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Keynote Nuclear: A Perspective on Development in NDE to Enable a Safe & Reliable Nuclear Power IndustryBaldev Raj and T.Jayakumar
Indira Gandhi Centre for Atomic Research, Kalpakkam 603102, India.
After giving a brief introduction to the directions of growth in NDE in nuclear industry in general, a few developments and applications of conventional and advanced non destructive evaluation methodologies for inspection of a few important components of Indian nuclear reactor systems for their safe and reliable performance are discussed. Special NDE techniques developed for on-line and off-line inspection of end cap welds, bearing pad welds and spacer pad welds of fuel bundles of pressurised heavy water reactors (PHWR), fuel cladding tubes for fast breeder reactor (FBR), inspection of ferromagnetic tubes of heat exchangers and steam generators, and tube to tube sheet weld joints of steam generators of FBR are some of the examples given in this paper. NDE procedures have been successfully developed and implemented for unique problems such as detection and location of leaks in pressure tubes and end shield of PHWR, for inspection of coolant channel systems of PHWR, core shrouds and feed water nozzles of boiling water reactors and ring beam of containment building of PHWR. The trends in NDE towards development and implementation of data fusion concepts, knowledge based systems, artificial intelligence, robotics intelligent processing and life extension methodologies are highlighted in the paper.
Keywords: Nuclear, inspection, life assessment, Non destructive evaluation (NDE), intelligent processing of materials (IPM)
Stringent safety requirement of nuclear industry is the primary driving force for the development of NDE science and technology. Over the years, a number of NDE techniques have been developed and implemented in nuclear industry for pre service and in service inspection requirements as per the specified codes, standards and procedures. Continuous developments are taking place in development and utilization of automated devices and robots to cut down costs and man rem consumption and also for inspection in inaccessible areas. Efforts are being made to use increasingly the probability of detection (POD) concepts consequent to the experience with the evaluation of defects in Sizewell reactor. Efforts are being made for getting general agreement on the use of POD concepts for quantitative defect evaluation in nuclear systems.
An area of rapid development taking place is the use of simulations and solutions to inverse problems for comprehensive defect characterisation, as this approach reduces the demand on the necessity for large scale mock ups and the cost and time for in service inspection and life management would be considerably reduced. With the growing demand for life assessment and extension of plants without compromising the safety, comprehensive defect characterisation using signal and image processing approaches and evaluation of changes in microstructures and residual stresses in components and structures have become essential. Extensive efforts are being made to develop NDE methodologies to meet these demands. While pockets of excellence in development and application of NDE techniques exists with different groups internationally, there is a necessity to bring synergism among these groups for the benefit of overall international community in the nuclear industry. The forums such as Asia-Pacific Conference on NDT, International Committee on NDT and World Federation of NDT Centres should take up this issue. In this paper, the developments and applications of NDE techniques in the Indian nuclear industry are highlighted.
The NDE technology has advanced considerably in India, in tune with the trend world over to meet the stringent requirements in ensuring the quality of fabricated components and reliable performance of plant components in service in nuclear industry to ensure safety and reliability. Conventional and advanced NDE techniques and procedures including special sensors, instrumentation and software for signal analysis and image processing have been developed in-house to improve the sensitivity and reliability for quantitative characterization of defects and leak detection and the requisite expertise has been established. Special NDE techniques developed for on-line and off-line inspection of end cap welds, bearing pad welds and spacer pad welds of fuel bundles of pressurised heavy water reactors (PHWR), fuel cladding tubes for fast breeder reactor (FBR), inspection of ferromagnetic tubes of heat exchanges and steam generators, and tube to tube sheet weld joints of steam generators of FBR are some of the examples that would be presented in the paper to demonstrate the challenging demands met by the NDE experts. NDE procedures successfully developed and implemented for unique problems such as detection and location of leaks in pressure tubes and end shield of PHWR, for inspection of coolant channel systems of PHWR, core shrouds and feed water nozzles of boiling water reactors and ring beam of containment building of PHWR would also be discussed. The trends in NDE towards development and implementation of data fusion, knowledge based systems, artificial intelligence, robotics, intelligent processing and life extension methodologies are also highlighted. Even though, extensive studies have also been carried out on developing NDE methodologies for evaluation of microstructures, creep and fatigue damage in different structural materials of interest to nuclear industry (Baldev Raj and Jayakumar, 1999), due to paucity of space, these aspects are not discussed in this paper.
