Heat Transfer Tube In Steam Generator Research Articles

The effect of NaCl concentration on impact-sliding fretting corrosion behavior of Inconel 690TT heat transfer tubes

Purpose The purpose of this paper is to investigate the effect of NaCl solution with different concentrations on impact-sliding fretting corrosion behavior of Inconel 690TT steam generator heat transfer tubes. Design/methodology/approach The optical 3D profiler was used to measure the wear profile and calculated the wear volume. Corrosion behavior was studied using open circuit potential monitoring and potentiodynamic polarization testing. The morphologies and elemental distributions of wear scars were analyzed using scanning electron microscopy and energy-dispersive spectroscopy. The synergism of wear and corrosion was analyzed according to the ASTM G119 standard. Findings The corrosion tendency reflected by OCP and the corrosion current calculated by Tafel both increased with the increase of NaCl concentration. The total volume loss of the material increased with concentration, and it was known from the synergism that the volume loss caused by corrosion-enhanced wear accounted for the largest proportion, while the wear-enhanced corrosion also made a greater contribution to volume loss than tangential fretting corrosion. Through the analysis of the material morphologies and synergism of wear and corrosion, the damage mechanism was elucidated. Originality/value The research findings can provide reference for impact-sliding fretting corrosion behavior of Inconel 690TT heat transfer tubes in NaCl solution with different concentrations.

Numerical simulation of jet boiling characteristics of high-pressure water injected into high-temperature liquid lead–bismuth in a confined space

The steam generator heat transfer tube rupture (SGTR) accident is potentially devastating and is one of the most serious design basis accidents for lead–bismuth cooled fast reactor (LFR). It is important to study the mechanism of water jet injecting into liquid lead–bismuth eutectic (LBE) for the safety analysis of SGTR accidents. In this paper, based on the volume of fluid (VOF) model, the large eddy simulation (LES) turbulence model and the Lee phase transition model, a three-dimensional numerical model is developed to investigate the typical characteristics of a high-pressure sub-cooled water jet injecting into high-temperature LBE in a confined space. The study focuses on examining the impact of water temperature, inlet pressure, and nozzle diameter on jet development and its surroundings, including the cover gas layer and LBE phase. The results indicate that the central region of the jet is primarily subject to flash boiling, particularly at the nozzle position, whereas the jet interface undergoes heat-transfer boiling. The residual liquid phase migrates upward along the interface and gradually boils. Based on the phase distribution characteristics, a typical jet can be categorized into four zones: the water-steam transition zone, the multiphase flow zone, the end water phase zone, and the steam block zone. Under the conditions analyzed in this paper (water temperature 140–260 °C, injection pressure 2–10 MPa, nozzle diameter 2–10 mm, LBE temperature 400 °C), the jet volume expands approximately exponentially, correlating positively with the three factors studied. It is also found that at high injection pressures and large nozzle diameters, the steam at the top of the jet is prone to destabilization and fragmentation, with steam volume and pressure oscillations being the dominant factors. Steam migration triggers pressure oscillations in the LBE region, and the intensity of these oscillations is closely linked to the migration rate. Under the conditions studied, the maximum pressure oscillation reached 5.5 times the initial pressure. The expansion of the jet steam drives the pressurization of the cover gas layer, which correlates positively with all three factors studied. The nozzle diameter has the greatest effect, increasing the maximum pressure by 168 % from the initial pressure.

Three-dimensional numerical simulation of the interaction between high-pressure sub-cooled water jet impacting high-temperature liquid lead-bismuth metal

The rupture accident of the steam generator heat transfer tube of the lead-bismuth cooled fast reactor causes a high-pressure sub-cooled water jet impacting high-temperature liquid lead-bismuth, and the interaction mechanism between water and liquid lead-bismuth needs to be studied clearly. In this paper, based on the volume of fluid method, the LES turbulence model and the Lee phase transition model, a three-dimensional numerical model is developed to study the multiphase flow and heat transfer mechanism of water and lead-bismuth interaction. The effects of sub-cooled water temperature, inlet pressure, and nozzle diameter on the distribution of phase state, pressure, temperature, and generated steam volume are studied in detail. The results indicate that the water jet can be divided into the jet water zone, the water-steam mixed zone, and the multiphase flow zone, respectively. Among them, the appearance of water-steam mixed zone depends on the internal boiling of the jet, and this zone tends to appear under the condition of low inlet pressure, small nozzle diameter and high sub-cooled water temperature. For multiphase flow zone, the temperature increases with sub-cooled water temperature, decreases with nozzle diameter and is almost not influenced by inlet pressure; The pressure increases with sub-cooled water temperature, inlet pressure and nozzle diameter. When nozzle diameter is large enough, however, it no longer has any effect on the temperature and pressure distribution within the jet. The volume of steam generated by water boiling is positively correlated with all three factors. Overall, the steam migrates upward through the outer side of the jet, absorbing heat and expanding during the migration process, and forming a high-temperature area at the top of the jet.

