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欧洲铅冷快堆水平螺旋管式蒸汽发生器热工水力性能分析

张巍 李净松 施慧烈 乔鹏瑞 王聪 张天清 何莹钊

张巍, 李净松, 施慧烈, 乔鹏瑞, 王聪, 张天清, 何莹钊. 欧洲铅冷快堆水平螺旋管式蒸汽发生器热工水力性能分析[J]. 核动力工程, 2022, 43(3): 38-45. doi: 10.13832/j.jnpe.2022.03.0038
引用本文: 张巍, 李净松, 施慧烈, 乔鹏瑞, 王聪, 张天清, 何莹钊. 欧洲铅冷快堆水平螺旋管式蒸汽发生器热工水力性能分析[J]. 核动力工程, 2022, 43(3): 38-45. doi: 10.13832/j.jnpe.2022.03.0038
Zhang Wei, Li Jingsong, Shi Huilie, Qiao Pengrui, Wang Cong, Zhang Tianqing, He Yingzhao. Thermal-Hydraulic Performance Analysis of Horizontal Spiral Tube Steam Generator for European Lead-Cooled Fast Reactor[J]. Nuclear Power Engineering, 2022, 43(3): 38-45. doi: 10.13832/j.jnpe.2022.03.0038
Citation: Zhang Wei, Li Jingsong, Shi Huilie, Qiao Pengrui, Wang Cong, Zhang Tianqing, He Yingzhao. Thermal-Hydraulic Performance Analysis of Horizontal Spiral Tube Steam Generator for European Lead-Cooled Fast Reactor[J]. Nuclear Power Engineering, 2022, 43(3): 38-45. doi: 10.13832/j.jnpe.2022.03.0038

欧洲铅冷快堆水平螺旋管式蒸汽发生器热工水力性能分析

doi: 10.13832/j.jnpe.2022.03.0038
基金项目: 中核集团青年英才项目(CNPO-K200556)
详细信息
    作者简介:

    张 巍(1985—),男,高级工程师,现主要从事蒸汽发生器热工水力研究,E-mail: zhangwei15@cnnp.com.cn

  • 中图分类号: TL425

Thermal-Hydraulic Performance Analysis of Horizontal Spiral Tube Steam Generator for European Lead-Cooled Fast Reactor

  • 摘要: 以欧洲铅冷堆(ELSY)水平螺旋管式蒸汽发生器(HST-SG)为研究对象,结合其结构参数和运行参数,选取了合适的传热阻力模型开发了一维稳态热工水力计算程序,采用该程序首先对ELSY HST-SG进行校核计算,以验证程序计算的准确性,再结合计算结果,对ELSY HST-SG热工水力性能进行详细分析,并针对不同运行参数开展对比分析研究。分析结果表明,ELSY HST-SG各项参数选择合理,热工水力性能优良,结构紧凑。因此,该程序可用于ELSY HST-SG的设计开发和性能分析。

     

  • 图  1  SG-PP一体化设计示意图[8]

    Figure  1.  Schematic Diagram of Integrated Design of Main Pump and SG

    图  2  传热管微元传热示意图

    T1,jT2,j分别为节点j处一、二次侧流体温度;T1,j+1T2,j+1分别为节点j+1处一、二次侧流体温度

    Figure  2.  Schematic Diagram of Heat Transfer in Tube Microelement

    图  3  水平螺旋管示意图

    R为质点轨迹半径;θ为微元所在位置弧度角;dl为微元传热管管长,dl=[(Rdθ)2+(dR)2]0.5

    Figure  3.  Schematic Diagram of Horizontal Spiral Tube

    图  4  稳态计算流程图

    Figure  4.  Flowchart of Steady State Calculation

    图  5  节点数敏感性分析

    Figure  5.  Sensitivity Analysis of the Number of Nodes

    图  6  ELSY HST-SG温度沿管长分布

    Figure  6.  ELSY HST-SG Temperature Distribution along the Tube Length.

