Персона: Харитонов, Владимир Степанович
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Институт ядерной физики и технологий
Цель ИЯФиТ и стратегия развития - создание и развитие научно-образовательного центра мирового уровня в области ядерной физики и технологий, радиационного материаловедения, физики элементарных частиц, астрофизики и космофизики.
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Руководитель научной группы "Теплогидравлика реакторов с водой сверхкритического давления"
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Харитонов
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Владимир Степанович
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- ПубликацияТолько метаданныеNumerical Simulation of Thermal–Hydraulic Processes in Liquid-Metal Cooled Fuel Assemblies in the Anisotropic Porous Body Approximation(2019) Chudanov, V. V.; Aksenova, A. E.; Pervichko, V. A.; Korsun, A. S.; Merinov, I. G.; Kharitonov, V. S.; Bayaskhalanov, M. V.; Корсун, Александр Сергеевич; Меринов, Игорь Геннадьевич; Харитонов, Владимир Степанович; Баясхаланов, Михаил Валерьевич© 2019, Pleiades Publishing, Inc.Abstract—: The article presents an anisotropic porous body model in which the transfer anisotropy is taken into account through determining—by means of tensor analysis techniques—the drag force, effective viscosity, and thermal conductivity. The model is intended for describing heat-and-mass transfer in fuel assemblies and tube bundles. For closing the system of anisotropic porous body equations, the integral turbulence model developed by the authors is used. To verify how correctly the hydrodynamics and heat transfer are described, a few hydrodynamic and thermal–hydraulic processes in water- and liquid-metal-cooled fuel rod assemblies are simulated in the anisotropic porous body approximation. The results from simulating the flow patterns of lead–bismuth eutectics in the experimental 19-rod assembly and water in a 61-rod nonheated assembly with its flow cross-section locally blocked in the central and corner parts are presented. The thermal–hydraulic processes in the BREST reactor fuel assembly’s heated 19-rod fragment with its flow cross-section locally blocked in the central part were also simulated using the CONV-3D DNS code in the framework of model cross-verification activities. The numerical analysis was carried out using the developed APMod software module implementing the anisotropic porous body model jointly with the integral turbulence model. It was demonstrated from a comparison of the numerical analysis results with both experimental data and simulation results obtained using the CONV-3D computer code that the APMod software module adequately describes the 3D fields of coolant velocities, pressure, and temperature arising in fuel rod assemblies with a locally blocked part of their flow section. The obtained results testify that the anisotropic porous body model can be used for simulating thermal–hydraulic processes in the cores and heat-transfer equipment of prospective reactors.
- ПубликацияТолько метаданныеSimulation of Heat and Mass Transfer in Wire-Wrapped Fuel Assemblies in the Anisotropic Porous Body Approximation(2020) Chudanov, V. V.; Aksenova, A. E.; Pervichko, V. A.; Korsun, A. S.; Merinov, I. G.; Kharitonov, V. S.; Bayaskhalanov, M. V.; Корсун, Александр Сергеевич; Меринов, Игорь Геннадьевич; Харитонов, Владимир Степанович; Баясхаланов, Михаил Валерьевич© 2020, Pleiades Publishing, Inc.Abstract: Results of the simulation of heat and mass transfer in wire-wrapped fuel assemblies in the anisotropic porous body approximation using the developed APMod software package are presented. The modifications introduced into the porous body model to make it suitable for wire-wrapped fuel assemblies are described. The predictions of thermal and hydraulic characteristics in the liquid-metal cooled experimental and model fuel assemblies according to this updated model are presented. An isothermal sodium flow in a Bundle 2A experimental 19-rod wire-wrapped assembly and uniform or nonuniform heating of the rods was studied. The predictions were compared with the experiments using the pressure difference across the assembly versus the coolant flowrate and the coolant temperature distribution in the assembly’s outlet section. The thermal–hydraulic characteristics in the BN-1200 reactor fuel assembly’s heated 19-rod fragment with its flow cross-section locally blocked in the central part calculated by the porous body model were compared with the predictions by the CONV-3D DNS code. Before their comparison, the distributions of local velocities, pressure, and temperature in an assembly cross-section calculated by the CONV-3D code were averaged over the averaging cells in the APMod software package. It is demonstrated that the APMod software package may be used to calculate parameters, which are averaged over a representative averaging cell, in a liquid-metal coolant flow in wire-wrapped fuel assemblies with an accuracy adequate for engineering applications.
