Comparison of Subcooled Flow Boiling in a Full-Length 5×5 Rod Bundle between Uniform and Non-Uniform-Axial Power Distribution

Zhang Junyi; Yan Xiao

doi:10.13832/j.jnpe.2021.05.0008

Volume 42 Issue 5

Sep. 2021

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Zhang Junyi, Yan Xiao. Comparison of Subcooled Flow Boiling in a Full-Length 5×5 Rod Bundle between Uniform and Non-Uniform-Axial Power Distribution[J]. Nuclear Power Engineering, 2021, 42(5): 8-14. doi: 10.13832/j.jnpe.2021.05.0008

Citation:

Zhang Junyi, Yan Xiao. Comparison of Subcooled Flow Boiling in a Full-Length 5×5 Rod Bundle between Uniform and Non-Uniform-Axial Power Distribution[J]. Nuclear Power Engineering, 2021, 42(5): 8-14. doi: 10.13832/j.jnpe.2021.05.0008

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Comparison of Subcooled Flow Boiling in a Full-Length 5×5 Rod Bundle between Uniform and Non-Uniform-Axial Power Distribution

doi: 10.13832/j.jnpe.2021.05.0008

Zhang Junyi,
Yan Xiao^,

CNNC Key Laboradtory on Reactor Thermal Hydraulics Technology, Nuclear Power Institute of China, Chengdu, 610213, China

Received Date: 2020-08-31
Rev Recd Date: 2021-07-05
Publish Date: 2021-09-30

Abstract

Abstract

The structure of PWR fuel assembly is in form of square rod bundle. In this study, Computational Fluid Dynamics(CFD) method is verified to compare thermal-hydraulic(T-H) characteristics between Uniform-Axial Power Distribution(U-APD) and Non-Uniform-Axial Power Distribution(Non-U-APD) under subcooled flow boiling condition. It is shown that CFD method has the ability to achieve a goog agreement on the void fraction prediction while reasonalbe wall boiling model, interface force model and bubble size distribution model are implemented. The comparison shows that the onset of significant void point of Non-U-APD bundle appear in advance and the averaged void of Non-U-APD along the axial direction has a more higher increasing rate rather than that of the U-APD. At the end of the heated length, the corner channel-avereged void of Non-U-APD is higher than that of the U-APD while the central channels are nearly the same, although the same inlet condition and heated power is applied in both the simulation.At the downstream of 5th and 6th Mixing Vane Grid(MVG), the liquid mass flux of corner and central channel in Non-U-APD are lower than that of U-APD due to the phase change.
- CFD,
- Sub-cooled flow boiling,
- Full-length rod bundle,
- U-APD,
- Non-U-APD

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References(17)

References

[1]	LO S, OSMAN J. CFD modeling of boiling flow in PSBT 5×5 bundle[J]. Science and Technology of Nuclear Installations, 2012(2012): 795935.
[2]	GOODHEART K, ALLEBORN N, CHATELAIN A, et al. Analysis of the interfacial area transport model for industrial 2-phase boiling flow application[C]. Italy: Proceedings of the 15th International Topical Meeting on Nuclear Reactor Thermal-Hydraulics, NURETH-15, Pisa, 2013.
[3]	LUTSANYCH S, MORETTI F, D’AURIA F. Validation of NEPTUNE CFD two-phase flow models against OECD/NRC PSBT subchannel experiments[J]. Nuclear Engineering and Design, 2017(321): 82-91. doi: 10.1016/j.nucengdes.2017.02.005
[4]	CONG T L, ZHANG R, CHEN L J, et al. Studies on the subcooled boiling in a fuel assembly with 5 by 5 rods using an improved wall boiling model[J]. Annals of Nuclear Energy, 2018(114): 413-426. doi: 10.1016/j.anucene.2017.12.058
[5]	张蕊,干富军,左巧林,等. 压水堆燃料棒束通道内过冷沸腾分析[J]. 原子能科学技术,2015, 49(9): 1579-1585. doi: 10.7538/yzk.2015.49.09.1579
[6]	韩斌,杨保文,张汇,等. 过冷沸腾工况下不同刚凸结构对定位格架热工水力性能影响的数值模拟分析[J]. 核动力工程,2017, 38(3): 158-163.
[7]	李松蔚,李仲春,杜思佳,等. 带7道格架的5×5棒束两相性能CFD分析[J]. 核动力工程,2019, 40(3): 185-190.
[8]	杜利鹏,陈晓龙,张鹏飞,等. 带定位格架的5×5棒束通道内过冷沸腾流动传热数值研究[J]. 动力工程学报,2019, 39(8): 679-685. doi: 10.3969/j.issn.1674-7607.2019.08.012
[9]	LI L, WANG M, ZHANG D, et al. A subcooled boiling model developed for narrow rectangular channels based on the CFD method[C]. U.S.: Proceedings of the 18th International Topical Meeting on Nuclear Thermal Hydraulics (NURETH-18). Portland, Oregon, 2019.
[10]	KURUL N, PODOWSKI M Z. Multidimensional effects in forced convection subcooled boiling[C]. Israel: Proceedings of the 9th International Heat Transfer Conference. Jerusalem, 1990: 21-26.
[11]	KOCAMUSTAFAOGULLARI G. Pressure dependence of bubble departure diameter for water[J]. International Communications in Heat and Mass Transfer, 1983, 10(6): 501-509. doi: 10.1016/0735-1933(83)90057-X
[12]	KOCAMUSTAFAOGULLARI G, ISHII M. Interfacial area and nucleation site density in boiling systems[J]. International Journal of Heat and Mass Transfer, 1983, 26(9): 1377-1387. doi: 10.1016/S0017-9310(83)80069-6
[13]	COLE R. A photographic study of pool boiling in the region of the critical heat flux[J]. AIChE Journal, 1960, 6(4): 533-538. doi: 10.1002/aic.690060405
[14]	TOMIYAMA A, KATAOKA I, ZUN I, et al. Drag coefficients of single bubbles under normal and micro gravity conditions[J]. JSME International Journal Series B Fluids and Thermal Engineering, 1998, 41(2): 472-479. doi: 10.1299/jsmeb.41.472
[15]	LO S, ZHANG D S. Modelling of break-up and coalescence in bubbly two-phase flows[J]. The Journal of Computational Multiphase Flows, 2009, 1(1): 23-38. doi: 10.1260/175748209787387106
[16]	RUBIN A, SCHOEDEL A, AVRAMOVA M, et al. OECD/NRC Benchmark based on NUPECPWR sub-channel and bundle tests (PSBT): volume Ⅰ: experimental database and final problem specifications[R]. Knoxville: US NRC and OECD Nuclear Energy Agency, 2010.
[17]	张君毅,闫晓,肖泽军,等. 均匀加热全长棒束过冷沸腾工况子通道参数场计算分析[J]. 原子能科学技术,2018, 52(1): 48-55.