Steam Generator Tubes Research Articles

A New Repair Criterion for Inconel690 Alloy Steam Generator Tube with Fretting Wear

Structural integrity assessment of steam generator tubes(SGT) is important for the safety operation of pressurized water reactor. In this work, a repair criterion for Inconel690 SGT with fretting flaw was proposed. Firstly, experiments were conducted to address material strength, eddy current examination and bursting pressure of defected SGT. Then, a numerical method was proposed to predict bursting pressure. At last, all factors were concluded in a simplified statistical method of structural integrity assessment. For anti-vibration bar fretting, Condition Monitoring criterion, Operational Assessment criterion as well as the repair criterion for different inspection interval were proposed. The present criterion can provide reference for the inspection interval considering long-term temporary shutdown or standby.

The Effect of Heat-and-Mass Transfer and Flow Hydrodynamics on the Flow Accelerated Corrosion Rate in Evaporators of Combined-Cycle Unit Heat-Recovery Steam Generators

Flow accelerated corrosion (FAC) is considered one of the major causes of damage to heat-recovery steam generators (HRSG) of combined-cycle units. In the last few decades, a great number of papers devoted to this problem have been published in many countries around the world. The results of investigations were analyzed as a rule, on the basis of the flow-averaged thermochemical parameters of the flow, while the features of the near-wall liquid film determining the FAC rate were not taken into account. The indicators in the near-wall region depends on the heat-and-mass transfer conditions, flow hydrodynamics, and the specific of chemicals used to correct the water chemistry (WC). For ammonia and oxygen water treatment chemistries, an equilibrium model is proposed and its validity for HRSG evaporators is substantiated. The rate of the exchange of gas reagents between steam and water is determined by the distribution coefficient Kd. To simplify calculations, simple approximating correlations of Kd for ammonia and oxygen is recommended. For the water chemistry with solid reagents, a diffusion model has been developed to calculate mass transfer on the basis of data on heat transfer in steam-generating tubes. The concentration of reagents near the wall is determined by both the turbulent transport of the liquid between the flow core and the near-wall layer and the coefficient of reagent distribution between the phases. Simple approximating correlations of distribution coefficients for Na3PO4, NaOH, and 90H Turb helamine are proposed. An approximate correlation between the hydraulic resistance coefficient Kh and the geometric parameter Kс of the FAC process has been established. The hydrodynamic fundamentals of the standard HRSG structural members have been studied in more detail than FAC. To estimate Kс, one can use the known value of Kh and the correlation between them. However, this correlation should be used with care since it has been obtained for certain conditions of unambiguity (i.e., specific construction material, water chemistry, flow history, etc.). Under other conditions, this correlation will change somewhat; therefore, it can only be used for rough estimation.

Experimental study on leak flow rate characteristics of high pressure subcooled water through axial and circumferential microcracks of steam generator tubes under high back pressure conditions

Leak-before-break analyses on tube microcracks are beneficial for preventing small loss of coolant accident in nuclear power plants. Specially, the tube leakage of steam generator may lead to steam generator tube rupture. Currently, experimental investigations on leak flow rate characteristics of normal temperature water and high pressure subcooled water through axial and circumferential microcracks of steam generator tubes were carried out, respectively. Cold Test results indicate that leak flow rates of circumferential microcracks for 0.13 ≤ W ≤ 0.16 mm and 3.6 ≤ δ/Dh ≤ 4.4 increase with a decreasing back pressure (Pout) first and then can be regarded as a constant value for a given working pressure in steam generator tubes (Pin). Additionally, each of Pin has a critical differential pressure between Pin and Pout (ΔPc). The ratio of ΔPc to Pin is equal to 0.55, which is smaller than 0.64 for axial microcracks. Hot Test results show that leak flow rates decrease significantly with an increasing working fluid temperature (Tf, in) when Tf, in is larger than a certain value due to the flashing of subcooled water near exits of microcracks. Besides, discharge coefficients of high pressure subcooled water through axial and circumferential microcracks under various back pressure conditions decreases with an increasing Pin for a given subcooled temperature. In general, new correlations were developed to give excellent predictions for the leak flow rate data of normal temperature water and high pressure subcooled water through axial and circumferential microcracks of steam generator tubes under various back pressure conditions with a maximum deviation of ±6%.

Molar Ratio Effect of Sodium to Chloride Ions on the Electrochemical Corrosion of Alloy 600 and SA508 in HCl + NaOH Mixtures.

