An experimental result on flaw sizing of steam generator tube at nuclear power plants using ultrasonic testing

A steam generator at a nuclear power plant consists of thousands of thin tubes, and is a highly important component in operation. Also, steam generator tubes play a critical role in maintaining pressure boundaries of the primary and secondary sides, and can be easily damaged due to operation conditions caused by high temperature and pressure. Therefore, considerable amount of efforts are being committed to evaluating structural integrity of steam generators during in-service inspection. Eddy current testing is the commonly used inspection technique to evaluate a steam generator tube's integrity, but it has limitations in accurately sizing flaws due to the nature of the technique which determines size based on the entire volume of a flaw. In this study, experiments were performed by using ultrasonic testing instead of eddy current testing for the inspection of steam generator tubes to detect various kinds of flaws and to see if the detected flaws can be sized accurately. Consequently, the ultrasonic testing technique could detect various types of flaw, and the flaw sizing results were reliable in length and depth.     

Development of probability of detection functions for evaluating integrity of axial outside diameter stress corrosion cracks in steam generator

The reliability of an eddy current testing (ECT) inspection system depends upon the inspection technique and quality of analyst. In evaluating the integrity of a steam generator (SG) tube, degradation detection and sizing accuracy are considered performance measures of the nondestructive evaluation (NDE) system. A probability of detection (POD) model serves as a functional measure of the ability of an NDE system to detect degradation. It is one of the inputs in the operational assessment, and it is used to estimate the degradation during service via ECT of the SG tube. In this study, the POD functions of the inspection technique and analyst were obtained to quantitatively analyze the ECT bobbin probe for axial outside diameter stress corrosion cracks in SG tubes. This should serve to evaluate the integrity of the SG tubes. The depth and amplitude of defects were used as parameters of the POD model. Hit (detection) and miss (no detection) binary data obtained from destructive and nondestructive inspection of cracked tubes were also used.     

Inspection techniques for primary pressurized water cooler tubes in the high temperature engineering test reactor

A primary pressurized water cooler (PPWC) with 136 inverse-U-tubes is installed in the primary cooling system of the high temperature engineering test reactor (HTTR). The HTTR is the first high temperature gas-cooled reactor in Japan with an outlet gas temperature of 950 °C and thermal power of 30 MW. The heat transfer tubes form the reactor pressure boundary of the primary coolant. Inspection techniques for the tubes should be established to carry out the in-service inspection efficiently. An automatic inspection system for the tubes uses probes for eddy current testing and ultrasonic testing. Defect detecting characteristics of the inspection probes and the application of the automatic inspection system to nondestructive test of the tubes were examined by a mockup test utilizing artificially degraded tubes. The automatic inspection system could smoothly insert and withdraw the probe at its moving velocity in the fixed positions of the defected tube. Nondestructive test of the tubes using the automatic inspection system was conducted during reactor shutdown period of the HTTR after test operation of about 55% of the full power. Through the nondestructive test, it was concluded that there was no defect on the outer surface of the heat transfer tubes of the PPWC inspected.     

Detection and Mode Identification of Axial Cracks in the Steam Generator Tube of the Nuclear Power Plant Using Ultrasonic Guided Wave

For those people who are involved in NDE, there is a growing concern regarding the significant traveling distance of a guided wave in a structure, which ensures the inspection of a large area of the structure from a single location. A significant number of studies on the guided wave have therefore been made to apply the foregoing to a nondestructive evaluation in many different industries and resulted in an increase in the efficiency of practical guided wave inspection. Unlike the previous studies based mainly on the detection of circumferential flaws, this study is focused on the axial flaw detection in the steam generator tubes of Korean standard nuclear power plants by generating the guided wave by changing frequency and selecting the applicable mode from the dispersion curve for the steam generator tube calculated in this study, where the dispersion-based short-time Fourier transform (D-STFT) algorithm is used to enhance mode identification. In conclusion, the L (0,1) mode at 2.25MHz is found to be most sensitive in detecting axial flaws in a steam generator tube.     

