Authors & Affiliations
Zubkov A.G., Oleksyuk D.A., Vertikov E.A., Noskov A.S.
National Research Centre “Kurchatov Institute”, Moscow, Russia
Zubkov A.G. – Researcher. Contacts: 1, pl. Akademika Kurchatova, Moscow, Russia, 123182.Tel.: +7 (926) 282-56-58; e-mail:
Oleksyuk D.A. – Head of the Department, Cand. Sci. (Tech.).
Vertikov E.A. – Junior Researcher.
Noskov A.S. – Head of Laboratory.
Abstract
This paper analyzes conventional methodologies used in domestic and international practice for calculating Critical Heat Flux (CHF) during Departure from Nucleate Boiling (DNB) in subcooled liquid flow, for subsequent determination of the Departure from Nucleate Boiling Ratio (DNBR) when assessing thermal-hydraulic reliability of VVER and PWR reactor cores. The study describes empirical correlations, look-up CHF tables, and three well-known phenomenological DNB models: bubble crowding model (BC), liquid sublayer dryout model (LSD), and dry spots area enlarging. An analytical comparison is performed between classical correlations across wide operational parameter ranges, along with benchmarking of VVER/PWR design correlations against phenomenological models using representative experimental datasets. Results demonstrate both qualitative and quantitative agreement between CHF values obtained through different calculation methods and experimental data. The paper presents a modified BC model and its implementation methodology in the subchannel code SC-INT, previously validated for local coolant parameter calculations (temperature, velocity, and pressure drop) based on experimental studies in VVER rod bundle. The upgraded model is compared against the Bezrukov correlation using 2300 data points from the Kurchatov Institute's CHF database for rod bundles.
Keywords
heat transfer crisis, Departure from Nucleate Boiling, Critical Heat Flux, flow boiling, Departure from Nucleate Boiling Ratio, DNB models, local parameters of the coolant, DNB, DNBR, VVER, PWR
Article Text (PDF, in Russian)
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