Nizhniy Novgorod, Russia
Nizhniy Novgorod, Russia

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Frick P.,RAS Institute of Continuous Media Mechanics | Khalilov R.,RAS Institute of Continuous Media Mechanics | Kolesnichenko I.,RAS Institute of Continuous Media Mechanics | Mamykin A.,RAS Institute of Continuous Media Mechanics | And 3 more authors.
EPL | Year: 2015

Turbulent convective heat transfer in a closed cylinder with aspect ratio L = 5D (D is the diameter and L is the cylinder length) filled with liquid sodium, heated at one end face and cooled at the other, is studied experimentally for three different positions: vertical, inclined at 45 degrees to the vertical and horizontal. The Rayleigh number, which is determined by the superimposed temperature difference and the cylinder diameter, varies within the range . It is shown that the convective heat transfer along the cylinder is most effective in the inclined cylinder, where an intense large-scale circulation exists on a background of developed small-scale turbulence. In the horizontal cylinder, the turbulence is weak, but the large-scale circulation provides moderate heat transfer. In the vertical cylinder, the large-scale circulation is absent, the turbulent fluctuations are most active, but the heat transfer is the weakest. The dependence of the Nusselt number on the Rayleigh and the Prandtl numbers, and the dependence of the Reynolds number on the Grashof number are shown and discussed. © Copyright EPLA, 2015.


Frick P.,RAS Institute of Continuous Media Mechanics | Khalilov R.,RAS Institute of Continuous Media Mechanics | Kolesnichenko I.,RAS Institute of Continuous Media Mechanics | Mamykin A.,RAS Institute of Continuous Media Mechanics | And 3 more authors.
Magnetohydrodynamics | Year: 2015

Turbulent convective heat transfer in a long closed cylindrical tube with L ≈ 20D (D = 96 mm is the diameter and L is the tube length) filled with liquid sodium, heated at one end face and cooled at the other, is studied experimentally for three different positions: vertical, inclined at 45 degrees to the vertical plane and horizontal. The Rayleigh number, which is determined by the superimposed temperature difference and tube diameter, varies within the range of Ra = (1-6)·106. It is shown that convective heat transfer along the tube is most effective in the inclined tube, where intense large-scale circulation (LSC) exists against the background of developed small-scale turbulence. In the horizontal position, turbulence is weak, but the LSC provides moderate heat transfer. In the vertical tube, LSC is absent, turbulent fluctuations are most active, but heat transfer is the weakest. The dependence of the Nusselt number on the Rayleigh and Prandtl numbers in the form Nu ~ (RaPr)x gives x ≈ 1 for the horizontal tube and x ≈ 0.8 for the vertical and inclined ones, which is essentially above the values known for turbulent convection in short vertical cylinders at "hard" (x= 2/7) and "ultrahard" (x = 1/2) convection.


Sukharev Yu.P.,JSC Afrikantov OKBM | Fomichenko P.A.,RAS Research Center Kurchatov Institute
Nuclear Engineering and Design | Year: 2014

The scope and trends of analytical and experimental activities aimed at studying neutronic characteristics of the GT-MHR high-temperature gas-cooled reactor are determined in compliance with the requirement for reliable validation of the GT-MHR design applying certified software tools. Risk factor existing in analytical validation of the GT-MHR neutronic characteristics consists in that the applicable neutronic analysis codes, which have been used up to the present time, are aimed at the HTGR designing and they have been tested mostly within analytical and experimental studies of HTGR with uranium fuel. Therefore, at the early stages of the GT-MHR nuclear design development it has already been determined that the experiments should be performed using critical test facilities and analytical benchmark-studies which are becoming quite relevant considering that plutonium fuel and erbium burnable poison are used. The report contains the validation program of the GT-MHR neutronic properties, the stages of its implementation and the main results obtained by now. © 2013 Elsevier B.V.


Bazhenov V.G.,Lobachevsky State University of Nizhni Novgorod | Zhestkov M.N.,Lobachevsky State University of Nizhni Novgorod | Zamyatin V.A.,JSC Afrikantov OKBM | Kibets A.I.,Lobachevsky State University of Nizhni Novgorod
PNRPU Mechanics Bulletin | Year: 2015

The process of nonstationary deformation of construction of the fast reactor with liquid metal coolant under postulated ULOF beyond design basis accident is examined. This type of accident includes core melting caused by disconnection of the main circulation pump arrangements of the primary circuit with the associated failure of the emergency protection. As a result of core melting, the area with a high energy level pressure is created. It is filled with sodium vapor. The progressive expansion of the energy area in the coolant leads to an increase of stress-strain state level of the reactor vessel and may lead to its destruction. The reactor facility must save integrity, provide localization of consequences of beyond design basis accident inside of the pressure vessel and avoid dangerous radiation effects on personnel of nuclear power station and the environment in these conditions. Current Lagrangian formulation is used for the description of coolant motion and structural elements of the reactor. The equation of motion derives from the balance of virtual capacity. Equations of the theory plastic flow are used in physical relations for metals. Deviatoric stress components are assumed to be equal to zero in the coolant; and the equation between hydrostatic pressure and density is taken as a state equation of quasiacoustic type. The contact between the coolant with structural elements of the reactor is simulated by the conditions of non-penetration. The problem solution is based on the method of moment schema of FEM and explicit finite-difference time integration scheme of the "cross" which are implemented in the computing system "Dynamics 3". The deformation of the fast reactor vessel is investigated numerically in the ULOF beyond design basis accident. The possibility of localizing effects consequences of beyond design basis accident inside of the pressure vessel of the reactor is analyzed. © PNRPU.


Golovko V.F.,JSC AFRIKANTOV OKBM | Dmitrieva I.V.,JSC AFRIKANTOV OKBM | Kodochigov N.G.,JSC AFRIKANTOV OKBM
2010 14th International Heat Transfer Conference, IHTC 14 | Year: 2010

The NPP design that integrates a high temperature helium cooled nuclear reactor with a gas-turbine power conversion unit requires investigations and development of high-efficiency heat-exchange equipment operating in the closed primary circuit. The equipment must be very compact, which implies highly efficient heat transfer at minimum pressure loss. This paper presents an analysis of optimal heat-exchange surface selection, as well as design and layout features of recuperators, precoolers and intercoolers. Considered are tube (made of straight, helical, including those with the small bending radius, finned tubes etc.), plate-and-fin and matrix heat-exchange surfaces combined as separate modules or as a single bundle. Suggested are methods and criteria to select rational heat-exchange surfaces with account of critical factors and limitations. Given are results of the comparative analysis and computational and experimental investigations of surfaces; design and layout solutions for heat-exchange apparatuses arranged in the vertical high-pressure vessel with limited dimensions. © 2010 by ASME.


Bakhmetyev A.M.,JSC Afrikantov OKBM | Bylov I.A.,JSC Afrikantov OKBM | Morev A.V.,JSC Afrikantov OKBM | Baklanov A.V.,JSC Afrikantov OKBM
ASME 2014 Small Modular Reactors Symposium, SMR 2014 | Year: 2014

Probabilistic methods allow to evaluate nuclear plant safety level systemically during plant lifetime and to develop recommendations on foreground directions in improving technical measures and operational procedures. Probabilistic assessment of plant safety level requires developing and introduction of appropriate methods and software. During last years JSC "Afrikantov OKBM" has developed different methods and software allowing to perform probabilistic study of nuclear plants safety and has implemented substantial amount of safety analysis and safety monitoring studies for different purposed small modular reactors. The paper describes methods and software developed by JSC "Afrikantov OKBM" and some results of its utilizing for small modular reactors. Copyright © 2014 by ASME.

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