NPP Power Units Research Articles

Methods and means of ensuring the accuracy of bending deformation measurements of connecting pipe between the VVER reactor and the steam generator in the conditions of a fuel campaign

The welded joint zone (WJ) No. 111 is a welding unit for the "hot" collector of the primary coolant to the DN1200 steam generator nozzle. The experience of operation of steam generators PGV-1000M of NPP power units with VVER-1000 shows that the WJ zone No. 111 is prone to the formation and development of operational defects. According to the results of periodic non-destructive testing of the WJ No. 111 zone performed at the PG of various power units of the VVER-1000 NPP, cases of detection of extended planar-type discontinuities, including through ones, were repeatedly recorded. Thus, this zone is one of the most critical zones of the reactor plant (RU) with VVER -1000. The root cause of the formation and development of operational defects has not yet been identified. Solving the problem of cracking of WJ No. 111 steam generators of nuclear power plants with VVER-1000 is one of the priority areas for improving the safety of operation of the power unit during the over-design service life. This problem is determined by a complex combination of temperature conditions, mechanical and corrosive effects, is relevant and has not yet been definitively solved. Therefore, it is very important to ensure the collection of reliable data on the actual stress-strain state of the WJ No. 111 zone in various operating modes (pressure loading of steam generators, heating/cooling of the power unit, hydraulic testing of the 1st and 2nd circuits). The article reveals the approaches and stages of creating and applying a strain gauge installation to ensure metrological suitability in the conditions of a campaign fuel company at an NPP power unit by using an adequate physical model to obtain additive and multiplicative correction coefficients.

Prolongation of Safe Operation of the K-1000-60/3000 Turbine Power Unit after Damage to the HPC Rotor

Currently, when large-scale military actions are taking place on the territory of Ukraine, the inclusion of the domestic energy system in the European one is a reliable component of providing electricity to the country's energy market. However, according to experts, uninterrupted and safe operation of nuclear power plants is still considered a prerequisite for the stable operation of the energy sector of Ukraine. The purpose of the paper is to assess the damageability and individual resource of the rotor of high-pressure cylinder (HPC) of the K-1000-60/3000 turbine power unit of the LMZ after damage to the blades in order to prolong the operation of the power unit in the conditions of a stressed state of the power system. One of the most effective ways to partially solve the problem of replacement of generating capacities is to extend the operating periods of NPP power units after the end of the project operating period, provided that nuclear and radiation safety standards are met. The review of the previously established service life of the energy equipment of NPP power units involves the assessment of the residual resource of the energy equipment in accordance with the regulatory documents. After the accidental damage of the blades of the last stage of the HPC rotor of the K-1000-60/3000 turbine power unit of the LMZ, there was a need to study the cyclic and static damage, the individual residual resource of the HPC rotor. In the process of achieving the goal, studies were carried out for three design options: the original option (five stages of the HPC rotor), the option without the blades of the last stage and the option without the fifth stage (with four first stages). The calculation of the resource indicators of the rotor in the execution of the HPC without blades of the 5th stage shows that the static damage accumulated in the main metal is 52%, the cyclic damage is 5% when applying the standard strength reserves for the number of cycles and for deformations at the level of nN=10 and nε=1.5 according to the recommendations of SOU-N MEV 40.1-21677681-52:2011. Thus, the total damage to the base metal is 57%, which sets the residual resource of the HPC rotor at the level of 88.4 thousand hours. The calculation of the resource indicators of the rotor in the execution of the HPC without entire 5th stage shows that the static damage accumulated in the base metal is 52%, the cyclic damage is 6% when applying the standard strength reserves for the number of cycles and deformations at the above-mentioned level. The total damage to the base metal is 58%, which determines the residual resource of the HPC rotor at the level of 85.6 thousand hours.

