内螺纹管临界热流密度预测与数值计算

doi:10.7538/yzk.2018.youxian.0149

摘要/Abstract

摘要：

基于壁面汽泡壅塞理论，针对近临界压力区两相流动沸腾的偏离泡核沸腾(DNB)现象，对垂直上升内螺纹管的DNB型临界热流密度(CHF)进行了数值计算研究。以内螺纹管为分析对象改进已有的汽泡壅塞模型，计算了汽泡层区与主流区的极限传递质量流量、湍流速度分布、汽泡层区临界截面含气率等本构关系，汽泡脱离直径的计算考虑了汽泡接触角的影响。本文模型还根据大量CHF实验数据拟合得到了新的α_b关联式。最后，基于Fortran语言编制了CHF的理论预测数值模型程序，研究分析了压力、质量流速、热平衡干度及进口欠焓对CHF的影响，并根据CHF查表值对本文模型进行评估，同时将实验得到的内螺纹管CHF数据与采用Bowring模型、Katto模型、Shah模型和本文模型计算的CHF进行比较，发现本文模型的误差最小，与实验值吻合结果较好，说明本文模型能较好地对垂直上升内螺纹管DNB型CHF进行预测。

关键词: 汽泡壅塞, 内螺纹管, 偏离泡核沸腾, 临界热流密度, 数值计算

Abstract:

Based on the theory of bubble coalescence in the wall bubbly layer, for a deeper understanding of departure from nucleate boiling (DNB) in two-phase flow boiling under near-critical pressures, critical heat flux (CHF) of DNB type in a vertical upward rifled tube was numerically studied. Some improvements of the existing bubble coalescence model were proposed for the rifled tube. Meanwhile, the limiting transverse interchange of mass flux crossing the interface of the bubbly layer and core region, the turbulence velocity distribution, critical void fraction of the bubbly layer and other constitutive relations were determined. The influence of the bubble contact angle was taken into account in the detached bubble diameter calculation. The new α_b empirical correlation was proposed through fitting a large number of CHF experimental data. Finally, based on Fortran language, the theoretical prediction numerical model of CHF was developed. The effects of pressure, mass flux, equilibrium quality and inlet subcooling on CHF were analyzed. The model was evaluated by the values of CHF Lookup Table. And the comparison between the predictions by Bowring model, Katto model, Shah model as well as the present CHF model and the experimental CHF data shows the best agreement of the proposed model, indicating that the proposed model can predict CHF of DNB type well for the vertical upward rifled tube.

Key words: bubble coalescence, rifled tube, departure from nucleate boiling, critical heat flux, numerical computation

谢海燕;蒋慧卿;赵云杰;杨冬*. 内螺纹管临界热流密度预测与数值计算[J]. 原子能科学技术, 2018, 52(11): 1968-1976.

XIE Haiyan;JIANG Huiqing;ZHAO Yunjie;YANG Dong*. Prediction and Numerical Computation of Critical Heat Flux in Rifled Tube[J]. Atomic Energy Science and Technology, 2018, 52(11): 1968-1976.