Fretting wear behavior test and numerical simulation of Inconel 690 alloy

ABSTRACT A steam generator heat transfer tube is important for energy exchange between the primary circuit and secondary circuit . The changes in friction and wear properties of Inconel 690 alloy and 403 stainless-steel anti-vibration strips in different media (air, deionized water, and boric acid solution) were compared and analyzed. Friction and wear tests were conducted in different media environments and under different loads, and finite element simulation was performed in different media environments.Results show that the main wear mechanisms in the air are delamination wear and oxidation friction. The oxygen content in the water is reduced, and the oxide formed is reduced. Because the solution replaces the abrasive particles to form the third body, which can reduce friction and lubricate, the abrasive particles and abrasive particles can be easily removed or dissolved. The wear mechanism in a water environment is delamination wear and abrasive wear. In boric acid, a passive film is formed on the contact wear surface of the Inconel 690 alloy after the previous fretting wear. The passivation film is worn first in the next fretting wear, which slows down wearing. The wear mechanism in a boric acid environment primarily includes abrasive wear and corrosion wear.

Fretting Fatigue Test and Simulation Analysis of Steam Generator Heat Transfer Tube

In a steam generator, the heat transfer tubes are supported by the contact with support plates and anti-vibration bars. The two-phase flow flows over the tubes and causes a vibration when operating. In fatigue analysis, the heat transfer tube is simplified to a beam model, and the contacts with the support plates and the anti-vibration bars are simplified as simple-supported boundary conditions. This linear simplification improves the computational efficiency but cannot simulate the actual situation of the contact area. In consideration of this situation, in the actual analysis, a downwards modified S–N fatigue curve is used to envelop the fretting. For different materials and contact pressure, this modification needs to be obtained through experimental and computational analysis. In this paper, the effect of fretting on fatigue performance of heat transfer tube material 690 alloy is discussed by means of high cycle fretting fatigue test of tube specimen in room temperature air, low cycle fretting fatigue test of sheet specimen in high temperature water environment, and SWT (Smith–Watson–Topper) fretting fatigue predicting simulation, and the conservatism of design fatigue curve is discussed. It is shown that, in range of low cycle and high cycle, the fatigue strength is lower than the mean curve, but it can still be enveloped by the design curve of ASME (the American Society of Mechanical Engineers). However, under the condition of ultra-high cycle, the design curve of ASME can no longer envelop the effect of fretting on fatigue performance, so a further downward modification is necessary to ensure the safety of design.

A coupled model of corrosion of the steam generator heat transfer tube

In this paper, a mechanism model of the coupling of ion migration and fluid dynamics of the heat transfer tube on the secondary side of the steam generator is established. The accuracy of the model was verified, the sensitivity of the migration process of impurity ions to metal corrosion was analyzed and discussed. The results show that migration distance of impurity ions (MDIIs) is highly sensitive to metal corrosion. MDIIs determines the electrochemical reaction (ER) rate of the metal surface. The controlling factor of ER on metal surface changes at 14 m/s, and does not change with pH. The corrosion of the metal surface is more affected by H+ concentration than the overpotential. In this study, a coupling mechanism model based on the corrosion of the heat transfer tube of a SG is proposed for the first time..

The high temperature corrosion of Incoloy 800H alloy at three different atmospheres

ABSTRACT Incoloy 800 H alloy is one of the candidate materials for high-temperature gas-cooled reactor steam generator heat transfer tubes. Previous studies have shown that corrosion phenomena such as oxidation and carburization occurred in high-temperature impure helium gas. On this basis, the corrosion behavior of Incoloy 800 H alloy in air, impure helium, and vacuum was investigated. After the experiment, electronic balance, scanning electron microscope, X-ray energy spectrum analysis, X-ray diffraction analysis, and gas chromatograph were used to characterize the corrosion behavior of the alloy. The results showed that CO is the main oxidation and carburization gas of Incoloy 800 H alloy in impure helium. Contents in ppm CO2 can supplement the reduction of CO and prevent the destruction of oxidation layer and carburization layer; The mass change of the alloy after corrosion in air at 950°C met the oxidation parabola law, the corrosion layer on the alloy surface fell off, and Si migration and accumulation in different degrees occurred in air and impure helium experiments; Carbon transfer and decarburization occurred in the alloy in a vacuum.

