An analytical review of the evolution of attempts to create, schematic and constructive solutions for energycooled nuclear reactors with nuclear steam overheating and supercritical parameters of the working fluid in the conditions of the former USSR and the Russian Federation was made. A comparison of a number of major technical and economic characteristics of main developments of tube and tank reactors is made, the available information on the results of experimental and industrial operation of AMB reactors structures is considered, their advantages and disadvantages are evaluated in terms of technical perfection, reliability, technical and economic performance as well as environmental safety. The expected reduction in capital costs of 40% during the transition of nuclear power units with a capacity of 1,000 MW to single-circuit promising SCWR reactors is achievable only if the steam temperature rises to 625 °C, which has not yet been mastered even in traditional power engineering. The specific energy intensity of reactor’s active zones promising Russian developments under the SCWR program is in most cases extremely high, which will have a negative impact on the characteristics of nuclear safety. The conclusion is made concerning the high probability of a significant increase in the accident rate of the SCWR reactor cores, due to the insufficient study of the heat transfer process characteristics on the heat transfer surface of the TVEL under the conditions of supercritical parameters of the coolant, in particular, such phenomena as pseudo film boiling and thermoacoustic vibrations. In general, insufficient level of completeness of the latest Russian developments and lack of final conceptual projects were noted, which does not provide sufficient grounds for choosing promising schematic and constructive decisions necessary for making reasonable forecasts about the possibility of using supercritical parameters of the coolant in the modern nuclear power industry in the near future.

In the context of the actual problems of the physics of operational damage of modern reactor steels produced in the leading countries of the world (USA, Russia, Western Europe) and used for the manufacture of nuclear reactor vessels and other equipment of the first circuit of nuclear power plants, the characteristic features of possible dynamic damage in the responsible elements of this are considered. The mentioned problems are systematized from the standpoint of analyzing the effects of radiation embrittlement, as well as physical and chemical processes that, under certain conditions, are capable of developing in the operating equipment of Ukrainian NPPs, which are already working out their design operational resource. The characteristic features of possible dynamic damage in the operating reactor equipment of Ukrainian and foreign nuclear power plants are considered. The problem is systematized, first of all, from the standpoint of analyzing the operational stability of domestic and foreign reactor steels in relation to their radiation embrittlement. The peculiarities of the course of this physical process have been analyzed, which should be taken into account when determining the maximum possible terms of extension of safe operation of nuclear power units with reactors of the VVER type at the NPP of Ukraine. The main metal-physical properties of reactor steels of various types and possible problems caused by neutron irradiation, physical and chemical processes, vibrational and thermomechanical fatigue, which threaten the unexpected sudden destruction of reactor vessels, are considered. Special attention is paid to mechanical damage and processes accompanying the operation of reactor housings under conditions of cyclic and dynamic loads. A warning has been given regarding the unjustified extension of the period of reactors safe operation. The significant technological lag of the former Soviet, and now Russian, metallurgy from the level of metallurgy of the leading Western countries was noted. Data are provided on the high operational properties of the latest American steels, from which modern reactors of the AR1000 type are manufactured in the USA, and the safety, technical, economic and environmental advantages of using these reactors in Ukraine in comparison with new models of reactors of the VVER-1000 and VVER-1200.

The prospects and possibilities of replacing the capacities of nuclear power units of Ukrainian nuclear power plants, which are currently almost completely depleting their operational life, as well as the possibility of extending the guaranteed service life are analyzed and generalized. Based on the study of reactor construction trends and proposals on the world market, a basic promising modern model of a high-capacity 3+ AP1000 nuclear water reactor manufactured by Westinghouse El. Corp. was selected and recommended as a basic promising for use in Ukraine. The results of new studies on the dynamics of strength loss of reactor steels of WWER reactors and their welding joints under the action of radiation are considered. The main attention is paid to the presence and processes of migration and segregation in the crystal structure of reactor steels of harmful impurities, in particular phosphorus and nickel. It is concluded that there are significant advantages of new foreign metallurgical technologies, the use of which in vessel steels provides reliability and increased guaranteed life of safe operation of PWR reactors, in particular, AP1000 type made in the USA. The urgent problems of commissioning additional shunting capacities in the National Unified Energy System of Ukraine were assessed and the conclusion was made that they can be solved by improving the shunting characteristics of existing high-capacity nuclear power units and (mainly) by accelerating construction and commissioning of small modular reactors with high shunting characteristics, in particular the SMR-160 model manufactured by SMR LLC (USA).

The fundamental thermophysical features of the heat exchange process between the heated wall of a vertical channel and the light-water coolant of supercritical parameters concerning the conditions of heat-generating assemblies channels and cores of perspective energy nuclear reactors are considered. The available methods and recommendations for determining the limits of thermal load are analyzed. It is a guarantee the absence of the characteristic dangerous mode possibility of deteriorated heat exchange in these conditions and corresponding sharp rise in the channels wall temperature, which threatens their destruction. The physical nature of the occurrence of degraded heat transfer regimes remains unclear, and the existing approaches to the implementation of thermohydraulic calculation in such conditions are not sufficiently justified. The complex nature of intercellular heat and mass transfer in the fuel assembly and the presence of individual thermohydraulic cells with reduced levels of heat transfer intensity indicate that the existing method of determining the area of degraded heat transfer in the reactor core channels with supercritical parameters of the coolant is significantly simplified. Insufficient data and research results have been revealed to create adequate methods of heat-hydraulic calculation, suitable for taking into account the peculiarities of the heat transfer process complex flow under conditions of supercritical parameters of the coolant. The application of such methods should be the basis for ensuring the safe operation of prospective reactors and minimizing potential losses of a different nature from accidents caused by the destruction of cores through unacceptable heat transfer modes. To this end, the main direction of further research is identified.

The analysis of the current state of research and developments in the field of creation of thermal-hydraulic computer codes has been performed. The experience of creation of foreign versions of best-estimate codes was analyzed. Considerable attention is paid to the issue of critical heat flux calculation of nuclear reactors channels. It is demonstrated that now the efficiency of application of modern computer codes for estimating of the heat transfer crisis in the water-cooled nuclear reactors requires further improvement. Calculation methods for accuracy increase of predicting this thermal-hydraulic phenomenon in reactor channels are considered. The well-known methods of critical thermal flux in nuclear reactors channels have been analyzed. Peculiarities of determination of the heat transfer crisis in the forced of the vapor-water steam motion have been reviewed. Adequacy of software computer codes designed to calculate the main safety parameters of water-cooled nuclear reactors was analyzed. The idea of the physical mechanism of the heat transfer crisis under forced motion of a two-phase flow in heated channels was considered. Particular attention has been paid to analysis of experimental and calculated data on conditions of initiation of a heat transfer crisis in fuel assemblies rods.