- ПубликацияТолько метаданныеHeat transfer in rod bundles cooled by supercritical water – Experimental data and correlations(2020) Churkin, A. N.; Deev, V. I.; Kharitonov, V. S.; Baisov, A. M.; Харитонов, Владимир Степанович; Баисов, Ахмед Магомедович© 2019 Elsevier LtdThe paper presents the results of a comparative evaluation of existing correlations for the prediction of heat transfer to supercritical pressure water against experimental data for bundles of heated rods. It is shown that the method, developed by the authors for the engineering calculation of the heat transfer coefficient in the supercritical region of the coolant parameters, makes it possible to describe the experimental data for smooth rod bundles with an error of less than 15%. In the case of rod bundles with wrapped wire, the same method gives values of the heat transfer coefficient, which are on average about 8% lower compared to the experimental data obtained. This is connected with the intensification of heat transfer due to swirling the flow and can be considered as a margin when estimating the limiting temperature of the heat transfer wall. Based on the comparison of the calculation results by different methods, the problems of choosing the ratio for determining the heat transfer coefficient in fuel assemblies of reactors with supercritical pressure water are discussed.
- ПубликацияТолько метаданныеA New Approach to Generalization of Experimental Data on Heat Transfer to Fluids in Supercritical Region(2020) Deev, V. I.; Kharitonov, V. S.; Baisov, A. M.; Churkin, A. N.; Харитонов, Владимир Степанович; Баисов, Ахмед МагомедовичThe paper contains the results of analysis of heat transfer regimes in the case of forced turbulent flow of water and modeling fluids in channels of various configurations at supercritical pressures. Two new complex criteria were proposed for describing heat transfer in the pseudo-phase transition region. A system of equations suitable for the engineering calculation of heat transfer in fuel assemblies of nuclear supercritical water reactors is presented. A comparison of the calculations of heat transfer coefficient (HTC) with experimental data for the regimes of normal, deteriorated and mixed heat transfer is made.
- ПубликацияТолько метаданныеDetermination of integral turbulence model parameters as applied to the calculation of flows in fuel assemblies of fast reactors in porous-body approximation(2020) Vlasov, M. N.; Korsun, A. S.; Maslov, Yu. A.; Merinov, I. G.; Kharitonov, V. S.; Корсун, Александр Сергеевич; Маслов, Юрий Александрович; Меринов, Игорь Геннадьевич; Харитонов, Владимир Степанович© Published under licence by IOP Publishing Ltd.This work aimed to correct the integral turbulence model developed earlier for assemblies of smooth rods. Two variants of the fuel assembly design were considered. In the first variant, the fuel rods were spaced using spacer grids. The presence of a spacer grid does not require a change in the form of the system of equations but leads to a change in the form of the resistance tensor and the generation of turbulence in the spacer grid region. In the second variant, a wire-wrapped fuel bundle was analyzed. The presence of a wire-wrapped fuel bundle requires an additional term in the equation for the conservation of momentum and change in the form of the resistance tensor. The simulations were obtained by CFD code ANSYS CFX and aimed at the determination of parameters involved in an integral model of turbulence being developed for modeling nuclear-reactor cores and heat exchangers in anisotropic porous-body approximation.