NDE methodologies have been developed and standardised for stringent quality control of the fuel cladding tubes of fast reactors. ASTM A 771 gives the guidelines on the mandatory use of ultrasonic testing for inspection of austenitic stainless steel tubing for breeder reactor core components. However, studies made by us have clearly demonstrated that both UT & ECT should be considered as complementary techniques and are essential to be employed on finished tubes. Tubes having no defects or defects less than 7% wall thickness were accepted (Barat et al., 1982). The significant percentage of unacceptable tubes by eddy current testing after their acceptance by UT underscored the importance and the essential nature of eddy current testing for the pre-service quality control. Random checks by metallography indicated that defects other than inclusions detected by ECT but missed by UT were short in length, but posed no problem for detectability by ECT. ECT also detected localised dispersion of fine inclusions, which were missed by UT. Studies on the feasibility of defect detectability in tubes with periodic wall thickness variations (banded) have been carried out with various approaches including, conventional single frequency eddy current testing (SFECT), multi frequency ECT (MFECT), phased array ECT (PAECT), deconvolution of conventional signals and the use of artificial neural networks (ANNs). PAECT detected all the artificial defects with a good SNR. Since the use of PAECT requires special instrumentation and probes, alternative approaches using deconvolution methods and ANN approaches were studied and found to perform very satisfactorily for reliable inspection of banded cladding tubes with scope for automation on the shop floor.
In PHWR, uranium dioxide fuel pellets are encapsulated in Zircaloy-2 cladding tubes of 0.37 mm thickness and sealed with end caps. Resistance welding of the end caps with this cladding tubes leaves a material upset both inside and outside the joint. The current quality assurance procedures using helium leak testing does not give sufficient confidence when the target is zero failure rate with the objective of keeping the coolant of nuclear reactors as less radioactive as possible. Ultrasonic testing of the end-cap weld joints poses several difficulties due to:
Intelligent processing approach is being pursued for fabrication of nuclear components with high reliability. Studies using both acoustic emission and infrared thermography techniques for on-line assessment of end plug welds have been carried out (Baldev Raj et al., 1998). Imperfections such as end squareness and ovality of tubes, graphite impurities etc. were considered. The total AE counts for tubes welded with impurities (graphite) is much higher. Analysis of the AE data has indicated that it is possible to detect formation of defects during welding process due to different imperfections. Thermal imaging was also carried out on these elements after the welding process was completed. Analysis of the cooling patterns and the cooling rates from the thermal images have indicated that it is possible to detect imperfections such as graphite impurities and the sparking that occurs during the welding process. The good weld reveals a thermal pattern with uniform isothermal width while a thermal pattern with non-uniform isothermal width reveals the bad weld.
The importance of high integrity welds in steam generators of fast reactors arises due to risks of sodium- water reaction. Tube to tube weld joints are the regions of a tube where the possibility of a leakage path is highest. Thus, extreme care is taken in the quality assurance of the joints. In this regard, two aspects are important:
The configuration of the weld joint is such that micro focal radiography with rod anode is the only solution for examination of such welds. A backward throw probe with a diameter of 10 mm and with a beam spread of - 5 x 55 x 360 degrees was inserted from the tube sheet side and the radiography of the weld was carried out with a projective magnification of 3X (Baldev Raj and Jayakumar, 1999). A special radiographic cassette was designed and developed for this purpose. Results on the welds of these tubes have shown that it is possible to resolve a 30 - 40 micron diameter steel wire placed on the tube ID and corresponds to 1.3% - 1.6% of the wall thickness of the tube. The feedback information from the micro focal radiography studies were helpful to arrive at the correct weld parameters for an optimum weld joint.
Magnetic Barkhausen noise (MBN) technique has also been developed to assess the adequacy of PWHT in releasing the residual stresses (Baldev Raj and Jayakumar, 1999). MBN measurements were made so as to cover the weld, HAZ and base metal regions. After the measurements in the as welded condition, the tubes were post weld heat treated at 973 K for 1 h followed by air cooling. The results showed that, in the as welded condition, there is maximum MBN peak height at both ends (base metal region) and there is gradual reduction in peak height with decreasing distance from weld. The weld shows the minimum MBN peak height. After the PWHT, the MBN peak height becomes more or less same at all the locations. These results provide the possibility to evolve an acceptance criterion based on the MBN peak height values to ensure the effectiveness of the PWHT.
