Nuclear Power Plant Equipment Research Articles

Evolution mechanism of internal flow in the hump region and hump optimization of axial-flow reactor coolant pump

The Nuclear Reactor Coolant Pump (RCP) serves as a pivotal equipment in nuclear power plants, and its operational reliability is directly linked to the overall safety of the plant. When operating in the hump region, axial-flow reactor coolant pumps inevitably experience significant vibration and noise. Therefore, it is necessary to study the internal flow characteristics in the hump region to achieve effective control and increase the hump margin, keeping it away from the normal operating region. This study conducted analyses of pressure pulsation characteristics under various flow rates through experimentation. Additionally, numerical simulations were employed to scrutinize the internal flow of the pump under corresponding flow rates. Research has found that the disturbance of intermediate frequency signals is the main reason for the increase in pressure fluctuations in the hump region, rooted in the annular flow at the impeller inlet and vortex phenomena in the flow channel. Subsequently, the performance of the hump region was improved by optimizing the design of the vane inlet attack angle. The conclusions drawn from this study provide insights and a foundation for the optimal design of reactor coolant pumps.

Assessment of short-term supply disruption impacts for the Swiss mobility, energy and ICT sectors – Application of the SPOTTER approach

In this paper, supply disruption impacts along supply chains within the Swiss mobility, energy and ICT sectors are assessed in the short-term (i.e. the next 5 years) using the recently developed SPOTTER approach with the objective to identify hotspots and technologies associated with comparably low supply risks. SPOTTER has been chosen because, compared to other approaches, it is the only one that allows for an assessment along the full supply chain and it is integrated into the Life Cycle Sustainability Assessment framework. To decrease the respective data acquisition efforts for the here-performed supply risk analysis of entire sectors, in a first step, material/product flows that are most influential for an assessment with SPOTTER are identified and considered for the calculation of related impact scores. Our study analyzes critical material and product flows between trade partner countries along relevant supply chains. It thus identifies new hotspots and adds relevant country-specific information to current knowledge. In particular, the identified hotspots refer to volatile prices of solar panels, interrupted trade of oil and gas from Russia, Nigeria and Niger as well as disrupted imports of electronic devices from China. Suggestions for mitigation strategies comprise hedging and/or diversifying the supply chains. Other identified hotspots refer to the critical supply of cobalt and natural graphite along battery supply chains, which could for example be addressed by supporting and/or incentivizing the construction of a circular economy. A third category of identified hotspots indicates risks related to the trade of hafnium and nuclear power plant equipment that could be mitigated by shifting to less risky energy technologies. Our comparisons of overall impacts between relevant technologies suggest higher supply risks for conventional cars, solar panels and Chinese laptops compared to battery electric cars, wind turbines and German laptops. In these comparisons, more details regarding disruptions along full product and fuel supply chains are considered than in current studies.

Experimental Study of the Influence of Supplementary Reinforcement on Tensile Breakout Capacity of Headed Anchors in Nuclear Power Plant Equipment Foundations

Anchor bolts are often used in nuclear power plants to connect equipment and equipment foundations. Under a severe earthquake, tensile breakout failure is prone to occur in the anchor bolts. As the total amount of installed machines rises, the inertial forces transferred to the anchor bolts under seismic loads also increase significantly. Therefore, the capacity is no longer satisfied by concrete alone, and specialized supplementary reinforcement needs to be installed around the bolts. The study analyzed the tensile behavior of anchor bolts in foundations with supplementary reinforcement experimentally. A total of 16 single-headed anchors in RC foundations with various diameters, yield strengths, and forms of supplementary reinforcement were tested under monotonic tensile loading. The results show that supplemental tie bars and supplemental U-shaped bars, respectively, rely on the bond with the concrete and their own tensile strength to increase the tensile breakout capacity. Furthermore, based on the failure mechanism, a new model considering the terms of concrete resistance and reinforcement resistance for the tensile breakout capacity of headed anchors around with supplementary reinforcement was proposed. Compared with the strut–tie model by EN 1992-4:2018, the predicted results of the model proposed by this study are relatively consistent with the experimental results, while the results by EN 1992-4:2018 are overly conservative.

