Steam Generators In Power Plants Research Articles

Research on flow induced vibration characteristics of heat transfer tube plugs in nuclear power plant steam generators

Research on flow induced vibration characteristics of heat transfer tube plugs in nuclear power plant steam generators

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.

Comparison of thermal hydraulic performance between horizontal and vertical steam generators in nuclear power plants

Abstract Steam generators (SG) for Nuclear Power Plants (NPPs) are important large-sized metal consuming equipment of nuclear power installations. Efficiency of steam generator operation determines the overall service life of the whole nuclear facility. Vertical steam generators are used normally in North America, Europe and South East Asia, but horizontal steam generators are currently of great importance to reduce building heights and construction costs that govern the overall cost of new nuclear power plants. The objective of the current study is to measure the performance of nuclear steam generators in NPPs through comparative assessments of vertical and horizontal configurations. For this purpose, the vertical U-tube steam generators of a typical 4-loop Pressurized Water Reactor (PWR) was replaced by a VVER-1000 horizontal steam generators to evaluate the horizontal SG performance, at the same plant conditions and characteristics, in comparison with the vertical SG. The thermal hydraulic system code RELAP5 was used in the current work to perform the required analysis, during steady and transient scenarios.

CNN-based automated approach to crack-feature detection in steam cycle components

Periodic manual inspection by trained specialists is an important element of asset management in the nuclear industry. Detection of cracks caused by stress corrosion is an important element of remote visual inspection (RVI) in power plant steam generator components such as boilers, superheaters and reheaters. Challenges exist in the interpretation of RVI footage, such as high degree of concentration for reviewing lengthy and disorienting footage due to narrow field of view offered by endoscope. Deep learning is considered useful to automate crack detection process for improved efficiency and accuracy, and has enjoyed success in related applications. This article utilises a new application of automated crack feature detection in steam cycle components to demonstrate a transferrable data-driven framework for a variety of anomaly inspections in such structures. Specifically, a case study of superheater (a type of reactor pressure vessel head) anomaly inspection is presented to automatically detect regions of crack-like features in inspection footage with a good accuracy of 92.97 % using convolutional neural network (CNN), even in challenging cases. Due to the black-box nature of the CNN classification, the explicability of the classification results is discussed to enhance the trustworthiness of the detection system.

Analysis of the Heat Exchanger Energy Efficiency of Variable Cross Section with an Inhomogeneous Coolant

Abstract One of the main priorities in the modern thermal power engineering development is the problem of energy saving due to the economical use of fuel and energy reserves. Increasing energy consumption with a simultaneous increase in energy prices and widespread environmental degradation necessitates the development and implementation of energy efficient technologies to save fuel, materials and labour costs. The object of study is tubular heat exchangers of variable cross section, which are widely used in steam generators of nuclear power plants, gas turbines and transport plants. Scale deposit properties and the composition of the heat coolant were studied, and their influence on the energy efficiency of heat exchangers was analysed. To study the scale deposits and coolant influence on the energy efficiency of heat exchangers, their properties were examined using an atomic emission analysis with the help of a TESCAN electron microscope. The principles of implementing technologies aimed at intensifying heat transfer, reducing hydraulic and heat losses in heat exchangers were formulated.

Fluid Flow in Helically Coiled Pipes

Helically coiled pipes are widely used in many industrial and engineering applications because of their compactness, larger heat transfer area per unit volume and higher efficiency in heat and mass transfer compared to other pipe geometries. They are commonly encountered in heat exchangers, steam generators in power plants and chemical reactors. The most notable feature of flow in helical pipes is the secondary flow (i.e., the cross-sectional circulatory motion) caused by centrifugal forces due to the curvature. Other important features are the stabilization effects of turbulent flow and the higher Reynolds number at which the transition from a laminar to a turbulent state occurs compared to straight pipes. A survey of the open literature on helical pipe flows shows that a good deal of experimental and theoretical work has been conducted to derive appropriate correlations to predict frictional pressure losses under laminar and turbulent conditions as well as to study the dependence of the flow characteristics and heat transfer capabilities on the Reynolds number, the Nusselt number and the geometrical parameters of the helical pipe. Despite the progress made so far in understanding the flow and heat transfer characteristics of helical pipe flow, there is still much work to be completed to address the more complex problem of multiphase flows and the impact of pipe deformation and corrugation on single- and multiphase flow. The aim of this paper is to provide a review on the state-of-the-art experimental and theoretical research concerning the flow in helically coiled pipes.

