Circuit Of Nuclear Power Plant Research Articles

Precipitation of oxide dispersion strength steels after long-term annealing at temperature of 475 °C

Oxide dispersion strengthened steels are key alloys used in nuclear installations. Steels with chromium content of up to 10 wt % are suitable due to their advantageous properties and these alloys are used for the construction of technological devices in the primary circuit of nuclear power plants. From other studies, it can be concluded that chromium has anti-corrosive properties due to the formation of a passivation layer, which results in reduced activation of the material by thermal neutrons. Macroscopic properties are determined by their microstructure, and therefore, the description of the microstructure is important. Transmission Mössbauer spectroscopy and atom probe tomography were used to characterise the material. This provides information about the physical and/or chemical environment of the resonant atoms can be obtained. Obtained spectral parameters reach saturation values from which the solubility limit of chromium in iron can be determined. In Cr-rich phase, the solubility limit can be estimated from the value of spectral parameters of the single-line in the spectrum annealed for the longest time. The suggested procedures are subsequently applied to the case studies of stainless steels suitable for the construction of various components of the III+/IVth generation of nuclear reactors (including fast and fusion reactors).

The whole-life ratcheting behavior of internal pressurized elbow subjected to force and displacement cycling with a damage-coupled kinematic hardening model

In this paper, a study was conducted on the ratcheting-fatigue tests of internal pressurized elbow in the primary circuit of nuclear power plant under force and displacement cycling, and the ratcheting effects at the critical locations of the elbow were evaluated. Based on thermodynamics and continuous damage mechanics theory, the damage evolution rule of the material was derived according to the derivative relationship between damage dissipation potential and damage related variable. A cyclic plastic constitutive model with damage was established by combining the damage evolution rule with Chen-Jiao-Kim (CJK) kinematic hardening rule. The subroutine of this constitutive model was embedded in ANSYS finite element software to numerically evaluate the evolution law of the whole-life ratcheting of the elbow. The results indicate that this damage-coupled kinematic hardening model can describe the whole-life ratcheting behavior of the elbow, and the numerical calculation results and experimental data are basically identical.

Research on the removal of charged colloidal filter materials from coolants in nuclear power plants

In this paper, nano-grade aluminum powder was used as raw material to prepare glass fiber filter material with positive Zeta potential by hydrothermal method. Since colloids of activated corrosion products in primary circuit of nuclear power plant are mainly electronegative, the filter material prepared in this experiment was used to remove colloidal particles in coolant by combining traditional mechanical filtration with electrostatic adsorption, thus improving the removal efficiency of colloidal particles. The leaching problem of soluble silicon in glass fiber filter material was studied. The leaching amount of soluble silicon in glass fiber filter material was decreased while the filtration efficiency was improved. Consequently reduce the risk of safe operation of nuclear power plants.

Study on the Effect of Structural Parameters of Volume Control Tank on Gas–Liquid Mass Transfer

The volume control tank (VCT) is an important facility in the primary circuit of nuclear power plants. During the normal operation of nuclear power plants, the mass transfer between the gas and liquid phases occurs in the VCT at all times. It is driven by submerged jets, which may cause potential risks to the operational safety of nuclear power plants. It is necessary to conduct an in-depth study to gain a deeper understanding of the gas–liquid mass transfer behavior in the VCT. In this paper, a new gas–liquid mass transfer model is developed that combines a surface divergence model with a CFD model to accurately simulate the mass transfer process of the gas phase into the liquid phase. The simulation data were verified by the experimental results. The deviation between the simulation results and experimental results is less than 6.55%. Based on this model, a simulation study was carried out for the effect of structural parameters of the VCT on gas–liquid mass transfer. The results show that the double-vortex structure above the jet inlet, the surface jet at the gas–liquid interface, and the vortex at the end of the jet are the three factors dominating the gas–liquid mass transfer in the VCT. The gas–liquid mass transfer can be influenced by the jet diameter since the jet diameter has a remarkable effect on the Kolmogorov scale and the macroscopic flow field structure. Moreover, both the Kolmogorov scale and the macroscopic flow field structure can be affected by the jet height. However, these two effects cancel each other out. Thus, the influence of the jet height on the gas–liquid mass transfer rate is negligible.

