Pipelines In Nuclear Power Plants Research Articles

Effects of fluid dynamics parameters on flow-accelerated corrosion at elbow of carbon steel pipeline

Flow-accelerated corrosion (FAC) has always posed a significant threat to the safe operation of the secondary circuit in nuclear power units. In this study, we investigated typical carbon steel elbow pipe sections susceptible to FAC failure using fluid dynamics software to analyze the hydrodynamic characteristics at varying inlet velocities (2 m s−1, 4 m s−1, and 6 m s−1). The distribution of the FAC rate was monitored in real time using an array electrode. The results revealed that the outermost side of the elbow pipe section was the most susceptible location to FAC. By comparing different fluid dynamic parameters with the FAC rate, we identified radial velocity as an effective parameter for characterizing the FAC rate. Additionally, we established an empirical formula for predicting flow-accelerated corrosion in elbow pipe sections using the least squares method. The implications of this research are pertinent to the design and operation of pipelines in nuclear power plants.

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.

Modern approaches to determining the standardised parameters of vibration state of NPP steam pipelines

During the operation of NPP steam pipelines the main damaging factors are not only corrosion and erosion leading to thinning of the pipe wall but also fatigue damage resulting not only from temperature self-compensation efforts, but also high vibration load. As practice shows, the most promising direction of ensuring the vibration resistance of steam pipelines at present is the computational and experimental study of the stress-strain state of pipelines under vibration loading. This work shows that the strength of the most commonly used standard sizes of steel pipelines under the influence of vibration loads is determined in accordance with Russian State Standard R 59115.9-2021. Taking into account the stress concentration, the amplitudes of the conditional elastic reduced stresses were determined for the 4 variants of the calculation schemes accepted for consideration. The permissible amplitude of stresses from an operating condition of 60 years is determined. It is shown that the permissible voltage amplitude for the design temperature t = 350°C will be no more than 46.4 Mpa for a pipe made of 12X18H10T. As for NPP pipelines, the normalized vibration parameters in the regulatory documentation were not established until 2022, when the following vibration velocity limits were adopted for pipelines of nuclear power plants in Russian State Standard R 59115.11-2021, vibration resistance testing is not required: nmax up to 15 mm/s, nmsv up to 7 mm/s. Thus, currently in the Russian regulatory documentation there are no normalized values of vibration parameters (as a rule, vibration velocity) for pipelines selected depending on the frequency of vibration exposure. It is proposed to develop and introduce into the regulatory documentation frequency-dependent criteria for limiting vibration parameters of NPP pipelines, specified in accordance with the actual operating conditions of these pipelines.

Nuclear power plant pipeline detection robot based on a new radiation-proof material

Soft robots are gradually gaining recognition in recent years for their ability to work in extreme environments. However, radiation shielding is a major issue when designing soft robots for use in nuclear environments. In this paper we present an effective soft robot design and develop new radiation shielding materials that enable the robot to perform a variety of tasks in the complex pipelines of a nuclear power plant while being sufficiently radiation shielded. To achieve this, we used a mixture of silicone and lead powder with excellent radiation shielding capabilities and designed and built a soft robot prototype consisting of multiple pneumatic actuators and a resin frame. It is capable of moving through complex pipelines and performing inspection tasks, while also providing radiation shielding for the internal components of the robot. Finite element simulations and prototype experiments were then carried out to evaluate the robot's performance in various scenarios. Simulation results show that the soft robot is able to perform inspection of nuclear power plant pipelines efficiently and effectively. Furthermore, the radiation shielding provided by the radiation-proof material is effective in protecting the internal components of the robot from radiation exposure. In conclusion, we have successfully developed a soft robot prototype that is capable of performing pipeline inspections in extreme environments such as nuclear power plants, while providing adequate radiation shielding. The proposed approach can be extended to other soft robotics applications in the nuclear environment and has the potential to improve the safety and efficiency of nuclear power plant operations.

