Plant Piping Research Articles

The Casson Dusty Nanofluid: Significance of Darcy–Forchheimer Law, Magnetic Field, and Non-Fourier Heat Flux Model Subject to Stretch Surface

This work aims to offer a mathematical model for two-phase flow that investigates the interaction of Casson nanofluid and dust particles across a stretching surface. MHD Darcy–Forchheimer porous medium and Fourier’s law through Cattaneo–Christove thermal flux are also considered. The governing equations for the two phases model are partial differential equations later transmuted into ordinary ones via similarity transforms. The Runge–Kutta method with the shooting tool is utilized numerically to solve the boundary layer equations computed in MATLAB to obtain numerical results for various pertinent parameters. The numerical outcomes of momentum, temperature, and concentration distribution are visible for both phases. The results of the skin friction, heat transfer coefficients, and the Sherwood number are also visible in the graphs. Furthermore, by comparing the current findings to the existing literature, the validity of the results is confirmed and found to be in good agreement. The fluid velocity is reduced against increasing strength of Casson fluid parameter, enhanced the fluid phase and dust phase fluid temperature. The temperature declines against the growing values of the relaxation time parameter in both phases. Dusty fluids are used in various engineering and manufacturing sectors, including petroleum transportation, car smoke emissions, power plant pipes, and caustic granules in mining.

Effects of the Alloy Design and Fabrication Process on Mechanical Properties of Mo + V-Added SA508 Gr.1A Steel for Main Steam Line Piping in Nuclear Power Plants

Effects of the Alloy Design and Fabrication Process on Mechanical Properties of Mo + V-Added SA508 Gr.1A Steel for Main Steam Line Piping in Nuclear Power Plants

High-Temperature Ultrasound NDE Systems for Continuous Monitoring of Critical Points in Nuclear Power Plants Structures

High temperature pipe cracks are the root of a steam power failure in the EU typically every 4 years, resulting in loss of human life, serious accidents and massive financial losses. According to IAEA’s Reference Technology Database, such an event on a nuclear power plant has an average cost of €120 million, including outage costs, emergency repair costs, insurance and legal costs. Since only one growing crack is needed to cause a major failure, they have to be inspected and monitored thoroughly. Breakdowns at extreme conditions (e.g. 580°C, 400 bar) are a result of two major weld failure modes: a) creep cracks near pipe welds; b) fatigue cracks on pipe welds. Current maintenance practice is to proceed with repairs on a detected crack according to its severity. For cost reasons, cracks that are not judged as severe enough will not be repaired. Crack severity judgement is based on its probability to cause a failure and this probability is derived taking into account the crack size and operational lifetime. More variables such as operating temperature and vibrations may rarely be found in other studies. Recent data from fracture mechanics statistical studies shows this connection between the size of a crack on a nuclear power plant pipe and its probability to lead to a failure. To deal with the above problems two Structural Health Monitoring (SHM) systems have been developed and they are presented in this work. These systems are able to achieve continuous operation for an extended time period at operating temperatures of nuclear power plants. The developed systems employ novel phased array (PA) ultrasonic and ultrasound guided wave (UGW) probes able to withstand and continuously operate even up to 580 °C. The systems are designed to be permanently mounted on superheated steam pipes, at locations of known defects and to continuously monitor their size. However, this supposes that defects will have already been detected by a traditional method during an outage. The PA transducers are placed according to the Time-of- Flight Diffraction (TOFD) technique’s topology, thus creating a novel configuration, while the UGW transducers are placed on a stainless steel ring in a circular array configuration. These configurations can enable continuous tracking of cracks growth with high accuracy, enabling maintenance crews to estimate the severity directly and not through statistics.

