Distribution of coolant flow rate in a steam generator in natural coolant circulation regimes

Abstract

The last few decades since the assimilation of atomic energy have been characterized by the formulation and implementation of reactor reliability and safety concepts. The necessity of removing the residual heat release in the core without damaging fuel elements in the case of serious accidents (specifically, with complete electric-power cutoff) has led to a wide swing in research work for adoption of natural coolant circulation regimes in the first loop and other heat-transfer media in nuclear power engineering. Cooldown in a natural coolant circulation regime during stoppage of the main circulation pumps during accidents is provided for in nuclear power plants with VVI~R reactors. However, not all questions concerning these regimes have been solved. Regimes based on natural circulation of the coolant require comprehensive investigations of the thermohydrodynamic characteristics, which can play the main role in practical implementation. This pertains both to operating nuclear power plants and to new plant designs (for example, plants with VVI~R-640 reactors), in which natural circulation of the coolant is the main component of reliability and safety. Passive heat removal systems are being developed based on it. A characteristic feature of nuclear power plant designs in Russia, as compared with foreign designs, is the use of horizontal steam generators. It is of scientific and practical interest to study the characteristics of the distribution of the coolant flow rate over the horizontal rows of pipes in the steam generators under conditions of natural circulation of the coolant. Among the latest investigations, we call attention to [1]. However, investigations are mainly directed toward studying the thermohydraulic processes directly in steam generators neglecting other factors, referring to the general loop circulation and at the same time influencing the flow-rate distribution. In the present paper the interrelation of these factors is examined.Basic Relations. Proceeding from the problem posed, we shall examine the main factors determining the conditions of natural circulation of coolant in the first loop. Figure I displays schematically the first loop of a nuclear power plant with a VVI~R reactor and horizontal steam generators. Here AH is the vertical displacement of the center of the pipe part of the steam generator relative to the center of the core, m; HsgP is the height of the pipe bundle of the horizontal steam generator, m. The moving head of natural circulation can be defined, to an adequate degree of accuracy, as Pdr = gApAH (Pa), where g = 9.81 m/sec 2 is the acceleration of gravity and Ap = ,o c --,% is the change in the coolant density in the core (kg/m 3) when the coolant is heated by AT (K). In the steady-state regimewhere APt is the hydraulic resistance in the first loop. The condition (1) determines the coolant flow rate (G, kg/sec), since APt = f(G). Using a linear approximation for the variation of p as a function of T and equating the average mass specific heat of the coolant Cp = const and...