Risk-based method for substantiating reduction of pressure of hydraulic strength tests and its application on steam line of nuclear power plant with WWER (PWR) reactor plant

Hydraulic tests are used in various sectors of the national economy to check the strength and tightness of vessels and pipelines, their parts and assembly units loaded with pressure. The hydraulic test pressure is a calculated value and is determined on the basis of regulatory documentation for the relevant control objects. Analysis of current trends in the field of hydraulic testing both in the Russian Federation and abroad has shown a trend to reduce the damageability of test objects by choosing the most optimal hydraulic test pressure. The main purpose of the research was to reduce the operational damage to equipment and pipelines of nuclear power plants by reducing the pressure of hydraulic strength tests during maintenance. The research is based on probability theory, mathematical statistics, strength physics and fracture mechanics, taking into account the possibility of occasional discontinuities or defects that might be observed in operation. The method is based on a risk-based approach, which makes it possible to determine the permissible risk variation based on recommendations on the safety of nuclear power plants in the Russian Federation. The objects of research are the equipment and pipelines of nuclear power plants. The result of the research is the development of a method substantiating the reduction of the pressure of hydraulic strength tests for equipment and pipelines of nuclear power plants based on a risk-based approach. The developed method can be used by organizations involved in the production of electricity (at nuclear or thermal power plants), transportation of oil, gas or other substances, at facilities where pipelines are an essential part of their structures, affecting the reliability and efficiency of operation of these facilities.

The Evolving Technological Framework and Emerging Trends in Electrical Intelligence within Nuclear Power Facilities

This paper thoroughly explores the feasibility of integrating a variety of intelligent electrical equipment and smart maintenance technologies within nuclear power plants to enhance the currently limited level of intelligence of these systems and better support operational and maintenance tasks. Initially, this paper outlines the demands and challenges of intelligent electrical systems in nuclear power plants, highlighting the current state of development of intelligent electrical systems, including new applications of artificial intelligence and big data technologies in power grid companies, such as intelligent defect recognition through image recognition, intelligence-assisted inspections, and intelligent production commands. This paper then provides a detailed introduction to the architecture of intelligent electrical equipment, encompassing the smart electrical equipment layer, the smart control system layer, and the cloud platform layer. It discusses the intelligentization of medium- and low-voltage electrical equipment, such as smart circuit breakers, smart switchgear, and low-voltage distribution systems, emphasizing the importance of intelligentization in improving the safety, reliability, and maintenance efficiency of medium- and low-voltage distribution equipment in nuclear power plants. Furthermore, this paper addresses issues in the intelligentization of nuclear power plant electrical systems, such as information silos, the inefficiency of traditional manual inspection processes, and the lack of comprehensive intelligent design and evaluation standards, proposing corresponding solutions. Additionally, this paper presents the trends in intelligent operation and maintenance technology and applications, including primary and secondary fusion technology, intelligent patrol system architecture, intelligent inspection based on non-destructive testing, and a comprehensive solution based on inspection robots. The application of these technologies aids in achieving automated inspection, real-time monitoring, and the intelligent diagnosis of electrical equipment in nuclear power plants. Finally, this paper proposes basic principles for the development of intelligent electrical systems in nuclear power plants, including intelligent architecture, the evolutionary path, and phased goals and key technologies. It emphasizes the gradual transition from automation to digitization and then to intelligentization and presents a specific implementation plan for the intelligentization of the electrical systems in nuclear power plants. This paper concludes with a summary of short-term and long-term goals for improving the performance of nuclear power plant electrical systems through intelligent technologies and prospects for the application of intelligent technologies in the operation and maintenance of nuclear power plants in the future.

Enhancing equipment qualification testing facility for nuclear power plants: Achieving rapid temperature and pressure increase during design basis events

The present study focuses on the development of an equipment qualification (EQ) testing facility for Class 1E equipment in nuclear power plants (NPPs), emphasizing the need to ensure safety functions under design basis events (DBEs). The Republic of Korea (ROK) has implemented international safety standards to support large NPP projects, necessitating the development of domestic EQ testing facilities to reduce the dependence on foreign facilities. To address this need, herein, a specialized facility capable of simulating harsh DBE conditions that may occur in NPPs in ROK was constructed. Through exhaustive research, target temperature and pressure profiles were developed for pressurized water reactors, and a superheated steam injection technique was devised for rapid temperature and pressure changes. The present study contributes to the advancement of domestic EQ testing capabilities, achieving up to 230 °C and 700 kPa within 30 s.