A Coupled Model of Corrosion of the Steam Generator Heat Transfer Tube

A Coupled Model of Corrosion of the Steam Generator Heat Transfer Tube

Numerical simulation of micro-crack leakage on steam generator heat transfer tube

The pressure relief rate and loss rate of coolant, which affect largely safety of fission reactor systems, are determined by the flashing process inside cracked passages. Information about flashing phenomenon is of great significance for the leakage online monitoring system of SG (Steam Generator), which ensures the normal operation of SG and safety of the reactor system during early stage of tube rupture accidents. In this research, flashing flow inside a thin micro-crack in heat transfer tube of SG has been studied using ANSYS-FLUENT. The classic cavitation model and evaporation–condensation model, in combination with both the mixture two-phase flow model and Eulerian two-fluid model, are adopted to simulate flashing phenomenon. The real geometry and operation conditions of AP1000 nuclear system are adopted to reflect the reality leakage process in SG. Two types of micro-cracks including axial crack and circumferential crack, which both could happen in the reality operations, are considered. The CFD (Computational Fluid Dynamic) results gained from this research have been compared with test data, and good agreement is demonstrated. The results show that the evaporation–condensation model under Eulerian two-fluid model framework behaves superior to the classic cavitation models in simulating the flashing phenomenon after evaluating accuracy of leakage rate and rationality of thermal hydraulic parameters distribution. In addition, two-phase choking flow phenomenon is obtained by decreasing backpressure of cracked tubes. The simulation in this research reveals the flashing process in short and micro-crack and the results could be good reference for leakage prediction of SG under LBB (Leak Before Break) operations to improve operation performance of SG and safety of the whole nuclear power system.

Разработка методики прогнозирования глушения теплообменных трубок парогенератора атомных электростанций с водо-водяным энергетическим реактором

The main feature of an NPP steam generator (SG) is that in addition to steam generation, it shall reliably and constantly cool the NPP reactor’s core. The SG shall also comply with a stringent requirement of ensuring tightness of the primary and secondary circuits with respect to each other; i.e., its design shall exclude the possibility of damage to the heat-transfer and other elements loss of tightness of which entails ingress of radioactive primary coolant into the NPP steam--water circuit with the subsequent possibility of radioactive products releasing into the environment. Thus, ensuring reliable operation and design life of SG heat-transfer tubes (HTTs) is among top-priority objectives for different types of NPPs used both in the domestic nuclear power industry and abroad. Thin-walled SG HTTs constitute an important part of the primary circuit boundary, and in order to perform the function of an efficient barrier, HTTs shall not have through defects or defects generating the need to blank off them. The article describes methods used to predict blanking off heat-transfer tubes and the residual technical life of the SG tube bundle used at an NPP equipped with a water-cooled water-moderated power-generating reactor (VVER) and presents the results from calculating the permissible average value of chloride ion concentration during the operation. To assess the SG operation time to failure, a computer program has been written in the MathCad environment, using which the above-mentioned assessment can be performed within an acceptable timeframe and with sufficient accuracy. A conclusion has been drawn about the need to adopt more stringent requirements for the chloride ion concentration in the SG blowdown water to avoid the need of blanking off leaky tubes damaged as a result of stress corrosion cracking.

The criterion for blanking-off heat-transfer tubes in the steam generators at VVER-based nuclear power plants based on the results of eddy-current examination

The problem of defining the criterion for blanking off heat-transfer tubes in the steam generators at nuclear power plants on the basis of signals obtained from the standard multifrequency eddy-current examination is considered. The decision about blanking off one or another tube is presently made with reference to one parameter of the relevant signal at the working frequency, namely, with reference to its phase, which directly depends on the depth of the flaw being detected, i.e., a crack in the tube. The crack depth equal to 60% of the tube wall thickness is regarded to be the critical one, at which a decision about withdrawing such a tube out from operation (blanking off) must be taken. However, since mechanical tensile rupture tests of heat-transfer tubes show the possibility of their further use with such flaws, the secondary parameter of the signal, namely, its amplitude, must be used for determining the blanking-off criterion. The signals produced by the standard flow-type transducers in response to flaws in the form of a longitudinal crack having the depth and length within the limits permitted by the relevant regulations were calculated using 3D finite-element modeling. Based on the obtained results, the values of the eddy-current signal amplitude were determined, which, together with the signal phase value, form a new amplitude-phase criterion for blanking off heat-transfer tubes. For confirming the effectiveness of this technique, the algorithm for revealing the signal indications satisfying the proposed amplitude-phase criterion was tested on real signals obtained from operational eddy-current examination of the state of steam generator heat-transfer tubes carried out within the framework of planned preventive repair.

Dynamic Characteristics Analysis of Steam Generator Tube of Nuclear Power Plant by Finite Elements

In nuclear power plants, the steam generator heat transfer tube is the weakest part of the primary circuit pressure boundary. Flow induced vibration is one of the main reasons for the failure of the heat transfer tube. In this paper, an ANSYS finite element software is used to carry out the modal analysis of the heat transfer tube, and to simulate the dynamic response of the heat transfer tube in the harmonic load based on the modal analysis.