The oldest power reactor in India is the Tarapur Atomic Power Station (TAPS) which has experienced more than 25 years of service (the economic life of the reactor). Many studies were undertaken to assess the residual life and thus to extend the reactor life. Various components like reactor dry well, reactor vessel, vessel internals, piping, pump casing, feed water nozzles, core shrouds etc. undergo degradation in service and thus decide on the life extension of the BWRs. A typical example of inspection of core shrouds of Indian boiling water reactors is given below:
Inspection of Core Shrouds in Boiling Water Reactors:
The core shroud in boiling water reactors (BWR) is a stainless steel cylinder which partitions feed water in the reactor vessel's downcomer annulus region from cooling water flowing through the core . The core shroud also provides a refloodable volume under postulated accident conditions and maintains the core geometry. Intergranular stress corrosion cracks (IGSCC) have been observed in the HAZ regions of the welds in the shrouds made of stainless steels due to residual stresses and operating stresses, sensitisation and the corrosive environment. Additionally, the radiation environment enhances the susceptibility for IGSCC of the shroud and this is referred to as irradiation assisted stress corrosion cracking (IASCC), which normally occurs above a fluence value of about 5x1020 n/cm2. Visual examination for location of the cracks and UT for detection of the cracks (which showed good correlation with the visual examination results) and for the determination of the depth of the cracks by remote handling devices have been developed and implemented (Srinivasan and Prasad,1994b).
Both analytical methodologies and condition monitoring tools have been developed for assessing the safe operating life of the coolant channels, which are the life limiting components of a PHWR (Srinivasan and Prasad, 1994a). Approaches to extend the life whenever feasible and for replacement of the coolant channels at the end of their safe operating life have also been developed. Subsequent to the failure of a pressure tube (PT) in Pickering-2 reactor, in Canada in 1983, it was realised that the garter springs could get displaced during hot conditioning and/or during the operation of the reactor. Special eddy current and ultrasonic testing probes and procedures have been developed with capabilities for performing ISI to obtain the following information:
A highly sensitive focussed eddy current probe has been developed to detect extremely fine defects of the order of scratches in the range of 50 - 75 microns. This special probe was successfully utilised to detect the presence of scratches greater than 50 microns deep in the Calandria tube of a PHWR that had undergone en-masse coolant tube replacement. Indigenous codes, systems and methodologies have been developed for assessing safe operating life of pressure tubes under the degrading effects of creep, hydriding and embrittlement. Continuous updating of the codes and methodologies based on ISI and post irradiation examination (PIE) is also adopted.
Leak detection in pressure tubes of PHWRs:
In a condition assessment campaign undertaken, in the unit 1 of Madras Atomic Power Station (MAPS) two possibly leaking pressure tubes among the 306 pressure tubes were detected using acoustic emission technique (AET) coupled with advanced signal analysis (Baldev Raj and Jayakumar, 1999). Conventionally adopted time domain methodology did not yield satisfactory results because of the poor SNR of the leak signal. The characteristics of the signals from different channels due to background noise were not similar and hence it was not easy to identify the leaking channel. Therefore, the first task was to segregate the channels with similar signal characteristics into different groups and to select the group containing the suspected leaking channel. Accordingly, a group of 15 channels were short listed as the possible leaking channels from the 306 channels using the criterion that the leaking channel should have signals with frequencies above 200 kHz. The ratio of the spectral energy between two different frequency bands, namely 700 to 1000 kHz and 40 to 175 kHz, and its variation with an increase in pressure were used to identify the suspect channels. For the two identified channels, this ratio increased with an increase in the pressure. Subsequent investigations by the plant personnel using vacuum testing and hydro testing confirmed that one of the two channels identified by the AET had the leakage of heavy water from the pressure tube. Similarly, AET has been successfully employed for detection of leaks in one of the end shields of a PHWR.