Non-uniform boiling heat transfer characteristics and calculation evaluation in U-tube steam generator tube bundle

U-tube steam generator, as one of the core equipment in nuclear power plants, has lateral non-uniform heating on the secondary side, which complicates the progress of heat transfer on the secondary side of the evaporator by causing fluid cross mixing. However, there is relatively little research on the properties of transverse non-uniform heat transfer in the secondary side tube bundle of the evaporator under natural circulation conditions with the free liquid surface. Therefore, to conduct comparative experiments and studies between uniform and non-uniform heating, an experimental setup with the natural circulation of the free liquid surface and the ability of transverse non-uniform heating was constructed, considering the characteristics mentioned above for the secondary side of the evaporator. The results indicate that higher proportions of gas-phase distribution are frequently found in the tube bundle on the high-power side, and corresponding positions usually have higher water temperature, heat transfer capacity, and heat transfer coefficient. Besides, four heat transfer correlations that are suitable for pool boiling are estimated, and the results show that the Rohsenow correlation has the best applicability to compute the heat transfer coefficient for uniform heating experiments while performing worse in non-uniform heating experiments. The main reason is the lack of consideration for the factors of fluid cross mixing in traditional correlations.

Technical Language Processing of Nuclear Power Plants Equipment Reliability Data

Operating nuclear power plants (NPPs) generate and collect large amounts of equipment reliability (ER) element data that contain information about the status of components, assets, and systems. Some of this information is in textual form where the occurrence of abnormal events or maintenance activities are described. Analyses of NPP textual data via natural language processing (NLP) methods have expanded in the last decade, and only recently the true potential of such analyses has emerged. So far, applications of NLP methods have been mostly limited to classification and prediction in order to identify the nature of the given textual element (e.g., safety or non-safety relevant). In this paper, we target a more complex problem: the automatic generation of knowledge based on a textual element in order to assist system engineers in assessing an asset’s historical health performance. The goal is to assist system engineers in the identification of anomalous behaviors, cause–effect relations between events, and their potential consequences, and to support decision-making such as the planning and scheduling of maintenance activities. “Knowledge extraction” is a very broad concept whose definition may vary depending on the application context. In our particular context, it refers to the process of examining an ER textual element to identify the systems or assets it mentions and the type of event it describes (e.g., component failure or maintenance activity). In addition, we wish to identify details such as measured quantities and temporal or cause–effect relations between events. This paper describes how ER textual data elements are first preprocessed to handle typos, acronyms, and abbreviations, then machine learning (ML) and rule-based algorithms are employed to identify physical entities (e.g., systems, assets, and components) and specific phenomena (e.g., failure or degradation). A few applications relevant from an NPP ER point of view are presented as well.

Multiobjective optimization of laser welding parameters for P92 steel based on MFO/MOEAD-Kriging

P92 steel is used in steam equipment of thermal power plants and nuclear power plants. It needs to withstand high temperature and high pressure during service, so high-quality welds are required. Laser welding can easily obtain good weld quality and reduce the probability of coarse grains and welding defects in the heat-affected zone. Post-weld deformation, residual stress, and weld hardness are important indexes for evaluating weld quality, which largely depends on the combination of laser welding process parameters. The multiobjective optimization methods of moth-flame optimization (MFO) and multiobjective evolutionary algorithm based on decomposition under the Kriging model were compared in this research. It was found that the MFO algorithm has a faster optimization speed and better overall effect. The Optimal Latin hypercube sampling method was used to design the simulation test. The welding seam-forming process was simulated by the SYSWELD software. The Kriging model was used to make the nonlinear relationship between the process parameters and the weld quality indexes significant. A new data set was designed to evaluate the accuracy of the model. On this basis, the efficiency of the combination of process parameters optimized by the two algorithms was compared. The simulation and experimental results were in good agreement with the multiobjective optimization results of the MFO. In the experimental verification, the optimized weld quality was analyzed from the aspects of microstructure and mechanical properties. The results showed that the multiobjective optimization method quickly and effectively reduces the probability of weld quality defects in actual welding, thereby improving weld quality. This multiobjective optimization method provides valuable research ideas for engineering research and development to effectively shorten its development cycle.

