Influence of pipeline geometry on hydrodynamics and heat transfer processes by an example of a ship steam generator

The theoretical and computational analysis proposed in this work is aimed at identifying the features of thermal and hydrodynamic processes carried out in the steam-generating channels of the ship type water-moderated nuclear power installations. It is shown that the complex geometry of the thermohydraulic tract curvilinear channels of the steam generating system has a significant effect on the efficiency of the transport nuclear power installation. In addition to the formation of large-scale vortex structures and swirling flow in the pipeline, the phenomenon of the swirling flow crisis is revealed, under which the low-frequency component of the acoustic spectrum is enhanced. The scientific and applied significance of the proposed research is associated with the need to ensure a wide range of operational changes in efficient and safe operation power modes of icebreaker nuclear power installations. The research, aimed at developing the principles of physical and mathematical modeling of complex vortex flows, is necessary to optimize the design parameters of the thermal power equipment elements of new generation ship nuclear power installations in order to ensure increased safety and reliability of their operation.

Enhancing the Regulatory Support of Occupational Radiation Protection during Decommissioning and Dismantling of Ships with Nuclear Power Installations and Nuclear Service Vessels

Results: Information is provided on the developed regulatory and methodological documents including health physics and organizational requirements for the radiation safety and protection supervision in the course of work on decommissioning and dismantling of nuclear submarines, nuclear service ships and surface vessels with nuclear installations, as well as during storage of reactor compartments of decommissioned nuclear submarines and block packs of nuclear service ships. Conclusion: The developed documents are aimed at further improving the regulatory legal and methodological support of citizens’ rights to safe working conditions, health protection and social protection.

Arguments for tightness testing of nuclear system pipeline parts, attached to pressureless tanks

Actual standard, namely Rules for design and safe operation of equipment and pipelines of nuclear power installations (2015) does not regulate strength and tightness testing of pipeline parts, located between valves and pressureless tanks of nuclear systems at nuclear power plants. This paper introduces arguments for strength and tightness test conduction of such areas of nuclear power plant pipelines, designed in accordance with Russian standards. It was also proposed some amendments for the norm aimed to improve safety and reliability of nuclear power plant systems with radioactive contamination. It was suggested to apply the result of current study to Leningradskaya-2 in order to improve safety and decrease the number of failures of the liquid radioactive waste storage system.

Methodology for Verification of the Heat Transfer Crisis in the Nuclear Fuel Assemblies

The present-day stage of nuclear-power engineering development raises sharply a number of complicated questions regarding the guarantee of safety operation of nuclear power-generating units of operating and designed Nuclear Power Plant (NPP). The most important of these unsolved technological problems were considered in [1] on the base of analysis of ways of operation reliability improvement for Nuclear Power Installations (NPI) with WWER and RBMK reactors. In connection with the priorities formulated in [1], in papers [2–6] the main aspects of approaches available for solution of most complex problem are considered: the development of methods of early identification of initial phases of emergency operation regimes in such nuclear power-generating units which are critically important for NPI trouble-free operation. It is necessary to stress, that reliable identification of the anomalies mentioned, especially of thermal-hydraulic nature ones in core region of nuclear reactor, must be provided under conditions when such operating troubles can not yet be detected by the issued supervisory instruments of NPI. Taking into account the requirement to prospective diagnostic provision of NPP equipment, in papers [2–5] are underlined that at present time the development of effective methods of anomalies identification in NPI equipment and development of mathematical software support on base of these methods for computer-aided diagnostic systems on base of AI conceptions in structure of hardware of operator support tools of new generation NPP are considered as the main condition which determine the development of diagnostic means with mentioned functional possibilities.