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.
- 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 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.
- ПубликацияТолько метаданныеThermal conductivity of lead in the temperature range of 350–1000 °C(2022) Kruglov, A. B.; Rachkov, V. I.; Merinov, I. G.; Kharitonov, V. S.; Paredes, L. P.; Круглов, Александр Борисович; Рачков, Валерий Иванович; Меринов, Игорь Геннадьевич; Харитонов, Владимир СтепановичThe article presents the results of measuring the coefficient of thermal conductivity of lead in the temperature range of 350–1000 °C using the pulse heating method. The methodology of processing experimental data is described. The estimates of the experimental data error are given. The difference in the content of impurities in the lead samples is shown to have an in significant effect on the thermal conductivity coefficient of the lead melt. The deviation of the experimental data on the thermal conductivity of lead from the proposed approximating function does not exceed ±2 %. The obtained data are compared with the known recommended dependences for calculating the thermal conductivity coefficient of lead. © 2022, A.B. Kruglov, V.I. Rachkov, I.G. Merinov, V.S. Kharitonov, and L.P. Paredes.
- ПубликацияТолько метаданныеTesting the Modified Subchannel TEMPA-SC Code in Comparison With Experiments and Other Computer Codes(2022) Churkin, A. N.; Baisov, A. M.; Deev, V. I.; Kharitonov, V. S.; Баисов, Ахмед Магомедович; Харитонов, Владимир СтепановичCopyright © 2022 by ASMEThe paper describes a modified version of the TEMPA-SC computer program designed to calculate temperature fields in bundles of rods cooled by a supercritical pressure (SCP) fluid. This version of the program is based on the subchannel method that was used in the TEMPA-1F program, developed earlier in the OKB “GIDROPRESS” for calculating heat and mass transfer in the core of VVER-type reactors cooled by single-phase water at subcritical pressure. As the relations that close the system of equations of mass, momentum, and energy conservation, the new version of the program includes correlations for calculating heat transfer and friction resistance, taking into account the strong dependence of the properties of the coolant on temperature and pressure. In particular, the use of the universal calculation model of heat transfer, developed by the authors of this paper, allows us to perform calculations in a wide range of flow parameters of various fluids, including the modes of normal, improved and deteriorated heat transfer. The results of tests of the TEMPA-SC program are presented in comparison with the available experimental data for water and modeling fluids (carbon dioxide, freons R-12 and R-134a) at SCPs, as well as with the published data of calculations by using similar subchannel programs (COBRA-SC, ASSERT-PV) and CFD codes. A qualitative agreement between the calculated and experimental data is shown.
- ПубликацияТолько метаданныеHeat transfer characteristics of water under supercritical conditions(2022) Churkin, A. N.; Deev, V. I.; Kharitonov, V. S.; Baisov, A. M.; Харитонов, Владимир Степанович; Баисов, Ахмед Магомедович© 2021 Elsevier Masson SASThe paper deals with the problems of describing heat transfer modes in a turbulent flow of supercritical water based on generalization of known experimental data for vertical round tubes. Due to the danger of overheating of heat transfer surfaces, special attention is paid to modes with deteriorated heat transfer. Later, the authors developed a universal model that includes the well-known Dittus–Boelter correlation with corrections for changes in the thermophysical properties of water with temperature and the effect of thermal acceleration occurring in the flow. It is shown that the variation of the coefficients in the correction functions allows us to take into account individual features of changes in characteristics of the process in modes with deteriorated heat transfer. Three forms of displaying the modes of heat transfer to water of supercritical parameters with conditional boundaries of the modes of normal, improved and deteriorated heat transfer are proposed. The estimates of the limiting heat flux obtained in this paper for normal modes are compared with the data of other authors.
- ПубликацияТолько метаданныеThermal conductivity of Pb-Na and Pb-Bi-Na alloys in the temperature range of 350–800 °C(2023) Kruglov, A. B.; Konovalov, I. I.; Tarasov, B. A.; Kharitonov, V. S.; Paredes, L. P.; Круглов, Александр Борисович; Коновалов, Игорь Иванович; Тарасов, Борис Александрович; Харитонов, Владимир Степанович
- 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 метаданныеInvestigating the contact thermal resistances of lead in a heat conducting sublayer(2024) Kruglov, A. B.; Rachkov, V. I.; Merinov, I. G.; Kharitonov, V. S.; Paredes, L. P.; Круглов, Александр Борисович; Рачков, Валерий Иванович; Меринов, Игорь Геннадьевич; Харитонов, Владимир Степанович