Iron hydrolysis and lithium uptake on mixed-bed ion exchange resin at alkaline pH

The use of ion exchange resins to remove ionic impurities from solution is prevalent in industrial process systems, including in the primary heat transport system (PHTS) purification circuit of nuclear power plants. Despite its extensive use in the nuclear industry, our general understanding of ion exchange cannot fully explain the complex chemistry in ion exchange beds, particularly when operated at or near their saturation limit. This work investigates the behaviour of mixed-bed ion exchange resin, saturated with species representative of corrosion products in a CANDU (Canadian Deuterium Uranium) reactor PHTS, particularly with respect to iron chemistry in the resin bed and the removal of lithium ions from solution. Experiments were performed under deaerated conditions, analogous to normal PHTS operation. The results show interesting iron chemistry, suggesting the hydrolysis of cation resin bound ferrous species and the subsequent formation of either a solid hydrolysis product or the soluble, anionic Fe(OH)3-.

Real-time monitoring of stress corrosion cracking in 304 L stainless steel pipe using acoustic emission

A stress corrosion cracking (SCC) monitoring indicator (MI) based on the acoustic emission (AE) technique is proposed for the real-time detection and monitoring of SCC initiation in 304 L stainless steel (SS) pipes used for the primary circuits of nuclear power plants (NPPs). A device is used to accelerate the initiation of SCC to shorten the experiment timeframe. A non-crack AE signal criterion was established, which was conducted by injecting an etchant and distilled water into 304 L SS pipe specimens as experimental and control groups, respectively. Furthermore, to characterize the AE signal responsible for cracking, the characteristics of the AE signal measured in the section between necking and fracture (where cracks occur and propagate) among the stress curves obtained from the 304 L SS tensile test using AE sensors were investigated. Subsequently, the maximum amplitude, energy, frequency characteristics (range and magnitude), and waveform of the remaining AE signals were investigated after removing some AE signals by applying the non-crack AE signal criterion in the experiment where the SCC penetrated the pipe. These were compared with the maximum amplitude, energy, frequency characteristics, and waveform which were investigated in the tensile test to derive common characteristics for the two sets of crack signals to establish an SCC MI. To evaluate the performance of the developed SCC MI, the SCC monitoring experiment was halted when AE signals corresponding to the SCC MI criteria were measured, to fabricate a pipe wherein non-penetrating SCC was formed. In this manner, by applying the SCC MI to detect SCC in real time, we verified that the SCC MI had excellent SCC detection performance. Furthermore, we performed scanning electron microscopy (SEM) image analyses and SEM-energy dispersive X-ray spectrometry analyses to investigate the causes of SCC propagation.

Influence of organic dispersant on nuclear grade anion exchange resin and their interaction behavior

ABSTRACT Polyacrylic acid (PAA) was used as an organic dispersant in the secondary circuit of nuclear power plants, which could interact with anion exchange resin and affect its properties in the condensate polishing treatment system. This study investigated the kinetics, isotherms, and thermodynamics characteristics of PAA adsorption on anion exchange resin. The results suggested that the inhibition of PAA adsorption on resin can be achieved by lowering the PAA concentration and temperature. In addition, the adsorption reversibility and influence of PAA on anion exchange resin were discussed. The chemical desorption agent (NaOH) can desorb 94.10% of the adsorbed PAA on resin. The physicochemical properties of anion exchange resin changed little after adsorbing and desorbing PAA. Furthermore, the interaction feature between PAA and anion exchange resin was revealed through analyzing the stoichiometry and ATR-FTIR spectra, which was partial ion exchange and most attachment. This feature caused the weak binding force between PAA and resin. Therefore, PAA was easily desorbed from the resin and the exchange capacity of resin can be restored after the chemical desorption process. In conclusion, PAA had little influence on the properties of anion exchange resin and did not cause nonreversible organic fouling.