Predictive maintenance system for high-end equipment in nuclear power plant under limited degradation knowledge

To ensure the safe operation of high-end equipment, the three-stage maintenance strategy comprising unplanned shutdown, temporary shutdown, and scheduled shutdown is currently employed in nuclear power plants. However, this strategy hinders the acquisition of degradation knowledge (run-to-failure data or degradation mechanism), thereby impeding the application of traditional predictive maintenance systems. Hence, the responsibility for determining the maintenance stage primarily lies with experienced field engineers, and an incorrect decision could potentially result in an unplanned shutdown. To this challenge, we integrate the three-stage maintenance strategy and prognosis methods to form a predictive maintenance system for nuclear power plants. The system framework is first established, and prognosis methods, including sensor selection, degradation trend prediction for short-term prognosis, and remaining useful life (RUL) prediction for long-term prognosis, are then developed under limited degradation knowledge. Finally, the system is deployed in the circulating water pump of a nuclear power plant utilizing an Internet of Things (IoT) architecture. An industry case study verifies that the proposed system can provide decision-making support and further achieve predictive maintenance for high-end equipment in nuclear power plants.

Unbalanced force analysis on valve core clamping in hydraulic spool valves

Hydraulic systems are widely used in nuclear power plant equipments. The reliability and stability of hydraulic systems can be seriously impacted by the valve core clamping resulting from unbalanced forces. This paper investigates the unbalanced force characteristics of hydraulic spool valve cores to identify the mechanisms behind such failures. Numerical models are established for hydraulic spool valve cores using sliding mesh technology with varying control face structures and space angles between the inlet and outlet, which are validated by experiments. The turbulent flow is modeled using the RNG k-ε model. Simulation results reveal that larger openings weaken the unbalanced force distribution while the larger inlet flow rates exacerbate the unbalanced force distribution, and control face structures significantly impact force distribution by altering the throttling effect of the throttle orifice. Furthermore, the impact of the space angles between the inlet and outlet is little compared with other parameters. These findings provide a reference for designing and optimizing hydraulic spool valves, contributing to higher reliability and stability of hydraulic systems.

Relevance of monitoring residual stresses and the need to reduce them

The objective of the study is to show the significant impact of taking into account the actual values of residual stresses on the strength and resource of safe operation of nuclear power facilities. Generally accepted approaches of strength physics and fracture mechanics were used to determine strength and service life. The degree of influence of residual stresses on the strength and service life of pipeline elements in the case of a cyclically acting load, under static load, when there is a danger of brittle fracture, and in the case of corrosion is investigated. The objects of the study were pipelines of nuclear power plants. It is noted that the current state of science and technology has made it possible to create a technique for non-destructive testing of mechanical stresses acting in metal structures based on the use of the acoustoelasticity method. Currently, on the basis of this method, a method for measuring residual welding and installation stresses in pipelines has been developed and certified for use in the field of atomic energy use. The measurement technique provides for the determination of membrane and bending stresses in annular sections located on rectilinear sections of pipelines. The values of stresses in the structural elements of pipelines and residual welding stresses in the deposited metal of welded joints are determined using ratios based on the use of the balancing principle. The relevance of introducing into the procedure for monitoring the technical condition of pipelines and equipment of nuclear power plants the procedure for determining the actual level of residual stresses, which significantly increases the reliability of the assessment of their strength and service life, is shown. It is proved that in order to implement the procedure for determining the actual level of residual stresses, it is necessary to use a non-destructive testing system for residual welding and installation stresses using the acoustoelasticity method based on a measurement technique that allows determining the actual values of residual stresses with established error characteristics.

Ланцюги постачання в атомній промисловості світу та участь у ньому українських підприємств

The article discusses the supply chain of power equipment (in particular, nuclear power plant equipment), which will allow Ukraine to rebuild its economy more efficiently and integrate into the global market for the production of equipment that will help fulfill the commitments to decarbonize the world's economies. The author analyzes the structure of the supply chain in the global nuclear industry and considers the challenges faced by the industry in connection with the introduction of the latest small modular reactor (SMR) technologies and the climate commitments made by countries in accordance with the goals of sustainable development. The expediency of integrating Ukrainian nuclear industry enterprises into the global supply chain is substantiated and the priority steps for this integration are identified.