This study aims to investigate the molar ratio effect of sodium to chloride ions on the corrosion of an Alloy 600 steam generator tube and an SA508 tubesheet material. The corrosion behavior was evaluated in solutions with three different molar ratios of sodium to chloride ions using a potentiodynamic polarization method. The corrosion potentials and corrosion rates of both the two materials were significantly decreased as the molar ratio increased from 0.1 to 10. Therefore, it is recommended that the molar ratio control to a value of 1 is beneficial only when the crevice chemistry has a low molar ratio with an acidic pH. The corrosion potentials and corrosion rates were little affected by the total sodium and chloride ion concentrations. SA508 acted as an anode and its corrosion rate was accelerated by galvanic coupling with Alloy 600.

High cycle fatigue behavior and numerical evaluation of Alloy 690TT steam generator tube

High cycle fatigue behavior of Alloy 690TT steam generator (SG) tube under different maximum stresses was investigated at stress ratio R = 0.1 through experimental and numerical evaluation. The morphology of fatigue fracture surface was observed by SEM, and the crack initiation and propagation mechanism of Alloy 690TT SG tube was analyzed. The fatigue crack of Alloy 690TT SG tube initiated at the intrados surface edge, then propagated along the direction of maximum shear stress and penetrated through the side of the specimen, and finally transient fracture occurred in the direction perpendicular to the loading axis. With the increase of maximum stress, the number of fatigue crack initiation sources increased, but they were all distributed on the edge of the intrados surface, and a large number of persistent slip bands (PSBs) were formed on the intrados surface. The fracture mechanism of the tubular specimen was different from that of the bar specimen, and its fracture mode affected the life evaluation of the tubular specimen. The finite element analysis software ABAQUS and advanced fatigue durability analysis software FE-SAFE were first used to evaluate the fatigue life of Alloy 690TT SG tube. Compared with the experimental results, the error in predicting fatigue life by numerical method was acceptable.

Development of in-bore magnetostrictive transducer for ultrasonic guided wave based-inspection of steam generator tubes of PFBR

An in-bore magnetostrictive transducer is designed for the steam generator tubes of Prototype fast breeder reactor for the generation of L(0,2) modes of frequencies in the range of 250–350 kHz. Towards this, axi-symmetric finite element models are developed to optimize the coil parameters. The optimized length of the transmitter and the receiver coils turns out to be 10 mm (~half the wavelength) for the frequency of 300 kHz. The optimized width of the coils turns out to be 0.46 mm. FE models also show the generation, propagation and reception of L(0,2) modes in the frequency range of 250–350 kHz. The role of skin effect in the magnetostrictive based-generation of L(0,2) modes with frequency is also discussed. A transducer is designed based on the FE results. The transducer is tested for the generation of L(0,2) mode in the frequency range of 250–350 kHz in a 1 m long steam generator tube segment. A good agreement is observed between FE and experimental normalized amplitudes and the times of flight for different frequencies. L(0,2) modes are found to generate and propagate and received, as predicted by the finite element simulations. An excellent agreement is observed between the experimentally measured group velocities with those obtained from the dispersion curves in this frequency range. Experiments show the signal to noise ratio to be better than 15 dB. To ascertain the utility of the transducer in steam generator tubes for the long range testing, L(0,2) mode at 300 kHz frequency is propagated in a 1.5 m long tube. The resulted multiple end reflections amount to the propagation of 51 m distance. To check the capability of detection of defects, a short tube with a full circumferential defect of depth 0.46 mm (20%WT) and a short tube with a pin hole of 1.5 mm diameter are considered. Further, FE results for the case of the axi-symmetric circumferential defect are validated experimentally. For the case of the pinhole (non-axi-symmetric), the experimental signal to noise ratio turns out to be 6 dB, which is only 6 dB lower as compared to that obtained using a piezo based ultrasonic transducer of frequency 300 kHz coupled to the end of the tube.

Effects of pH control agents on magnetite deposition on steam generator tubes

Magnetite deposition on the secondary side of steam generators is a major concern in nuclear power plants. Therefore, it is necessary to mitigate the deposition using optimized water chemistry control. The purpose of this study is to investigate the effects of three pH control agents, ammonia, morpholine, and ethanolamine, on the magnetite deposition onto an Alloy 690TT tube under conditions similar to operating steam generators. The deposited magnetite particles were polyhedral or spherical in shape and numerous micropores were clearly observed in the deposits regardless of pH agents. The amount of deposited magnetite was lowest when ammonia was used, and increased approximately 1.7 times and 2.6 times when using morpholine and ethanolamine, respectively. The effects of the pH agents were discussed based on the difference of zeta potentials between magnetite particles and the bare surface of the Alloy 690TT tube, particle size, and magnetite solubility.

