Periodic In-service Inspection Research Articles

Enhanced Eddy Current Techniques for Detection of Surface-Breaking Cracks in Aircraft Structures

Abstract Well-timed detection of the defects (e.g., fatigue cracks and corrosive damages) in aircraft structures is the question of vital importance. Periodic in-service non-destructive inspection (NDI) is necessary to prevent expensive aircraft breakdowns during in-service life. Eddy current (EC) NDI method have many advantages for in-service aircraft inspection due to high mobility and the possibility to detect fatigue cracks without direct contact with the inspected surface (even through a protective coating without removal). Last decades some enhanced EC techniques were developed for aircraft in-service inspection in the Karpenko Physico-Mechanical Institute of National Academy of Sciences (Lviv) in collaboration with Ukrainian and foreign aircraft companies. Proposed earlier EC efficiency coefficient was applied for the selection of the ferrite-core EC probe parameters. Investigations concerned with the crack detectability for single-coil EC probes of different sizes were carried out to improve the inspection procedures. The sensitivity of the developed single-coil EC probe was experimentally investigated. The possibility to distinguish the signals associated with detected defect and lift-off was shown. The new inspection procedure was proposed for full suppression of the noise concerned with the rivet edge influence. The high-frequency EC flaw detector of Leotest VD 3.03 type based on self-generator mode and miniature single-coil EC probes application was developed for detection of the surface-breaking fatigue cracks. The original scheme of double-circuit self-generator with intermittent oscillations was invented and investigated. The pulse-repetition frequency of high-frequency oscillations was used as an informative parameter. The important feature of the developed self-generating scheme is the effective suppression of the lift-off influence (the changes of the clearance between the EC probe and the inspected object surface during the scanning). Developed EC flaw detectors of Leotest VD 3.03 type and inspection procedures were successfully implemented into the maintenance practice for detection of the fatigue cracks initiated in the wing and the fuselage structures and aviation engines in many companies such as ANTONOV AIRCRAFT, Lviv State Aircraft Repair Plant, Konotop aircraft repair plant «AVIAKON», Ukrainian International Airlines, MOTOR-SICH, State Enterprise “IVCHENKO-PROGRESS”, etc.

Estimation of Flow-Accelerated Corrosion Rate in Nuclear Piping System

Abstract Flow-accelerated corrosion (FAC) is a life-limiting factor for the piping network of the primary heat transport system (PHTS) in CANDU® reactors. The pipe wall thinning caused by FAC is monitored by carrying out periodic in-service inspections (ISI) to ensure the fitness-for-service of the piping system. Accurate prediction of the lifetime of various components in the PHTS piping network requires estimation of FAC thinning rate. The traditional Bayesian inference techniques commonly employed for parameter estimation are computationally costly. This paper presents an inexpensive and intuitive simulation-based Bayesian approach to FAC rate estimation, called approximate Bayesian computation using Markov chain Monte Carlo (ABC-MCMC). ABC-MCMC is a likelihood-free Bayesian computation scheme that generates samples directly from an approximate posterior distribution by simulating data sets from a forward model. The efficiency of ABC-MCMC is demonstrated by presenting a comparison with a likelihood-based Bayesian computation scheme, Metropolis-Hastings (MH) algorithm, using a practical data-based example. Furthermore, an innovative step has been proposed for reducing the Markov chain burn-in time in the proposed scheme. To indicate the need of a Bayesian approach in quantifying the uncertainties related to the FAC model parameters, results from the linear regression method, a common industrial approach, are also presented in this study. The numerical results show a notable reduction in computational time, suggesting that ABC-MCMC is an efficient alternative to the traditional Bayesian inference methods, specifically for handling noisy degradation data.

