Modeling of void fraction covariance and relative velocity covariance for upward boiling flows in subchannels of a vertical rod bundle

被引:5
作者
Hibiki, Takashi [1 ]
Ozaki, Tetsuhiro [2 ]
机构
[1] City Univ Hong Kong, Dept Mech Engn, Kowloon, 83 Tat Chee Ave, Hong Kong, Peoples R China
[2] TEPCO Syst Corp, 2 37 28 Eitai, Tokyo 1350034, Japan
关键词
Subchannel analysis; Covariance; Distribution parameter; Rod bundle; Boiling flow; 2-PHASE FLOW;
D O I
10.1016/j.ijheatmasstransfer.2023.124277
中图分类号
O414.1 [热力学];
学科分类号
摘要
Accurate simulation of boiling two-phase flows in a rod bundle is indispensable for the robust, economical, and safe design of various heat transfer systems using the rod bundle configuration. Subchannel analysis codes are used for this purpose. Interfacial drag force modeling significantly affects the prediction accuracy of the void fraction. The void fraction and relative velocity covariances constitute the interfacial drag force. However, the covariances are currently not considered in existing subchannel codes due to a lack of reliable constitutive equations to calculate the void fraction and relative velocity covariances. This study aims to model the subchannel-average void fraction and relative velocity covariances for subcooled and saturated boiling flows in three types of subchannels in a rod bundle. The considered subchannels are interior, edge, and corner subchannels. The subchannel-average void fraction and relative velocity covariances for saturated boiling flow are modeled by the data obtained from local void fraction data collected for saturated boiling flow in an 8 x 8 rod bundle under pressures from 1.0 to 8.6 MPa. The subchannel-average void fraction and relative velocity covariances for subcooled boiling flow are modeled based on the bubble-layer thickness model. The modeled subchannel-average void fraction and relative velocity covariances are well validated with the experimental data. The modeled subchannel-average void fraction and relative velocity covariances are expected to be implemented in subchannel analysis codes to improve the void fraction prediction accuracy in each subchannel type.(c) 2023 Elsevier Ltd. All rights reserved.
引用
收藏
页数:14
相关论文
共 30 条
[1]  
Aydogan F., 2006, THERMAL HYDRAULICS A, V2, P133, DOI [10.1115/ICONE14-89174, DOI 10.1115/ICONE14-89174]
[2]  
Borkowski J.A., 1992, TRAC BF1MOD1 MODELS
[3]   Effect of void fraction covariance on relative velocity in gas-dispersed two-phase flow [J].
Brooks, Caleb S. ;
Liu, Yang ;
Hibiki, Takashi ;
Ishii, Mamoru .
PROGRESS IN NUCLEAR ENERGY, 2014, 70 :209-220
[4]   CFD methodology and validation for single-phase flow in PWR fuel assemblies [J].
Conner, Michael E. ;
Baglietto, Emilio ;
Elmahdi, Abdelaziz M. .
NUCLEAR ENGINEERING AND DESIGN, 2010, 240 (09) :2088-2095
[5]   Modeling of void fraction covariance in two-phase flows with phase change [J].
Dandekar, Akshay V. ;
Brooks, Caleb S. .
INTERNATIONAL JOURNAL OF HEAT AND MASS TRANSFER, 2016, 100 :231-242
[6]   NEPTUNE:: A new software platform for advanced nuclear thermal hydraulics [J].
Guelfi, Antoine ;
Bestion, Dominique ;
Boucker, Marc ;
Boudier, Pascal ;
Fillion, Philippe ;
Grandotto, Marc ;
Herard, Jean-Marc ;
Hervieu, Eric ;
Peturaud, Pierre .
NUCLEAR SCIENCE AND ENGINEERING, 2007, 156 (03) :281-324
[7]   Modeling of bubble-layer thickness for formulation of one-dimensional interfacial area transport equation in subcooled boiling two-phase flow [J].
Hibiki, T ;
Situ, R ;
Mi, Y ;
Ishii, M .
INTERNATIONAL JOURNAL OF HEAT AND MASS TRANSFER, 2003, 46 (08) :1409-1423
[8]   One-dimensional drift-flux model and constitutive equations for relative motion between phases in various two-phase flow regimes [J].
Hibiki, T ;
Ishii, M .
INTERNATIONAL JOURNAL OF HEAT AND MASS TRANSFER, 2003, 46 (25) :4935-4948
[9]   VOID FRACTION DISTRIBUTION IN BWR FUEL ASSEMBLY AND EVALUATION OF SUBCHANNEL CODE [J].
INOUE, A ;
KUROSU, T ;
AOKI, T ;
YAGI, M ;
MITSUTAKE, T ;
MOROOKA, S .
JOURNAL OF NUCLEAR SCIENCE AND TECHNOLOGY, 1995, 32 (07) :629-640
[10]   2-FLUID MODEL AND HYDRODYNAMIC CONSTITUTIVE RELATIONS [J].
ISHII, M ;
MISHIMA, K .
NUCLEAR ENGINEERING AND DESIGN, 1984, 82 (2-3) :107-126