The strategy for the development of nuclear power presupposes the construction of thermal water-cooled reactors and fast reactors [1][2][3]. At present, water moderated and cooled power reactors are the foundation of our country's nuclear power. The organizations of the State Corporation Rosatom have developed designs for new-generation reactor setups -AES-2006 and VVER-TOI, which are the result of evolutionary development and improvement of operating water moderated and cooled power reactors, which proved their reliability in thousands of reactor-years of accident-free operation. VVER are characterized by elevated power, the best economic and operational performance as well as enhanced safety compared with currently operating reactors. The thermal power is increased by, for example, using fuel assemblies with a new design incorporating heat-exchange intensifiers. The safety of the new designs of NPP with VVER is enhanced by means of the principle of technological diversity, which consists in combining active and passive safety systems. The passive systems ensure shutdown and prolonged removal of residual heat in the presence of a sealed loop or a depressurized loop and do not require operator intervention or an external long-time power source.To secure the possibility of increasing the power and validating the serviceability of new passive safety systems for VVER, a large-scale program of thermophysical and experimental studies has been completed.Increasing the Power Density and Efficiency of the VVER Core. The new NPP designs with VVER presuppose an improved design of the core as a whole and the fuel assemblies in particular. One promising avenue for increasing the capacity of power-generating units and the efficiency of the fuel cycle is to use fuel assemblies with improved thermohydraulic characteristics, which is accomplished by using mixing or intensifying lattices to increase heat transfer. Acting on the coolant flow these setups decrease the nonuniformity of the heating of the coolant (enthalpy) in the transverse section of a fuel assembly. They also make it possible to increase the turbulence properties of the flow, which causes moisture to settle on the walls of the fuel elements and promotes the removal of a large quantity of heat.In order to validate the heat-engineering reliability of a VVER core with fuel assemblies with mixing lattices, it is necessary to determine the effect of their design on the hydrodynamics and mass-transfer processes in the core. Important problems are the choice of the arrangement of the fuel assemblies and optimization of the design of the mixing lattices, which must possess the optimal combination of parameters such as the hydraulic resistance, the intensity of mixing of the coolant and a safety margin to crisis of boiling. The main method for studying the hydrodynamics and mass transfer in reactor fuel assemblies is the experimental investigation of fuel-assembly models on thermohydraulic stands: SVD-2, STF, and the TRASSER setup [4]. The main characteristics of the tripl...
This paper is devoted to the construction of a tabular method for calculating critical heat fluxes in fuel-element assemblies with a square arrangement. The method consists of a basic table and a system of correctional functions, describing the dependence of the critical heat flux on the important independent parameters. Experience in developing such a method for triangular assemblies is used. The method is compared with experimental data and other computational methods. The advantages of the new method are shown: maximum width of the range of parameters, possibility of taking account of the influence of the most important parameters, possibility of using in cell-by-cell (channelwise) calculations, and high accuracy of the description.The publication in 1997 of the international version of the tables for critical heat fluxes in pipes [1] laid the foundation for the development of similar tables for critical heat fluxes in other channels. The tables for pipes are the core in describing a crisis in channels and are the basis for developing tabular methods for describing and calculating a crisis in many channels with a complicated cross-section and in fuel-element assemblies. Subsequent works performed at the Physics and Power-Engineering Institute developed methods for tabular description and calculation of critical heat fluxes in ring [2] and other single channels [3], in triangular assemblies of cylindrical fuel elements [4,5], and in the assemblies with a mixed arrangement of fuel elements [6].Tabular methods for calculating critical heat fluxes in channels have substantial advantages over conventional methods and are widely used in practical calculations using approaches based on an analytical (in the form of formulas) description of experimental data. Tabular methods make it possible to describe a wider range of parameters, and are based on a wider experimental base, give a continual description of critical heat fluxes in the entire field of parameters, and are convenient for estimating the aggregate influence of the parameters and for monitoring the results of calculations.The Tabular Method of the Physics and Power Engineering Institute for Assemblies with a Square Arrangement of Fuel Elements. The following have been used in developing a method for PWR, BWR, RBMK, and other assemblies with a square arrangement of fuel elements: 1) experience in constructing tables for pipes, ring channels, and assemblies with a triangular arrangement of fuel elements [4,5] and 2) developments, presented in [7], where it is shown that critical heat fluxes in assemblies with triangular and square arrangements of fuel elements differ substantially from one another. Experimental data arrays for assemblies from the database at the Physics and Power Engineering Institute were used [8]. These data are published in [9][10][11][12][13][14][15][16][17]. Unfortunately, the extensive data of Columbia University [18-20] are inaccessible. The following parameters characterize the array used in the present paper: the number of experimenta...
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