Assessment of ring beam of containment structure using impact echo technique:
Containment structures of some of the PHWRs are made of prestressed concrete with pretensioned cables. One of the important components of this containment structure is the ring beam. The inner containment dome of one of the PHWRs got delaminated during construction. Due to sudden release of force in prestressing cables during collapse of the dome and during subsequent detensioning/slackening of the cables, delaminations in the ring beam were suspected. Impact echo testing has been carried out for assessment of the structural integrity of the ring beam (Anish Kumar et al., 2000). In order to develop the test procedure for carrying out impact echo testing, mock up calibration blocks having a size of 4000 X 4335 mm and representing a circumferential length equivalent to about 4 degrees of the ring beam containing simulated flaws, viz. voids of sizes 50, 100 and 200 mm at a depth of 500 mm, surface opening cracks of 25, 50 and 75 mm depth, reinforced bars of diameter 20, 32 and 45 mm at a depth of 50 mm etc., was made. In order to study the response of the delamination of reinforced rod, various rods of diameter 20, 32 and 45 mm each at a depth of 50 mm were kept and shaken before the settling of the concrete to produce disbond at the steel-concrete interface. The results also indicated that the voids of 100 mm and 200 mm diameter could be detected at a depth of 500 mm, the reinforced rods of diameter 20, 32 and 45 mm could be detected at a depth of 50 mm and the depths of the surface opening cracks could be measured with the accuracy of about 10%. The delamination of the rods also could be detected if the position of the rod is known. Sometimes, even if the information on the position of the rod is not known, by observing the additional peak corresponding to flexural vibrations, the delamination can be detected reliably.
The detectability of the impact-echo system in terms of the depth and the lateral dimension of the defect has also been established. It has been found that a void can be detected, if the ratio of the depth of location to the lateral dimension of the void is less than 5. A reinforced rod can be detected if the ratio of the depth of location to the diameter of the reinforced rod is less than 3. For detection of all these types of defects, the basic criterion, that the impact echo signal should contain waves having a wavelength less than the size of the defect to be detected, must be satisfied. Based on the optimized test parameters identified with the help of studies carried out on the mock up blocks, impact echo testing was successfully carried out on the ring beam of the reactor containment structure for assessing its structural integrity. To the best of our knowledge, this is internationally the first time that impact echo technique has been employed for integrity evaluation of a critical component of the containment structure of a nuclear power plant.
Every nuclear power plant consists of a large number of heat exchangers and steam generators. Pitting corrosion and stress corrosion cracking are two major failure modes observed in the case of heat exchanger and steam generator tubes, since they are subjected to continuous flow of high temperature fluids, steam, and other aggressive environments. ECT is the most popular technique for periodic monitoring of these tubes because of its ease of operation, versatility and reliability. Multi frequency eddy current testing procedures have also been developed to eliminate the interference of signals, due to probe wobble and presence of support plates and sodium deposits, during ECT of heat exchangers of fast breeder reactors.
ECT of ferromagnetic heat exchanger tubes is difficult due to their high and continuously varying magnetic permeability. The use of magnetic saturation can overcome these difficulties to a large extent. The ferromagnetic tubes can be satisfactorily inspected by ECT, if they are magnetically saturated such that the material starts behaving as non-ferromagnetic. In this area, IGCAR has successfully designed an eddy current probe using a high strength permanent magnet of Nd-Fe-B and the performance of the probe was evaluated using an ASME calibration tube (made of 2.25Cr-1.0Mo steel) of a PFBR steam generator, with artificial defects (Baldev Raj and Jayakumar, 1999). For ISI of ferromagnetic tubes, a new technique that has great promise and potential is remote field eddy current testing (RFECT). The primary advantage of this technique is the ability to inspect tubular products with equal sensitivity to both internal and external metal loss or other anomalies, linear relationship between wall thickness and the measured phase lag and absence of lift-off problems. The technique features the ability to inspect both ferromagnetic and non-ferromagnetic materials with equal sensitivity. Pioneering work has been carried out at IGCAR with respect to the development of an RFECT instrument and the computer simulation of the technique. Wall loss down to 15% was detected using an indigenously developed RFECT instrument. The presence of transition and remote field zones and the influence of tube diameter and wall thickness on the changes in the transition and remote field zones have been studied using a 2D-FEM code.