Current in-service inspections of austenitic stainless steel and dissimilar metal welds in light water nuclear power plants

A tendency towards growing requirements for the inspection of austenitic piping can be observed in several countries. In Germany the revised KTA rule demands the UT inspection of austenitic and dissimilar metal welds in piping with diameters of 200 mm or more.On the basis of experience gained from austenitic piping with integranular stress corrosion cracking (IGSCC), longitudinal waves and mode conversion techniques are used. Depending on the geometry, material and grain orientation, spurious signals can be observed which require additional evaluation or analysis measurements.A promising new technique is based on horizontally polarized shear (SH) waves generated by electromagnetic acoustic transducers (EMATs). Investigations in the laboratory and field inspections showed that SH waves are well suited for the detection of longitudinal flaws, especially where the weld can be examined from one side only.For the complete solution of a given inspection problem SH waves can be combined with well-known standard techniques in order to provide redundant information for the characterization and sizing of indications.The investigation of possibilities of SH waves showed that the problem of cast austenitic steel inspection might not be solved using this technique. However, measurements using low frequency UT transducers showed promising results.     

Development of an ultra sonic phased array system for nondestructive tests of nuclear power plant components

A phased array ultrasonic inspection (PAULI) system is being developed to obtain electronically scanned ultrasonic images of the inside of nuclear power plant components for nondestructive evaluation. The development strategy of PAULI system was the modification of a medical ultrasound imaging system that had 64 individual transceiver channels. Optimization of array transducers has been also pursued based on the systematic investigation of the radiation beam field simulated by the use of the boundary diffraction wave models. 7.5 MHz phased array transducers was, then, fabricated and tested with the carbon steel specimen having side-drilled holes. For the nondestructive tests on power plant component, a sample mockup of turbine blade root with EDM notches was fabricated and the detection capability was demonstrated. The developed system can provide electronically scanned ultrasonic images in real time fashion and greatly enhance the efficiency and reliability in the flaw detection and location in comparison with the classical ultrasonic testing (UT) using A-scan signals. For the flaw classification, the analysis of the electronically scanned ultrasonic images was not sufficient at this moment but analysis of features obtained from A-scan signals of flaws at the various steering angles showed the potential capability.     

Potential steam generator tube rupture in the presence of severe accident thermal challenge and tube flaws due to foreign object wear

This study develops a methodology to assess the probability for the degraded PWR steam generator to rupture first in the reactor coolant pressure boundary, under severe accident conditions with counter-current natural circulating high temperature gas in the hot leg and SG tubes. The considered SG tube flaws are caused by foreign object wear, which in recent years has emerged as a major inservice degradation mechanism for the new generation tubing materials. The first step develops the statistical distributions for the flaw frequency, size, and the flaw location with respect to the tube length and the tube's tubesheet position, based on data of hundreds of flaws reported in numerous SG inservice inspection reports. The next step performs thermal-hydraulic analysis using the MELCOR code and recent CFD findings to predict the thermal challenge to the degraded tubes and the tube-to-tube difference in thermal response at the SG entrance. The final step applies the creep rupture models in the Monte Carlo random walk to test the potential for the degraded SG to rupture before the surge line. The mean and range of the SG tube rupture probability can be applied to estimate large early release frequency in probabilistic safety assessment.     

Computation and measurement of calandria tube sag in pressurized heavy water reactor

Calandria tubes and liquid injection shutdown system nozzles in a pressurized heavy water reactor are to sag due to irradiation creep and growth during plant operation. When the sag of calandria tube becomes bigger, the calandria tube possibly comes in contact with liquid injection shutdown system tube crossing beneath the calandria tube. The contact subsequently may cause the damage on the calandria tube resulting in unpredicted outage of the plant. It is therefore necessary to check the gap between the two tubes in order to periodically confirm no contact by using a proper measure during the plant life. An ultrasonic gap measurement probe assembly which can be inserted into two viewing ports of the calandria was developed in Korea and utilized to measure the sag between both tubes in the reactor. It was found that the centerlines of calandria tubes and liquid injection shutdown system tubes can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. Based on the irradiation creep equation and the measurement data, a computer program to calculate the sags was also developed. With the computer program, the sag at the end of plant life was predicted.     