Studies of thermal stratification stresses and its impact on fatigue design of NPP piping

Thermal stratification is identified as one of the causes of failure in many nuclear power plant piping. It causes high circumferential temperature gradient in horizontal pipe portions and results in high thermal stresses. In general, most of the piping design codes do not consider these stresses. This paper presents a detail study of thermal stratification stresses and its impact on fatigue life of horizontal portions of NPP piping. The finite element evaluations of thermal stratification stresses are carried out for different constraint conditions at pipe end and for several idealizations of circumferential temperature variation. Simplified equations are developed for calculation of thermal-stratification induced thermal axial and tangential stresses. The thermal stresses evaluated using these developed formulas are compared with the FEM and the test data available in literature. A case study was carried out on a thermally stratified typical NPP surge line. Here the thermal stresses and their impact on fatigue damage was evaluated.

Oxidation Damage Evolution in Low-Cycle Fatigue Life of Niobium-Stabilized Austenitic Stainless Steel.

Austenitic stainless steel is a vital material in various industries, with excellent heat and corrosion resistance, and is widely used in high-temperature environments as a component for internal combustion engines of transportation vehicles or power plant piping. These components or structures are required to be durable against severe load conditions and oxidation damage in high-temperature environments during their service life. In this regard, in particular, oxidation damage and fatigue life are very important influencing factors, while existing studies have focused on materials and fracture behavior. In order to ensure the fatigue life of austenitic stainless steel, therefore, it is necessary to understand the characteristics of the fracture process with microstructural change including oxidation damage according to the temperature condition. In this work, low-cycle fatigue tests were performed at various temperatures to determine the oxidation damage together with the fatigue life of austenitic stainless steel containing niobium. The characteristics of oxidation damage were analyzed through microstructure observations including scanning electron microscope, energy-dispersive X-ray spectroscopy, and the X-ray diffraction patterns. In addition, a unified low-cycle fatigue life model coupled with the fracture mechanism-based lifetime and the Neu-Sehitoglu model for considering the influence of damage by oxidation was proposed. After the low-cycle fatigue tests at temperatures of 200–800 °C and strain amplitudes of 0.4% and 0.5%, the accuracy of the proposed model was verified by comparing the test results with the predicted fatigue life, and the validity by using the oxidation damage parameters for Mar-M247 was confirmed through sensitivity analysis of the parameters applied in the oxidation damage model. As a result, the average thickness of the oxide layer and the penetration length of the oxide intrusion were predicted with a mean error range of 14.7% and 13%, respectively, and the low-cycle fatigue life was predicted with a ±2 factor accuracy at the measurement temperatures under all experimental conditions.

Thermal Fatigue of Pressurized Water Reactor Coolant System Loop Drain Lines Due to Outflow Activities

Abstract Thermal cycling induced fatigue is widely recognized as one of the major contributors to the damage of nuclear plant piping systems, especially at locations where turbulent mixing of flows with different temperature occurs. Thermal fatigue caused by swirl penetration interaction with normally stagnant water layers has been identified as a mechanism that can lead to cracking in dead-ended branch lines attached to pressurized water reactor (PWR) primary coolant system. Electric Power Research Institute (EPRI) has developed screening methods, derived from extensive testing and analysis, to determine which lines are potentially affected as well as evaluation methods to perform evaluations of this thermal fatigue mechanism for the U.S. PWR plants. However, recent industry operating experience (OE) indicates that the model used to predict thermal fatigue due to swirl penetration is not fully understood. There are limitations with the EPRI generic evaluation. In addition, cumulative effects from other thermal transients, especially those resulted from outflow activities, may also contribute to the failure of reactor coolant system (RCS) branch lines. In this paper, we report direct OE from one of our PWR units where thermal fatigue cracking is observed at the RCS loop drain line close to the welded region of the elbow. A conservative analytical approach that takes into account the influence of thermal stratification, in accordance with ASME section III class 1 piping stress method, is also proposed to evaluate the severity of fatigue damage to the RCS drain line, as a result of various transients, particularly the transients from outflow activities. Finally, recommendations are made for future operation and inspection based on results of the evaluation.