An analysis of the “Fast” noise measurements of the dynamic VVER processes

Studies of maneuverable modes of VVER to confirm the possibility of participation of nuclear power plants in the mode of daily carrying capacity have been conducted for quite a long time. Tests at various nuclear power plants with VVER-1000 (Zaporizhzhia NPP in 1998, Khmelnitsky NPP in 2005, Tianwan NPP in 2007) have shown the practical possibility of NPP participation in the daily schedule of carrying capacity, however, the commissioning of nuclear power plants with VVER-1200 requires similar work on all new units with VVER-1200: NVAES-2, LNPP-2 Belarusian NPP. The article presents some aspects of the use of noise control methods for analyzing the condition of equipment and the core. Since the emergence of the technology of noise analysis of signals from VVER reactor installations, researchers have formulated several criteria for obtaining results of appropriate quality. The fundamental requirement for conducting noise experiments was the registration of data in stationary modes of operation of power units, since any non-stationarity made significant changes in spectral estimates, which ultimately complicated the work and “distorted” the results obtained. This requirement was included in the operating instructions of various diagnostic systems using noise signal analysis methods (the SUS system, manufactured by Siemens). For a long period of time, the current situation suited both developers of various diagnostic systems and NPP personnel operating them at power units. On the one hand, this was due to the imperfection of the technical means used (low speed of analog-to-digital converters, limited storage capacity, bulky equipment, etc.), on the other hand, the use of domestic NPP power units only in the base load mode without tracking daily power fluctuations in the power system. The standard archives of the upper block level system, the in-reactor control system and additionally produced multi-channel “fast” measurements with a frequency of 1 kHz for the analysis of maneuverable mode 95-55-95% of the VVER-1200 reactor plant were analyzed. Global disturbances of the core have been detected after one step of the regulatory body of the control and protection system, which attenuates within one second if the next step of the control and protection system has not occurred during this time. Such fast neutron processes can be controlled only by neutron-noise measurements with an upper frequency of at least 20 Hz.

Feasibility study of measures to master a new technology for manufacturing thin-walled large-size bottoms by flanging (in conditions of «Atommash» the branch of «AEM-Technologies» JSC in Volgodonsk)

The object of research in this article is the «Atommash» the branch of «AEM-Technologies» JSC in Volgodonsk, an organisation that was designed and commissioned in 1976 as the head machine-building enterprise for the production of a complete set of hull and heat-exchange equipment for NPP power unit in the volume of the nuclear island. The subject area chosen is the organisation of production of large-size products, which the enterprise intends to diversify through the acquisition of a line for the manufacture of thin-walled bottoms of large-size pressure vessel housings for NPPs and oil and gas chemistry. The purpose of the research is to develop a project for the development of new technological capacities based on the analysis of existing technologies and methods of production, which will increase the efficiency of production activities at the «Atommash» the branch of «AEM-Technologies» JSC in Volgodonsk. The main methods in writing the article were theoretical analysis of scientific literature on the research problem, graphical method and study of local documentation of the research object in terms of modernisation of production facilities. The main result of the research is the conclusion about the necessity of introducing a new technology – cold rolling (flanging) in order to expand the technological capabilities of the enterprise, to increase the efficiency of production in this area, to gain profit from the supply of thin-walled vessel bottoms to other enterprises to replace European suppliers, and as a consequence an increase in competitive advantage.

Ways of conversion of nuclear power plants that have exhausted their service life at existing sites

A large number of NPP power units designed and built on the territory of Russia and Ukraine have exceeded or are close to exceeding the design operating life, even taking into account their extension. The terms of the extension of operation also have their limits, especially taking into account the types of reactors of the plants being extended. The complete dismantling of old nuclear power plants is an extremely costly undertaking. The article formulates the problem setting for converting double-circuit NPPs into thermal power plants with the possible use of the infrastructure and equipment of the turbine hall. The existing experience of NPP decommissioning is analyzed. Variants of application of boiler equipment, reconstruction of existing turbines and use of gas turbine plants for the implementation of options with a steam-gas cycle are considered. A comparison has been made of various options for completing fossil-fueled power units, their expected power, parameters and efficiency. Qualitatively, additional capital investments in new equipment, environmental impact and cost-effectiveness of the considered options for completing new TPP power units have been assessed. It is noted that the implementation of any of the considered options should be preceded by analysis of the condition of thermal mechanical equipment remaining in operation and a feasibility study of the proposed conversion option.

IMPROVING OPERATIONAL SAFETY OF NPP STEAM GENERATORS BY CONTROLLING STEAM HUMIDITY

The article considers the possibility of increasing the safety of NPP power units operation by improving the quality of steam generator water supply control. The quality of control is improved by introducing a new information signal about steam humidity before the turbine. The necessity of introducing this signal is substantiated. The article describes the properties of the object of regulation by various channels, the initial coordinate for which is steam humidity.

Scenario analysis loss of heat removal from the spent fuel pool on nuclear power plant

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».