The design and thermohydraulics study of the HTR-10 High Temperature Helium Experimental Loop

A High Temperature Helium Experimental Loop (HTHEL) for the purpose of studying the transportation and deposition behavior of solid fission products in high-temperature helium coming from the steam generator (SG) in the 10 MW High Temperature Gas-cooled Test Reactor (HTR-10) is studied and designed. Through the optimal design based on thermohydraulics analysis, the three-sleeve structure of deposition sampling device (DSD) could realize full-length temperature control evenly and simulate the physical environment of the heat transfer tube of SG in the HTR-10 in the sense of thermohydraulics, which could be used to study the original source term in the primary circuit. The simulation of the graphite dust particle trajectories in the DSD are shown and it is elucidated that DSD could also be used to study the behavior of graphite dust in the high temperature helium in the SG of HTR, which will provide deeper understanding for the analysis of source terms of HTR-10.

Research on Fluid-Structure Interaction Dynamic Characteristics of Steam Generator Heat Exchanger Tubes

For heat exchanger tube of steam generator, the relation between heat exchanger tube and fluid is typical fluid-structure interaction problem. Flow induced vibration has been found so far to be responsible for fatigue damage and failure of steam generator tubes, which will result in large economic loss and radioactive pollution. So the steam generator tubes are the weakest link in the primary coolant loop. Based on the synthesis of all sorts of factors influencing the dynamic characteristics of steam generator heat transfer tubes, establishing the heat transfer tube model, research on the weakening effect of fluid hole on fluid, the natural frequencies of the heat transfer tubes are analyzed under different fluid holes and fluid hole distance by numerical simulation.

Research on pitting corrosion of steam generator heat transfer tubes based on acoustic emission

Abstract Corrosion of steam generator heat transfer tubes (SGHTT) is one of the important problems which affect safety operation of nuclear power plants (NPP), and the hazard of pitting corrosion of heat transfer tubes is the most serious. With an acoustic emission device, the signals during a corrosion test on SGHTT were collected and analyzed, and the corrosion points in the tubes were located accurately. The results show that pitting corrosion of heat transfer tubes has passed through three periods in its development: expansion phase, stationary phase and rapid developing phase. The corrosion damage of HTT can be found earlier with acoustic emission than any other non-destructive testing methods. Acoustic emission can be used for on-line and real-time monitoring of the safety and operation of the steam generator and has therefore a great significance.

Thermal Influence on Steam Generator Heat Transfer Tube During Sodium-Water Reaction Accident of Sodium-Cooled Fast Reactor

Sodium-water reaction is a design-basis accident of a sodium fast reactor. A breach of the heat transfer tube in a steam generator (SG) results in contact of liquid sodium with water. The typical phenomenon is that the pressurized water blows off and is mixed with the liquid sodium surrounding SG tubes. The design and safety concern is a possibility of the secondary failure of nearby heat transfer tubes that could cause undesirable development of the accident. One needs to evaluate the temperature transients of the heat transfer tubes in the reaction region for safety evaluation. In the present study, a computational method is developed for this purpose. It solves the sodium thermal hydraulics and the heat conduction in the adjacent heat transfer tubes. An experiment performed at the Japan Atomic Energy Agency is analyzed with the method developed in this study. It is found that analyzed temperatures are in good agreement with the experimental data. Based on the experimental and computational results, multiphase multicomponent flow characteristics are depicted. Furthermore, the heat transfer coefficient is evaluated using the instantaneous heat flux and temperature obtained from the numerical simulation.

Nuclear power development in Japan

As of September 1982, 24 nuclear power units were in operation in Japan with a total output of 17,177 MWe, 11 units were under construction, and 8 units were in the preparation stage. It is expected that nuclear power will provide 30.1% of the total of 840 TWh of generated electricity and 11.3% of the total energy supply in FY 1990. The average annual capacity factor has been raised to over 60% since 1980, due to joint efforts of the government, power utilities, and reactor suppliers to improve reliability, shorten periodic inspections, and extend operating cycle. At present, nuclear power is supplying electricity at a reasonable cost compared with other sources of power. A program of improvement and standardization of light-water reactors was initiated in 1975. In the first and second phases of this program, the following typical improvements were made: 1. (1) boiling-water reactor containment vessel design; 2. (2) pressurized-water reactor containment design; 3. (3) automatic and remote operation in control rod drive replacement; and 4. (4) steam generator heat transfer tube inspection and repair. The third phase is now in progress and will be completed in 1986, aiming at developing advanced boiling-water and pressurized-water reactors.