Analog regulators of the energy-liberation distribution in nuclear reactors

Translated from Atomnaya Énergiya, Vol. 66, No. 2, pp. 81–85, February, 1989.     

Physical-mathematical model of volume condensation and transport of aerosols inside the containment shell of a vver reactor

Main Science Center of the Russian Federation — Physics and Power Engineering Institute. Translated from Atomnaya énergiya, Vol. 79, No. 5, pp. 376–381, November, 1995.     

Use of liquid metals in nuclear and thermonuclear engineering,and in other innovative technologies

By now, a good deal of experience has been gained with using liquid metals as coolants in nuclear power installations; extensive knowledge has been gained about the physical, thermophysical, and physicochemical properties of these coolants; and the scientific principles and a set of methods and means for handling liquid metals as coolants for nuclear power installations have been elaborated. Prototype and commercialgrade sodium-cooled NPP power units have been developed, including the BOR-60, BN-350, and BN-600 power units (the Soviet Union); the Rapsodie, Phenix, and Superphenix power units (France), the EBR-II power unit (the United States); and the PFR power unit (the United Kingdom). In Russia, dedicated nuclear power installations have been constructed, including those with a lead-bismuth coolant for nuclear submarines and with sodium-potassium alloy for spacecraft (the Buk and Topol installations), which have no analogs around the world. Liquid metals (primarily lithium and its alloy with lead) hold promise for use in thermonuclear power engineering, where they can serve not only as a coolant, but also as tritium-producing medium. In this article, the physicochemical properties of liquid metal coolants, as well as practical experience gained from using them in nuclear and thermonuclear power engineering and in innovative technologies are considered, and the lines of further research works are formulated. New results obtained from investigations carried out on the Pb-Bi and Pb for the SVBR and BREST fast-neutron reactors (referred to henceforth as fast reactors) and for controlled accelerator systems are described.     

Formalization of the process of directional composition of structures of autonomous power-supply systems during design

The morphological method of synthesis of the structures of autonomous power-supply systems is presented, which is a further development of the sequential algorithms. A procedure of creating variants of the structure is given in the form of the initial information block, as well as a series of the sequentially performed operations in the formation of a tuple of electrical units with imposed restrictions being checked. The method is based on simulation of the main design rules followed by development of the design documentation. The theoretical content of the method is disclosed in the form of the initial information (axioms), rules of information processing, and logical formulas obtained as a result of applying the logical operations to the initial premises. It is recommended to use the calculation of the predicates as a mathematical basis making it possible to present the main theoretical premises of synthesis as the deductive axiomatic theory with a standard formalization. The main advantages of the proposed method are shown as specified by reduction in the time of finding of the functionally necessary structures by eliminating the need to carry out an exhaustive search for variants and using computers.     

Thermohydraulic studies of sodium-cooled reactor facilities

Thermohydraulic studies of reactor facilities with fast reactors are complex experimentally and computationally. Extensive experimental data are obtained on the velocity and temperature profiles, hydrodynamic resistance and heat emission, initial heat section, and interchannel mixing of the coolant in the fuel assemblies. These are used to develop engineering methods of performing thermohydraulic calculations of fuel assemblies as well as computational compute codes. The particulars of the hydrodynamics and heat transfer in intermediate heat exchangers and steam generators of reactor facilities with fast reactors are studied. This has made it possible to validate their thermohydraulic characteristics.     

Basic Questions Concerning the Thermophysics of Nuclear Power

The term thermophysics encompasses three areas – hydrodynamics, heat transfer, and technology of coolants. These areas are closely interrelated and influence reactor physics, corrosion processes, and the reliability and safety of a nuclear power system. At the present level of knowledge, when developing prospective fourth-generation reactors and other power systems, the thermophysical processes occurring in the loops must be considered not separately but together, taking account of their mutual influence on the operation of the loops in the nuclear power system. These processes also must be taken into account in order to develop systems for controlling them.     

Heat transfer under natural convection of liquid metal during its boiling in a system of channels

We describe an experimental installation, a model, a test procedure, and results from experimental investigations into hydrodynamics and heat transfer during the boiling of an eutectic sodium-potassium alloy in a system of two parallel channels that simulate the operating conditions of channels in a typical fast-neutron reactor.