Shaking table test and numerical analysis of nuclear piping under low- and high-frequency earthquake motions

A nuclear power plant (NPP) piping is designed against low-frequency earthquakes. However, earthquakes that can occur at NPP sites in the eastern part of the United States, northern Europe, and Korea are high-frequency earthquakes. Therefore, this study conducts bi-directional shaking table tests on actual-scale NPP piping and studies the response characteristics of low- and high-frequency earthquake motions. Such response characteristics are analyzed by comparing several responses that occur in the piping. Also, based on the test results, a piping numerical analysis model is developed and validated. The piping seismic performance under high-frequency earthquakes is derived. Consequently, the high-frequency excitation caused a large amplification in the measured peak acceleration responses compared to the low-frequency excitation. Conversely, concerning relative displacements, strains, and normal stresses, low-frequency excitation responses were larger than high-frequency excitation responses. Main peak relative displacements and peak normal stresses were 60%–69% and 24%–49% smaller in the high-frequency earthquake response than the low-frequency earthquake response. This phenomenon was noticeable when the earthquake motion intensity was large. The piping numerical model simulated the main natural frequencies and relative displacement responses well. Finally, for the stress limit state, the seismic performance for high-frequency earthquakes was about 2.7 times greater than for low-frequency earthquakes.

Water hammer and cavitational hammer in process plant pipe systems

Abstract Fast acting valves are often applied for quick safety shut-down of pipelines for liquids and gases in the chemical and petrochemical industry as well as in power plants and state water supplies. The fast deceleration of the liquid leads to water hammer upstream the valve and to cavitational hammer downstream the fast closing valve. The valve characteristics given by manufacturers are usually measured at steady state flow conditions of the liquid. In comparison, the dynamic characteristics depend on the initial liquid velocity, valve closing velocity, the absolute pipe pressure and the pipe geometry. Fraunhofer UMSICHT conducts various test series examining valve dynamic characteristics in order of the dynamic analysis of pressure surges in fast closing processes. Therefore a test rig is used which consists of two pipelines of DN 50 and DN 100 with an approximate length of 230 m each. In this paper the results of performed pressure surge experiments with fast closing and opening valves will be compared to calculations of commercial software programs such as MONA, FLOWMASTER 2. Thus the calculation software for water supply, power plants, oil and gas and chemical industry can be permanently improved.

High sensitivity ultrasonic NDT technique for detecting creep damage at the early stage in power plant steels

Failures of welded main steam pipes in power station plants are related to cracks forming and propagating in the fine-grained zone, so called heat-affected zone (HAZ). Experience shows that Type IV cracks are indeed some of the most common causes of creep failure for low-alloy ferritic steel pipes and headers in power stations. Typically, crack damage will be characterised by high strain rates in the HAZ and local grain boundary separation, which leads to crack growth. In this paper, the high sensitivity technique for detecting creep damage (Type IV cracking), based on focused ultrasound was implemented. Signal processing algorithms were proposed that enable associating the properties of backscattered signals to the internal characteristics of the metal structure. As the result, the boundaries of the parent metal, HAZ and weld metal in an ultrasonic image can be indicated. The approach was validated on the samples exposed to different levels of creep damage and verified with metallographic analysis. It demonstrated that technique allowed detecting clouds of early stage creep damage.

Influence of 12Cr1MoV Material on Tissue Properties at High Temperature and Long Operating Time

12Cr1MoV is commonly used for pressure pipes in thermal power plants. However, its service life has always prevented the development of such metallic materials. This experiment applies a tensile and impact experiment to investigate the metal cluster of 12Cr1MoV low-alloy and heat-resistant steel with 60,000 h service at 550 °C. Results indicate that, after 60,000 h of high-temperature exposure, the metal cluster of Cr, Mo, and V elements may gradually decrease. First, the decreasing elements will precipitate out of the solid solution. Then, the precipitated elements transform into carbides that accumulate and grow on the grain boundaries. The continuous growth of the precipitated carbide of alloy elements may also create pearlite in the cluster, which results in severe pearlite periodization and tensile fracture due to plastic, through-crystal fracture. Then, the solid solution-strengthened tissue disappears, which severely decreases the thermal strength of 12Cr1MoV low-alloy and heat-resistant steel. At the same time, the brittleness of the steel will increase. The end of life of the metal occurs after 60,000 h of high-temperature use at 550 °C. This result may also provide a basis for future life assessment of 12Cr1MoV steel.