    图  7  管束各环节传热系数沿管长分布

    Figure  7.  Distribution of Heat Transfer Coefficient of Each Link of Tube Bundle along Tube Length

    图  8  传热管热流密度与流体传热温差沿管长分布

    Figure  8.  Distribution of Heat Flux Density and Fluid Heat Transfer Temperature Difference along Heat Transfer Tube Length

    图  9  一、二次侧介质流速沿管长分布

    Figure  9.  Velocity Distribution of the Media on the Primary and Secondary Sides along Tube Length

    图  10  管内沿程压力/压降分布图

    Figure  10.  Pressure/Pressure Drop Distribution along the Tube

    图  11  不同蒸汽压力下温度分布曲线

    Figure  11.  Temperature Distribution Under Different Steam Pressures

    图  12  不同给水温度情况下二次侧流体温度分布

    Figure  12.  Temperature Distribution of Secondary Side Fluid under Different Feed Water Temperatures

    图  13  不同堵管率时的功率和管内阻力对比

    Figure  13.  Comparison of Power and in-tube Resistance at Different Tube Blocking Rates

    表  1  ELSY 堆主热传输系统参数[8]

    Table  1.   Main Parameters of ELSY Heat Transfer System[8]

    参数参数值参数参数值
    反应堆
    电功率/ MW
    600SG模块数量8
    单台SG
    热功率/MW
    175蒸汽压力/MPa18
    铅进口温度/℃480.0给水温度/℃335
    铅出口温度/℃400.0蒸汽出口温度/℃450
    下载: 导出CSV

    表  2  SG主要结构参数[8]

    Table  2.   Main Structural Parameterrs of SG

    参数参数值参数参数值
    SG内筒外径/mm1240传热管长度/m55
    SG外筒内径/mm2420每层的传热管
    数量
    2
    螺旋管束区高度/mm2620传热管外径、壁厚/mm22.22、2.5
    传热管数量219轴向、径向间距/mm24、24
    下载: 导出CSV

    表  3  传热经验关系式汇总

    Table  3.   Summary of Heat Transfer Relationship

    传热区采用的经验关系式
    单相对流Mori-Nakayama公式[10]
    过冷沸腾Chen公式[11]
    泡核沸腾Chen公式[11]
    蒸干区Manabe公式 [12]
    过热蒸汽Mori-Nakayama公式[10]
    下载: 导出CSV

    表  4  程序校核计算结果

    Table  4.   Program Check and Calculation Results

    参数功率/
    MW
    铅进口
    温度/℃
    铅出口
    温度/℃
    给水温
    度/℃
    蒸汽出口
    温度/℃
    管长/m
    设计
    参数
    175.00 480.0 400.00 335.0 450.0 55.0
    校核
    计算
    179.68 480.0 397.88 335.0 462.5 55.0
    误差 2.67% −0.53% 2.78%
      “—”表示无误差
    下载: 导出CSV

    表  5  不同压力下额定功率设计参数对比

    Table  5.   Comparison of Design Parameters of Rated Power under Different Pressures

    压力/MPa14161820
    给水温度/℃335335335335
    蒸汽出口温度/℃463.3461.5459.8459.1
    功率/ MW186.9183.0179.1174.7
    蒸汽流速/(m·s−1)45.5738.6633.1228.79
    管内阻力/ kPa1392.11169.6965.5768.2
    下载: 导出CSV