Isolated slug traveling in a voided line and impacting at an end orifice

Driven by upstream high-pressure steam, liquid slugs in nuclear power plant pipelines impact the end orifice at high speed, leading to bursting pipelines and threatening the plant's safety. This research aimed to accurately and efficiently assess the dynamic behavior of an isolated slug driven by pressurized air in a voided line with an end orifice. An improved one-dimensional (1D) model for the slug motion and impact was established. The dynamic variation of the pressure at both the slug's tail and front, the variation of the slug length, and the frictional resistance coefficient in the model was obtained by three-dimensional (3D) computational fluid dynamics (CFD). Based on 27 cases with different pipeline diameters and tank pressures, it was observed that the driving air pressure had a quadratic relationship and that the slug length had a constant rate of decrease vs the slug tail displacement. Finally, the decrease in the driving air pressure behind the slug, the increase in the air pressure ahead of the slug, the holdup coefficient, and the friction factor obtained from the 3D CFD results were interpreted in the 1D model, and the velocity histories of the liquid slug were found to be in excellent agreement with the 3D CFD solutions.

Applications of the FEM to pipe whip analysis using coupled modelling technique

Pipelines in Nuclear Power Plants (NPPs) work under high pressure and temperature which may cause a guillotine rupture on the highly pressurized pipe during a sudden fracture due to a failure of any working component. Owing to the serious safety concerns resulting from such failures, it becomes important to understand the behaviour of these pipe structures under dynamic loading, particularly, the pipe whipping effect. Therefore, this research study deals with understanding the structural response of pipe whip with a computationally efficient Finite Element (FE) model.The complete three-dimensional (3D) FE model is modified and coupled with a one-dimensional (1D) FE model (coupled 1D/3D FE model) and solved under the exact transient conditions, and compared with the complete 3D FE model. The computational time of the complete 3D approach was reduced to about 10 times using the coupled 1D/3D model without compromising on the accuracy of the numerical results. Furthermore, the numerical results were found to be in close agreement with the referenced experimental data. This research provides justifiable results from the FE models where the dynamic behaviour of high-pressure pipes under transient conditions can be simulated by using the proposed coupled 1D/3D modelling approach.

Influence of decontamination process parameters on ultrasonic decontamination efficiency of pipeline

Many radioactive pollutants will be deposited inside the nuclear power plant pipelines after long-term service, which will form radioactive hot spots and increase the radiation dose rate level of operators. The ultrasonic decontamination efficiency was studied through the establishment of ultrasonic decontamination dynamic test bench, under different decontamination process parameters. It has also been applied in the engineering filed. The results show that: Ultrasonic decontamination efficiency increases with the increase of decontamination power, and the maximum decontamination efficiency of pollutants is up to 62.69%. The main reason is that the mechanical effect and cavitation effect of ultrasonic are significant with the increase of power, which is conducive to the removal of pollutants in the pipeline. When the decontamination time is 60min, the ultrasonic decontamination efficiency is more than 90%. Meanwhile, after ultrasonic decontamination of the pipeline hot spots, the contact dose rate at the pipeline hot spots is reduced by 75.20% at the highest, and the environmental dose rate is reduced by more than 90%, which reduces the collective dose level of operators working near the pipeline hot spot greatly, and ensures the health of the operators and environmental safety.

Modeling of Pipe Whip Phenomenon Induced by Fast Transients Based on Fluid–Structure Interaction Method Using a Coupled 1D/3D Modeling Approach

The sudden increase in the operating pressure of nuclear power plants (NPPs) is due to the water hammer phenomenon, which tends to produce a whipping effect that causes serious damage to the pipes and their surroundings. The mechanical response of these pipelines under the influence of such fast fluid transients can be estimated using the fluid–structure interaction (FSI) method. The computational time and expense are predominantly dependent on the number of finite elements developed in the model. Hence, an effective modeling technique with limited and efficient nodes and elements is desired to obtain the closest possible results. A coupled 1D/3D finite element modeling approach using the FSI method is proposed to determine the influence of fast transients on the mechanical pipe whipping behavior of gas pipelines in NPPs. The geometric coupled modeling approach utilizes the presence of both the 3D solid elements and the 1D beam elements sharing a local conjunction. The computational model is modelled for a pipe-to-wall impact test scenario taken from the previously conducted French Commissariat a l’Energie Atomique (CEA) pipe whip experiments. The results of displacement, stresses, and impact velocity at the 3D section featuring the elbow are compared for the change in the 3D solid length varied at the juncture of the elbow. The computed results from the Ansys FSI coupling method using the Fluent and Transient Structural modules provides fair validation with the previously conducted experimental results and correlates with the CEA pipe whip tests on pipe-to-wall impact models. Thus, the 1D/3D coupled modeling approach, which minimizes the area of the solid region by constricting it to the impact area with appropriate contact modeling at the junctures, can be considered in the future for decreasing the computational time and the creation of finite elements.