Analysis of Material Loss Behavior According to Long-Term Experiments on LDIE-FAC Multiple Degradation of Carbon Steel Materials

Recently, damage caused by liquid droplet impingement erosion (LDIE) in addition to flow-accelerated corrosion (FAC) has frequently occurred in the secondary side steam piping of nuclear power plants, and the damage-occurring frequency is expected to increase as their operating years’ increase. In order to scrutinize its causes, therefore, an experimental study was conducted to understand how the behavior of LDIE-FAC multiple degradation changes when the piping of nuclear power plants is operated for a long time. Experimental results show that more magnetite was formed on the surface of the carbon steel specimen than on the low-alloy steel specimen, and that the rate of magnetite formation and extinction reached equilibrium due to the complex action of liquid droplet impingement erosion and flow-accelerated corrosion after a certain period of time. Furthermore, it was confirmed at the beginning of the experiment that A106 Gr.B specimen has more mass loss than A335 P22 specimen. After a certain period of time, however, the mass loss tends to be the opposite. This is presumed to have resulted from the magnetite formed on the surface playing a role in suppressing liquid droplet impingement erosion. In addition, it was confirmed that the amount of erosion linearly increases under the conditions in which the formation and extinction of magnetite reach equilibrium.

THE USE OF THERMAL ENERGY OF VENTILATION AND HEATING SYSTEMS IN LIVESTOCK PREMISES

In order to save energy resources when managing microclimate in livestock premises, installations and systems based on renewable energy sources are being developed, in particular, regenerative heat exchangers made using heat pipes. Among the well-known designs of heat exchangers using heat pipes, static designs of heat exchangers using heat pipes have become widespread, since they do not have moving parts, their warranty life is 15-20 years. They are easy to manufacture, have a higher thermal and energy efficiency than heat exchangers with intermediate operating fluid, can be used for both heating and cooling without changing their design. (Research purpose) The research purpose is analyzing the use of heat pipes (thermosiphons) in heat recovery plants of agricultural technological processes and developing a heat exchanger circuit using heat pipes with a thermoelectric heat pump. (Materials and methods) Patent search has been made using the websites of the Eurasian Patent Organization, the World Intellectual Property Organization, Rospatent and Google patents. (Results and discussion) The article presents the analysis of the use of heat pipes (thermosiphons) in heat recovery plants of technological processes of agricultural production. The article presents the developed functional and technological schemes of a heat exchanger using heat pipes for livestock premises. The use of a thermoelectric heat pump for additional heating of the supply air was justified in order to overcome freezing of the surface of the heat exchanger instead of using the electric heater. Smooth-walled gravitational heat pipes (thermosiphons) are the most promising in agriculture. (Conclusions) The process of heat utilization from the removed ventilation air is effectively carried out using a regenerative heat exchanger with heat pipes. The usage of thermal energy from the air removed from the room is possible using the proposed design of a heat exchanger on heat pipes with a thermoelectric heat pump.

Finite Element Analysis of Internally Circumferential Surface Cracked Pipe Stainless Steel under Bending and Torsion Loadings

Early Light Water Reactor Plants (LWR plants) in Japan are nearly 30 years old and utilities have been conducting technical evaluations of facility aging. Accurate prediction of fracture behavior for ductile materials is one of the most important issues to guarantee the integrity of the structure such as pipes in plants. If a crack is detected during the in-service inspections, the structural rational maintenance of the cracked pipe has been promoted based on the defect evaluation methods specified ASME Boiler and Pressure Vessel Code and JSME Rules on Fitness-for-Service for Nuclear Power Plants. Recently, experiments and simulations have been performed on pipes subjected to bending and torsion loadings. However, the failure criterion is still under investigation, and it is significant to propose a method to evaluate it appropriately. In this study, a finite element model of a large circumferentially cracked stainless steel pipe was generated and elastic-plastic analysis of stationary crack and crack extension were performed to validity of the results.

