The spent nuclear fuel storage system is designed to store and cool spent fuel for several years taking into account scheduled reloads and unloading of the entire core accumulated after use in a nuclear reactor. It consists of special pools or containers where spent fuel is placed for temporary storage before final treatment or disposal. These systems provide safe and efficient storage of spent fuel to prevent radioactive material from leaking into the environment and minimize risks to human health and the natural environment. The events that occurred during the Fukushima nuclear disaster on March 11, 2011, underscored the importance of safe storage of spent fuel in the spent fuel storage pool. Storage safety has therefore become a key aspect in this area. This article describes the heat sink loss calculations for the analytical substantiation of the emergency response instructions for the shutdown state of the Armenian NPP power unit No.2 using the RELAP5/Mod3.2 computer code. The initiating event in the event of loss of heat removal from the spent fuel pool is considered. The analysis of nuclear safety in the course of the development of a beyond design basis accident with a long NPP blackout was carried out in relation to the spent fuel pool of a power unit with a reactor plant WWER-440 (project V-270). The radiation consequences are estimated. The article provides calculations of the following accidents to determine the necessary actions of the operator: loss of heat removal from the spent fuel pool without operator action and loss of heat removal from the spent fuel pool from the organizations of subsequent make-up of the spent fuel pool with a boron cleaning pump (2NBO-2). The calculations are based on boundary and initial conditions corresponding to the assumptions of the «better estimate». 

One of the characteristics of energy strategy, which currently implementing is atomic energy elaboration and NPP development with rising part of atomic generation in national energy system. Of course one of the most important property of atomic generation is safety (Chernobyl, Three Mile Island, Fukushima Daiichi accidents). To maintain the operation of security systems in conditions of loss of external power supply, autonomous power sources - diesel-generator sets are used, but their use is associated with a number of difficulties of operation and maintenance, and currently developing technologies allow to realize new processes of energy transformation. And on the other hand modern technologies are providing access to new energy saving technologies. In this article we will try to compare energy storage systems, which use different physical principles and try to review their advantages and particularities. Also we especially explore operational advantages of battery energy storage systems and their influence on atomic safety and maintenance cycles and also integrating to digital control system and more effective generation and flexibility during NPP operation with conclusion about new business for «Rosatom» company.

Modern information and measuring and control systems (IMCS) usually have a modular structure. Convenient means of organizing intermodule interaction are the CAN interfaces (CAN FD) and the CANopen protocol. The standard tools provided by the CANopen specification are focused on the description of individual modules and do not ensure the integrity and consistency of the IMCS configuration as a whole. This article discusses the causes of possible configuration errors and suggests a method for describing the IMCS based on the CANopen protocol, which ensures consistent configuration of the system modules. The goal is achieved by supplementing the standard description of modules with an explicit description of the messages transmitted by modules to each other, indicating the source, receivers and a list of parameters. A scheme of information flows in a modular IMCS is developed, a general structure of the system description is proposed, including module parameters and message parameters. Data structures describing modules and messages are considered in detail, special attention is paid to ensuring the integrity of intermodule information links. It is shown that the proposed method makes it possible to detect or exclude configuration errors such as messages without a source or without recipients, duplication of parameter transmission, an attempt to transmit a non-existent parameter, the absence of transmission of some parameters, the absence of a source for the received parameter. The use of the XML language for the implementation of the IMCS description method is proposed, which provides automatic verification of the system configuration by standard language means based on the document schema. It is noted that the use of the method in the modeling and configuration program of the modular vibration monitoring system provided the identification of errors of inconsistent or incomplete configuration of individual modules at the earliest stages of design.

Modern requirements for the safe operation of nuclear power plants dictate the need to introduce systems for early reactor plants state diagnosis. Vibrations have always been a threat to the safe nuclear power plants operation, therefore, solving the issues of improving the nuclear power plants vibronoise diagnostics accuracy, especially in the field of ultra-low frequencies, is an urgent task. The most promising way to solve the problem of flexible guides in electromechanical vibration transducers is the levitation effect, which makes it possible to completely eliminate mechanical contact and, accordingly, reduce the sensitivity threshold of the transducer. The article presents an analysis of existing vibration transducers and offers analogue and digital mathematical models of the magnetic levitation system, taking into account the influence of the electromagnetic force nonlinearity. The characteristics of such a nonlinear system are studied using the method of harmonic linearization, which makes it possible to obtain an equivalent linear system. At the same time, the third order terms of the electromagnetic force expansion in the Taylor series are taken into account, which made it possible to obtain a more accurate analogue the vibration transducer model. For the developed vibration transducer digital model, using the z-form method, a system function and the corresponding difference equation were obtained. Expressions and graphs of transient, pulse and frequency characteristics are derived for both analogue and digital vibration converter models, which allow us to conclude about the accuracy and adequacy of the developed digital model. Using obtained vibration transducer digital models will make it easier to model its operation and reasonably select the vibration transducer initial parameters.

It is experimentally proven that the main cause of vibration excitation of equipment and internals of the main circulation circuit (MCC) are acoustic standing waves (ASW) and main circulation pumps (MCP). Usage of an interdisciplinary approach made it possible to create a digital acoustic model of pressurizer system with pipelines attached to it and prove that it is a self-oscillating system capable of generating acoustic standing waves (ASW) similar to simultaneous operation of several Helmholtz resonators. Each Helmholtz resonator in pressurizer is able to suppress a certain frequency of ASW generated by the reactor, which depends on coolant temperature. Determination of natural frequencies of coolant pressure oscillations (NFCPO) in pressurizer using Thomson formula is based on method of electroacoustic analogy. Acoustic compliance of pressurizer and acoustic mass of various combinations of pipelines connected to it make it possible to calculate NFCPO, in the system under consideration, according to the formula presented in the form of acoustic analogues of inductance and capacitance. Acoustic parameters of a system of Helmholtz resonators formed by a pressurizer and a connecting pipeline with hot leg of the third loop in nominal operating mode of NPP with WWER-1000 are calculated. A technology has been developed and verified, advantage of which is usage of a pressurizer for damping ASW, frequencies of which fall within vibration bandwidth of primary circuit structural elements. Calculation and experimental proof of previously unknown ability of pressurizer to damp ASW generated by the reactor in a wide frequency range is given, and possibility of preventing resonances with vibrations of equipment and NPP structures is substantiated.