Mechanism and kinetics of hematite reduction under typical PWR secondary circuit condition

Hydrazine is added into secondary circuit in order to eliminate oxygen and to impose a reducing potential. Another possible role of hydrazine is to reduce hematite potentially present and thus reduce risk of stress corrosion cracking of steam generator tubes. However, the latter effect of hydrazine is still not clear. In order to better understand the mechanism and the kinetics of hematite reduction by hydrazine under secondary circuit conditions (275 °C, slightly alkaline, diluted hydrazine), an experimental study has been carried out. It was found that the reduction of hematite proceeds through the adsorption of hydrazine, surface reduction and dissolution of hematite, and coprecipitation of well crystallized octahedral magnetite.

Eddy Current Probe With Integrated Tunnel Magnetoresistance Array Sensors for Tube Inspection

This article presents an eddy current probe with array tunnel magnetoresistance (TMR) sensors for inspection of the steam generator tube of the nuclear power plant. The TMR sensors are integrated on a circular printed circuit board (PCB) with the sensitive axis along the radius directions. The dimensions of each TMR sensor in the array are length (0.45 mm) $\times $ width (0.45 mm). Fine spatial resolution magnetic field imaging can be generated as the size of the sensor is small. The operating principle of the probe is based on Faraday’s law, where eddy current is induced in the tube wall by an excitation coil carrying ac current, and then, the disturbance of the eddy current is monitored in terms of the magnetic field measured by the TMR sensors. Thanks to the high sensitivity and fine spatial resolution of the TMR sensors, the probe is capable of detecting outer diameter (OD) defect with an effective size 0.127 mm $\times \,\, 0.436$ mm. Moreover, when the probe scans in a defect-free area, the variation of lift-off of the probe to the inner surface of the tube wall does not bring in noises, which is important for maintaining a high signal-to-noise ratio in the practical inspection. The operation principle of the probe was studied based on a finite-element method (FEM) model, and then, a prototype unit was developed and tested. The experimental results demonstrated the feasibility of the concept and validated that the probe has excellent ability to detect various defects.

Carbide Faceting in Alloy 690: HRTEM Study

Abstract Alloy 690 (Unified Numbering System [UNS] No. 6690) in thermally treated (TT) condition is an advanced steam generator tubing material that offers high corrosion resistance with tailored carbide morphology and chromium concentration adjacent to the grain boundary. The morphology of the intergranular carbides was found to have strong dependence on the grain boundary character distribution (GBCD) from an electron backscattered diffraction (EBSD) study. The effect of thermal aging on crystallography and faceting of carbides was studied with the aid of high-resolution transmission electron microscopy (HRTEM). Carbides formed at 600°C, 700°C, and 800°C were structurally consistent with face centered cubic (fcc) M23C6. The orientation relationship of the carbides with grain allowing maximum atomic correspondence was found to be (111)C ‖ (111)M and <110>C ‖ <110>M. HRTEM analysis revealed that the carbides were partially faceted with multiple facet initiation at lower temperature and become fully faceted with fewer stable facets at 800°C. The non-faceted parts of carbides had lattice structure discontinuity from the faceted portion creating new energy interface within the carbides. It is suspected that the growth of carbides, formation of facets, and facet transformation takes place via ledge growth and coalescence. Stable, low-energy faceted interfaces of 800°C remain inactive and hold the carbides firmly, preventing dissolution or dislodging in corrosive environments and thus providing superior corrosion resistance.

Numerical investigation on LBE-water interaction for heavy liquid metal cooled fast reactors

