基于漂移流模型的蒸汽发生器稳态热工水力分析
Steady Thermal-hydraulic Analysis of Steam Generator Based on Drift Flux Model
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摘要: 为研究蒸汽发生器的稳态热工水力特性，建立了四方程漂移流模型，并开发了一维计算程序。对蒸汽发生器U型管管束空间考虑为由一次侧通道、二次侧通道和传热管构成，对一次侧通道和二次侧通道的过冷段采用单相流模型，二次侧通道的沸腾段采用四方程漂移流模型，建立基于交错网格的一阶迎风差分方程，通过热平衡-自然循环压降的交叉迭代计算得到稳态热工水力参数。利用程序计算了秦山300 MW核电厂100%、75%、50%、30%、15%功率稳定运行工况下的热工水力特性，并与RELAP5的计算结果进行比较，两组结果一致性较好。Abstract: In order to study the steady thermal-hydraulic characteristics of a steam generator, a 1D simulation code based on the 4-equation drift flux model was developed. The U tube bundle space was considered to consist of the primary channel, the secondary channel and the tube wall. The flowing characteristics in sub-cooling part of the primary and secondary channels were simulated with single phase flow model, while the boiling part in the secondary channel was simulated with 4-equation drift flux model. The first-ordered upwind differencing equations were derived based on the staggered grid. An alternate iteration method of heat balance and driving force of natural circulation was implemented then. By the proposed method, the steady-thermal-hydraulic characteristics for steam generator of Qinshan 300 MW NPP, under 100%, 75%, 50%, 30%, 15% power condition, were analyzed and compared with simulated results of RELAP5. Two sets of results are in good agreement.
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[1] JAMES C S, JAMES K A. Appling experience in an advanced design[J]. Nucl Eng Inter, 1986, 31(393): 83-86. [2] KALAR S P. Modelling transients in PWR steam generator units[J]. Nuclear Safety, 1984, 25(1): 17-23. [3] 薛汉俊，严锦泉. 适用于微机的核蒸汽发生器热工水力分析程序——SGTH2[J]. 核动力工程，1989，10（6）：47-50.XUE Hanjun, YAN Jinquan. SGTH-2: A computer program for thermal-hydraulic analysis of steam generators[J]. Nuclear Power Engineering, 1989, 10（6）: 47-50(in Chinese). [4] 解衡，张金玲，贾斗南，等. 一体化压水堆蒸汽发生器的热工水力瞬态特性分析[J]. 核动力工程，1998，19（5）:413-418.XIE Heng, ZHANG Jinling, JIA Dounan, et al. Thermal-hydraulic transient behavior calculation of once-through steam generator of intergrated PWR[J]. Nuclear Power Engineering, 1998, 19（5）: 413-418(in Chinese). [5] 玉宇，居怀明. 螺旋管蒸汽发生器热工流体力学二维模型[J]. 清华大学学报：自然科学版，2004，44（9）：1202-1204.YU Yu, JU Huaiming. 2-D model for thermal-hydraulic analysis of helical-tube generators[J]. Journal of Tsinghua University: Science and Technology, 2004, 44（9）:1202-1204(in Chinese). [6] KAZIMI M, MASSOUD M. A condensed review of nuclear reactor thermal-hydraulic computer codes for two-phase flow analysis, MIT-EL 79-018[R]. USA: Department of Nuclear Engineering, MIT, 1980. [7] IAPWS. Revised release on the IAPWS industrial formulation 1997 for the thermodynamic properties of water and steam[EB/OL]. [2007-08]. http:∥www.iapws.org. [8] 于平安. 核反应堆热工分析[M]. 3版. 上海：上海交通大学出版社，2002：208-215. [9] TAITEL Y, BORNEA D, DUKLER A E. Modeling flow pattern transitions for steady upward gas-liquid flow in vertical tubes[J]. AIChE Journal, 1980, 26: 345-354. [10] 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. [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: 531-535. [12] BERGLES A E, ROHSENOW W M. The determination of forced convection surface boiling heat transfer[J]. Journal of Heat Transfer, 1964, 86: 363. [13] RELAP5 MOD3.3 code manual, Vol. Ⅳ: Models and correlations, NUREG/CR-5535/Rev[R]. US: Nuclear Regulatory Commission, 2006.
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