EDN: MSMICP
Authors & Affiliations
Vertikov E.A., Oleksyuk D.A., Malyutin M.A., Zubkov A.G.
National Research Centre “Kurchatov Institute”, Moscow, Russia
Vertikov E.A. – Engineer. Contacts: 1, pl. Akademika Kurchatova, Moscow, Russia, 123182. Tel.: +7 (985) 191-60-16; e-mail: This email address is being protected from spambots. You need JavaScript enabled to view it..
Oleksyuk D.A. – Head of Department, Cand. Sci. (Tech.).
Malyutin M.A. – Engineer.
Zubkov A.G. – Researcher.
Abstract
In Russia and abroad, subchannel codes are the main means for carrying out thermal-hydraulic calculations to substantiate the thermal reliability of the cores of water-cooled nuclear reactors. The paper presents an analysis of sets of closing relations used in various domestic subchannel codes based on a single-fluid model of two-phase coolant flow. It is concluded that there is no generally accepted position regarding the recommended closing relationships for calculating friction coefficients within bundles of rods and transverse turbulent flows. The analysis of problems related to the validation of subchannel codes for various parameters has been carried out: pressure drop across the height of the fuel rod bundle, local values of the coolant velocity and temperature in the cells, as well as local coolant parameters in the two-phase region. It is concluded that there is a lack of the required amount of experimental data on the local characteristics of two-phase flow in rod bundles of triangular geometry. The need for validation of subchannel codes in the region of two-phase coolant flow is shown in order to be able to separate the error in calculating the value of the critical heat flux into components associated with the inaccuracy of the program when calculating local parameters and the inaccuracy of the empirical correlation for determining the value of the critical heat flux.
Keywords
subchannel analysis, rod bundle, local coolant parameters, critical heat flux, validation, VVER
Article Text (PDF, in Russian)
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Problems of Atomic Science and Technology. Series: Nuclear and Reactor Constants, 2025, no. 1, 1:20