Pool type Reactor Pressure Vessels (RPVs) containing steam generators with high secondary coolant pressure are adopted for a number of heavy liquid metal reactor conceptual designs. The primary side heavy liquid metal, e.g., liquid Lead-Bismuth Eutectic (LBE) with low pressure and secondary water with high pressure can come into contact in the event of Steam Generator Tube Rupture (SGTR) scenarios. Due to the pressure wave propagation and resultant sloshing of liquid LBE, the vigorous interaction between water and melt LBE can threaten the structural integrity of the reactor internals. Moreover, Reactor Pressure Vessel (RPV) pressurization, reactivity feedback as a result of steam entrance into the core as well as primary system pollution are also key concerns for SGTR accidents.The article aims to evaluate the capability of the MC3D code for LBE–water interaction in the event of SGTR scenarios. MC3D is a multidimensional numerical tool developed by IRSN for multi-phase flow studies and assessments in nuclear safety. Its field of application is mainly on Fuel-Coolant Interaction (FCI), especially the stages of premix and explosion in case of severe accidents. In order to evaluate the capability of MC3D for the Coolant-Coolant Interaction (CCI) of Liquid metal cooled Fast Reactors (LFRs), simulation of a small-scale experiment investigating water jet plunging into a pool of molten lead-bismuth alloy is carried out. Afterwards, experiments on LIFUS5/Mod2 facility at ENEA studying LBE-water interaction in the event of SGTR, for which a series of experiments have been conducted including the injection of sub-cooled water into LBE from the bottom of the test vessel, is simulated with MC3D as well. The simulation results for both facilities are then compared with those of the tests. The results predicted with MC3D for LIFUS5/Mod2 is also compared against the simulation results with SIMMER. The comparison impliesthat MC3D is generally capable for the phenomena of CCI in the event of SGTR for LFRs. Nevertheless, the potential employment of MC3D for intensive investigation of the LBE-water interaction will still need further experimental studies and improvement of physical models.

Analytic springback prediction in cylindrical tube bending for helical tube steam generator

This paper newly proposes an efficient analytic springback prediction method to predict the final dimensions of bent cylindrical tubes for a helical tube steam generator in a small modular reactor. Three-dimensional bending procedure is treated as a two-dimensional in-plane bending procedure by integrating the Euler beam theory. To enhance the accuracy of the springback prediction, mathematical representations of flow stress and elastic modulus for unloading are systematically integrated into the analytic prediction model. This technique not only precisely predicts the final dimensions of the bent helical tube after a springback, but also effectively predicts the various target radii. Numerical validations were performed for five different radii of helical tube bending by comparing the final radius after a springback.

Human reliability analysis for operators in the digital main control rooms of nuclear power plants

ABSTRACT As the main control room of nuclear power plants (NPPs) has been gradually digitized, new human reliability problems may emerge because of a series of new changes in the cognitive processes, behavioral patterns, and error mechanisms of operators. Aiming to address this situation, this paper proposes a method as guidance for human reliability analysis (HRA) of different cognitive Stages. This method first constructs the influencing factors of three cognitive processes, including monitoring, decision-making, and execution of actions, and then evaluates the weights of these influencing factors through an analytic hierarchy process (AHP). In this study, the parameters used in the proposed HRA method were determined by analyzing the test data obtained from a simulation model, and the results demonstrated the rationality and feasibility of the proposed method. A case example using this HRA method was given in which the human error probabilities at three stages in a nuclear power plant (NPP) steam generator tube ruptures (SGTR) accident were obtained. In summary, the proposed method is a simple and feasible HRA tool that can be applied in digital NPP main control rooms (MCRs).

Revealing and Assessing the Deposit Amount on Heat-Exchanger Tubes of NPP Steam Generators According to the Operational Data of Routine Eddy Current Monitoring

One of the measures aimed at preventing unscheduled shutdowns of reactor units at nuclear power plants owing to inter-circuit leaks in steam generators consists in monitoring the formation of deposits on the surfaces of heat exchangers. The azimuthal and axial extension of the electrically conductive layer in the deposits, as well as its thickness, can be used to judge the intensity of corrosion. In addition, assessing the contamination of the heat exchanger tubes from the side of the second circuit is important to prevent the accumulation of deposits on the heat exchange surfaces and to plan measures for their removal in the course of chemical washing. Monitoring conductive deposits is based on the existence of a quantitative relationship between the energy parameters of a low-frequency signal of the axial eddy-current transducer on the absolute channel and the volume of the conductive layer in the deposit. The results of studies aimed at revealing and numerically assessing the thickness (volume) of deposits on the outer surface of heat-exchanger tubes in steam generators at nuclear power plants with VVER-440 and VVER-1000 reactor units are presented. For this purpose, operational data based on standard eddy-current monitoring have been used to assess the condition of the metal in heat-exchanger tubes obtained using the PIRATE software package. At the first stage, characteristic diagnostic parameters of eddy-current signal readouts from electrically conductive deposits on heat-exchanger tubes have been determined. Criteria were then formulated according to which readouts corresponding to the sites with deposits can be automatically detected. The next stage consisted in determining the parameters of the eddy-current signal readouts that make it possible to estimate the thickness of the deposits. An algorithm has been developed for constructing a calibration curve for assessing the volume and average thickness of electrically conductive deposits on the surface of heat-exchanger tubes.