Актуальні проблеми фізики динамічних пошкоджень в елементах першого контуру реакторів типу ВВЕР

In the context of the actual problems of the physics of operational damage of modern reactor steels produced in the leading countries of the world (USA, Russia, Western Europe) and used for the manufacture of nuclear reactor vessels and other equipment of the first circuit of nuclear power plants, the characteristic features of possible dynamic damage in the responsible elements of this are considered. The mentioned problems are systematized from the standpoint of analyzing the effects of radiation embrittlement, as well as physical and chemical processes that, under certain conditions, are capable of developing in the operating equipment of Ukrainian NPPs, which are already working out their design operational resource. The characteristic features of possible dynamic damage in the operating reactor equipment of Ukrainian and foreign nuclear power plants are considered. The problem is systematized, first of all, from the standpoint of analyzing the operational stability of domestic and foreign reactor steels in relation to their radiation embrittlement. The peculiarities of the course of this physical process have been analyzed, which should be taken into account when determining the maximum possible terms of extension of safe operation of nuclear power units with reactors of the VVER type at the NPP of Ukraine. The main metal-physical properties of reactor steels of various types and possible problems caused by neutron irradiation, physical and chemical processes, vibrational and thermomechanical fatigue, which threaten the unexpected sudden destruction of reactor vessels, are considered. Special attention is paid to mechanical damage and processes accompanying the operation of reactor housings under conditions of cyclic and dynamic loads. A warning has been given regarding the unjustified extension of the period of reactors safe operation. The significant technological lag of the former Soviet, and now Russian, metallurgy from the level of metallurgy of the leading Western countries was noted. Data are provided on the high operational properties of the latest American steels, from which modern reactors of the AR1000 type are manufactured in the USA, and the safety, technical, economic and environmental advantages of using these reactors in Ukraine in comparison with new models of reactors of the VVER-1000 and VVER-1200.

Membrane Covered Probes for O<sub>2</sub> and H<sub>2</sub> Determination - Diffusion through Liquid Films and Solid Membrane in Practice

In the era of the expansion of hydrogen use, its concentration measurement becomes more important. We further focus on one of the H2 concentration measurement purposes, where the hydrogen diffusion in a solid membrane and in a liquid electrolyte play the key role. To keep optimal process conditions in the primary cooling circuit of nuclear power plants, various chemical species are dosed in. Among the species the concentration of which is monitored in primary coolant, belong oxygen and hydrogen. While plenty of companies offer oxygen sensors suitable for the measurement in the primary coolant, the hydrogen sensor, really selective to H2 concentration, is offered by only one company. It is worth, therefore, accomplishing the development of a hydrogen sensor, which began at UCT Prague in the 1990's and, after several successful measurements in nuclear power plant, interrupted due to fateful events in the research team. We introduce here the results of the first part of contemporary work of the Mass Transfer Laboratory based on new technologies but using the experience from 1990's. Having at disposal modern functional samples to measure both oxygen and hydrogen concentrations, we verified a fair long-term stability of the sensors and, further, we would like to cooperate with an industrial partner to finalize the development of prototypes and start the production of monitoring units.

Simulation of mixing flows with different temperatures in a T-junction using OpenFOAM code

This paper aims to evaluate the frequency of velocity and temperature fluctuations in the mixing region using OpenFOAM code. Turbulent mixing of fluids at different temperatures can lead to temperature fluctuations at the pipe material. These fluctuations, or thermal striping, inducing cyclical thermal stresses and resulting thermal fatigue, may cause unexpected failure of pipe material. Therefore, an accurate characterization of temperature fluctuations is important in order to estimate the lifetime of pipe material. Thermal fatigue of the coolant circuits of nuclear power plants is one of the major issues in nuclear safety. To investigate thermal fatigue damage, the OECD/NEA-Vattenfall T-Junction Benchmark was initiated to test the ability of state-of-the-art Computational Fluid Dynamics (CFD) codes to predict the important parameters affecting high-cycle thermal fatigue in mixing tees. In this study, to simulate the standard problem described above, the OpenFOAM code is used, which is an open integrated platform for numerical simulation of problems in continuum mechanics. At the first with Salome-meca code, a computational grid was created, consisting of about 450,000 nodes, and k-eps model and RANS models were used to simulate turbulence. OpenFOAM code results were compared with the available experimental results. The results were found to be in well-agreement with the experimental results in terms of amplitude and frequency of temperature and velocity fluctuations.