ANALYSIS OF THE MAIN APPROACHES TO THE ASSESSMENT OF THE CONDITION OF TPP AND NPP EQUIPMENT

The materials of the article analyze the current state of the energy industry in Ukraine and the countries of the world and analyze the ways to increase the efficiency of thermal power plants and nuclear power plants. The elements of the power unit can be affected by aging processes that occur over time or as a result of long-term use. The negative impact of aging can lead to the degradation of the element, namely to the deterioration of its reliability and, as a result, the safety of the power unit as a whole. The development and implementation of measures to mitigate aging processes and to ensure reliable operation of elements during the reassigned resource period is a mandatory type of activity for managing the aging of thermomechanical equipment and pipelines.Monitoring of the technical condition of the elements is performed by monitoring the established parameters and characteristics that determine the technical condition of the elements, during their operation, during tests, measurements, operational control of metal, repairs and maintenance of the elements. Monitoring should be carried out systematically throughout the entire life of the power unit. In turn, technical diagnostics monitors the current state of all elements, detects abnormal conditions, determines the causes of their appearance, which allows you to assess the situation and take measures to eliminate them. The purpose of the article is to analyze the main approaches to assessing the state of energy equipment of thermal and nuclear power plants.

Seismic Response Effect on Base-Isolated Rigid Structures by Mass Eccentricity in Nuclear Plants

The purpose of this paper is to analyze the seismic response effect caused by the mass eccentricity of individual equipment when conducting base isolation for the improvement of the seismic performance of a nuclear power plant. Recent research has interpreted and confirmed through analysis and testing that base isolation for safety-related equipment in nuclear power plants is an efficient alternative to designing for excessive seismic loads. Depending on the equipment, unavoidable mass eccentricity can occur, which necessitates verification of the response impact caused by eccentricity. In this paper, we analyze the seismic response impact of equipment with mass eccentricity using small base isolators. To do so, sensitivity analysis of the seismic response due to mass eccentricity is conducted for a base-isolated concentrated mass model. Furthermore, three efficient mass eccentricity models suitable for testing are designed and manufactured. Simulation analyses using the finite element method (FEM) models are performed, followed by three-axis shake table tests to validate the seismic response impact of mass eccentricity. In conclusion, it is confirmed that applying small base isolators to equipment with mass eccentricity can affect seismic response impact to some extent when compared for a beyond-design-basis earthquake (BDBE).

Health state identification of circulating seawater pump-unit in nuclear power plant based on multi-virtual vibration source fusion in the presence of strong data imbalance

The circulating seawater pump-unit (CRF pump-unit) is a crucial equipment that provides cooling water for the secondary circuit of a nuclear power plant (NPP). Its long-term, fault-free operation ensures nuclear power safety and economy. In this study, considering the limited fault-state samples in the historical database and the random fault locations in NPP equipment, a health state identification framework of the CRF pump-unit is proposed, which combines multiple sources of network input and a feature fusion network. The proposed kernelled independent component analysis with reference (KICA-R) is utilized with reference signals to separate virtual sources from multiple raw acceleration signals. It enhances the fault characteristic frequencies related to several subsystems of a complex mechatronic system in the training samples of the network. Based on the newly designed multi-source channel attention module (MS-CAM) that can adaptively allocate fault-sensitive weight coefficients, a health state identification network incorporates the enhanced input features through the virtual sources obtained from KICA-R. By conducting the benchmark analysis on two public datasets and various types of fault simulation tests on the CRF pump-unit simulator, training sets with different levels of data imbalance are constructed, and the optimal network parameters are determined. The comparisons against cICA validate the ability of KICA-R in virtual source separation, and the strong performance of the MS-CAM-based network in identifying different health states of the CRF pump-unit is further demonstrated through detailed comparative analysis.