Technical reliability of active fire protection features – generic database derived from German nuclear power plants

For Probabilistic Fire Safety Analysis fire event and fault trees specific to the conditions of the nuclear power plant need to be established to estimate the corresponding branch point probabilities and end states for core damage frequency. That also requires applying technical and human reliability data for fire specific event sequences. The technical reliability of fire detection systems, fire and smoke extraction dampers, fire doors and fire extinguishing systems and equipment including extinguishing media supplies has been estimated by means of their failure rates. To account for epistemic uncertainties Bayesian estimation methods have been used, depending on homogeneity of the studied components.The data has been evaluated by analyzing the documentation of periodic in-service inspections as well as additional information and reports which resulted from the inspection findings. For more complex systems, in addition to the components’ failure rates, fault trees are presented to calculate the systems reliability.This type of data now covers up to 111 plant operational years of six power reactor units of different age and type. The generic data may also be applied as a-priori information for estimating the reliability of components with similar design and equivalent inspection and maintenance practice for other nuclear power plants.

Nonparametric Estimation of the Probability of Detection of Flaws in an Industrial Component, from Destructive and Nondestructive Testing Data, Using Approximate Bayesian Computation.

We consider the problem of estimating the probability of detection (POD) of flaws in an industrial steel component. Modeled as an increasing function of the flaw height, the POD characterizes the detection process; it is also involved in the estimation of the flaw size distribution, a key input parameter of physical models describing the behavior of the steel component when submitted to extreme thermodynamic loads. Such models are used to assess the resistance of highly reliable systems whose failures are seldom observed in practice. We develop a Bayesian method to estimate the flaw size distribution and the POD function, using flaw height measures from periodic in-service inspections conducted with an ultrasonic detection device, together with measures from destructive lab experiments. Our approach, based on approximate Bayesian computation (ABC) techniques, is applied to a real data set and compared to maximum likelihood estimation (MLE) and a more classical approach based on Markov Chain Monte Carlo (MCMC) techniques. In particular, we show that the parametric model describing the POD as the cumulative distribution function (cdf) of a log-normal distribution, though often used in this context, can be invalidated by the data at hand. We propose an alternative nonparametric model, which assumes no predefined shape, and extend the ABC framework to this setting. Experimental results demonstrate the ability of this method to provide a flexible estimation of the POD function and describe its uncertainty accurately.

Design and Development of a Control and Drive System for Cable Pusher Module for PFBR Steam Generator Tube Inspection

Abstract The Steam Generator of the PFBR is a vertical heat exchanger with sodium flowing in the shell side, and water in tube side. The tubes have an expansion bend to accommodate the thermal expansions and periodic in-service inspection is to be carried out to check the reduction in tube wall thickness and crack using Eddy Current Probes. Cable Pusher Module is used to insert and retract the probe in to the tube having 23 meter length. This paper presents the detailed design and development of CPM which includes the control system, software and mechanical aspects. CPM transmits the parameters like odometer reading, slip, velocity and time of inspection through serial interface. Graphical User Interface is developed using VB 6.0 Compiler for data logging. Slip of the motor is plotted for the tube length and this suggests an optimum design for the mechanical pulley arrangement of CPM.

In-service Inspection and Structural Health Monitoring for Safe and Reliable Operation of NPPs