参考文献

［1］周磊，刘祥锋，闫晓，等. 矩形窄缝通道临界热流密度数值预测［J］. 核动力工程，2011，32(4)：46-51. ZHOU Lei, LIU Xiangfeng, YAN Xiao, et al. Numerical prediction of critical heat flux in narrow rectangular channels［J］. Nuclear Power Engineering, 2011, 32(4): 46-51(in Chinese).
［2］沈植，杨冬，姜勇，等. 环形炉膛锅炉亚临界及近临界压力区低质量流率内螺纹管传热特性［J］. 中国电机工程学报，2015，35(8)：1947-1953. SHEN Zhi, YANG Dong, JIANG Yong, et al. Heat transfer characteristics of internally ribbed tube for a boiler with annular furnace at sub-critical and near critical pressure［J］. Proceedings of the CSEE, 2015, 35(8): 1947-1953(in Chinese).
［3］WEISMAN J. Current status of theoretically based approaches to the prediction of the critical heat flux in flow boiling［J］. Nuclear Technology, 1991, 99(1): 1-21.
［4］WEISMAN J, PEI B S. Prediction of critical heat flux in flow boiling at low qualities［J］. International Journal of Heat and Mass Transfer, 1983, 26(10): 1463-1477.
［5］CHANG S H, LEE K W. A critical heat flux model based on mass, energy, and momentum balance for upflow boiling at low qualities［J］. Nuclear Engineering and Design, 1989, 113(1): 35-50.
［6］KWON Y M, CHANG S H. A mechanistic critical heat flux model for wide range of subcooled and low quality flow boiling［J］. Nuclear Engineering and Design, 1999, 188(1): 27-47.
［7］SAHA P, ZUBER N. Point of net vapor generation and vapor void fraction in subcooled boiling［C］∥5th International Heat Transfer Conference. Tokyo, Japan: ［s. n.］, 1974: 175-179.
［8］潘杰，杨冬，肖荣鸽，等. 低干度流动沸腾临界热流密度预测模型［J］. 核动力工程，2013，34(4)：58-63. PAN Jie, YANG Dong, XIAO Rongge, et al. Critical heat flux prediction model for low quality flow boiling［J］. Nuclear Power Engineering, 2013, 34(4): 58-63(in Chinese).
［9］喻娜，张渝. 汽泡壅塞改进模型对矩形通道临界热流密度(CHF)的预测［C］∥中国核学会2009年学术年会. 北京:中国原子能出版社,2009.
［10］LAHEY R T, MOODY F J. The thermal hydraulics of a boiling water nuclear reactor, second edition［R］. La Grange, America: American Nuclear Society, 1993.
［11］TONG L S, HEWITT G F. Overall view point of flow boiling CHF mechanisms［R］. US: ASME, 1972.
［12］LEVY S. Forced convection subcooled boiling: Prediction of vapor volumetric fraction［J］. International Journal of Heat and Mass Transfer, 1967, 10(7): 951-965.
［13］KOHLER W, KASTNER W. Heat transfer and pressure loss in rifled tubes［C］∥Proceedings of 8th International Heat Transfer Conference. San Francisco: ［s. n.］, 1986.
［14］STAUB F W. The void fraction in subcooled boiling-prediction of the initial point of net vapor generation［J］. Journal of Heat Transfer, 1968, 90: 151-157.
［15］PAN J, LI R, YANG D, et al. Critical heat flux prediction model for low quality flow boiling of water in vertical circular tube［J］. International Journal of Heat and Mass Transfer, 2016, 99: 243-251.
［16］ZUBER N, FINDLAY J A. The effects of non-uniform flow and concentration distributions and the effect of the local relative velocity on the average volumetric concentration in two-phase flow［J］. Journal of Heat Transfer, 1965, 87(4): 453-468.
［17］BECKER K M, DJURSING D, LINDBERG K, et al. Burnout conditions for round tubes at elevated pressures［C］∥Proceedings of the International Symposium on Two-Phase Systems. ［S. l.］: ［s. n.］, 1972: 55-73.
［18］THOMPSON B, MACBETH R V. Boiling water heat transfer burnout in uniformly heated round tubes: A compilation of world data with accurate correlations［M］. ［S. l.］: ［s. n.］, 1964.
［19］ZENKEVICH B A. The Generalization of experimental data on critical heat fluxes in forced convection of sub-cooled water［J］. Journal of Nuclear Energy, 1959, 1(2): 130-133.
［20］GROENEVELD D C, SHAN J Q, VASIC＇A Z, et al. The 2006 CHF look-up table［J］. Nuclear Engineering and Design, 2007, 237: 1909-1922. 
［21］谢海燕，李耀德，蒋慧卿，等. 近临界压力区垂直内螺纹管临界热流密度实验研究［J］. 原子能科学技术，2017，51(9)：1571-1577. XIE Haiyan, LI Yaode, JIANG Huiqing, et al. Experimental investigation on critical heat flux in a vertical rifled tube at near-critical pressures［J］. Atomic Energy Science and Technology, 2017, 51(9): 1571-1577(in Chinese).
［22］BOWRING R W. Simple but accurate round tube, uniform heat flux, dryout correlation over the pressure range 0.7-17 MN/m² (100-2500 psia)［R］. UK: United Kingdom Atomic Energy Authority, 1972.
［23］KATTO Y, OHNO H. An improved version of the generalized correlation of critical heat flux for the forced convective boiling in uniformly heated vertical tubes［J］. International Journal of Heat and Mass Transfer, 1984, 27(9): 1641-1648.
［24］SHAH M M. Improved general correlation for critical heat flux during upflow in uniformly heated vertical tubes［J］. International Journal of Heat and Fluid Flow, 1987, 8(4): 326-335.

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