Персона: Баисов, Ахмед Магомедович
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Hydraulic resistance of supercritical pressure water flowing in channels – A survey of literature
© 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.
Testing the Modified Subchannel TEMPA-SC Code in Comparison With Experiments and Other Computer Codes
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.
Heat transfer in rod bundles cooled by supercritical water – Experimental data and correlations
© 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.
Heat transfer characteristics of water under supercritical conditions
© 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.
A New Approach to Generalization of Experimental Data on Heat Transfer to Fluids in Supercritical Region
The 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.