536.24Modern computer codes for analyzing the safety of WER reactors include blocks for calculating the conditions for the appearance of heat-flux crisis in bundles of rod fuel elements. In each code the critical heat-flux density is calculated differently using empirical correlations most suitable for the specific fuel assembly construction. However, the existing recommendations are, as a rule, suitable only in a limited range of geometric and regime parameters. For this reason, in the RETRAN code, for example, several correlations possessing the best statistical characteristics in a definite pressure range are employed [I]. For p > 10.3 MPa, a B&W2 correlation is realized. In the pressure range 8.96-10.3 MPa interpolation of the data using the B&W2 and Barnett formulas are used to interpolate the data. The Barnett formula is used for calculations in the range 6.9-8.96 MPa. Interpolation of data using Barnett's formulas and a modified Barnett formula are usedin the range 5-6.9 MPa. For pressures p < 5 MPa, a modified Barnett formula is employed. The B&W2 correlation is based on experimental data on heat-flux crisis in bundles of rod fuel elements with uniform energy release over the height of the bundle. A Barnett correlation has been developed for ring-shaped channels in a limited pressure range near 6.9 MPa and a modified Barnett formula has been developed for vertical cooling of bundles of rod fuel elements.A Biazi correlation, obtained by generalizing data on heat-flux crisis in pipes and ring-shaped channels, is used in the TRAC code and a Canadian variant of "skeletal tables" for pipes is used in the RELAP5/MOD3 and CATHARE codes [2].Of course, before this choice is made the relations are verified by comparing with a database. Specifically, in [3] the results of a comparison of six of the best known foreign correlations for calculating the conditions for beat-flux crisis in PWR and BWR fuel assemblies are presented (11,000 points of a Columbia University database are used).As an analog in our own domestic practice for substantiating the thermotechnical safety of VVER reactors, we can site the reports of the Scientific-Research and Design Institute of Eleetrotochnical Apparatus (SRDIETA) [4] and the Institute of Nuclear Reactors of the Russian Science Center "Kurchatov Institute" [5]. The results of the first publication must be updated, and the second work was performed on a limited experimental material with "local" values of the mass velocity and equilibrium steam content (in all, 1675 points in the database of the Russian Science Center "Kurehatov Institute'). Of the four formulas studied in the work mentioned, only the recommendation of V. S. Osmachkin is used in Russia. Therefore only limited information on this question has been published in journals. The aim of the present paper is to eliminate this deficiency. We start from the correlations developed in their time by three leading organizations for substantiating the thermoteehnical safety of VVER reactors.Method of the Russian Science Center "Kur...
536.24Considerable attention has been given recently to verifying the thermal codes of an improved estimate, and determining their possibilities from the description of local and integral experiments which model the development of the process in the active region and the circulation system of a reactor in emergency situations.Improving the theoretical procedure for making a realistic estimate of the maximum permissible heat release of the active region of different kinds of water-cooled reactors under normal operating conditions is directly related to estimating the reliability of the theoretical parameters in situations where imperfections in the models of the processes, random and systematic errors in the databases, and uncertainties in specifying the initial parameters characterizing the state of the nuclear power plant all act together.The complexity of the processes modeled by thermal codes, imposes increased requirements on the correct choice of the system of closing equations. An insufficient study of the features of the hydrodynamics and heat exchange and the types of flow in two-phase media leads to the need to use numerous empirical constants and relations. Among the most important phenomena which determine the temperature state of the fuel elements is the heat-emission crisis and after-crisis heat exchange, and up to recently empirical correlations, developed and optimized for a certain type of reactor and a specific cooling system, have usually been employed.The three most well-known Russian correlations used to justify the thermal safety of the VVI~R reactor were analyzed in [1], and it was shown that none of them has any particular advantages when describing the experimental data. For this reason it is better to estimate the potential possibilities of modern analytic models as alternatives to empirical correlations. There are specific examples of the realization of this approach. In particular, the Japanese code algorithm FIDAS [2] is based on a three-liquid model and, according to the designers, the unit for calculating the heat-release crisis does not require empirical constants. A similar class of programs, judging by the publications, have become familiar in the USA, France and Germany.A review of research on the heat-release crisis during the past ten years can be found in [3, 4]. At the beginning of the 1990's models were proposed on the modern understanding of the phenomena that occur in the region of the walls when boiling occurs. The results of theoretical research and data on the hydrodynamics of the flow of two-phase media were used. Two approaches are used in the analytical procedures [3]. In the first, using the law of conservation of mass, momentum and energy, a closed system of partial differential equations is introduced for the vapor, the flow of droplets and a liquid film. The complete mathematical model of dispersed-ring flow developed at the Physico-Energy Institute is of this type [5]. But, because of its complexity, it has not so far been used as an engineering instnmaent, desp...
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