Numerical Study of the Thermo-mechanical Behavior of 304L Stainless Steel Pipeline Junctions

The present paper is focused on the behavior of different junctions under the effect of internal pressure and temperature variation. The main junctions studied are of the T-junction and Y-junction type. The goal being to calculate the stress fields coupled to the temperature variation. This coupling realistically defines the thermo-mechanical behavior of the junction and in particular at the intersection where high stress concentrations are observed. The junction is an element widely used in industrial installations such as the cooling circuits of nuclear power plants which are subject to temperature fluctuations due to the mixture of hot fluids and cold these are called mixing zones. These fluctuations can lead to thermal or mechanical fatigue damage and cracking in the circuits causing leaks. Numerical modelling is carried out on the ANSYS calculation code based on the finite element method.

H2 generation at metal oxide particle surfaces under γ-radiation in water

ABSTRACT The formation of hydrogen gas in the water circuits of nuclear power plants has implications for the safe and efficient operation of the plant and so requires thorough investigation. Therefore, typical metal oxides found on water circuit structural material surfaces have been subjected to γ-radiation in water, and the production of H2 by radiolysis has been analysed as a function of oxide composition and water content. The observed radiolytic hydrogen yields, G(H2), were analysed as a function of structural material oxide showing that the presence of the metal oxides increased the rate of hydrogen production relative to neat water radiolysis due to energy absorption by the oxides. The increase in the rate of H2 production was dependent on the oxide composition. The G(H2) value increased at low water weight fractions indicating surface effects are important in the H2 production mechanism.

Методические особенности конечно-элементного исследования прочности трубопроводов водо-водяных энергетических реакторов с учётом локальных утонений от воздействия эрозионно-коррозионных процессов

The methodological features pertinent to strength analysis of the pipelines used in the secondary coolant circuit of nuclear power plants (NPPs) with taking into account local thinning caused by flow accelerated corrosion processes are addressed. The geometric features of pipelines are essential not only for assessing the flow accelerated corrosion intensity in metal, but also for the final strength assessment, in particular, in analyzing the strength of pipelines with walls containing local thinning spots. The pipeline structure lifetime prediction accuracy depends directly on the strength analysis of pipeline elements in which local thinning spots have occurred, especially if these elements themselves have a complex shape. Elbows, sockets, adapters (transitions), and shell rings can be referred to such elements. All of them are thin-walled structures having a complex shape. Software systems like ERCO and PELBOW are best suited for numerically studying them. These computer programs are based on the shell finite element method, using which the thickness at each finite element node can be specified either in the form of certain dependencies or (in some cases) manually. In accordance with the 3D finite element theory of thick-walled shells, a pipeline section can be approximated by a mesh of finite isoparametric 3D elements of a thick shell. The flow accelerated corrosion wear (FACW) depends essentially on the chemical composition of steels. Domestically produced Grade 10Kh9MFB steel is compared with foreign Grade P91 steel (that is close to the former in chemical composition and in performance characteristics), which has positively proven itself and widely applied for making NPP pipelines, with a view to recommend Grade 10Kh9MFB steel for use.

Методика определения времени безопасной эксплуатации оборудования и трубопроводов на основе результатов неразрушающего контроля