Overview of steam generator tube degradation and integrity issues

The degradation of steam generator tubes in pressurized water nuclear reactors, and, in particular, the stress corrosion cracking (SCC) of Alloy 600 tubes, continues to be a serious problem. Primary water SCC is commonly observed at the roll transition zone (RTZ), at U-bends and tube denting locations, and occasionally in plugs and sleeves. Outer-diameter SCC (ODSCC) and intergranular attack (IGA) commonly occur near tube support plate (TSP) crevices, near the tube sheet in crevices, or under sludge piles, and occasionally in the free span. A particularly troubling recent trend has been the increasing occurrence of axial and circumferential cracking at the RTZ on both the primary and secondary sides. Outer-diameter stress corrosion cracking in TSP crevices, commonly consisting of segmented axial cracks with interspersed uncracked ligaments, is also becoming more common. Despite recent advances in inservice inspection (ISI) technology, a clear need still exists for quantifying and improving the reliability of ISI methods with respect to the probability of detection of the various types of flaws and their accurate sizing. These improvements are necessary to permit an accurate assessment of the consequences of leaving degraded tubes in service over the next reactor operating cycle. Degradation modes such as circumferential cracking, intergranular attack, and ODSCC at the TSP have affected a large number of tubes. New regulatory guidance is being developed that requires the development and implementation of a steam generator management program that monitors tube condition against accepted performance criteria to ensure that the tubes will perform the required safety function over the next operating cycle. In formulating new guidance for the implementation of alternate repair criteria, the U.S. Nuclear Regulatory Commission is also evaluating the contribution to overall plant risk from severe accidents. Preliminary analyses are being performed for a postulated severe-accident scenario that involves station blackout and loss of primary feedwater, in order to determine the probability of failure for degraded tubes.     

The newest two-phase flow control devices in LWR equipment based on ultrasonic and WAT-technology

This paper deals with new acoustic methods of two-phase flow diagnostics used to carry out research in the fields of nuclear power thermophysics and nuclear power plant (NPP) technologies equipment control. All the designs are to be used under extreme conditions, characteristic for water coolant, with temperature up to 350°C and pressure 20 MPa. All the safety and reliability requirements are met. The methods use waveguide ultrasonic transducers for longitudinal and bending waves, made according to specially designed technology (waveguide acoustic transducers — WAT technology). This paper deals with the operating principles of transducers and processing device physical models as well as some results on the practical use of this equipment. The method of acoustic impedance is based on measuring attenuation of a longitudinal or bending ultrasonic wave in a thin-walled tube diameter vapour fraction or the level of the coolant in the tank. The waveguide transducers, designed by the centre, use bending waves of a surface type. They enable us to carry out diagnostics of the liquid film on the inner surface of the tube or discover gas inclusions in the liquid flow. The paper touches upon the method of acoustic emission for measuring moisture content in a steam flow.     

Modeling of eddy current probe response for steam generator tubes

Sample calculations were performed with a three-dimensional (3D) finite-element model to describe the response of an eddy current (EC) probe to defects in steam generator (SG) tubing. Such calculations could be very helpful in understanding and interpreting EC probe response to complex tube/defect geometries associated with the inservice inspection (ISI) of SG tubes. The governing field equations are in terms of coupled magnetic vector and electric scalar potentials in conducting media and of total or reduced scalar potentials in nonconducting regions. To establish the validity of the model, comparisons of the theoretical and experimental responses of an absolute bobbin probe are given for two types of calibration standard defects. Simulation results are also presented on the effect of ligament size in axial cracks on bobbin probe response.     

Measurements of the acoustic field on austenitic welds: a way to higher reliability in ultrasonic tests

In nuclear power plants many of the welds in austenitic tubes have to be inspected by means of ultrasonic techniques. If component-identical test pieces are available, they are used to qualify the ultrasonic test technology. Acoustic field measurements on such test blocks give information whether the beam of the ultrasonic transducer reaches all critical parts of the weld region and which transducer type is best suited. Acoustic fields have been measured at a bimetallic, a V-shaped and a narrow gap weld in test pieces of wall thickness 33, 25 and 17 mm, respectively. Compression wave transducers 45, 60 and 70° and 45° shear wave transducers have been included in the investigation. The results are presented: (1) as acoustic C-scans for one definite probe position, (2) as series of C-scans for the probe moving on a track perpendicular to the weld, (3) as scan along the weld and (4) as effective beam profile. The influence of the scanning electrodynamic probe is also discussed.     