Evaluation of the seismic performance of butt-fusion joint in large diameter polyethylene pipelines by full-scale shaking table test

High-density polyethylene (HDPE) pipelines in nuclear power plants (NPPs) have to meet high requirements for seismic performance. HDPE pipes have been proved to have good seismic performance, but joints are the weak links in the pipelines, and pipeline failures usually initiate from the defects inside the joints. Limited data are available on the seismic performance of butt-fusion joints of HDPE pipelines in NPPs, especially in terms of defects changes inside the joints after earthquakes. In this paper, full-scale shaking table tests were performed on a test section of suspended HDPE pipelines in an NPP, which included straight pipes, elbows, and 10 butt-fusion joints. During the tests, the seismic load-induced strain of the joints was analyzed by strain gauges, and it was much smaller than the internal pressure and self-weight-induced strain. Before and after the shaking table tests, phased array ultrasonic testing (PA-UT) was conducted to detect defects inside the joints. The locations, numbers, and dimensions of the defects were analyzed. It was found that defects were more likely to occur in elbows joints. No new defect was observed after the shaking table tests, and the defects showed no significant growth, indicating the satisfactory seismic performance of the butt-fusion joints.

Automation of the procedure for determining the admissibility of erosion- corrosion wear in pipelines of nuclear power plants made of carbon steels

Many pipelines in Ukrainian nuclear power plants have been in operation for over 30 years, and the determination of the admissibility of erosion-corrosion wear (ECW) defects in the pipeline wall metal is currently one of the pressing issues in the nuclear energy sector of Ukraine. A comparative analysis of determining the admissibility of pipe wall thinning due to ECW was conducted using the adopted methodology “MT-T.0.03.224-18” by the State Enterprise “Energoatom” and the guidance based on the use of regulatory documents for designing pipeline elements. The use of the methodology demonstrated significant advantages in terms of reducing the conservatism of ECW assessment, especially for small-scale thinning and pipelines operating at pressures below 60-80 kgf/cm2, which allows for a substantial reduction in repair work. However, the implementation of the methodology in practice may encounter difficulties due to the complexity and labor-intensiveness of the calculation assessment procedure for nuclear power plant personnel. Therefore, to automate the ECW admissibility assessment in accordance with the methodology requirements, software (SW) was developed for use by nuclear power plant personnel, enabling the prompt generation of a protocol for the express assessment procedure of ECW admissibility for a straight section of the pipeline based on the measurement results of wall thinning parameters in the axial direction. The software allows for concluding whether the section can be put into operation, sent for repair, or requires further refined assessment. Additionally, software based on the finite element method (FEM) was developed, which, when necessary, allows for a refined assessment of a straight section, taking into account the depth and angle of thinning in the circumferential direction, and calculates the stress of the limit plastic state of the pipeline section with ECW from the perspective of viscous fracture. The devel- opment is applicable in the field of nuclear energy in Ukraine.

Statistical model for assessing the reliability of non-destructive testing systems by solving inverse problems

Objectives. The wear monitoring of metal structural elements of power plants—in particular, pipelines of nuclear power plants—is an essential means of ensuring safety during their operation. Monitoring the state of the pipeline by direct inspection requires a considerable amount of labor, as well as, in some cases, the suspension of power plant operation. In order to reduce costs during monitoring measures, it is proposed to use mathematical modeling. This work aimes to develop a mathematical model of a diagnostic system for assessing the probability of detection of defects by solving inverse problems.Methods. A binomial model for assessing the reliability of monitoring, comprising the Berens-Hovey parametric model of the probability of detection of defects and a parametric model based on studying test samples, was analyzed. As an alternative to this binomial model, a computational method for assessing the reliability of non-destructive testing systems by solving an inverse problem was proposed. To determine the parameters of the defect detection probability curve, the model uses data obtained by various monitoring teams over a long period of power plant operation. To serve as initial data, the defect distribution density over one or more of the following characteristics can be used: depth, length, and/or cross-sectional area of the defect. Using the proposed mathematical model, a series of test calculations was performed based on nine combinations of initial data. The combinations differed in the confidence coefficient of the initial monitoring system, the parameters of the distribution of defects, and the sensitivity of the monitoring system.Results. The calculation data were used to construct curves of the probability density of detected defects as a function of the defect size, recover the values of the defect distribution parameters under various test conditions, and estimate the error of recovering the parameters. The degree of imperfection of the system was estimated using the curve of the detection probability of a defect by a certain monitoring system.Conclusions. Under constraints on the data sample size, the proposed methodology allows the metal monitoring results to be applied with greater confidence than currently used methods at the same time as evaluating the efficiency of monitoring carried out by individual test teams or laboratories. In future, this can be used to form the basis of a recommendation of the involvement of a particular team to perform diagnostic work.