Effect of Oxygen Content on Heat Transfer in Steam-Generating Tubes in Transverse Flow of Heavy Liquid-Metal Coolant

A series of works was conducted on determining the coefficient of heat transfer in a bundle of heat-exchange tubes in transverse flow of lead in the base oxygen regime and with oxygen concentration on the saturation line. The object of investigation was an experimental model consisting of a horizontal bundle of six rows of smooth tubes with four tubes per row. The investigation was conducted with sub- and supercritical pressure in the water loop. It was confirmed that the oxygen content in lead significantly affects the coefficient of heat transfer. In the experimental range of velocities up to 0.3 m/sec, the coefficients of heat transfer in the base oxygen regime are two to three times higher than for lead with oxygen concentration on the saturation line. In agreement with other studies, it was found that for lead with oxygen concentration on the saturation line the coefficient of heat transfer is velocity-independent in a narrow section.

Code improvement, separate-effect validation, and benchmark calculation for thermal-hydraulic analysis of helical coil once-through steam generator

Helical coil once-through steam generators (HCOTSGs) have been widely used in the design of small modular reactor (SMR) during the past several decades. As the widely accepted system analysis code, RELAP5 underestimates the heat transfer capability and pressure drop of helical coil tube steam generators using the original geometrical parameters of HCOTSG, because the built-in thermal-hydraulic empirical correlations are only suitable for straight tubes. In this study, thermal-hydraulic models for helical coil tubes and tube bundles are implemented in RELAP5 while the original functions of the code are not affected. A new flag for helical component is proposed and heat transfer boundaries for tube side and shell side are developed. The separate-effect validations of friction factor and heat transfer in helical coil tube are carried out based on experimental data and code-to-code verification. Then, the original RELAP5 and developed RELAP5-HCOTSG are used to simulate helical coil tube steam generators of two SMRs. One is the integral reactor IRIS, and the other is MRX. The calculated results are compared to the design parameters. It shows that the HCOTSG module developed in this study can improve the capability of RELAP5 to predict the thermal-hydraulic characteristics in SMR once-through steam generators with helical coil tubes.

A comparison of failure assessment diagram options for Inconel 690 and Incoloy 800 nuclear steam generators tubes

In the context of structural integrity assessment of cracked Steam Generator Tubes (SGTs), the Failure Assessment Diagram (FAD) has been proposed as a modern tool able to reduce the conservativeness of traditional plastic limit load analyses. As different FAD failure curves are available in the literature, this paper presents a comparison between the so-called Options 1, 2 and 3 FAD curves derived for Inconel 690 and Incoloy 800 nuclear SGTs. Detailed Option 3 curves are obtained from elastoplastic finite element analyses and experimental data from fracture tests. Material-specific Option 2 curves are drawn from stress vs. strain data of uniaxial tests of SGTs. It is shown that the simplest generic Option 1 curve is conservative but appropriate for structural integrity assessments of cracked nuclear SGTs in operation. Alternatively, Option 3 FAD curves must be used when an excessive conservatism shall be avoided.

Novel array ECT probe with three-phase excitation for steam generator tube inspection

Eddy current testing (ECT) is commonly used in steam generator tube inspection. Among multiple probes applied in this area, array probe shows the merit of the capability of obtaining a C-scan image in a simple linear scan. In this work, a novel array ECT probe with three-phase excitation is proposed. The probe consists of two rows of excitation coils and a row of pickup coils. The induced eddy current shifts electrically in the testing sample so that the necessity of multiplexer is eliminated, resulting in advantages of lower cost, less noise and faster inspection speed. Theoretical analysis indicates that the background signal of the array sensor would be small. A 3D finite element method (FEM) model is developed to study the operating principle and to predict the inspection performance of the probe. A prototype probe has been developed and tested, by which a steam generator tube sample with different kinds of defects was inspected. The experiment results validate the simulation model and further demonstrate the feasibility of the sensor.

Analysis of eddy-current testing and metallographic examinations for corrosion damages of heat exchanging tubes in NPP steam generators

Analysis of eddy-current testing and metallographic examinations for corrosion damages of heat exchanging tubes in NPP steam generators

Experiments of Self-wastage Phenomena Elucidation in Steam Generator Tube of Sodium-cooled Fast Reactor

Experiments of Self-wastage Phenomena Elucidation in Steam Generator Tube of Sodium-cooled Fast Reactor