Secondary side corrosion in steam generator tubes: lessons learned in France from the in-service inspection results

Non-destructive testing (NDT) has proved to be very important in the maintenance of steam generator tubing. This is particularly true in the case of secondary side corrosion, because this type of degradation leads to various morphologies which are often complex (intergrranular attack) (IGA), intergranular stress corrosion cracking (IGSCC), or a mixture of both. Their detection and characterization by the usual NDT techniques have been achieved through numerous laboratory studies, which were conducted in order to determine the performance and limitations of NDT. Pulled tube examination in a hot laboratory was very valuable, for both NDT and fracture mechanics aspects. The eddy current bobbin coil probe, used for multipurpose inspection of tubes, allows the detection of IGA-SCC at the tube support plate elevation. In France, the use of rotating probes is not required for that type of degradation, since the repair criterion is based on bobbin coil results only. The bobbin coil is also used for detection of IGSCC occurring in free spans, within sludge deposits. The eddy current rotating probe allows, in that case, characterization of main cracks. Concerning the outer diameter initiated circumferential cracks which occur at the top of the tube sheet, only the rotating probe is used. An ultrasonic (UT) inspection was performed several times, in order to obtain information on UT capabilities. The goal of tube inspection is obviously knowledge of the status of steam generators, but also to follow up degradations and to estimate their revolution, and to verify the beneficial effect of some corrective measures, e.g. boric acid injection.     

UT digitized data processing for in service inspection of pressurized water reactor vessels

Pressurized reactor vessels in France have been examined from the inside with ultrasonic focused transducers since the very first inspection (Gagnor and Levy, 1993, Proc. 7th Asian-Pacific Conference on Nondestructive Testing, Shangaï, China, 867 pp.). The developments carried out in collaboration with the French Atomic Energy Commission (CEA) to improve the characterization of flaws detected in the body of the vessels or in the nozzles, in the vicinity of the inner or the outer surfaces now have application throughout the CIVAMIS software. The processing modules of CIVAMIS, which are implemented on site since 1994 and used by INTERCONTROLE during the in service inspections of the French PWR vessels, allow full characterization of these specific flaws. The first module is devoted to the characterization of defects located near the outer surface of the vessel or the bottom head welds (OSD module). It includes the modeling software MEPHISTOMIS which predicts the echoes coming from the interaction between the ultrasonic beam and the defects. The second module of CIVAMIS (inner surface defect module called ISD), applied to the analysis of flaws expected near the inner surface of the vessels, has been used during performance demonstration exercises on qualification mock-ups, and also on-site in five expert appraisals since its qualification in 1995. The third module available on the system has been developed and qualified for the analysis of flaws likely to appear near the inner surface of the nozzles. This module, named ‘Undercladding Crack Defect' (UCD) module, provides the operators with a set of pre-defined processing configurations well adapted to the characteristics of the transducers and of the digital acquisitions triggered in this examination case. The last new module (SAFE-END module) has been developed for the evaluation of defects located close to the bimetallic weld in the nozzles. The capacity of CIVAMIS to be adapted to each examination area and the characterization tools included in the different modules are developed in the present paper.     

Inspection procedure assessment using modelling capabilities

Before carrying out inspections on technical components, criteria have to be met to qualify the inspection procedure. The inspection qualification can be performed by technical justification or performance demonstration. The qualification of non-destructive testing methods can be checked on full scale mock-ups where real or realistic defects are implemented. To minimize costs it will become mandatory to model the inspection procedure. In the case of ultrasonic testing, the modelling includes the ultrasonic probe, the pulse-defect interaction and the probe geometry. The applied modelling code is the elastodynamic finite integration technique (EFIT) which includes mode conversion effects. The results of the code are either wave fields, A-scans at different probe positions or complete r.f. data fields. Examples demonstrate the efficiency in modelling angle beam probes which transmit both pressure, shear and subsurface longitudinal waves. These waves are scattered by horizontal or surface breaking cracks. The modelled r.f. data fields are used within the synthetic aperture focusing approach to predict the ultrasonic image which would be obtained in performing the experiment. In addition it will be shown that the EFIT wave fronts for anisotropic homogeneous media are explicitly related to the group velocity in these media.     

A case study on detection and sizing of defects in steam generator tubes using eddy current testing

Eddy current testing (ECT) method is widely used to detect various types of defects occurring in nuclear steam generator tubes. Therefore, the reliability of its detection and sizing accuracy for defects should be validated. For this purpose, two tubes with defect signals were pulled from an operating steam generator and destructively examined. The defect type was a circumferential crack for one tube and an intergranular attack (IGA) for the other tube. The plus point coil probe showed a better capability to detect and size both a circumferential crack and a volumetric IGA than pancake and bobbin coil probe. The destructive results are correlated with the ECT results obtained during the in-service inspection.     