In nuclear power plants, continuous efforts are made to ensure that plants are operated in safe, reliable and economical manner. While utmost care is taken during design, construction and commissioning of the structures, systems& components (SSCs), continued healthiness has to be ensured during operation phase in accordance with the design intent. This is primarily achieved through the establishment of a comprehensive life management programme of surveillance, condition monitoring, periodic In-service inspections (ISI) and maintenance, the purpose of which is to ensure that required safety margins are maintained for all important SSCs throughout plant service life. In any industry, the important mechanisms which lead to SSCs degradation are general corrosion, Flow accelerated corrosion, erosion, thermal effects, fatigue and mechanical wear & fretting. In a Nuclear Power Plant, in addition to these, degradation also occurs due to irradiation and creep. Therefore SSCs need to be designed with due consideration of all such degradation mechanisms. During operation phase, the limiting conditions for operation and requirements of surveillance & condition monitoring for all components important to safety are documented in Technical Specification. A plant specific ISI manual defines and elaborates the ISI internal inspection methods and acceptance criteria. The ISI requirements are decided taking into account best industry practices and operational experience and meet the requirements of existing codes and standards. Both these documents are approved by AERB. In pressurized heavy water reactors, major components covered in ISI programme are coolant channels, feeder pipes, steam generators, heavy water heat exchangers, pressure boundary components, relief valves etc. The inspection methods adopted include use of UT, ECT, DPT and visual inspection. In addition to functional checks for some of the components, ferrography, vibration monitoring and thermography etc are also utilized for rotating equipment. The SSCs where unacceptable indications are revealed by ISI are repaired, replaced or isolated. The base line data generated during Pre-Service inspection is used for comparing and trending of observations. The observations made during ISI are subjected to thorough review and analysis by qualified experts to obtain assurance that unacceptable degradation in component quality is not occurring and it remains fit for service. The containment system of a NPP plays a crucial role in minimizing dose to the public in case of an accident situation. The integrity of Primary and Secondary Containment is assessed by conducting integrated leak rate tests in every Biannual Shut Down. The concrete structure is also subjected to NDT checks on a specified frequency to ensure continued healthiness. Indian nuclear power plants have accumulated an operating experience of more than 380 reactor years of operation. Over the years, the programmes for surveillance, condition monitoring and in-service inspections have been improved significantly and match with best in the industry. As a result, there has been no major age related failure in any of the important SSCs and thus continued safe and reliable operation of NPPs is assured. This paper brings out the ISI & health monitoring methodology adopted at Indian NPPs, to ensure safe and reliable operation.

Technical requirements and applications for ISI of prestressed concrete nuclear containments with ungrouted tendons

With change of times, nuclear containment systems have been evolved, and nowadays dominant type in Korea is trending toward “prestressed concrete containments with ungrouted tendons”. Of many codes and standards for periodic inservice inspection of prestressed concrete nuclear containments with ungrouted tendons, ASME Section XI IWL and Regulatory guide 1.35 are widely used and applied as basic application technical requirements. However some conflicting and different contents exist to some degree even though the greater parts of those were mostly identical. In this paper, overall status of inservice inspections, application codes and standards and technical requirements of the prestressed concrete nuclear containments with ungrouted tendons, were reviewed. And the issues of use for inservice inspection application codes and standards were researched. Also, for inservice inspections of nuclear containments, Korean situations were presented and compared briefly.

Statistical Assessment of Integrity in Steam Generator Tubes Considering Uncertainty of NDE

A statistical assessment model for structural integrity of steam generator tubes with axial cracks at the top of the tubesheet was proposed using Monte Carlo method. In the model, a method for estimating the number of "real cracks" from in-situ inspection (ISI) data was used. Based on the estimated "real cracks", the number of newly detected cracks and growth of cracks during arbitrary operating period were simulated using the Monte Carlo method. The flaw growth rate used in the simulation was statistically calculated from the periodic in-service non-destructive inspection data. The number of cracks, the probabilistic distribution of crack sizes at the end of next operating interval and the probability of burst during operation were calculated from numerously repeated simulations using the proposed model.

Estimation of the Number of Physical Flaws from Periodic ISI Data of SG Tubes Using Effective POD

It is necessary to know the number of flaws and their size distribution in order to calculate the probability of failure or to estimate the amount of leakage through the tube wall of steam generators. But in-service inspection (ISI) flaw data is different from the physical flaw data. In case of a single inspection, it is easy to estimate the number of physical flaws using the POD curve. However, we may be faced with some difficulties in obtaining the number of physical flaws from the periodic in-service inspection data. In this study a simple method for estimating the number of physical flaws from periodic in-service inspection data is proposed. In order to obtain the flaw growth history, the flaw growth is simulated using the Monte Carlo method and the flaw size and the corresponding POD value are obtained for each flaw at each periodic inspection time. The flaw growth rate used in the simulation is statistically calculated from the in-service inspection data. By repeating the simulation numerous flaw growth data can be generated and the effective POD curve can be obtained as a function of flaw size. From the effective POD curve the number of physical flaws and their size distributions are obtained. The effective POD value is affected by flaw growth rate and POD of the used inspection method. The usefulness and convenience of the proposed method is evaluated from several applications and satisfactory results are obtained.

