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绕丝交混模型对钠冷快堆组件子通道分析的影响

方闻韬 佟立丽 曹学武

方闻韬, 佟立丽, 曹学武. 绕丝交混模型对钠冷快堆组件子通道分析的影响[J]. 强激光与粒子束, 2023, 35: 096001. doi: 10.11884/HPLPB202335.230051
引用本文: 方闻韬, 佟立丽, 曹学武. 绕丝交混模型对钠冷快堆组件子通道分析的影响[J]. 强激光与粒子束, 2023, 35: 096001. doi: 10.11884/HPLPB202335.230051
Fang Wentao, Tong Lili, Cao Xuewu. Influence of wire wrap mixing model on sub-channel analysis of sodium-cooled fast reactor assembly[J]. High Power Laser and Particle Beams, 2023, 35: 096001. doi: 10.11884/HPLPB202335.230051
Citation: Fang Wentao, Tong Lili, Cao Xuewu. Influence of wire wrap mixing model on sub-channel analysis of sodium-cooled fast reactor assembly[J]. High Power Laser and Particle Beams, 2023, 35: 096001. doi: 10.11884/HPLPB202335.230051

绕丝交混模型对钠冷快堆组件子通道分析的影响

doi: 10.11884/HPLPB202335.230051
基金项目: 国家自然科学基金项目(U1967202)
详细信息
    作者简介:

    方闻韬,120020910365@sjtu.edu.cn

    通讯作者:

    佟立丽,lltong@sjtu.edu.cn

  • 中图分类号: TL364

Influence of wire wrap mixing model on sub-channel analysis of sodium-cooled fast reactor assembly

  • 摘要: 钠冷快堆燃料棒表面缠绕的绕丝能够强化通道间的冷却剂横向流动,降低组件盒内温度分布的不均匀性,提升反应堆安全性。现有的子通道程序通过采用不同类型的绕丝交混模型,模拟了绕丝对组件盒内各类参数计算结果的影响。为了研究不同绕丝交混模型对钠冷快堆组件盒内流动与传热模拟的影响,基于Mikityuk对流传热模型以及Cheng-Todreas流动压降模型,分别采用强迫横流模型以及带绕丝湍流交混模型建立了子通道分析方法,并与美国ORNL开展的FFM-2A实验数据以及其他子通道程序针对该实验的分析结果进行了对比验证。结果表明在低流量条件下两种模型均能较好模拟带绕丝组件的流动与传热情况;在高流量条件下使用强迫横流模型分析结果与实验符合较好,使用带绕丝湍流交混模型的分析结果高估了靠近中心通道的出口冷却剂温度。
  • 图  1  燃料棒与绕丝相对位置对横向流动的影响

    Figure  1.  Influence of relative positions of fuel rods and wrapping wires on lateral flow

    图  2  子通道与燃料棒编号

    Figure  2.  Numbering of sub-channels and rods

    图  3  通道间隙编号

    Figure  3.  Numbering of gaps

    图  4  绕丝交混模型对出口温度预测的影响

    Figure  4.  Influence of models on outlet temperature prediction

    图  5  通道间隙横向流量对比

    Figure  5.  Comparison of cross flow rate of channel gap

    图  6  模拟结果与其他子通道程序对比

    Figure  6.  Comparison of simulation results with other sub-channel codes

  • [1] Ninokata H, Efthimiadis A, Todreas N E. Distributed resistance modeling of wire-wrapped rod bundles[J]. Nuclear Engineering and Design, 1987, 104(1): 93-102. doi: 10.1016/0029-5493(87)90306-2
    [2] Jeong H Y, Ha K S, Chang W P, et al. Modeling of flow blockage in a liquid metal-cooled reactor subassembly with a subchannel analysis code[J]. Nuclear Technology, 2005, 149(1): 71-87. doi: 10.13182/NT05-A3580
    [3] 陈选相, 吴攀, 单建强. 钠冷快堆分析程序ATHAS-LMR的子通道模型[J]. 原子能科学技术, 2012, 46(6):695-700