АЭС малой мощности с аккумулятором теплоты фазового перехода

Nuclear power plants (NPP) are characterized by low maneuverability. This fact exacerbates the problem of the balance of generated and consumed power when nuclear power plants are used as an energy source. This problem is especially relevant for isolated power systems. One of the ways to increase the maneuverability of nuclear power plants is to use heat accumulators directly in the NPP circuit. The study is carried out for two most attractive options of including a heat accumulator in the NPP circuit: 1) parallel to a high-pressure heater to heat feed water; 2) to generate additional steam for a low-pressure cylinder. The efficiency of each of the options is determined based on the calculation of thermal schemes of nuclear power plants with a heat accumulator. The authors have calculated the power of the turbine unit and electricity generation when a heat accumulator is included in the nuclear power plant circuit. The analysis of the data obtained shows that when using a heat accumulator in the NPP circuit in parallel with high pressure heater to heat feed water, the power of the turbine unit is increased by 7,2 % in the accumulator discharge mode, and additional electricity generation due to accumulated heat is 0,63 MJ for each kilogram of waste steam. When using a heat accumulator to generate additional steam for a low-pressure cylinder, the power of the turbine unit is increased by 6,6 % in the accumulator discharge mode, and the additional power generation due to the accumulated heat is 0,40 MJ per kilogram of waste steam. Application of heat accumulators in the NPP circuits will allow using the accumulated heat to generate additional electricity during high demand hours. Thus, the use of heat accumulators in NPP circuits is a way to increase the maneuverability of nuclear power plants, to reduce the negative impact of maneuverable modes on NPP equipment, and to reduce the number of liquid radioactive waste. The most effective and simple way to use the heat accumulator in the NPP circuit is to install it in parallel to the high-pressure heaters. In future, it is planned to determine the design of the heat accumulator, to model heat exchange processes in the heat accumulator

Study on Unsteady Flow Characteristics of Cooling Water Pump for Nuclear Power Plant Equipment under Low Flow Rate Conditions

During the operation of cooling water pumps, it is necessary to operate them under conditions of low flow rate. In order to improve the unstable performance of the cooling water pump under low flow rate conditions. Taking the cooling water pump as the research object, the internal flow and pressure pulsation characteristics of the cooling water pump under 0.4Q to 0.6Q conditions were investigated, and the influence of different operating conditions on the performance and vibration of the cooling water pump was analyzed. The ANSYS CFX 2022 software and the SST k-ω turbulence model were used to perform a three-dimensional numerical simulation of the cooling water pump. After analyzing the simulation results, the velocity and pressure cloud and streamline diagram within the semi-spiral suction casing and impeller were obtained. The internal flow state of the cooling water pump was then analyzed in detail under low flow rate conditions. At the same time, a series of monitoring points were set up within the impeller, and the pressure pulsation within the impeller was analyzed using the frequency domain diagram and the radial force polar coordinate diagram. The results show that at flow rates between 0.4Q and 0.6Q, a certain amount of vortex has been generated in the suction casing, which affects the flow state when entering the impeller. Furthermore, significant vortices have been generated in the middle and back part of the blade and mainly concentrated in the pump cover of the mid-open pump. At the same time, when in low flow rate conditions, the primary frequency of pressure pulsation is mainly the axial frequency, with three times the axial frequency and blade frequency following. The amplitude of the pressure surface (PS) of the blade is greater than that of the suction surface (SS) and increases as the flow rate decreases. The internal radial force corresponds with the result of pressure pulsation and exhibits a certain pattern. This study outlines the coolant pump’s internal flow and pressure pulsation characteristics under low-flow conditions. It proposes a solution to stabilize the cooling water pump at low flow rates and provides theoretical guidance for optimizing its design.

Improvement of methods for analyzing the technical condition of electromechanical equipment of nuclear power plants

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.

Degradation analysis of horizontal steam generator tube bundles through crack growth due to two-phase flow induced vibration

A correct understanding of vibration-based degradation is crucial from the standpoint of maintenance for Steam Generators (SG) as crucial mechanical equipment in nuclear power plants. This study has established a novel approach to developing a model for investigating tube bundle degradation according to crack growth caused by two-phase Flow-Induced Vibration (FIV). An important step in the approach is to calculate the two-phase flow field parameters between the SG tube bundles in various zones using the porous media model to determine the velocity and vapor volume fraction. Afterward, to determine the vibration properties of the tube bundles, the Fluid-Solid Interaction (FSI) analysis is performed in eighteen thermal-hydraulic zones. Tube bundle degradation based on crack growth using the sixteen most probable initial cracks and within each SG thermal-hydraulic zone is performed to calculate useful lifetime. Large Eddy Simulation (LES) model, Paris law, and Wiener process model are considered to model the turbulent crossflow around the tube bundles, simulation of elliptical crack growth due to the vibration characteristics, and estimation of SG tube bundles degradation, respectively. The analysis shows that the tube deforms most noticeably in the zone with the highest velocity. As a result, cracks propagate more quickly in the tube with a higher height. In all simulations based on different initial crack sizes, it was observed that zone 16 experiences the greatest deformation and, subsequently, the fastest degradation, with a velocity and vapor volume fraction of 0.5 m/s and 0.4, respectively.