Development of void fraction-quality correlation for two-phase flow in horizontal and vertical tube bundles. (May 2017)

• Record Type: Journal Article
• Authors: Hibiki, Takashi; Mao, Keyou; Ozaki, Tetsuhiro
• Is Part Of: Progress in nuclear energy. Volume 97(2017:May)
• Page Start: 38
• Page End: 52
• Publication Date: 2017-05
• Subjects: Void fraction -- Cross flow -- Horizontal bundle -- Rod bundle -- U tube -- Steam generator; Nuclear energy -- Periodicals; Nuclear engineering -- Periodicals; 333.7924
• DOI: 10.1016/j.pnucene.2017.01.003
• Languages: English
• ISSNs: 0149-1970

Abstract

Abstract: A steam generator thermal-hydraulic code based on homogeneous flow model has been useful based on its numerical stability and simpler formulation. One of key parameters for a steam generator thermal-hydraulic analysis is void fraction which determines two-phase mixture density and affects two-phase mixture velocity. These parameters are important for a heat transfer tube vibration analysis. A void fraction-quality correlation is very important to accurately convert the quality into the void fraction. The void fraction-quality correlation should preferably be applicable to parallel and cross flows in rod or tube bundles since two-phase flow in the steam generator encounters flow configuration change from the parallel flow along the tube bundle in the riser section of the steam generator to the cross flow in the U-bend section of the steam generator. A set of correlations depending on flow configuration such as parallel and cross flows, rod or tube array pattern and mass flux is developed based on legacy Smith correlation. The correlation agrees with the parallel and cross flow data with the mean absolute error (or bias) of 0.117% and the standard deviation (random error) of 2.26% and with the mean absolute error (or bias) of 0.760% and the standard deviation (random error) of 6.21%, respectively. The correlations are further simplified to a single correlation applicable for parallel and cross flow in rod or tube bundles. The Smith correlation with a modified constant entrainment parameter e being 0.5 is recommended for predicting void fraction in the steam generators. The Smith correlation with e = 0.5 is expected to be applicable for parallel and cross flows with various rod or tube array patterns including normal square, parallel triangular and normal triangular arrays. Highlights: Performed literature review on basic two-phase flow porous media formulations. Performed literature review on void fraction correlations and available database of two-phase flow in tube bundles. Developed a new void fraction correlation for tube bundles based on legacy Smith correlation. Compared the newly developed void fraction correlation with experimental data and performed the error analysis. Simplified the newly developed void fraction correlation for parallel and cross flows in steam generators.