    Chen Xuanxiang, Wu Pan, Shan Jianqiang. Subchannel model of analysis code ATHAS-LMR for LMFBR[J]. Atomic Energy Science and Technology, 2012, 46(6): 695-700
    [4] Lodi F, Grasso G, Mattioli D, et al. ANTEO+: A subchannel code for thermal-hydraulic analysis of liquid metal cooled systems[J]. Nuclear Engineering and Design, 2016, 301: 128-152. doi: 10.1016/j.nucengdes.2016.03.001
    [5] 吴宗芸, 刘天才, 吴明宇. 基于双区域模型的钠冷快堆组件子通道分析程序的开发与验证[J]. 原子能科学技术, 2022, 56(4):672-683

    Wu Zongyun, Liu Tiancai, Wu Mingyu. Development and validation of subchannel analysis program based on two-region model for sodium cooled fast reactor assembly[J]. Atomic Energy Science and Technology, 2022, 56(4): 672-683
    [6] Wantland J L. ORRIBLE–A computer program for flow and temperature distribution in 19-ROD LMFBR fuel subassemblies[J]. Nuclear Technology, 1974, 24(2): 168-175. doi: 10.13182/NT74-A31473
    [7] Rowe D S. COBRA IIIC: digital computer program for steady state and transient thermal-hydraulic analysis of rod bundle nuclear fuel elements[R]. Richland, WA, USA: Battelle Pacific Northwest Labs. , 1973.
    [8] Sun R L, Zhang D L, Liang Y, et al. Development of a subchannel analysis code for SFR wire-wrapped fuel assemblies[J]. Progress in Nuclear Energy, 2018, 104: 327-341. doi: 10.1016/j.pnucene.2017.12.005
    [9] Liu X J, Scarpelli N. Development of a sub-channel code for liquid metal cooled fuel assembly[J]. Annals of Nuclear Energy, 2015, 77: 425-435. doi: 10.1016/j.anucene.2014.10.030
    [10] Mikityuk K. Heat transfer to liquid metal: review of data and correlations for tube bundles[J]. Nuclear Engineering and Design, 2009, 239(4): 680-687. doi: 10.1016/j.nucengdes.2008.12.014
    [11] Cheng S K, Todreas N E. Hydrodynamic models and correlations for bare and wire-wrapped hexagonal rod bundles—bundle friction factors, subchannel friction factors and mixing parameters[J]. Nuclear Engineering and Design, 1986, 92(2): 227-251. doi: 10.1016/0029-5493(86)90249-9
    [12] Castellana F S, Adams W T, Casterline J E. Single-phase subchannel mixing in a simulated nuclear fuel assembly[J]. Nuclear Engineering and Design, 1974, 26(2): 242-249. doi: 10.1016/0029-5493(74)90059-4
    [13] Pramuditya S, Takahashi M. Thermal–hydraulic analysis of wire-wrapped SFR test subassemblies by subchannel analysis method[J]. Annals of Nuclear Energy, 2013, 54: 109-119. doi: 10.1016/j.anucene.2012.11.011
    [14] Bogoslovskaya G P, Zhukov A V, Sorokin A P. Models and characteristics of interchannel exchange in pin bundles cooled by liquid metal[R]. Vienna, Austria: International Atomic Energy Agency, 2000.
    [15] Zheng S G. Constitutive correlations for wire-wrapped subchannel analysis under forced and mixed convection conditions[D]. Cambridge: Massachusetts Institute of Technology, 1984.
    [16] Fontana M H, MacPherson R, Gnadt P, et al. Temperature distribution in a 19-rod simulated LMFBR fuel assembly in a hexagonal duct (fuel failure mockup bundle 2A): Record of experimental data[J]. ORNL-TM-4113, 1973.
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出版历程
  • 收稿日期:  2023-03-10
  • 修回日期:  2023-07-01
  • 录用日期:  2023-06-12
  • 网络出版日期:  2023-07-07
  • 刊出日期:  2023-09-15

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