A neuro-fuzzy tool for CT-PT contact detection in a pressurized heavy water reactor

In a pressurized heavy water reactor (PHWR), contact between the calandria tube (CT) and the pressure tube (PT) makes them susceptible to delayed hydrogen cracking. Periodic inspection of the channels must be carried out to detect such contacts. As the number of channels in a PHWR is very large (306 in a 230 MW plant) periodic in-service inspection of all the channels leads to an unacceptable downtime. A non-intrusive technique that employs a system-identification method is presently used for contact detection. Attempts to identify all the contacting channels, without missing any, lead to overprediction of the number of channels in contact; i.e., many channels are diagnosed as contacting, while those channels are actually not in contact. This puts a large number of healthy channels in the at-risk list, reducing the efficacy of the method. Previously, the authors demonstrated the power of a neural post-processor in improving the strike rate of the system-identification tool. This paper demonstrates a stand-alone neuro-fuzzy tool for the detection of CT–PT contact. The network consists of a cascade of self-organizing artificial neural networks (ANNs), along with fuzzy processors. The performance of the network has been compared with that of the system-identification techniques. The noise tolerance of the network is also demonstrated.

Fracture mechanics investigation regarding the effects of cracks on the structural integrity of a BWR-core shroud

As a consequence of core shroud intergranular stress corrosion cracking (IGSCC) detected in the course of inservice inspections, a fracture mechanics analysis was carried out to evaluate the effects of postulated cracks on the structural integrity. In this study, critical crack sizes and crack growth were calculated. Due to the comparatively low stress acting on the core shroud during normal operation, the residual stresses in the welds make up the major proportion of the tensile stresses responsible for IGSCC. In order to consider residual stresses of the lower core support ring welds, a finite element analysis was performed at MPA Stuttgart using the FE-code ANSYS. The crack growth computed on the basis of USNRC crack growth rates d a/d t demonstrated that crack growth in depth direction increases quickly at first, then retards and finally comes almost to a standstill. The cause of this ‘quasi-standstill’ is the residual stress pattern across the wall, being characterized by tensile stresses in the outer areas of the wall and compressive stresses in the middle of the wall. Crack growth in circumferential direction remains more or less constant after a slow initial phase. As the calculation of stress intensity factors K I of surface flaws under normal conditions demonstrated, a ‘lower bound’ fracture toughness value is only exceeded in the case of very long and deep surface flaws. It can be inferred from crack growth calculations that under the assumption of intergranular stress corrosion cracking, the occurrence of such deep and at the same time long flaws is unlikely, regardless of the initial crack length. Irrespective of the above, the calculated critical throughwall crack lengths, which were determined using a ‘lower bound’ fracture toughness value, demonstrated that even long throughwall cracks will not affect the component’s integrity under full load. Moreover, it can be concluded from the studies of crack growth that—assuming intergranular stress corrosion cracking—a sufficiently long period will elapse before a crack which has just been initiated reaches a relevant size. Therefore, it can be stated that these cracks will likely be detected during periodic inservice inspections.

Artificial neural networks in CT–PT contact detection in a PHWR

In a pressurized heavy water reactor (PHWR), contact between calandria tubes (CT) and pressure tubes (PT) makes them susceptible to delayed hydrogen cracking. Periodic inspection of the channels must be carried out to detect the contact. As the number of channels in a PHWR is very large (306 in a 230 MW plant) periodic in-service inspection of all the channels leads to an unacceptable downtime. A non-intrusive technique that employs a system identification method is presently used for contact detection. The technique tends to overpredict the number of channels in contact, i.e. they diagnose many channels as contacting while the channels are in fact not in contact. This puts a large number of healthy channels on the at risk list reducing the efficacy of the method. This paper demonstrates the power of artificial neural networks (ANNs) in diagnosing the CT–PT contact. A counterpropagation neural network consisting of a Kohonen layer and a Grossberg layer has been employed. The noise tolerance of the network has been demonstrated.