A Method for Improving the Measurement Accuracy of 3D Point Cloud of a Geothermal Power Plant Piping by UAV

A Method for Improving the Measurement Accuracy of 3D Point Cloud of a Geothermal Power Plant Piping by UAV

ТЕПЛОВЫЕ ТРУБЫ В АТОМНОЙ ЭНЕРГЕТИКЕ

The paper provides an overview of technical solutions for using of heat pipes in nuclear power plants both developed and operating. The review based on the scientific, technical and patent literature shows wide application heat pipes as heat transfer devices. Using of them for small and super-small power plants seems to be especially effective, because of high specific cost of plants with circulating coolants. A heat pipe is a device transferrind the heat by means of evaporation and condensation of a coolant circulating automatically under the action of capillar or gravitation forces. Heat pipes are used rather widely, both abroad and in Russia. The first application of a heat pipe principle in nuclear power plants was published in 1957, even before the emergence of the term "heat pipe". Now, there are about 300 patents in the world related to heat pipes application in nuclear power plants. Theare are seweral thouthands articles on the development of nuclear reactors with heat pipes have been published in the scientific and technical literature. One should expect that fifth-generation nuclear reactors cooled by heat pipes without any mechanisms and machines for the circulation of the coolant, as well as without the consumption of mechanical and electrical energy, will be appeared in this decade.

Improvement on optimal design of dynamic absorber for enhancing seismic performance of nuclear piping using adaptive Kriging method

For improving the seismic performance of the nuclear power plant (NPP) piping system, attempts have been made to apply a dynamic absorber (DA). However, the current piping DA design method is limited because it cannot provide the globally optimum values for the target design seismic loading. Therefore, this study proposes a seismic time history analysis-based DA optimal design method for piping. To this end, the Kriging approach is introduced to reduce the numerical cost required for seismic time history analyses. The appropriate design of the experiment method is used to increase the efficiency in securing response data. A gradient-based method is used to efficiently deal with the multi-dimensional unconstrained optimization problem of the DA optimal design. As a result, the proposed method showed an excellent response reduction effect in several responses compared to other optimal design methods. The proposed method showed that the average response reduction rate was about 9% less at the maximum acceleration, about 5% less at the maximum value of the response spectrum, about 9% less at the maximum relative displacement, and about 4% less at the maximum combined stress compared to existing optimal design methods. Therefore, the proposed method enables an effective optimal DA design method for mitigating seismic response in NPP piping in the future.

Diagnosing nuclear power plant pipe wall thinning due to flow accelerated corrosion using a passive, thermal non-destructive evaluation method: Feasibility assessment via numerical experiments

Flow accelerated corrosion (FAC) in nuclear power plant pipes is one of the leading causes of accidents, fatalities, damage and outages. Current FAC identification methods employ expensive sensing technology and are “active” methods, where the response of the piping system to an externally-generated thermal, mechanical or optical excitation must be measured. As a result, these techniques require a disruptive and time-consuming setup. In this article, we propose a method that utilizes pipe surface temperature measurements to passively monitor for FAC-induced pipe wall thinning without the need for expensive equipment or post-installation setup time. This diagnostic method utilizes a simulation data-driven diagnostic model to estimate the amount of thickness reduction in a pipe based on changes in measured steady-state pipe temperatures. In order to reduce the computational burden of generating large, simulation-based datasets, the behavior of the insulation of the pipe was modeled using a suitably calibrated heat transfer boundary parameter. Additionally, global sensitivity analysis was performed to determine system parameter(s), such as the temperature of water flowing inside the pipe, which significantly affect the steady state pipe wall temperature and could cause errors in diagnosis. Two diagnostic models, one using only the change in steady-state temperature as an indicator for FAC-induced pipe wall thinning and the other using water temperature as an additional diagnostic model input were evaluated for their ability to estimate thickness reductions in a pipe using simulated pipe wall temperature data. For the numerical experiments conducted in this work, both models estimated wall thickness with errors within 0.5 mm, indicating that the proposed technique can potentially be used as a low-cost, first-pass method for FAC monitoring.

