2.0主要计算流程图

    ACC—蓄压安注箱

    Figure  3.  Calculation Flow Chart of LOCUST 2.0

    图  4  水自由落体节点图与计算结果

    Figure  4.  Nodalization Diagram and Calculation Results for Water Free Fall Case

    图  5  ANL竖直管试验空泡份额测量值与计算值的对比

    jf—液相表观速度

    Figure  5.  Comparison of Measured and Predicted Void Fractions for ANL Vertical Pipe Test

    图  6  FRIGG-2棒束试验轴向空泡份额分布

    Figure  6.  Axial Distribution of Void Fraction for Rod Bundle Test FRIGG-2

    图  7  Marviken临界流试验工况22喷放流量

    Figure  7.  Blowdown Mass Flow Rate of Marvike Critical Flow Test 22

    图  8  棒束试验THTF轴向棒壁面温度

    Figure  8.  Wall Temperature of Rod Bundle Test THTF

    图  9  FLECHT-SEASET电加热棒壁面温度

    Figure  9.  Heated Rod Wall Temperature of FLECHT-SEASET Test Cases

    表  1  不同系统程序结构的对比

    Table  1.   Structure Comparison of Different System Codes

    序号 程序结构 优点 缺点 典型代表
    1 两场五方程模型
    (一个混合物动量方程+滑速比模型)
    无条件适定;计算量相对更少 两相动力学描述偏于简单,对于
    非泡状流型的模拟能力不足
    RETRAN-02
    2 两场六方程模型 两相动力学描述较精细 存在不适定区域(可通过虚拟质量力等进行改善) RELAP5、
    TRACE
    3 两场七方程模型 无条件适定;两相动力学描述较精细 计算量相对六方程模型更大;成熟性尚有限 RELAP7、
    LOCUST 2.0
    下载: 导出CSV
  • [1] RETTIG W H, JAYNE G A, MOORE K V, et al. RELAP3: a computer program for reactor blowdown analysis: IN-1321[R]. Idaho Falls: Idaho Nuclear Corp. , 1970.
    [2] GRS. ATLET models and methods: GRS-P-1/Vol. 4, Rev. 5[Z]. 2019.
    [3] U.S. Nuclear Regulatory Commission. RELAP5/MOD3 code manual volume 1: code structure, system models, and solution methods: NUREG/CR-5535[R]. Washington: U.S. Nuclear Regulatory Commission, 1995.
    [4] Nuclear Regulatory Commission. TRACE V5.0 Theory manual: field equations, solution methods, and physical models: ML120060218[R]. Washington: U. S. Nuclear Regulatory Commission, 2008.
    [5] BESTION D. The physical closure laws in the CATHARE code[J]. Nuclear Engineering and Design, 1990, 124(3): 229-245. doi: 10.1016/0029-5493(90)90294-8
    [6] BERRY R A, PETERSON J W, ZHANG H, et al. RELAP-7 theory manual: INL/EXT-14-31366-Rev003[R]. Idaho Falls: Idaho National Lab, 2018.
    [7] EMONOT P, SOUYRI A, GANDRILLE J L, et al. CATHARE-3: a new system code for thermal-hydraulics in the context of the NEPTUNE project[J]. Nuclear Engineering and Design, 2011, 241(11): 4476-4481.
    [8] HA S J, PARK C E, KIM K D, et al. Development of the SPACE code for nuclear power plants[J]. Nuclear Engineering and Technology, 2011, 43(1): 45-62. doi: 10.5516/NET.2011.43.1.045
    [9] 徐财红. 两相流热工水力系统分析软件LOCUST-1.2开发概述[C]. 重庆: 中国核学会核反应堆热工流体力学分会第一届学术年会, 2021.
    [10] 单建强, 廖承奎, 苟军利, 等. 压水堆核电厂瞬态安全数值分析方法[M]. 西安: 西安交通大学出版社, 2016: 140-156.
    [11] BERRY R A, SAUREL R, LEMETAYER O, et al. The discrete equation method (DEM) for fully compressible, two-phase flows in ducts of spatially varying cross-section[J]. Nuclear Engineering and Design, 2010, 240(11): 3797-3818. doi: 10.1016/j.nucengdes.2010.08.003
    [12] CARISON K E, RIEMKE R A, ROUHANL S Z, et al. RELAP5/MOD3 Code manual, volume 3: developmental assessment problems (DRAFT): NUREG/CR-5535[R]. Washington: U. S. Nuclear Regulatory Commission, 1990.