Nickel‐based alloy corrosion in CANDU steam generators: E–pH diagrams of the Ni–NH3–H2O and Ni–CH3COO−–H2O ternary systems

AbstractNuclear power plant steam generator (SG) tubes contain nickel as the main alloying element and there is concern about their corrosion. We optimized models for calculating the high‐temperature and high‐pressure thermodynamic properties of common nuclear alloy elements. Subsequently, we calculated the E–pH diagrams for nickel in different concentrations of ammonia or acetate ion at temperatures ranging from 100°C to 260°C and a pressure of 4.7 MPa. This information is used to predict the corrosion behavior of nickel in the secondary circuit conditions of Canadian Deuterium Uranium (CANDU) SGs. Increasing the ammonia or acetate ion concentration resulted in the predominance of Ni(NH3)n2+ or Ni(Ac)n(2−n) at the expense of Ni2+ and NiO, indicating a higher risk of nickel corrosion. Calculations showed that under normal operating conditions with [NH3]tot = 5 × 10−5 and [CH3COO−]tot = 10−8 m at 260°C and 4.7 MPa, nickel will be passivated as NiO, preventing the rapid degradation of nickel‐based alloys. However, in the presence of a crevice that allows the acetate ion to concentrate, nickel would dissolve in the form of the Ni2+ ion, endangering safe operation of CANDU SGs.

Comparative Study on the Corrosion Behavior of Materials in Heat Recovery Steam Generators for Combined Cycle Power Plant, Based on Sulfuric Acid Concentrations: Carbon Steels vs. Duplex Stainless Steel (UNS S32205)

This study examined the effects of sulfuric acid concentrations on the corrosion behaviors of three types of commercial steels, conventional carbon steel (SA192), sulfuric acid dew resistant steel (S-TEN), and duplex stainless steel (S32205), considered to be potential materials for the heat recovery steam generator (HRSG) in combined cycle power plants. Based on the results of electrochemical polarization, immersion and wet-dry cycle tests, when exposed to a high sulfuric acid concentration (50%), SA192 and S-TEN exhibited higher corrosion resistance than S32205, which was due to the formation of a stable FeSO4 scale on the carbon steel materials. With prolonged exposure, S-TEN with slightly higher concentrations of Cu and Sb showed lower weight loss than SA192. Conversely, when subjected to low sulfuric acid concentrations (5 and 10%), S32205 exhibited passivation behavior, and showed significantly superior corrosion resistance (i.e., much smaller weight loss) than the other two steel samples. The examination of localized corrosion damages through cross-section observation using a scanning electron microscope revealed also that S32205 suffered less damages from the wet-dry cycling. The long-term superior corrosion resistance of duplex stainless steel in low sulfuric acid concentrations, compared to carbon steel and sulfuric acid due resistant steel, make it a promising candidate material for the HRSG in combined cycle power plants.

Noise resistant steam generator water level reconstruction for nuclear power plant based on deep residual shrinkage network

Nuclear power plant (NPP) steam generator (SG) water level has strong nonlinear pattern and lagging characteristic under control system and manual adjustment. SG water level signal accidental missing and frequent abnormity may lead to incorrect operational intervention or even emergency shutdown. Deep learning SG water level signal reconstruction method trained with historical sensor data can provide supportive information for operators’ decision-making to decrease the financial cost of maintenance and improve the safety of NPP assets. However, the sensor data transferred by electronic devices can be mixed up with noise in transients or complex conditions. This study proposed a noise resistant SG water level signal reconstruction method based on deep residual shrinkage network (DRSN). The dataset in this study is collected from a power-down process collected from a NPP in China. Results from the experiments under different levels of noise are conducted to illustrate the efficacy and robustness of the proposed approaches compared with other deep learning signal reconstruction methods.