Numerical investigation on similarity of isothermal and thermal flow mixing in a horizontal T-junction configuration

Turbulent and stratified mixing flows can cause thermal fatigue in nuclear power plant piping systems. An isothermal Mixed-Fluid-Interaction (MFI) test rig is used to investigate mixing phenomena related to High Cycle Thermal Fatigue in nuclear power plant piping systems. Here, the cold heavy branch pipe fluid of the thermal mixing Fluid Structure-Interaction (FSI) facility is modeled by a sugar or salt solution. Due to limited optical accessibility of the FSI facility a numerical similarity comparison of the flow phenomena occurring in both experimental setups (MFI/FSI) is essential. Thus, Large-Eddy Simulations are carried out which are experimentally validated by applying the Wire-Mesh-Sensor measurement technique (MFI) and temperature measurements (FSI). Numerical LES investigation on similarity between two isothermal cases, three thermal mixing cases, and one geometrically similar scaled-up thermal mixing case is performed when the momentum ratio MR and the Richardson number Ri is kept constant. A very good agreement in the spatial distribution of the time averaged mean and rms values as well as statistical fluctuations is observed between the isothermal and thermal cases. Non-dimensional power spectrum density (PSD) of the mixing scalar fluctuation showed an almost unique profile, indicating the same origin for the occurring periodical fluctuation in all cases. The non-dimensional magnitude of the dominant frequency of the mixing scalar fluctuations in terms of the Strouhal number is in the range of 0.38<Sr<0.40. In the context of similarity, it is shown that the momentum ratio and the Richardson number are mainly responsible for the jet-formation and thus for the down- and upstream flow behaviour. The influence of the inlet Reynolds numbers Rem and Reb as well as the density ratio ΓR on the normalized flow characteristics seem to be negligible in comparison.

Seismic performance limit of nuclear power plant piping system with seismic isolation system considering measurement points: Low-cycle fatigue life

This paper presents the first part of a two-part investigation on the seismic performance limits of nuclear power plant piping systems with seismic isolation systems considering the measurement points and proposes a scheme for low-cycle fatigue life. This part of the study focuses on pipe elbows, which are components vulnerable to seismic loads in crossover piping systems connecting structures with and without seismic isolation systems in nuclear power plants. The low-cycle fatigue life, which is the seismic performance limit of the pipe elbow, was determined by considering the cyclic loading of the low-frequency characteristics caused by the seismic isolation system. Three-inch test specimens of SCH40, which are typical standards for nuclear power plant piping systems, are fabricated herein consisting of a pipe elbow and straight pipes. In the test specimens, the pipe elbow where the load is concentrated and the lengths of the straight pipes that are connected to the elbow may differ from those in field conditions. Therefore, their effects on the seismic performance limit were analyzed using the relationship between the moment and deformation angle. Cyclic loadings were performed at varying amounts of a constant amplitude, and an imaging system was used to measure the moments and deformation angles. The deformed shapes of the test specimens caused by external forces were compared for each measurement point, and the leakage lines and low-cycle fatigue curves were presented using the relationship between the moment and relative deformation angle.

Seismic performance limit of nuclear power plant piping system with seismic isolation system considering measurement points: Damage index

This paper presents the second part of a two-part investigation on the seismic performance limits of nuclear power plant piping systems with seismic isolation systems considering the measurement points and aims to quantitatively express these limits using a damage index based on the dissipated energy. This part of the study focuses on a method to quantitatively express the seismic performance limits needed for seismic fragility evaluations of crossover piping systems connecting structures with and without seismic isolation systems in nuclear power plants. Test specimens for this study were manufactured by welding straight pipes to both ends of a 90° long pipe elbow. The dissipated energy was calculated using the relationship between the moment measured at the center of the pipe elbow and relative deformation angle at each of the straight pipe’s measurement points using an imaging system; this was compared to the dissipated energy calculated from the relationship between force and displacement, and a suitable length of the straight pipe for low-cycle fatigue tests was found. Further, a method of applying a weighting factor to the yield points, which are needed to calculate the damage index, are examined; thus, it was confirmed that the damage index can be used to quantitatively express the seismic performance limits of thin-walled steel pipe elbows.