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 Степанович

The 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.

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.

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.

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.

2017, Bochkarev, A. S., Korsun, A. S., Kharitonov, V. S., Alekseev, P. N., Бочкарев, Алексей Сергеевич, Харитонов, Владимир Степанович, Корсун, Александр Сергеевич

The 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.

2024, Bayaskhalanov, M. V., Merinov, I. G., Pisarevskiy, M. I., Kharitonov, V. S., Баясхаланов, Михаил Валерьевич, Меринов, Игорь Геннадьевич, Харитонов, Владимир Степанович

2022, Churkin, A. N., Baisov, A. M., Deev, V. I., Kharitonov, V. S., Баисов, Ахмед Магомедович, Харитонов, Владимир Степанович

Copyright © 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.

2015, Морозов, А. В., Ремизов, О. В., Маслов, Ю. А., Харитонов, В. С., Харитонов, Владимир Степанович, Маслов, Юрий Александрович

Составлено в соответствии с Государственным образовательным стандартом по дисциплинам «Основы проектирования и конструирования ЯЭУ», «Динамика и безопасность ЯЭУ». В пособии представлено описание пассивных систем безопасности, предназначенных для управления различными типами аварий на ЯЭУ. Рассмотрены системы, входящие в состав действующих, сооружаемых и проектируемых реакторов с различными теплоносителями и дана классификация этих систем. В помощь студентам, выполняющим курсовые проекты, демонстрируются разнообразные технологические и схемные решения, используемые в пассивных системах аварийного охлаждения различных типов ЯЭУ. Предназначено для студентов, обучающихся по направлениям 14.03.02 и 14.04.02 – Ядерные физика и технологии (программа «Теплофизика ядерных энергетических установок), 14.03.01 – Ядерная энергетика и теплофизика (программа «Атомные электростанции и установки»), а также специальности 141403 – Атомные станции: проектирование, эксплуатация и инжиниринг. Пособие также может полезно для аспирантов соответствующих специальностей и специалистов, работающих в атомной энергетике.

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.

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