A mechanistic study on dealloying-induced stress corrosion cracking of Alloy 800 in boiling caustic solutions

Alloy 800 (Fe-30Ni-20Cr) is used in steam generators of some nuclear power plants. In-service performance has been excellent, but stress corrosion cracking (SCC) can occur in certain off-chemistry environments. In this study, the SCC mechanism for Alloy 800 in boiling caustic solution is investigated. After exposure, a Ni-rich porous surface film is formed with a large density of cracks. Complementary electron microscopy techniques are applied to study the SCC mechanism at the nanoscale from chemical and mechanical perspectives. Results suggest SCC initiation occurs by transmission of a micro-crack from the Ni-rich layer into the substrate material via a cleavage mechanism.

Enhancing hydraulic stability in once-through steam generators of small nuclear power plants: A novel approach using throttling baffles and feed pipes.

Given the increasing global interest in small nuclear power plants (SNPPs), the proposed research aims to investigate the design of once-through steam generators (OTSGs) used in SNPPs, with a particular focus on ensuring hydraulic stability in the heat exchange tubes. The study will examine the theoretical issues of ensuring stability and propose a new approach that includes the installation of throttling baffles and feed pipes to achieve hydraulic stability in OTSGs. The research will use a 45 MW capacity OTSG as an example, based on the design of the ship nuclear power plant KLT-40S, whose heat exchange surface (HES) is a set of coaxial cylindrical coils, similar to 60% of SNPPs with pressurized water reactors (PWRs). The rest of the SGs have either straight or U-shaped vertical heat exchange tubes (HET). The study will also explore the results of calculating the additional hydraulic resistance in front of the heat exchange tubes and propose a new method for clarifying the existing inequality used to ensure stability. Ultimately, the goal of this research is to provide a novel design for OTSGs based on throttling baffles and feed pipes to ensure hydraulic stability in SNPPs and provide new insights and methods for research in this field.

Universal “Plug and Play” Real-Time Entire Automotive Exhaust Effluents, Industry Vents and Flue Gas Emissions Liquefiers: The Game Changer Approach-Phase Two Category

Universal “Plug and Play” Real-Time Entire Automotive Exhaust Effluents, Industry Vents and Flue Gas Emissions Liquefiers: The Game Changer Approach-Phase Two Category

RISK-INFORMED METHOD FOR QUALIFYING STRATEGIES FOR OPERATION CONTROL OF METAL OF SAFETY RELATED SYSTEMS OF NUCLEAR POWER PLANTS

The optimization of strategies for organizing operation control of metal of equipment and pipelines of safety related systems of nuclear power plants is a pressing issue in ensuring the operation safety and efficiency of nuclear power plants with VVERs. An analysis of regulatory and operational documentation, as well as well-known scientific developments, showed that the optimizing strategies for operation control of metal of safety related systems have not been sufficiently studied. Known strategies for organizing and planning operation control of metal of safety related systems do not take into account many years of experience in operating nuclear power plants with VVER (about 1000 reactor-years), the results of annual monitoring of the technical condition and reliability of equipment and pipelines of safety related systems (including continuous metal monitoring systems), as well as the influence of individual systems on the safety indicators of nuclear power plants and other relevant factors of reliable and safe operation. A risk-informed method has been developed for qualifying modernizations and optimizing strategies for organizing operation control of metal of safety related systems of nuclear power plants. The method is based on an analysis of the results of monitoring metal integrity of equipment and systems, as well as the probability of destruction (failure) of equipment and/or pipelines of safety related systems. Based on the developed risk-informed method for qualifying optimization of the strategy for organizing metal monitoring, the possibility to reduce the control frequency and scope for metal of the emergency feedwater pipeline of steam generators of nuclear power plants with VVER-1000 is substantiated. Further directions for the development of the practical application of the risk-informed method presented in this paper for the calculation qualification of strategies for operation control of metal of safety related systems of nuclear power plants can be related to experimental qualification of the method, as well as the improvement of deterministic methods for modelling processes of degradation/destruction of metal structures in transitional/emergency operating modes of nuclear power plants.

A hybrid sliding mode controller approach for level control in the nuclear power plant steam generators

This work presents a hybrid sliding mode controller approach for level control in the nuclear power plant steam generators. In a nuclear power plant, the steam generator is one of the essential pieces of equipment. Therefore, this paper aims at a robust hybrid scheme that merges internal model control concepts, sliding mode control methodology, and gain scheduling using Takagi–Sugeno multimodel fuzzy systems. Since the process presents integrating and inverse response with dead time and a highly dependent response associated with the operating power variability, this work considers process identification like an optimization problem. Hence, parallel processing algorithms such as particle swarm optimization are used. The performance achieved with the new proposal is suitable for set-point tracking and disturbance rejection.

