The article considers the possible ways to optimize the technological solutions of the recharge and boron control system of nuclear power plants under construction within the AES-2006 project. The possibilities for optimization of technological solutions of the system of recharge and boron regulation of the AES-2006 project, which will not affect the reliability and efficiency of its main functions: purge-recharge of the primary circuit and boron regulation, were studied. As a result of the analysis of technological solutions and analytical calculations carried out during the work, it was found that in the system of recharge and boron regulation of the NPP within the project AES-2006 it is possible to perform optimization basing on reduction the metal content of the heat exchange equipment by reducing the surface area of the heat exchangers of the coolant outlet, reducing the power of pumps, as well as reducing the diameter of a number of main pipelines. Implementation of the proposed optimization of technological solutions will allow a more rational arrangement of the system and reduce capital costs for the construction of nuclear power plants as a whole, while not adversely affect the safety of the system and its functions.

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...

The article presents the results of the study of neutron-physical characteristics of the container for storage of radioactive waste of nuclear power plants with uranium-graphite reactors. The interaction of gamma quanta (in the energy range from 0.1 to 2 MeV) with the structural materials of the container is simulated. The numerical values of the parameters determining the radiation characteristic of the container with the estimation of the calculation error are obtained. The following main characteristics of the container are determined: the attenuation coefficient of the equivalent dose, the numerical factor of gamma radiation accumulation. These characteristics can be used to justify the radiation safety of the container, in particular when selecting protection materials, as well as when building additional heterogeneous protection barriers.