    表  6  不同给水温度情况下参数变化

    Table  6.   Parameter Changes under Different Feed Water Temperatures

    给水温度/℃330335345355
    蒸汽压力/ MPa18181818
    蒸汽出口温度/℃458.3459.8463.0466.4
    两相区长度/m20.4220.4920.4420.34
    沿程压降/kPa951.5965.51000.01040.5
    功率/MW182.28179.15172.38164.08
    下载: 导出CSV
  • [1] ALEMBERTI A, LFROGHERI M. Lead-cooled Fast Reactor (LFR) risk and safety assessmentwhite paper[C]//GEN Ⅳ International Forum, 2014: 16-17
    [2] INGERSOLL D T, CARELLI M D. Handbook of small modular nuclear reactors[M]. 2nd ed. Duxford: Woodhead Publishing, 2020: 33-34, 40-41.
    [3] CACUCI D G. Handbook of nuclear engineering[M]. New York: Springer, 2010: 2400-2406.
    [4] GRABEZHNAYA V A, MIKHEEV A S, KALYAKIN S G, et al. Testing the steam generator model with helical coiled lead-heated tubes[J]. Thermal Engineering, 2014, 61(11): 777-784. doi: 10.1134/S0040601514110019
    [5] CINOTTI L, SMITH C F, SEKIMOTO H, et al. Lead-cooled system design and challenges in the frame of generation IV International Forum[J]. Journal of Nuclear Materials, 2011, 415(3): 245-253. doi: 10.1016/j.jnucmat.2011.04.042
    [6] 刘法钰,张小英,陈佳跃,等. 螺旋管直流蒸汽发生器一、二次侧耦合传热特性分析[J]. 核动力工程,2020, 41(5): 24-29.
    [7] NARCISI V, GIANNETTI F, MARTELLI E, et al. Steam generator mock-up preliminary design suitable for Pb-Li technology demonstration and code assessment[J]. Fusion Engineering and Design, 2019, 146: 1126-1130. doi: 10.1016/j.fusengdes.2019.02.022
    [8] ALEMBERTI A, CARLSSON J, MALAMBU E, et al. European lead fast reactor—ELSY[J]. Nuclear Engineering and Design, 2011, 241(9): 3470-3480. doi: 10.1016/j.nucengdes.2011.03.029
    [9] CHERNYSH A, IARMONOV M, MAKHOV K, et al. Experimental study of the characteristics of heat transfer in an HLMC cross-flow around tubes[J]. Journal of Nuclear Engineering and Radiation Science, 2015, 1(4): 041015. doi: 10.1115/1.4030365
    [10] MORI Y, NAKAYAMA W. Study of forced convective heat transfer in curved pipes (2nd report, turbulent region)[J]. International Journal of Heat and Mass Transfer, 1967, 10(1): 37-59. doi: 10.1016/0017-9310(67)90182-2
    [11] CHEN J C. Correlation for boiling heat transfer to saturated fluids in convective flow[J]. Industrial & Engineering Chemistry Process Design and Development, 1966, 5(3): 322-329.
    [12] MANABE F, KOSUGI T, TSUCHIYA T, et al. Experimental studies of heat transfer performance with the No.2 50 MW steam generator in Japan: PNC-TN941 82-27[R]. Ibaraki-ken, Japan: Power Reactor and Nuclear Fuel Development Corporation, 1982.
    [13] RIVAS E, MUÑOZ-ANTÓN J. Dryout study of a helical coil once-through steam generator integrated in a thermal storage prototype[J]. Applied Thermal Engineering, 2020, 170: 115013. doi: 10.1016/j.applthermaleng.2020.115013
    [14] ITŌ H. Friction factors for turbulent flow in curved pipes[J]. Journal of Basic Engineering, 1959, 81(2): 123-132. doi: 10.1115/1.4008390
    [15] LOCKHART R W, MARTINELLI R C. Proposed correlation of data for isothermal two-phase, two-component flow in pipes[J]. Chemical Engineering Progress, 1949, 45(1): 39-48.
    [16] FAZIO C, SOBOLEV V P, AERTS A, et al. Handbook on lead-bismuth eutectic alloy and lead properties, materials compatibility, thermal-hydraulics and technologies: 2015 edition[R]. Issy-les-Moulineaux, France: Organisation for Economic Co-Operation and Development, 2015.
    [17] 毕勤成,陈听宽,田永生,等. HTGR蒸汽发生器螺旋管内壁温分布和传热恶化规律的研究[J]. 核动力工程,1998, 19(5): 408-412.
    [18] XIAO Y, HU Z X, CHEN S, et al. Experimental investigation and prediction of post-dryout heat transfer for steam-water flow in helical coils[J]. International Journal of Heat and Mass Transfer, 2018, 127: 515-525.
    [19] HEJZLAR P, BUONGIORNO J, MACDONALD P E, et al. Design strategy and constraints for medium-power lead-alloy–cooled actinide burners[J]. Nuclear Technology, 2004, 147(3): 321-343. doi: 10.13182/NT147-321
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出版历程
  • 收稿日期:  2021-05-07
  • 录用日期:  2021-08-04
  • 修回日期:  2021-08-30
  • 刊出日期:  2022-06-07

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