Vertikov E.A.1, Zubkov A.G.1, Oleksyuk D.A.1, Baisov A.M.2, Bosenko S.V.2, Churkin A.N.2
1 National Research Centre “Kurchatov Institute”, Moscow, Russia
2 Experimental and Design Organization “GIDROPRESS”, Podolsk, Russia
Vertikov E.A. — Engineer. Contacts: 1, pl. Akademika Kurchatova, Moscow, Russia, 123182. Tel.: +7 (985) 191-60-16; e-mail:
Zubkov A.G. – Researcher.
Oleksyuk D.A. – Head of Department, Cand. Sci. (Tech.).
National Research Centre “Kurchatov Institute”.
Baisov A.M. – 1st Category Design Engineer, Cand. Sci. (Tech.).
Bosenko S.V. – 1st Category Design Engineer.
Churkin A.N. – Head of the Department, Cand. Sci. (Tech.), Experimental and Design Organization “GIDROPRESS”.
Abstract
At present, thermal-hydraulic codes using the subchannel method of modeling coolant flow within rod bundles are the main means of substantiating the thermal reliability of nuclear reactor cores under steady-state operating conditions. The article gives an overview of the existing Russian subchannel thermal-hydraulic codes developed and used in various organizations of the nuclear industry for reactors with water coolant: SC-1, SC-INT and SC-Core (NRC “Kurchatov Institute”), PUCHOK-1000, TIGRSP, TEMPA-1F and TEMPA-SK (OKB GIDROPRESS JSC), KANAL (JSC “Afrikantov OKB Mechanical Engineering”), VYAZ-M, MIF-SKD (IPPE JSC). For each code a brief history of development is given, as well as the key features in terms of numerical methods used, the field of application and availability of certification passport. Questions about the interaction of computational tools using the subchannel approximation with codes that have a different scale of modeling the computational domain are discussed. As a part of system thermal-hydraulic codes, the subchannel method can be used as an additional module for detailed calculation of the core, which will allow to reasonably reduces the degree of conservatism in determining the departure from nucleate boiling ratio within the framework of modeling transient and emergency processes in comparison with the “hot channel” approximation. Computational fluid dynamics codes, which allow three-dimensional modeling of coolant flow, can be used, firstly, to clarify existing and create new closing relations for calculation of friction coefficients, local hydraulic coefficients and turbulent mixing efficiency, which are put into subchannel codes, and, secondly, to obtain boundary conditions for distribution of coolant flow rate by fuel assemblies at the inlet of the core. In order to improve the quality of description of heat and mass transfer processes in smooth rod bundles taking into account the presence of spacer and intensifier grids as further ways to improve subchannel codes, the following necessities are emphasized: expansion of the validation database, transition to a two-fluid description of the flow of two-phase flows and realization of the possibility of full-scale calculation of cores.
Keywords
heat and mass transfer, subchannel method, calculation codes, VVER, core, fuel assembly, smooth rod bundles
Article Text (PDF, in Russian)
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