Experimental research on flow characteristics of boiling two-phase liquid sodium in annuli

被引:0
|
作者
Qiu Z.-C. [1 ,2 ]
Lan Z.-K. [2 ]
Qiu S.-Z. [1 ]
Gao X.-L. [1 ,3 ]
Lu X.-D. [2 ]
Sun D.-C. [2 ]
机构
[1] State Key Laboratory of Multiphase Flow in Power Engineering, Xi'an Jiaotong University, Xi'an
[2] CNNC Key Laboratory on Nuclear Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, Chengdu
[3] Nuclear and Radiation Safety Center, Ministry of Environmental Protection, Beijing
来源
| 2017年 / Atomic Energy Press卷 / 51期
关键词
Annuli; Boiling two-phase; Frictional multiplier; Liquid sodium;
D O I
10.7538/yzk.2017.51.01.0052
中图分类号
学科分类号
摘要
Experimental research on the flow characteristics of boiling two-phase liquid sodium in an annuli was done. Experimental conditions were G≤2 000 kg·m-2·s-1, p≤0.1 MPa and q≤550 kW·m-2. Two-phase frictional multiplier was calculated to investigate the boiling two-phase friction pressure drop. The experimental results show that the two-phase frictional multiplier decreases with the increase of the Martinelli parameter. On the base of the data of this work, Lurie et al.'s data and Kaiser et al.'s data, a correlation of the two-phase frictional multiplier for liquid sodium boiling in an annuli was obtained. RSDφ was used to evaluate the applicability of this correlation to the liquid sodium boiling in an annuli. It shows that this correlation is suitable for the flow characteristics of liquid sodium boiling in an annuli. © 2017, Editorial Board of Atomic Energy Science and Technology. All right reserved.
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页码:52 / 58
页数:6
相关论文
共 14 条
  • [1] Farmer F.R., Nuclear Reactor Safety, pp. 188-207, (1977)
  • [2] Hennies H.H., Cowking C.B., Griffith J.D., Et al., The fast-neutron breeder fission reactor: Safety issues in reactor design and operation, philosophical transactions, The Royal Society A, 331, pp. 409-418, (1990)
  • [3] Waltar E., Padilla A., Mathematical and computational techniques employed in the deterministic approach to liquid metal fast breeder reactor safety, Nucl Sci Eng, 64, pp. 418-451, (1977)
  • [4] Waltar A.E., Reynolds A.B., Fast Breeder Reactors, pp. 505-639, (1981)
  • [5] Maschek W., Struwe D., Accident analyses and passive measures reducing the consequences of a core-melt in CAPRA/CADRA reactor cores, Nucl Eng Des, 202, pp. 311-324, (2000)
  • [6] Lurie H., Noyes R., Boiling Studies for Sodium Reactor Safety, Part II: Pool Boiling and Initial Forced Convection Tests and Analyses, Tech. Rep. NAA-SR-9477, (1964)
  • [7] Lottes P., Flinn W., A method of analysis of natural circulation boiling systems, Nucl Sci Eng, 1, (1956)
  • [8] Ninokata H., Deguchi A., Assessment of the physical models in a two-fluid model code and interpretation of experiments, Nucl Energy, 28, pp. 161-170, (1989)
  • [9] Kottowski H., Savatteri C., Fundamentals of liquid metal boiling thermohydraulics, Nucl Eng Des, 82, 2-3, pp. 281-304, (1984)
  • [10] Kaiser A., Peppler W., Voross L., Type of flow, pressure drop, and critical heat flux of a two-phase sodium flow, Nucl Eng Des, 30, pp. 305-315, (1974)