Volume 124, Nº 6 (2018)

The design principles and parameters of a promising neutron source, based on the fission reaction of the isotope ²³⁷Np, for use in research in extracted beams are presented. Such a source can operate as a pulsing reactor or super-booster or as a super-booster multiplying the neutrons from the target of a proton accelerator by using a reactivity modulator. Owing to the threshold character of ²³⁷Np fission such a neutron source will possess significantly higher characteristics than a similar setup with plutonium, and in terms of utilization efficiency it will be the leading source in the world for research employing neutron spectroscopy.

This work is devoted to the development of a system of analytical tests for verification of the thermohydraulic codes used in modeling the flow of one- and two-phase flows of liquid metals. The change of temperature in constant and variable flow of coolant, effect of axial heat conduction of the coolant on the temperature distribution, development of natural circulation in a closed loop, motion of a two-phase mixture under gravity, and fluctuations of the gas volume and distribution of the true process steam content in the presence of boiling/condensation are studied. The developed system of tests was used to verify the HYDRA-IBRAE/LM thermohydraulic code and the thermohydraulic module of the SOKRAT-BN code. The computational errors in the individual thermohydraulic parameters are obtained.

A method of calculating and the computational results obtained for the time variation of the radial deformation of the fuel-element cladding of a single-element EGC based on cermet with a polycrystalline molybdenum matrix in which uranium dioxide particles are distributed are presented. It is shown that the maximum admissible deformation of cladding comprised of 1-mm thick, hardened, single-crystalline, tungsten-tantalum alloy with a 10 mm in diameter central channel in the kernel is reached in 10 years.

The aims and problems of research on the means for intensifying heat exchange in the equipment of nuclear power plants are examined: nuclear reactors, steam generators, turbine condensers, heat-exchangers of dry water cooling towers, intermediate separators-steam superheaters of turbines, heat-exchange apparatus of safety systems, turbo-electro-generators, oil coolers, feed-water heaters. Methods of identifying heat-exchange are analyzed: different types of finning, turbulence generators and swirlers of one- and two-phase flows, coating on heating surfaces, systems with mixed flow. Computational recommendations are made for determining the intensity of heat emission from promising types of heat-exchange surfaces: bundles of lowfinned tubes and spiral coils.

Analysis of the distribution of radionuclides in the water and bottom deposits in the bays of the Kara Sea attests that the radionuclide emission from the sources of radioactive materials dumped into the bays decrease over time. The equations of the transport of radionuclides in submerged objects into the marine environment beyond their limits were constructed on the basis of new results. Evaluations of the consequences of the long-term presence of submerged radiation-hazardous objects at the sea bottom depend on the temporal reduction of the time constant for the emission of radionuclides. The calculations show that taking account of this parameter with the formation of corrosion openings in closed submerged objects the emission of ²³⁹Pu from them into the ambient environment will almost cease after several hundreds of years (700 or more years). Neglecting this factor, the emission of ²³⁹Pu will continue to increase for 6000–12000 years, and the contamination of the bottom deposits by this isotope will be significant.