    [13] U. S. Nuclear Regulatory Commission. TRACE V5.0 assessment manual: appendix a: fundamental validation cases: ML120060187[R]. Washington: U. S. Nuclear Regulatory Commission, 2008.
    [14] KATAOKA I, ISHII M. Drift flux model for large diameter pipe and new correlation for pool void fraction[J]. International Journal of Heat and Mass Transfer, 1987, 30(9): 1927-1939. doi: 10.1016/0017-9310(87)90251-1
    [15] HIBIKI T, ISHII M. One-dimensional drift–flux model for two-phase flow in a large diameter pipe[J]. International Journal of Heat and Mass Transfer, 2003, 46(10): 1773-1790. doi: 10.1016/S0017-9310(02)00473-8
    [16] NYLUND O, BECKER K M, EKLUND R, et al. Hydrodynamic and heat transfer measurements on a full-scale simulated 36-rod marviken fuel element with uniform heat flux distribution: FRIGG-2[R]. Stockholm: Aktiebolaget Atomenergi, 1968.
    [17] HENRY R E, FAUSKE H K. The two-phase critical flow of one-component mixtures in nozzles, orifices, and short tubes[J]. Journal of Heat Transfer, 1971, 93(2): 179-187. doi: 10.1115/1.3449782
    [18] MOODY F J. Maximum flow rate of a single component, two-phase mixture[J]. Journal of Heat Transfer, 1965, 87(1): 134-141. doi: 10.1115/1.3689029
    [19] U.S. Nuclear Regulatory Commission. The Marviken full scale critical flow tests: summary report: NUREG/CR-2671[R]. Washington: U.S. Nuclear Regulatory Commission, 1982.
    [20] YODER G L, MORRIS D G, MULLINS C B, et al. Dispersed-flow film boiling in rod-bundle geometry: steady-state heat-transfer data and correlation comparisons: NUREG/CR-2435[R]. Oak Ridge: Oak Ridge National Lab. , 1982.
    [21] SHAH M M. Unified correlation for heat transfer during boiling in plain mini/micro and conventional channels[J]. International Journal of Refrigeration, 2017, 74: 606-626. doi: 10.1016/j.ijrefrig.2016.11.023
    [22] GROENEVELD D C, SHAN J Q, VASIĆ A Z, et al. The 2006 CHF look-up table[J]. Nuclear Engineering and Design, 2007, 237(15-17): 1909-1922. doi: 10.1016/j.nucengdes.2007.02.014
    [23] BJORNARD T A, GRIFFITH P. PWR blowdown heat transfer[C]. U.S.: Proceedings of the Winter Annual Meeting of the American Society of Mechanical Engineers. New York: American Society of Mechanical Engineers, 1977.
    [24] GROENEVELD D C, LEUNG L K H, VASIC’ A Z, et al. A look-up table for fully developed film-boiling heat transfer[J]. Nuclear Engineering and Design, 2003, 225(1): 83-97. doi: 10.1016/S0029-5493(03)00149-3
    [25] LOFTUS M J, HOCHREITER L E, CONWAY C E, et al. PWR FLECHT SEASET unblocked bundle, forced and gravity reflood task data report. Volume 1: NUREG/CR-1532[R]. Pittsburgh: Westinghouse Electric Corp. , 1981.
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出版历程
  • 收稿日期:  2022-12-20
  • 修回日期:  2023-06-26
  • 网络出版日期:  2023-12-11
  • 刊出日期:  2023-12-15

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