- ПубликацияОткрытый доступInherent Safety Characteristics of Advanced Fast Reactors(IOP Publishing Ltd, 2017) Bochkarev, A. S.; Korsun, A. S.; Kharitonov, V. S.; Alekseev, P. N.; Бочкарев, Алексей Сергеевич; Харитонов, Владимир Степанович; Корсун, Александр СергеевичThe study presents SFR transient performance for ULOF events initiated by pump trip and pump seizure with simultaneous failure of all shutdown systems in both cases. The most severe cases leading to the pin cladding rupture and possible sodium boiling are demonstrated. The impact of various features on SFR inherent safety performance for ULOF events was analysed. The decrease in hydraulic resistance of primary loop and increase in primary pump coast down time were investigated. Performing analysis resulted in a set of recommendations to varying parameters for the purpose of enhancing the inherent safety performance of SFR. In order to prevent the safety barrier rupture for ULOF events the set of thermal hydraulic criteria defining the ULOF transient processes dynamics and requirements to these criteria were recommended based on achieved results: primary sodium flow dip under the natural circulation asymptotic level and natural circulation rise time.
- ПубликацияТолько метаданныеHydraulic resistance of supercritical pressure water flowing in channels – A survey of literature(2021) Churkin, A. N.; Deev, V. I.; Kharitonov, V. S.; Baisov, A. M.; Харитонов, Владимир Степанович; Баисов, Ахмед Магомедович© 2021 Elsevier B.V.The survey is devoted to the problem of the hydraulic resistance of channels with water at supercritical pressures. The pressure drop in fuel assemblies is one of the main characteristics of water-cooled nuclear reactors. It is expected that pressure drops in SCWRs, which are considered now as a perspective of the IV Generation PWR, may be essentially less than in modern reactors. However, the flow resistance of the reactor equipment, as previously, is an important element among the various parameters of the future plants with SCWR. From 1968 up today, in spite of the obvious peculiarities of hydrodynamics and heat transfer processes in fluids at supercritical pressures, comparatively few experimental investigations of this problem have been carried out. Up to 2000, the experiments at supercritical pressure of water were conducted mainly with smooth round tubes, some data on pressure drops in annular channels were obtained, and it was known only one work in which the hydraulic resistance of a tight-lattice 7-rod bundle with helical fins was studied. Later, the similar investigations began intensively to develop in the People's Republic of China. On the whole, the experimental results, which have been published, are contradictory, and the correlations, based on these data, may be used in limited ranges of geometric and regime parameters. The conclusion was made that there is the urgent necessity to continue the investigations in the here examined field. The purpose of future researches must be the removal of existent contradiction in experimental results, receiving the new data on hydraulic resistance of channels simulating of the real fuel assemblies of SCWRs and the development of general correlations suitable for engineering calculations of hydrodynamic characteristics of the SCWR core in the range of operating parameters.
- ПубликацияТолько метаданныеAnalysis of numerical studies on the thermal-hydraulic and neutronic thermal-hydraulic stability of supercritical water reactors(2021) Sujyan, A. M.; Deev, V. I.; Kharitonov, V. S.; Харитонов, Владимир Степанович
- ПубликацияТолько метаданныеAnalysis of numerical studies into the thermal-hydraulic and coupled neutronic and thermal-hydraulic stability of supercritical water reactors(2021) Sudzhyan, A. M.; Deev, V. I.; Kharitonov, V. S.; Харитонов, Владимир Степанович© 2021 Obninsk Institute for Nuclear Power Engineering, National Research Nuclear University 'MEPhI'. All rights reserved.The paper presents a review of modern studies into the potential types of the supercritical reactor core coolant flow instabilities. Instabilities affect adversely the operating safety of nuclear power plants. Despite an impressive number of numerical studies on the subject, there are problems which remain unsolved. This is largely explained by drawbacks in numerical reactor models. The major of these are the use of one simulated channel instead of a system of two or more parallel channels, the lack of consideration of neutronic feedbacks, and a problem of choosing calculated ratios for the heat-transfer coefficient and the hydraulic resistance coefficient in conditions of a supercritical water flow. Based on this, a decision was made to undertake an analysis which will make it possible to identify these problems and to formulate, on their basis, general requirements to the model of a nuclear reactor with supercritical light-water coolant. The need has been noted for building improved numerical models for the integrated analysis of interlinked hydrodynamic, thermal and neutronic processes in the reactor plant's cooling channels with regard for the peculiarities of the flow and heat exchange in water with highly variable properties.