Abstract

In many power plants, boiling can result in two-phase flows during normal and accident conditions. Heat and mass transfers through the gas-liquid interface in a two-phase flow process are proportional to the interfacial area concentration (IAC). These calculations for nuclear systems are typically performed using advanced system analysis codes such as TRACE. TRACE uses a two-group (2G) model, where bubbles are divided into two groups according to their drag coefficients. Correlations are used to calculate the group-wise void fractions (VFs) and IACs. This study evaluates the accuracy of the TRACE 2G VF and IAC models for vertically upward dispersed flows in large-diameter (LD) pipes. 2G experimental data in a wide range of dispersed flows from a VF of 0.16–0.68 in LD pipes were used for the model evaluation. Results showed that the TRACE 2G VF and IAC models could be replaced with more advanced models. Therefore, the second part of this study aims to advance the TRACE 2G IAC model while the formulation framework of the model is preserved. Thus, the advanced relations can be implemented in the TRACE code. An approach based on the 2G drift-flux model (DFM) is proposed to calculate group-wise VFs. Furthermore, this study improves the TRACE 2G Sauter mean diameter (SMD) correlation for LD pipes. The mean absolute relative errors (MAREs) of proposed models for the group one (G1) and group two (G2) VFs and total IAC predictions were improved from 42.9 %, 52.0 %, and 40.1 % to 21.6 %, 22.8 %, and 19.2 %, respectively.

Department(s)

Nuclear Engineering and Radiation Science

Publication Status

Full Text Access

Keywords and Phrases

Drift-flux model; Interfacial area; Interfacial transfer; Large-diameter pipes; Void fraction

International Standard Serial Number (ISSN)

0149-1970

Document Type

Article - Journal

Document Version

Citation

File Type

text

Language(s)

English

Rights

© 2026 Elsevier, All rights reserved.

Publication Date

01 Jan 2026

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