One recent technique that has shown tremendous potential in detecting defects in expansion zones and rolled joints of heat exchangers with equal sensitivity to both longitudinal and transverse defects is phased array ECT (PAECT). This system uses a substantially different bridge circuit and probes that produce a rotating magnetic field of constant magnitude which is insensitive to expansion zone, tube sheet or support plate. At IGCAR, a novel tandem PAECT probe has been designed and developed that minimises the insensitive zones, in turn, leading to enhanced sensitivity for inspection of tubes.
The demand for automated inspection systems for heat exchangers and steam generators is ever increasing mainly because of
Imaging techniques are playing an important role in non-destructive evaluation. An eddy current impedance imaging (ECII) system has been built around a Personal Computer (PC) (Baldev Raj and Jayakumar, 1999). Many a time, it would be necessary to remotely inspect welds by ultrasonic testing. Identification of weld centre line is necessary for fixing the required skip distance and scan ranges for UT. In the case of austenitic stainless steel welds, by making use of the presence of delta ferrite in the material, eddy current inspection method has been developed for identification of weld centre line. The accuracy of detection of the weld centre line is found to be ± 0.1mm. These studies have been further extended to develop an intelligent imaging system for quantitative evaluation of defects (Rao et al., 2001). This scheme quickly detects defects in the presence of disturbing variables such as surface roughness, material property / microstructural variations, lift-off, edge-effect etc. and produces accurate three-dimensional pictures of the defects. This is accomplished through the synergistic use of artificial neural networks and morphological thinning based image processing methods. The EC probe for imaging has been optimised using an indigenously developed finite element model.
The portable X-ray diffraction techniques can rapidly measure applied and residual stresses in a small area on the surface of a component. The technique is limited to measurement of surface residual stresses only (up to about 20 microns thickness), thus enabling evaluation of fatigue damage in components, as part of condition assessment programme. Studies have been carried out to assess the progress of fatigue damage in steam turbine blades (Rai et al., 1997). The fifth stage low pressure turbine blades of PHWR have a 'fir tree' type root for fixing the blades to the rotor. The blades are fabricated in such a way that high compressive residual stresses (550MPa) are introduced in the 'fir tree' region. These compressive stresses offset the operating tensile stresses, thus the blades experience net stresses which are below the fatigue limit. Due to stress concentration in the 'fir tree' region resulting from fabrication tolerances and variation in the operating stresses, fatigue damage gets accumulated in the blades, leading to redistribution of the original compressive residual stresses. The FEM based stress analysis indicated that relatively high stress concentration exists at one of the radii of the fir tree region. Systematic residual stress measurements made in the as fabricated blades and also in the blades operated for different durations demos\demonstrated that changes occur in surface residual stresses due to their redistribution consequent to fatigue damage thus help to identify and quantify the accumulated fatigue damage and enable replacement or extension of life of the blades. Some of the blades operated in service showed reduction in the compressive stresses at the fir tree region due to accumulation of the fatigue damage. Further, the residual stress measurements made on the blades with service induced cracks showed considerable reduction in the compressive residual stresses ahead of the crack tip.
One of the emerging possibilities to effectively utilise the knowledge explosion is to explore the concepts of artificial intelligence (AI) wherever applicable. Successful implementations of AI concepts, in the form of verified and validated knowledge based systems (KBS) and knowledge-based inference mechanisms are currently being developed for various specific problems. A KBS is ideally suited
Multi sensor data fusion and integration can be defined as the synergistic use of supplementary and complementary information from multiple sources to assist in the accomplishment of a task. During NDT&E, these sources of information can originate from multiple identical sensors such as two (or more) eddy current sensors or from multiple different sensors such as one eddy current and one ultrasonic sensor. In the simple case of two (or more) identical sensors, one sensor can be used for decision making and the other for consultation, to compare or check information conveyed by the first one. The final decision can be made from combination of information from one or both sensors into a single feature.
The development and application of non destructive evaluation techniques and methodologies for the life management of nuclear power plants in India are described. The indigenous development carried out to meet the stringent quality requirements in evaluation of fabricated components and innovative methodologies using multidisciplinary approaches and advances for assessment of in-service performance of plants are highlighted.
Authors wish to thank many colleagues from Department of Atomic Energy whose work has been discussed in this paper.
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