Review of Creep-Thermomechanical Fatigue Behavior of Austenitic Stainless Steel

Research on the creep-thermomechanical fatigue (CTMF) behaviors of austenitic stainless steel for nuclear power plant pipelines is reviewed in the present paper. The stress response behavior, the main damage mechanisms, including thermomechanical fatigue, creep, oxidation, and dynamic strain aging (DSA), as well as the effects of strain dwell type, dwell time, and temperature-strain phase angle on fatigue life behavior of austenitic stainless steel under CTMF loading conditions are systematically discussed, and the coupled effects of various damage mechanisms are revealed. It is emphasized that CTMF is closer to the actual service condition of nuclear power plant pipes. It is pointed out that the traditional method of life design based on the isothermal fatigue test data is not conservative. Finally, the research on CTMF behaviors of austenitic stainless steel for nuclear power plant is summarized and prospected.

Investigation on fretting Wear performance of laser cladding WC/Co06 coating on 42CrMo steel for hydraulic damper

Hydraulic dampers are widely used in nuclear power plant pipelines and mechanical equipment as a kind of vibration reduction and energy consumption protection device. The connection part of the pin head (42CrMo steel) is prone to fretting wear. In this paper, the purpose is to prepare composite coatings for improving the wear resistance of pin heads by laser cladding. The micromorphology, phase, hardness and micro-friction wear properties of different WC content (5 wt%, 10 wt%, 15 wt%, 20 wt%) were fully studied. The results show that the elements W, C and Cr are distributed on the white network structure of the cladding layer and play a role in solid solution strengthening; the Co06 cladding coating mainly contains intermetallic compounds such as γ-Co solid solution and Cr23C6. New phases such as WC, Fe3W3C, CoCx, Cr7C3 are formed in the coating, and these new phases play an important role in improving the performance of the cladding layer; The hardness of the coating reaches the highest 810HV0.3 and the lowest is 425HV0.3, which is significantly higher than that of the matrix; The friction coefficient of the coating is significantly lower than that of the matrix, and the wear volume of the cladding layer is also greatly reduced. When the WC content is 15%, the minimum wear volume is 50.7% lower than that of the matrix; The wear mechanism of the matrix is changed from abrasive wear, spalling wear and oxidative wear for oxidative wear and adhesive wear. Taking into account comprehensively, this paper believes that the coating with WC content of 15 wt% has the best wear resistance.

Engineering Support for Thermomechanical Connection of Pneumatic and Hydraulic Pipelines in Nuclear Power Plants

Engineering Support for Thermomechanical Connection of Pneumatic and Hydraulic Pipelines in Nuclear Power Plants

Safety analysis of thermal equipment and piping systems in the CANDU nuclear power plant subjected to thermal shocks

Due to the complexity of nuclear power plants and the risks of damage that may occur during their existence, it is necessary to pay more attention to the safety features. An important influence is the operational temperature of thermal equipment and pipes, especially within the primary circuit, as the coolant is radioactive and in case of damage would cause natural disasters. An assessment of thermal reliability of nuclear power plant equipment and systems is necessary for optimization in the design stage. High thermal shocks with severe and sudden drop in coolant temperature represent a high potential for damage, especially in propagation of existing cracks, under stress due to high pressures. On the other hand, vibrations produced by variable fluid flow conditions or of mechanical origin may induce vibratory fatigue, which combined with thermal fatigue may influence the system’s safety. This study aims to develop computational models for thermal fatigue in pipelines in nuclear power plants, highlighting pressurized thermal shock analysis, temperature distribution diagram along the wall thickness, stress distribution diagram by wall thickness, factor distribution diagram stress intensity, critical crack size diagram, critical reference temperature of nil ductility transition (RTNDT) diagram. Evaluation of these diagrams were performed using the calculation programs Nuclear Piping Integrity Expert System (NPIES) and Reactor Vessel Integrity Evaluation System (RVIES).