Products of research and development in NDT

The Nondestructive Ultrasonic Testing and Evaluation is directed to describe flaws microstructure, structure residual stress, reliability or quality insurance. An overview is given on the scientific background, interpretation capability based upon mathematic-numerical models, new design of equipments and demonstrated at components containing relevant flaws. Emphasis is laid upon SNR-improvement in the UT of austenitic weldings. Examples are shown using the backscattering technique with linear polarized shear waves. Stress measurements are performed with electromagnetic ultrasonic probes (EMAT). The surface inspection with magnetic imaging of the eddy current field or the multi-frequency eddy current methods are described. The possibilities of detection, classification and reconstruction of flaws are demonstrated in the time- and frequency domain with the satellite pulse echo technique, focal probes and the synthetic aperture methods like ALOK, Phased Array, Holography and SAFT.     

Steam generator mock-up for assessment of inservice inspection technology

A steam generator mock-up has been assembled for round-robin (RR) studies of the effectiveness of currently practiced inservice inspection technology for detection of current-day flaws. The mock-up will also be used to evaluate emerging inspection technologies. The 3.66 m (12 ft)-tall mock-up contains 400 tube openings, each consisting of nine test sections that can be used to simulate current-day field-induced flaws and artifacts. Included in the mock-up are simulations of tube support plate (TSP) intersections and the tube sheet (TS). Cracks are present at the TSP, TS, and in the free span sections of the mock-up. For initial evaluation of the RR results, various eddy current methods, as well as multivariate models for data analysis techniques are being used to estimate the depth and length of defects in the mock-up. To ensure that the RR is carried out with procedures as close as possible to those implemented in the field, input was obtained from industry experts on the protocol and procedures to be used for the exercise. One initial assembly of the mock-up with a limited number of flaws and artifact has been completed and tested. A second completed configuration with additional flaw and artifacts simulations will be used for the round robin.     

Axial strain localization of CuCrZr tubes during manufacturing of ITER-like mono-block W/Cu components using HIP

Two forms of axial strain localization of CuCrZr tubes, i.e., cracking and denting, were observed during the manufacturing of ITER-like mono-block W/Cu components for EAST employing hot isostatic pressing (HIP). Microscopic investigations indicate that the occurrence of axial strain localization correlates to the heavily deformed Cu grains and elongated Cr-rich precipitates as well as highly anisotropic microstructures, which impair the circumferential ductility. Annealing the as-received tubes at 600 °C alleviates cracking due to partial recrystallization of Cu grains. However, the annealed tubes are still sensitive to wall thinning (caused by non-uniform polishing or tube bending), which results in denting. Denting may cause bonding flaws at CuCrZr/Cu interfaces and the underlying mechanisms are discussed. To some extent, denting seems do not affect the high heat flux performance of components. In this paper, we demonstrate that testing only the axial mechanical properties is not enough for manufacturers who use HIP or hot radial pressing technologies, especially for those anisotropic tubes.     

Ultrasonic measurement of gap between calandria tube and liquid injection shutdown system tube in PHWR

Sag of CT or liquid injection shutdown system tubes in pressurized heavy water reactor is known to occur due to irradiation creep and growth during plant operation. When the sag of CT is big enough, the CT tube possibly comes in contact with liquid injection shutdown system tube (LIN) crossing beneath the CT, which subsequently may prevent the safe operation. It is therefore necessary to check the gap between the two tubes in order to confirm no contacts when using a proper measure periodically during the plant life. An ultrasonic gap measuring probe assembly which can be fed through viewing port installed on the calandria was developed and utilized to measure the sags of both tubes in a pressurized heavy water reactor in Korea. It was found that the centerlines of CT and LIN can be precisely detected by ultrasonic wave. The gaps between two tubes were easily obtained from the relative distance of the measured centerline elevations of the tubes. But the measured gap data observed at the viewing port were actually not the data at the crossing point of CT and LIN. To get the actual gap between two tubes, mathematical modeling for the deflection curves of two tubes was used. The sags of CT and LIN tubes were also obtained by comparison of the present centerlines with the initial elevations at the beginning of plant operation. The gaps between two tubes in the unmeasurable regions were calculated based on the measurement data and the channel power distribution.