The reliability of nuclear power plants (NPPs) has an influence on power generation safety and stability. The reliability of NPP equipment and pipelines (E&P), and the frequency of in-service inspections are directly linked with damage mechanisms and their development rates. Flow accelerated corrosion (FAC) is one of significant factors causing damages to E&P because these components experience the influence of high pressure, temperature, and high flow velocity of the inner medium. The majority of feed and steam path components made of pearlitic steels are prone to this kind of wear. The tube elements used in the coils of high pressure heaters (HPH) operating in the secondary coolant circuit of nuclear power plants equipped with a VVER-1000 reactor plant were taken as the subject of the study. The time dependences of changes in the wall thickness in HPH tube elements are studied proceeding from an analysis of statistical data of in-service nondestructive tests. A method for determining the initial state of the E&P metal wall thickness before the commencement of operation is proposed. The article presents a procedure for predicting the distribution of examined objects' wall thicknesses at different times of operation with determining the occurrence probability of damages caused by flow accelerated corrosion to calculate the time of safe operation until reaching a critical state. A function that determines the boundary of permissible values of the HPH wall thickness distributions is obtained, and it is shown that the intervals of in-service inspections can be increased from 6 years (the actual frequency of inspections) to 9 years, and the next in-service inspection is recommended to be carried out after 7.5 years of operation. A method for determining the existence of FAC-induced local thinning in the examined object has been developed. The developed approaches and obtained study results can be adapted for any pipelines prone to wall thinning to determine the frequency of in-service inspections (including an express analysis based on the results of a single nondestructive in-service test), the safe operation time, and quantitative assessment of the critical value reaching probability.

Key Assembling Issues Relating to Mechanical Vibration of Fabricated Rotor of Large Induction Machines

In many applications, where rotating machines of high power are used, high demands on reliability and safety are laid. Precise manufacturing procedures has to be kept even in case of machines retrofits when e.g. rotors are newly assembled. Even small inaccuracies or misguiding the precise technological procedure can lead to improper running of the machine and it can result in shut-down of complete drive. In case of primary circuits of nuclear power plants, it further means the shut-down of the whole power plant. Consequently, it results in significant financial losses (foregone profit and cost given by problem fixing). The paper presents a methodology of vibration origin investigation in case of large rotating machines which are part of drives like pump, compressors etc. The methodology is based on the detection of undesirable vibration using a diagnostic system on-site and further it uses mathematical modelling of corresponding mechanical parts to reveal the vibration origin. The modelling along with the measurement showed that the detected dangerous vibration is caused by misguided assembling of the rotor of the machine.

Probability prediction of crack growth rate of environmentally assisted cracks of nickel-based alloys based on Latin hypercube sampling

Environmentally assisted cracking (EAC) of nickel-based alloys is one of the most significant potential safety hazards in the primary circuit of nuclear power plants. Considering the uncertainties of load and material properties of base metal and the oxide film, Latin hypercube sampling method is applied to predict the EAC crack growth rate combining the film slip-dissolution/oxidation model with the elastic-plastic finite element method. Meanwhile, to improve the efficiency of numerical analysis, MATLAB is employed in the secondary development for ABAQUS. With the help of existing experimental EAC data, the probability prediction of the growth rate at the tip of EAC in one-inch compact tension specimen is simulated and analyzed in this study. The results show that the randomness of material properties and load markedly affect the uncertainty of strain and EAC growth rate. Among the variables, The randomness of the load has the greatest influence on the uncertainty of the crack growth rate of EAC.

Research on the Micro-Mechanical State at Tip of Environmentally Assisted Cracking Based on Latin Hypercube Sampling Method

Environmentally assisted cracking (EAC) of nickel-based alloys is one of the most significant potential safety hazards in the primary circuit of nuclear power plants. To understand the influence of randomness on micro-mechanical state at tip of EAC, Latin hypercube sampling method is applied to analyze the uncertainty of stress-strain in the oxide film at the EAC tip considering the uncertainties of load and material properties of base metal and oxide film. Meanwhile, to improve the efficiency of numerical analysis, MATLAB is employed in the secondary development for ABAQUS. With the help of finite element numerical simulation and Latin hypercube sampling method, the uncertainty of mechanical properties at tip of EAC in one-inch compact tension specimen is simulated and analyzed in this study. The results show that the randomness of material properties and load markedly affect the uncertainty of micro-mechanical state. Among the variables, The randomness of load has the greatest influence on uncertainty of strain, and Poisson`s ratio of oxide film is the smallest effect.