OPTIMIZATION OF OPERATING MODES OF POWER UNITS OF NUCLEAR POWER PLANTS

The materials of the article consider the main methodological provisions of calculations and optimization of NPP power unit equipment parameters using mathematical modeling methods. For the effective implementation of tasks related to determining the optimal parameters and structures of NPP power unit equipment using mathematical modeling and well-developed multifactor optimization methods, it is necessary to fulfill a number of requirements at their productions. Practice shows that it is impractical to optimize with the help of a single mathematical (simulation) model the entire set of parameters characterizing a given power unit, because with such a formulation, the optimization tasks are often mutually incorrect due to a significant discrepancy in the accuracy of various source information, unequal influence of parameters on the target function, specific differences in the mathematical description of various units and elements of the power unit. In order to effectively optimize the parameters of NPP power units, it is necessary to create a system of interconnected mathematical models, which include: a group of detailed mathematical models of individual units and elements of power unit equipment; more generalized mathematical models for the main equipment of power units built on their basis; complete mathematical model of power units.

Input data to match model and prototype states of WWER NPP

The paper considers some possible input data of NPP power unit model for solving the problem of model state matching with prototype state. In the course of work on matching the state of NPP model with the prototype, a numerical experiment using full-scale modeling complex PROSTOR was conducted. Its essence was to obtain steady states of the model with capacity from 94% to 104% with a step of 0,05%. For modeling we used data of Kalinin NPP unit 3 loading 4 on 248 effective days. When analyzing the results of the numerical experiment, two effects were found. One of them, expressed in the bend of the derivative of the obtained graph, is connected with reaching the upper limit of the possibility to regulate the pressure in the main steam collector. The regulating valves on the turbine generator are the executing mechanism of pressure regulator in main steam collector. At the break of the graph they open at 100%, which causes further inability to regulate pressure. The pressure in the main steam collector starts to increase. This effect corresponds to the real processes occurring at the NPP. In this case, as one of the model tuning parameters, it is proposed to use the adaptive coefficient to the concentration of boric acid in the reactor, for the control of which, it is proposed to use a mathematical boric regulator. It was also shown that it is impossible to accurately control the pressure in the reactor due to the peculiarities of operation of the volume compensator electric heaters automatics. In this regard, it was proposed to use an additional mathematical volume compensator electric heaters regulator to match the state of the model with the prototype. These regulators have been successfully tested as part of the PROSTOR software package to match the states of the model and the prototype.

Conversion of CPS Channels at the Leningrad NPP Power Units 3 and 4 for the Production of Radionuclides

Conversion of CPS Channels at the Leningrad NPP Power Units 3 and 4 for the Production of Radionuclides

Calculation and Analysis of Technical and Economic Indicators of the NPP Power Unit

Technical and economic indicators (TEI) characterize the efficiency, reliability, and durability of power equipment during its operation. Based on automating the determination of TEI directly during the production of electricity, objective information can be obtained on the efficiency of the flow of technological processes, as well as on the actual state of power equipment. The conditions for optimal operation of power plants are determined in the process of analyzing its technical and economic indicators. Therefore, the analysis of indicators is one of the most important tasks of automated process control systems (APCS) of nuclear power plants, which largely determines the efficiency of the power unit.

ANALYSIS OF MODELS OF AN AUTOMATIC POWER CONTROL SYSTEM FOR A PRESSURIZED WATER REACTOR IN DYNAMIC MODE WITH A CHANGE IN THE STATIC CONTROL PROGRAM

The process of decarbonization of the industry was intensified by the decision to include nuclear energy in the EU Green Taxonomy. The increase in the capacity of nuclear power plants and the attractiveness of this industry for a long time causes interest in the use of nuclear power plants for electricity dispatching. Thus, the development of an automatic control system for changing the power of nuclear power plants in dynamic modes becomes relevant. The energy release along the height of the reactor core under the action of xenon fluctuations is one of the problems in ensuring the safe operation and stability of the NPP power unit when the power level changes. Violation of the energy release safety criteria occurs when non-stationary transient xenon processes with positive feedback occur in the core. Within the framework of the automatic control theory, the modeling of the system and the control object is compiled on the basis of the use of physical and mathematical models. Transfer functions, structure and parameters of the control system are determined. As an alternative, it is proposed to use an approximation model of the process in the core. The model is presented on the basis of transformations in such a way that the results of the calculation on a certain interval coincide or are close to the results of the calculation when using the analytical model. Such a model can approximate a process that depends on two parameters: time and degree of load change. The results of changing the regulation of technological parameters by the automated power control system of nuclear power unit in dynamic mode, the model of which is based on an approximation or physical and mathematical model, were analyzed. The results of changing the parameters are obtained by comparing the results during the transition of control from one static program to another. Switching between static control programs was studied for such process parameters as coolant temperature at the entrance to the reactor core, average coolant temperature and steam pressure in the secondary circuit. The studies were carried out under the condition that the reactor power was reduced to 80 % during the operation of one control program and increased to 100% of the power when another static program was used in reverse.