Experimental analysis of droplet impacting on inclined wall of the corrugated plate dryer in steam generator of nuclear power plants

Corrugated plate dryer (CPD) plays a vital role in the steam generator. In the CPD, the droplet impinging on the corrugated plate (CP) of the CPD, and the behavior after the impacting affects the separation efficiency of the CPD. The mechanical study of droplets movement on corrugated plate provides a theoretical basis for optimizing the design of the CPD. This paper analyzes the experimental phenomenon of droplet impacting on the dry inclined wall (DIW). The maximum spreading diameter (MSD) of droplet on the DIW is analyzed by combining theory with experiment, and a semi-empirical equation is obtained, which is fits well with the experimental data with error within 5%. The secondary droplet is analyzed by pixel analysis. The following conclusions were obtained that with the inclination angle (IA) increasing, the front spreading factor (FSF) will increases, the back spreading factor (BSF) decreases. When We increases, the SF of the droplet impacting the DIW increases, and when the Oh increases, the SF of the droplet impacting the DIW decreases. In the theoretical analysis of the separation performance of the dryer of a nuclear power plant, the influence of the Oh needs to be considered in the early stage of operation. The influence of We is mainly considered in the middle and late stages of the operation process. In the operation of nuclear power plants, artificial intelligence algorithms and control methods can be used to adjust the operating conditions of the CPD inlet droplets to ensure that the steam-water separation efficiency of CPD reaches the highest value. The above conclusion can provide guidance for the design of the CP of the steam generators in nuclear power plants.

MATHEMATICAL RELATIONSHIPS OF MODELS OF TWO-PHASE FLOWS IN CHANNELS OF REACTORS AND STEAM GENERATORS OF NPP’s FOR CALCULATION OF THEIR HYDRODYNAMIC CHARACTERISTICS

Two-phase flows of the working substance in the steam-generating channels of reactors of single-circuit NPP’s and steam generators of two- and three-circuit NPP’s represent the joint movement of liquid (feed water) and steam in the form of a steam-water mixture. The forms of movement of two-phase flows are much more diverse, and the laws of their movement are much more complicated than flows of single-phase media, and they are characterized by a large number of parameters. This is due to the fact that when the flow of the steam-water mixture moves in the steam-generating channels, mass exchange processes occur between the liquid and vapor phases of the mixture, in which the volumes of the phases change; the interfaces between the phases of the mixture vary along the length of the channel and often do not have a distinct character. The cross-sectional mean velocities of the liquid and vapor phases of the flow are not equal to each other, and the difference in the velocities of the phases in the steady-state mode of motion of the two-phase flow depends on the mass flow rate of the steam-water mixture, its pressure, direction of flow, heating intensity, and other indicators. All this complicates experimental studies and calculations of parameters and hydrodynamic (gas-dynamic) characteristics of two-phase flows of a steam-water mixture of a working substance in reactors and steam generators of nuclear power plants. The article discusses mathematical relationships for calculating the hydraulic characteristics of two-phase flows of different structures during the movement of a steam-water mixture in the steam-generating channels of reactors and steam generators of nuclear power plants, namely bubble, slug, rod, dispersed. The ratios are given for various designs of the channels of reactors and steam generators.

MATHEMATICAL RELATIONS OF MODELS OF SINGLE-PHASE FLOWS IN THE CHANNELS OF REACTORS AND STEAM GENERATORS OF NPP’s FOR CALCULATION OF HEAT TRANSFER PROCESSES AND HYDRODYNAMIC CHARACTERISTICS

The article presents the mathematical relationships of models of single-phase flows in channels of a other design for calculating the parameters of heat transfer processes and hydrodynamic (gas-dynamic) characteristics of flows. A comparative analysis of these ratios and their differences are indicated depending on the design of the channels. To calculate the processes of heat exchange in steam generators of nuclear power plants during the movement of single-phase flows of the coolant and working substance, the criterion equations of the similarity theory are used, which describe the generalized relationships between the quantities characterizing the processes that are similar to each other. These are the criteria for Re; Nu; Pr; Pe, et al. The article provides generalized relationships between similarity criteria for single-phase flows of different types of coolants (water, gas, liquid metals) and working substances that are used in the design and construction of reactors and steam generators.

