Experimental efficiency evaluation of the of various geometry twisted bands for heat transfer intensification in the heat-exchange channels of a nuclear power unit equipment

The paper is devoted to experimental study of heat exchange and pressure drop of channel with various geometry twisted bands. The studies were carried out in the range of operating modes parameters of the nuclear power units’ heat exchange equipment. The three different designs of intensifiers are presented in the paper. The values of heat transfer coefficient and pressure drop are obtained. The dependences of the Nusselt number (Nu) on the Reynolds number (Re) were calculated. The comparative analysis of intensifiers is made. The efficiency factor was also calculated on experimental data. The most optimal geometry form of intensifier was selected.

CREATION AND VISUALIZATION OF AN OPERATIONAL PARAMETERS ARCHIVE

The developed methodologies allows visualizing the behavior of limiting param- eters over time and identifying patterns in the dynamics of changes in the parameters of a nuclear power unit, as well as determining if any of the parameters intersects the safe opera- tion settings. The relevance of the work is due to the importance of high-quality processing and visualization of data from the nuclear reactor archive using modern effective methodolo- gies. Also, the development of new approaches to the analysis of archived data can improve the efficiency of power plant operators.

Analysis of the possibility for operating a floating nuclear power plant in conjunction with a desalination plantAnalysis of the possibility for operating a floating nuclear power plant in conjunction with a desalination plant

Object and purpose of research. The article discusses in comparison the methods of desalination of seawater and their energy features from the point of view of the feasibility of including a desalination plant in the complex with a floating nuclear power unit. Materials and methods. Based on the analysis of various literary sources, a review of the main methods of nuclear desalination is made. The IAEA DEEP program was used to compare different desalination technologies. Main results. Based on the results of simulating nuclear desalination in the IAEA DEEP program, using the example of the Persian Gulf, preliminary recommendations were drawn up on the use of desalination methods in the joint operation of a desalination plant with a floating nuclear power unit. Conclusion. The integrated complex allows for desalination by both membrane and thermal methods. For the optimal choice of technology, it is necessary to specify the area of deployment and the relative position of the floating nuclear power unit and the desalination plant, and further search for a compromise based on more accurate calculations.

Rapid preliminary modeling of transport reactor cores

At the present time, JSC Baltiskiy zavod has built and transported to the deployment site at Pevek Akademik Lomonosov, a floating nuclear power unit (FNPU), project 20870. There are also three multi-purpose nuclear icebreakers of project 22220 (Arktika, Sibir, Ural) under construction at Baltiskiy being at different readiness stages. A decision has been made to build a nuclear icebreaker, Lider, of even a higher power. Integral reactors developed by JSC OKBM Afrikantov are installed in the nuclear icebreakers using new assembly-type cores which have not been used earlier in floating facilities. A great deal of preliminary calculation is required to give these cores as advantageous characteristics as possible. The paper proposes a procedure for rapid modeling of floating cores with varied operating and design characteristics. This procedure can be used as part of preliminary modeling. The procedure is based on using a combined dimensionless parameter proposed by the author in (Korolev 2009). A chart is presented to model the key performance of cores for floating objects with a nuclear reactor NPPs. Eight assembly-type core options, which can be installed in transport reactors of a modular or integral design, are analyzed.

Modeling of flexible nuclear power unit operational modes in the mathematical model of the Ukraine’s power system daily electric load profile dispatching

The use of large amounts of existing baseload NPPs capacities with a significant increase in renewable generation in the mathematical model of optimal dispatching of generating capacities of Ukraine’s power system leads to a significant surplus of electricity during peak power generation at solar PV’s, which necessitates additional sources of flexibility of the power system, such as battery electricity storage systems. The projects of new advanced nuclear power units provide for their use in flexible load modes with a maximum unloading of up to 50% of the rated capacity. Advanced NPP power units with small modular reactors are designed for even greater more maneuverable operation with possible unloading of up to 20% of rated capacity. The article presents approaches to modeling the use of NPP power units in variable load modes in the mathematical model of the optimal daily load schedule dispatching of Integrated Power System of Ukraine. The first approach is to model the operation of NPP power units similar to modeling the participation of cycling TPP units in covering the daily electrical load profile of the power system, in particular, changes in generation power in the range from minimum to nominal load levels, load rate. The second approach is to determination for each nuclear power unit of variants of modes of their hourly loading, the choice of one of which is made as a result of optimization. The modeling results showed that the choice of optimal flexible loading modes of new nuclear power units allows to provide the load balance of the power system almost completely with available capacities, including cycling TPP units, and the use of pump-storage generating units to transfer excess PV generation at peak hours of electricity consumption, which avoids the use of battery storage systems. Keywords: mathematical model of dispatching, operational mode, nuclear power unit, daily electric load profile, power system

Optimization of Systems Repair Plans and Assessment of the Useful Life of Nuclear Power Plant Equipment

This paper presents a computer-aided method of planning the volumes of repairs of systems of nuclear power units and a method for calculating their gamma-percentile life. This planning is carried out on the basis of predicting the reliability indicator, the probability of no-failure operation for a certain time period, with the gamma-percentile life of the equipment being determined by solving the corresponding equations. The tasks considered are related to an important energy problem of extending the operation of nuclear power units. Its importance is determined mainly by economic feasibility: it is cheaper to assess the useful life of a nuclear power unit and, on this research basis, extend its operation, than create a new unit. It is also shown that the calculation of the probability of a radiation accident at a nuclear power unit is associated with the results of planning the repairs of its systems, with assessment of its useful life. An optimization problem is formulated: it is required to find such a plan for the volumes of repair of a system that, with limited repair costs, its reliability indicator for a given duration deviates least from the maximum permissible value. The solution to the problem is based on calculating the structural reliability of the system. A graphological image of the system is built in the form of a composition of graphological images of typical structures. After the reliability indicator of typical structures has been calculated, the structures are replaced with individual structural elements, which makes it possible to simplify the initial graphological image of the system in a computational scenario and calculate its reliability indicator. The determination of the repair volume is carried out by applying a version of the coordinate-wise optimization method. To assess the gamma-percentile life, a model is adopted, in which the recoverable equipment components have an unlimited life, although, of course, they "age", and the non-recoverable components spend their life up to the level when their replacement becomes conditioned by the violation of the requirement for the maximum permissible value of the system reliability indicator. Estimates of the gamma-percentile life of the equipment are calculated by planning system repairs on a sequence of intervals of annual energy production by a nuclear power unit.