Abstract
The Critical Heat Flux (CHF) prediction under high pressure condition, even close to the vicinity of the critical pressure of water, is an important issue. Although there are many empirical CHF correlations, most of them have covered the pressure under 15MPa. In this study, based on the CHF experiment database of upflow boiling in vertical round tube from 15MPa to the vicinity of the critical pressure of water, the Katto, Bowring, Hall-Mudawar, Alekseev correlations, and Groeneveld LUT-2006 are comparatively studied. With an error analysis of the predicted CHF to the experiment database, the prediction capability and the applicability of these correlations are evaluated and the parametric trends of CHF varying with pressure from 15MPa to critical pressure are proposed. Simultaneously, according to the characteristics of Departure from Nucleate Boiling (DNB) type CHF under high pressure condition, the constitutive correlations of Weisman & Pei model are proposed. The prediction results of three entrainment and deposition correlations of Kataoka, Celata, and Hewitt corresponding to the Dry-Out (DO) type CHF are analyzed. Based on the two improved models above, a comprehensive CHF mechanistic model under high pressure condition combining the DNB and DO type CHF is established. The verification based on the experiment database of upflow boiling in vertical round tube and the parametric trends analysis of CHF varying with thermal-hydraulic and geometric parameters are carried out. Findings of this study have a positive effect on further development of CHF prediction method for universal CHF mechanism, especially under high pressure region.
Highlights
The accurate prediction of critical heat flux (CHF) in flow boiling is important in the design and safety analysis of nuclear reactor
It demonstrates that the present mechanistic model is applicable for the CHF prediction of upflow boiling in vertical round tube under high pressure conditions, and the RMS
The verification of the present mechanistic model based on the experiment database and the parametric trends analysis of CHF varying with thermalhydraulic and geometric parameters have been carried out
Summary
The accurate prediction of critical heat flux (CHF) in flow boiling is important in the design and safety analysis of nuclear reactor. The occurrence of CHF results in a sharp degradation of the convective heat transfer between the fuel rod cladding and the reactor coolant which may result in cladding failure. The supercritical water cooled reactor (SCWR) has high operating pressure and temperature, and, during sliding pressure start-up procedure from subcritical pressure to supercritical pressure, the thermophysical properties and transport properties of the coolant in the core would change greatly [1]. The CHF prediction under high pressure condition, even close to the vicinity of the critical pressure of water, is an important issue for SCWR. The CHF phenomenon has been extensively investigated over the last five decades, knowledge of the physical nature of CHF is still incomplete and the mechanisms of boiling crisis are still not well understood
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