Static Analysis and Health Monitoring of Pipelines in Nuclear Power Plants With Noise Consideration

Abstract This study proposes a numerical methodology for the assessment of physical parameters which are critical to damage detection and structural health monitoring of pipelines propagating fluids under high temperatures. The thermal gradient of the fluid, self-weight of the piping system, and the pressure exerted on the pipe walls are taken into consideration for static analysis. The displacement, stresses, and strains are computed for the imposed loads. The boundary conditions for a pipeline element are assigned through translational and rotational springs whose stiffnesses are unknown. The expression for the boundary stiffnesses is derived by coupling the internal pressure due to fluid flow with thermal loading and weight. The dependent and independent parameters for deciphering the degradation of the pipe element are identified to minimize instrumentation for health monitoring. The primary aspects of damage being investigated are wall thinning due to corrosion resulting in stiffness degradation. In practical scenarios, it has been observed that excessive instrumentation of a structural component for health monitoring might result in measurement noise. Hence, artificial noise is introduced to the measurement stresses and strains in a Gaussian distribution form, which is corrected through a numerical algorithm.

Development of quadruped robot for inspection of underground pipelines in nuclear power plants

Structural integrity of seawater pipelines in nuclear power plants is a very important issue. In accordance with the operating technical guidelines, the human operators directly enter the pipe and inspect it at every maintenance test. However, in this regard, safety issues such as narrow space and harmful gas are emerging every year. In response to these needs, a quadruped robot that can inspect underground pipes and assist workers has been developed. The robot has an articulated robotic arm that can receive an impact sound of hammering a pipe wall to test pipe integrity. The state of the pipe was examined using a Convolutional Neural Network algorithm. On the other hand, moving in a plumbing environment requires stable walking ability. To determine the gait sequence, a hierarchical gait controller is proposed. The hybrid controller, which consists of joint impedance and torque control, calculated from Model Predictive Control, can switch the gait modes comparing the reference and the current foot contact condition at each control cycle.

Research on Application of Pulsed Eddy Current Testing in Nuclear Power Plant Pipelines

The condensate pipelines, main water supply pipelines, drain pipelines, and partial extraction pipelines of nuclear power plants are all coated with coatings on the outside of the pipelines to improve heat exchange efficiency. At present, the detection methods of ferromagnetic pipelines are mainly conventional ultrasound and ultrasonic guided waves. It is necessary to remove the insulation layer of the outer wall of the pipeline before the inspection, which leads to a prolonged inspection period and increased labor costs. This paper uses the measured and calculated values of the induced voltage to establish the optimal parameter inversion problem and combines the coupling relationship between the parameters to propose a reliable pulsed eddy current detection method for the relative wall thickness of ferromagnetic pipes. Using the pulsed eddy current detection method for the relative wall thickness of ferromagnetic pipelines proposed in this paper, the scanning detection of nuclear power plant steel pipes is compared with conventional ultrasonic testing. The detection results are reliable and suitable for non-destructive testing and evaluation of nuclear power plant ferromagnetic pipeline wall thickness corrosion thinning.

Influence of Mass Transfer and Physicochemical Processes in the Working Fluid on Erosion–Corrosion of Equipment and Pipelines in Nuclear Power Plants

Influence of Mass Transfer and Physicochemical Processes in the Working Fluid on Erosion–Corrosion of Equipment and Pipelines in Nuclear Power Plants

MATHEMATICAL MODELING OF THE PROCESS OF POSSIBLE DAMAGE TO NPP PIPELINES DUE TO THEIER LOCAL EROSION AND CORROSION SINKING

The article is devoted to computer modeling of the process of possible damage to the pipelines of nuclear power plants that received local thinning of the walls.The geometric features of pipelines are essential not only for assessing the intensity of erosion and corrosion of a metal, but also for the final strength assessment, in particular, when calculating local thinning of the walls of pipelines. The accuracy of forecasts of the life of pipeline structures directly depends on the ability to evaluate the strength of pipeline elements that have received local thinning, especially if these elements themselves have a complex shape. Such elements include elbows, nozzles, adapters (transitions), shells. All of these elements are thin-walled structures of complex shape. Computer modeling of the process of crack formation and growth requires solving several problems. A review and critical analysis of the available experimental information on the growth rates of cracks in specific structural materials and in the environment that is characteristic of given operating conditions is required. An assessment of the crack growth rate is impossible without determining the stresses and strains in the shell structure of complex spatial geometry. The development of routine criteria for crack growth and their attachment to a stress-strain state calculation program is the next task. To solve these problems, a complex of three-dimensional finite element methodologies and programs based on the theory of thick shells has been created. The article presents some results of calculating the stress-strain state of adapters and the growth of cracks in them.