Currently, a comprehensive approach is being developed for the integrity control of fuel rod claddings in pressurized water reactor (PWR) nuclear power plants. One of the key stages is the reliable detection of leaking fuel rods within the reactor's active zone during power operation. This study examines the general trends in the activity of fission product radionuclides in the primary coolant of various generations and designs of PWRs. The experience of Novovoronezh Nuclear Power Plant is considered regarding the method of sampling and separation of inert radioactive gases from the primary coolant. A comparison is made between different methods of measuring inert radioactive gases in the primary coolant: direct spectrometry, cryogenic method, and vacuum extraction method. The issue of reliable detection of fuel rod leakage during reactor operation is discussed, using the example of Unit 5 at Novovoronezh Nuclear Power Plant. Comparative results are presented for the analysis of the integrity of fuel rod claddings based on key reference radionuclides. The most reliable results indicating fuel rod leakage were obtained through the control of specific activities and analysis of changes in the ratio of inert radioactive gases using the vacuum extraction method. Control of the specific activity of inert radioactive gases in the primary coolant enables a highly reliable assessment of the condition of physical barriers. The proposed method of sampling and separation of inert radioactive gases demonstrates high reliability and convergence of results, serving as an additional tool for integrity control of fuel rod claddings during power operation. Having representative data on the specific activities of inert radioactive gases allows for a reliable assessment of the condition of fuel rod claddings within the reactor core.

The article analyzes the problem of diagnostics of NPP electric drive valves. Failure of this type of equipment can become the initial event for design or beyond design accident. Therefore, during the period of preventive maintenance at NPPs diagnostics of up to 2000 units of valves is carried out mainly by registration and analysis of the motor current signal during opening and closing. However, the applied statistical and frequency analysis methods do not always provide the necessary quality of diagnostics. The article points out the disadvantage of the applied methods – insensitivity to the dynamics of electromechanical equipment in terms of nonlinear manifestations of defects. As a tool for analyzing these manifestations the phase-plane method is proposed and its theoretical basis is outlined. The diagnostic signal is described as a vector diagram of a harmonic series subjected to amplitude and phase modulation. The most characteristic forms of phase trajectories - cardioids, nephroids, epicycloids – are shown. The main advantages of the phase-plane method over traditional frequency methods – sensitivity, possibility to identify the type of modulation, phase shift, dynamic and chaotic manifestations – are demonstrated by examples of processing diagnostic signals of NPP equipment. An important advantage of signal representation in the form of phase portrait is the visualization of dynamic features of the signal. The result of the work is a new method that allows to extract additional information about the object state by analyzing the non-stationary component of the initial data. It is proposed to introduce the phase-plane method into the practice of processing diagnostic signals of electric drive valves and other electromechanical equipment operated at NPPs.

As modern control systems engineering imposes limits on mathematical models of control objects, further development, and verification of the mathematical models suitable for power control system synthesis of nuclear power facilities is an ongoing problem. This article deals with the low-order dynamics model with lumped parameters of the Russian-designed pressurized water reactor, as well as its verification with experimental data from the full-size simulator of the WWER – 1200 nuclear reactor within two tests related to changing the position of Group 12 of CPS CR and changing the inlet coolant temperature. This model was created as a MATLAB S-function lv.2 model because of its capability to handle any type of signal.  Within this approach, the equation describing the coolant heating process is represented as a «two well-stirred tanks in series» model. The article demonstrates the upside of the given approach in comparison with the conventional approach, where the average temperature of the coolant is determined as an arithmetic mean of inlet and outlet coolant temperatures, respectively, on the basis of the test with a sudden change of the reactor inlet coolant temperature. In the frequency domain, the authors carry out stability analyses of the given model in a state-space form in relation to different external disturbances. Conclusions about the suitability of the given model as a fifth-order control object for parametric synthesis of a controller for the power control system of a nuclear power facility are drawn.

This paper aims to identify the problems that mediate the shortage of specialists in the energy sector in Russia. The article analyzes the situation related to the insufficient popularity of engineering education among applicants, which leads to the risk of under-enrollment of students by universities within the budget enrollment control figures, which mediates difficulties in providing the quantitative parameter of the required specialists. Reduction of points for admission to technical profiles by universities leads to the attraction of applicants who do not demonstrate a conscious choice of future profession and, with a high probability, may refuse to work in their specialty, which is demonstrated by the presence in the energy sector of only 23% of specialists under 35 years of age, which is below the national average. As a result of the study it was determined that in the course of joint efforts of the state, universities and energy companies themselves to attract young people to engineering profession and work in the energy sector, the most active role should belong to employers. In this regard, the authors consider the expediency of using HR-branding in the format of expanding the target audience of potential employees of energy companies with the inclusion of preschoolers, schoolchildren and students to whom material and moral value proposition is made. Expanding the audience of the HR-brand of the company to preschoolers, schoolchildren and students, applying real moral and material motivators for them will contribute to the growth of popularity of this area of training and conscious choice of profession, which, in the long term, will solve the issue of the shortage of qualified personnel in the energy sector of Russia.