The prediction of crack shape development for in-service cracks in offshore welded tubular joints

Abstract Due to the severe cyclic environmental loading experienced by offshore structures in the North Sea, fatigue has been identified as one of the most important causes affecting the long-term structural integrity of the welded tubular joints on these structures. Fatigue damage appears in the form of fatigue crack growth. Maintenance of structural integrity therefore concentrates on carrying out periodic in-service inspection and integrity assessment of the intact and cracked structures. The aims of the maintenance programme are to ensure that either no crack is present in the structures, or if cracks are discovered, that the remaining joints are still ‘fit’ for their purpose in terms of fatigue life and static capacity. The key information linking the in-service inspection and integrity assessment is the size of the cracks. Although the crack depth is the controlling variable for fatigue analysis, both depth and length of the surface breaking cracks are needed for the assessment of remaining ultimate capacity. In the past it was.found to be difficult to correlate the analytical prediction of crack aspect ratio (depth to half-length ratio) development with measured values from full-scale test results. Many integrity assessments therefore used a predefined aspect ratio pattern for the analysis. This paper reports an extension of a methodology in order to produce a better correlation between predicted and measured aspect ratio patterns. The methodology is based on fracture mechanics analysis with structural modelling.for cracked tubular joints. The paper reviews briefly the recent progress in the analysis of fatigue crack growth in tubular joints and outlines how the new methodology i.s formulated. The results showed that the methodology was adequate.for predicting aspect ratio growth for cracks deeper than 25% of the chord wall thickness.

Remote metallurgical investigations on pressure tubes removed from CANDU power reactors

As part of the periodic in-service inspection program CANDU reactors, pressure tubes are periodically removed for destructive examination. The procedures, equipment, and facilities used to perform metallurgical examinations on these highly irradiated components are described. The initial examinations of the tubes from the generating station are performed underwater in inspection bays. Detailed visual examination and metallography are subsequently performed in shielded hot-cell facilities; a description of the remote metallographic equipment and preparation techniques used is given. Examinations of two recently removed Zr-2.5%Nb pressure tubes containing fretting-wear flaws and a lamination flaw are used to highlight the techniques employed.

Effect of In-service Inspections on the Reliability Degradation of Structure Due to Fatigue Damage

The purpose of this paper is to set up a general method for estimating the reliability of ocean structural member with/without leak sensor through periodic in-service inspections. Emphasis is put on the classifications of all results which are obtained by the inspections into probability events and to make the probabilities of occurrences of each event clear at arbitrary inspection times.Choosing a connecting component of TLP tendon, which has a possibility of failure before and after the fatigue crack initiation, reliability analyses of inspections are carried out. And the effects of the fatigue properties of member, the failure rate of member, detectability of a crack by inspection and the shape of initiated crack on the relationship between structural reliability and inspection interval are discussed. Further, the reliability analyses of inspections under the condition of the leak sensor is attached inside of the pipe and under the condition of the crack initiation life and crack propagation life are random variables at the same time are developed.

A Logical Selection of Measures for Quality Assurance on the Integrity of Liquid-Metal Fast Breeder Reactor Primary Boundaries

An example of a logical approach to standardization of in-service inspection requirements is introduced, analyzing each measure according to risk reduction factors in order to develop a systematic configuration of safety measures. Comparison of a heavy pressure vessel section fabricated to resist pressure from the many nozzles in a light water reactor vessel with a thin-walled reactor vessel in a liquid-metal fast breeder reactor (LMFBR) operating at high temperatures and low pressure indicated no apparent difference in the failure probability of the reactor vessel. If both large failures and cracks are considered, 10/sup -5//vessel.yr is to be assumed. In assessing the unreliability of the guard vessel, dominant factors are common cause failures and subordinate modes. The coupling factor of the common cause failure of the guard vessel with reactor vessel failure is assumed 10/sup -3/. The subordinate mode failure of the guard vessel when the sodium leak from the reactor vessel is left unattended should be considered in the rate of 10/sup -2//reactor vessel failure. Especially in LMFBRs, visual tests are more practicable than volumetric tests for primary sodium boundaries, and the risk reduction factor by periodic in-service inspections is limited to only 5%.