Methods for reducing nitrogen oxide emissions in steam generators of thermal power plants

Main problem: Ensuring environmental safety of thermal power plants by reducing emissions of harmful substances, in particular nitrogen oxides. When all types of fossil fuel, including solid fuel, are burned, nitrogen oxides are formed in the boilers of TPPs. The sources of their formation are air nitrogen and nitrogen- containing components of the organic matter of the fuel. As you know, they adversely affect the health of humans, plants and animals. Therefore, it became necessary to consider and analyze methods to reduce these emissions. Purpose: To review and analyze various ways to reduce nitrogen oxide emissions and propose a new scheme for reducing these emissions by recirculating flue gases. Methods: This is achieved due to the fact that in the known method for purifying the flue gases of steam generators from nitrogen oxides by lowering the temperature in the furnace of the steam generator by supplying flue gases with a temperature below the temperature in the furnace of the steam generator, it is proposed that the flue gases be fed into the furnace of the steam generator after ash cleaning. At the same time, as a result of the supply of recirculated gas cooled after filtering and passing through the main smoke exhauster into the combustion chamber, having a temperature of 110-170 °C, in comparison with the initial version, a greater decrease in temperature in the furnace of the steam generator occurs, which in turn leads to a decrease in the flue gases of oxides nitrogen, since the chemical reaction of their formation goes with the absorption of heat. Results and their importance: The technical result at the proposed method of cleaning from nitrogen oxides is to reduce the consumption of electricity by eliminating the collateral wear of the recirculation gas duct, due to the cleaning of time gases from ash (a requirement of the rules of technical operation of PTE).

DFT studies on the interaction of Fe2+/Fe3O4(1 1 1) and OH-/Fe3O4(1 1 1) during the adsorption process in the steam generators of nuclear power plants

Based on the fouling problem on the outer surface of the steam generator tube in the secondary circuit condition of pressurized water reactors, the density functional theory was used to calculate the adsorption characteristics of Fe2+, OH− on Fe3O4 surface. Eight kinds of highly symmetric adsorption configurations of Fe2+ and OH− on Fe3O4 (1 1 1) surface under two kinds of Fetet1- and Feoct2-terminated surfaces were established. Meantime, the interaction mechanisms and electronic structures of Fe2+/Fe3O4(1 1 1), OH−/Fe3O4(1 1 1) before and after the adsorption process were studied. The results demonstrated that the most stable adsorption site of Fe2+/Fe3O4(1 1 1) is the Fe3O4(111) tet-b configuration. After Fe2+ adsorbed on Fe3O4(111) tet-b sites, the lengths of Fe-Fe and Fe-O bonds and the Fe-Fe-O angle changed. The parallel and vertical directions of top atoms in Fe3O4(1 1 1) surface changed and the shift of topmost atoms was the largest, indicating that the Fe3O4 crystal lattice parameters can be affected by the Fe2+ adsorption. The charge population, density of states, and electron local function results indicated that there were 0.5e to 0.81e electron transfer between Fe2+ and Fe3O4 (1 1 1). The electron transfer of OH-/Fe3O4(111) oct-Oh is from 0.18e to 0.63e. Both of the Fe2+ and OH- can form the chemical bond on the Fe3O4(1 1 1) surface. The results illustrate the nucleation and agglomeration of Fe3O4 scale and provide the foundation for the further research on Fe3O4 scale deposition.

Diverter damper force analysis and stress intensity factors for its double fillet welds by boundary element method

Mechanisms usually transmit power by welded links into a shaft. Should this weld be partial, like fillet welds, there will be an area of stress concentration that could behave like a crack due to sharp corners. These can lead to structural failures due to fatigue. Consequently, to properly design, inspect and maintain these mechanisms it is necessary to use fracture mechanics stress intensity factors. The crack mesh simplicity required by the boundary element method allows easy modeling even for complex geometry. This technique will be used to obtain such parameters for a double fillet weld found in a diverter damper mechanism responsible for controlling hot gas flow to heat recovery steam generator in power plants. But the solutions presented can be used for similar geometrical configurations, specially to shaft link mechanisms. Empirical equations were also obtained to further ease fracture mechanics and fatigue analysis for this kind of geometry.