Methodological Problems in Assessing the Availability Indicators of NPP Power Units at the Design Stage

Methodological Problems in Assessing the Availability Indicators of NPP Power Units at the Design Stage

Predictive evaluation of changes in the volumes of radioactive waste to be deeply dispositioned in activated reactor structures in the process of their storage after the final shutdown of the Belarusian NPP power units

This paper presents the results of computational studies of the amount of solid high-level and long-lived intermediate level radioactive waste (HLW and ILW-LL) generated during neutron activation of structural materials adjacent to the core of the VVER-1200 reactor of the Belarusian NPP, depending on the time after the final shutdown of the reactor. The assessment of the volumes of HLW and ILW-LL of activation origin, formed over 60 years of operation of the VVER-1200 reactor, was carried out on the basis of computational studies of the induced activity of structural and shielding materials using reactor and Monte Carlo program codes (SERPENT 2, TVS-M, DYN3D, MCU- PD). As a result of calculation studies, it was established that the masses and volumes of activated materials, according to the levels of pollution related to HLW and ILW‑LL, within 10 years after the final shutdown of the VVER-1200 reactor of the Belarusian NPP will be 273 tons and 43 m3 , respectively. In the interval of 30–70 years, the masses and volumes of activated materials of these categories of RW will amount to 262 t and 33 m3 , respectively. From 100 years to 200 years, the masses and volumes of activated materials of these categories of RW will be 118 t and 15 m3 , respectively. Within 10 years after the reactor shutdown, isotopes 55Fe (2.7 years), 60Co (5.27 years), 63Ni (96 years) will make the main contribution to the value of the integral specific activity of the materials of the structures of the fence, shaft, surfacing, reactor vessel, block of protective tubes (BPT), the space under the core, thermal insulation; after 10 years – 63Ni. The main contribution to the value of the integral specific activity of materials of absorbing elements (PEL) with Dy2TiO5 during the entire period of storage will be made by 63Ni.

Multicriteria investigation of the efficiency of combining a NPP power unit with a heat storage system

A heat storage system, including an additional low-power steam turbine, can provide effective participation of nuclear power plants in covering daily irregularities in energy consumption in the power system. In addition, the system provides an additional reserve for the NPP’s own needs, due to the possibility of using the energy of residual heat release from the reactor core to supply power to the station’s own needs. In the paper, the methodology developed by the authors is presented, which allows to research ways to improve the efficiency of nuclear power plants based on a comprehensive assessment of technical and economic efficiency, changes in plant safety and effects for the entire state energy system as a whole, achieved through the installation of new equipment. As an example, a thermal storage system, a steam-gas plant, and a system for passive heat removal from the reactor core have been investigated.

STUDY OF THE CONDENSER’S OPERATING CHARACTERISTICS OF THE NPP UNIT

The paper shows that the technical condition of condensing devices of steam turbines largely determines the amount of electricity losses, reliable and economical operation of NPP units, and the modernization of steam turbine capacitors will provide a significant increase in electricity generation with relatively low capital investment compared to construction of new NPP power units. Analysis of the perfection of the heat transfer process in the condenser according to the principles of thermoeconomic diagnostics to identify the causes of abnormal operation of the power conversion system showed that the main causes of load reduction are determined by rising cooling water temperature and deviation of steam pressure from normal assessment of contamination of the heat transfer surface, which significantly affects the reduction of electricity generation. The modernization main points of the capacitor of the Zaporizhzhya NPP power unit № 3 on the principle of "block-modular" design developed by PJSC "Turboatom" and the characteristics of the capacitor provided by the new design are considered. The method of thermal calculation of the condenser with the use of iterative methods, which takes into account the presence of air leakage in the space of the condenser and the appearance of contamination of the heat exchange surface, is given. The increase of electric power at the generator terminals at change of steam load factor and pollution thickness is calculated and analyzed. An analytical dependence of the investigated parameters of reducing the increase in electric power during the modernization of the capacitor is obtained and the interpretation of the results is carried out. The work emphasizes the importance of processing the results of diagnostics and monitoring of the technical condition of steam turbines condensing devices and information about their impact on the amount of electric power generation.