Prediction of Flow Accelerated Corrosion of NPP Pipeline Elements by Network Simulation Method

Based on a comprehensive approach that uses the computer simulation of the process of destroying structural materials and technology of self-learning neural networks, a methodology has been developed for predicting the rate of flow accelerated corrosion (FAC) of pipeline elements with a single-phase medium of the second circuit of nuclear power plants (NPPs). The neural network model has been implemented in the Delphi Integrated Development Environment. The neural network consists of an input layer containing seven elements and an output layer with two elements. As the input variables of the neural network, the parameters that have the greatest influence on FAC process are chosen. These are the medium temperature, the pipeline internal diameter, the oxygen content in the medium, the coolant flow velocity, the hydrogen index, the time of monitoring (or the start of operation), and the time for which the prediction is performed. For each of the network input parameters, intervals of possible values were chosen. At that, the factors that affect FAC rate, but not included in the feasible model (chromium, copper and molybdenum content in the pipeline material, amine type) are assumed to be permanent. As the output parameters of the neural network, FAC rate and the variation of the pipeline element wall thickness within the predicted time interval have been selected. As the activation function of the neural network the sigmoid function is used. As a method of training the neural network, the error back propagation method has been chosen, which assumes both a forward and reverse passage through the network layers. As the learning algorithm of the neural network, the one with a teacher has been chosen. As a test sample for the neural network, it is proposed, along with operational control data, to use the results of calculations based on a statistical model created in the framework of a special calculation-experimental method. The application of the developed methodology makes it possible to improve the prediction accuracy of FAC rate without determining all the dependencies between the many factors that influence FAC process. The low errors of the constructed models make it possible to use the results of calculations both to determine the resource characteristics of NPP pipelines and optimize operational control.

Effect of Hydrazine on Redistribution of Corrosion Product Deposits in the Primary Circuit of a Shutdown Reactor in Propulsion-Type Nuclear Power Plants

A redistribution process of corrosion product (CP) deposits in the primary circuit of nuclear power plants (NPP) with propulsion service is examined. It is demonstrated that, after a shutdown of the NPP, the activity of activated corrosion products (ACP) and the concentration of stable corrosion products in the coolant increase considerably due to their entering the coolant from the deposits on surfaces of the primary circuit equipment. The corrosion products enter the coolant principally in the form of insoluble compounds, and the suspended solid particles above 0.4 μm in size account for a minimum of 90% of the total content of CPs in the coolant. This increases the CP concentration in the coolant by a factor of approximately 10, and the total activity of ACPs becomes more than 300 times the initial value determined during reactor operation. It was established that injection of hydrazine facilitated redistribution of the corrosion products and radionuclides associated with them between the surface of construction materials and the coolant: the corrosion products went into the coolant and the rate of their redeposition decreased by a factor of ten. This enables us to increase the fraction of highly active ACPs removed from the circuit by the normal cleaning system. In addition, a shift in the equilibrium on injection of hydrazine gives a higher activity of ACPs in the coolant after completion of the deposition process and removal of CPs that entered the circuit in the normal treatment system. The equilibrium activity of ACPs is approximately ten times the value specific for the installation that is cooled down without introduction of hydrazine. A decrease in this characteristic to stable low values is observed only after the reactor is put into operation.

Large eddy simulation of thermal turbulent mixing and reduction of temperature fluctuations with swirl in T-junction

Turbulent mixing of two fluids having a temperature difference can lead to temperature fluctuations in the T-junctions, particularly in the primary cooling circuit of nuclear power plants. Temperature changes produce thermal stresses causing thermal fatigue. This is a key topic related to nuclear security. This article presents the numerical results of two thermal turbulent mixing simulations, applied to a T-junction (OECD-NEA-Vattenfall T-junction). These simulations apply, respectively, to the case of a flow with and without rotational movements into the cold side. The numerical predictions were made using the large eddy simulation with the commercial code Fluent. The numerical results obtained in the first case were compared with the available experimental data and were found in an agreement satisfactory to the profiles of average variables and the spectrum of the temperature near the wall. The comparison with RMS profiles was less satisfactory in the center of the main tube, but was of good quality in the near-wall region. In the second simulation, the main movement of the fluid is affected by the spiral paths created by the swirl velocity caused at the cold inlet, which improves the mixing of hot and cold fluids downstream of the T-junction. From the analysis of these results, it clearly appears that the temperature fluctuations decrease near the wall in this case which leads to less high thermal loads on the tube, and hence a lesser thermal fatigue.