DETERMINATION OF THE DEPENDENCE OF THE NPP UNIT POWER ON THE STEAM TEMPERATURE AT THE OUTLET OF THE SUPERHEATER SEPARATOR FIRST STAGE WITH AN UNCERTAINTY OF INFORMATION

The article discusses the application of the standard statistical approach and interval analysis procedures to the estimation of the parameters of the dependence of the power of an NPP power unit on the steam temperature at the outlet of the first stage of the superheater separator under conditions of measurements noise, unknown probabilistic characteristics of errors and with a short measurements sample. The influence of reducing the error in determining the power of the power unit on the choice of the experiment result processing method is investigated. The situation of measurement error limitation without reliable information about its distribution is considered. The correctness of the application of interval analysis numerical methods for processing experimental data under conditions of uncertainty and a significant decrease in the maximum value of the total error is shown. The data were approximated by the Levenberg-Marquardt minimization method when the relative error of the measurement results equaled 5%.

Influence of the heat transfer coefficient on the level of residual stress after heat treatment of the VVER-1000 reactor baffle

Background. Improvement of the methodology for the computational analysis of residual stresses in the structural elements of the reactor is an integral part of the work when extending the service life of NPP power units.
 Objective. Determine the value of residual technological stress arising in the baffle of a VVER-1000 reactor during welding and postweld heat treatment according to the austenitizing mode. To evaluate the effect of considering the dependence of the heat transfer coefficient on the temperature of the baffle surface at cooling in air during heat treatment.
 Methods. Numerical modeling of the stress-strain state of the baffle during welding and postweld heat treatment was carried out using the finite element method.
 Results. It was determined that in the process of heat treatment according to the austenitizing mode, the residual welding stress in the baffle are almost completely relaxed. Due to the high temperature gradient during rapid cooling in air after heating in the process of austenitization, new rather high residual stresses are formed in the zones of the baffle with the greatest metal thickness.
 Conclusions. Based on the results of the investigation, a high level of residual technological stress was determined, which should be considered when calculating the justification for extending the service life of the VVER-1000 reactor baffle.

Basic bases of calculations and optimization of NPP power unit equipment parameters methods of mathematical modelling

The materials of the article consider the optimization of certain parameters and characteristics of the equipment of NPP power units, which are closely related to the processes of their design and construction. Modern NPP power units are complex technical systems. They include a set of interconnected equipment for different technological purposes, which ensures the performance of power units of a complex function of electricity production and heat of the specified quality and according to a given load schedule. Complete mathematical models of the functional state of steam turbine power units are characterized by a large number of nonlinear connections and contain implicit functions. This complicates their widespread use to solve problems of systematic analysis of the quality of operation of power units. The aim of the work is to analyze the basic theoretical foundations, methods and approaches to the calculation and optimization of the parameters of the equipment of NPP power units by methods of mathematical modeling. The solution of the problem of optimization of NPP power unit parameters includes the following stages: selection of optimality criteria (objective functions); development of a system of interconnected mathematical models in accordance with the required hierarchical level of optimization research; selection of computational methods and optimization algorithms. Taking into account the above methodological provisions and approaches increases the efficiency of mathematical modeling to solve problems of calculations and optimization of NPP power unit parameters.

Simulating operation of power units 1 and 2 at Novovoronezh NPP II with two electrical feed pumps disabled and the backup pump not enabled

Introduction. The article analyzes the operation of Unit 1 and 2 of Novovoronezh Nuclear Power Plant II (equipped with VVER-1200 reactors) with two electrical feed pumps disabled and the backup pump not enabled. These operating conditions are subsequently simulated using the power unit model software-hardware package (PUM SHC) developed by LLC IF SNIIP ATOM. Research objectives. The objective of this work was to check the reliability of the forecasts of changes in the power unit parameters obtained using the PUM SHC, based on operational data. Methods. The simulated power unit parameter changes in transient conditions were in good agreement with the data collected in real tests. During the simulation, the power unit dynamic stability was preserved, i.e., the operational parameters were within the design limits and did not exceed the protection operation set points. Results. The results of the work suggest the possibility of using current NPP power unit simulations: for developing proposals for adjusting the operation control algorithms in case of malfunctions and emergency modes with the main equipment shutdown and power unit protection actuation; and for verifying design solutions for updating the power unit systems, which are associated with the use of new equipment or changes in flow diagrams. Conclusion. Current power unit models can be applied both for existing power units and for new ones that are being commissioned.