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NDE methods continue to be refined for India's reactor fleet. For its BWRs, a test of different UT methodologies shows a clear favourite; for its PHWRs, a recent method for detecting tricky hydride blisters is successful. By Paritosh Nanekar and Bijoy Kumar Shah
There is a comprehensive ageing management programme in place for all operating nuclear power plants in India. Periodic in-service inspection of critical components goes a long way in assuring their structural integrity, thereby guaranteeing the safe operation of nuclear plants.
Non-destructive examination (NDE) techniques play a crucial role in this regard. The two boiling water reactors (BWRs) at Tarapur, India’s first nuclear power plants, have been in operation for more than three decades. The in-service inspection programme for BWR components is based on the guidelines of ASME Boiler and Pressure Vessel Code Sec. XI. Some of the critical components which are periodically examined during refuelling outages include: primary pipelines, feedwater nozzles, turbine blades, steam generators, core internals, pressure vessel and its internals, etc. Intergranular stress corrosion cracking (IGSCC) is the generic problem in these reactors. The components prone to such attack are periodically monitored by NDE for initiation and growth of IGSCC.
India currently has 16 operating pressurised heavy water reactors (PHWRs) and few more under construction. In Indian PHWRs, zirconium alloys are predominantly used as the material of construction for core components. The integrity of pressure tubes, which carries the fuel and the high temperature, high pressure coolant, is central to the safety of PHWRs. The pressure tube, which is prone to hydrogen attack specifically by delayed hydride cracking, hydride blistering and hydride embrittlement, is periodically examined during in-service inspection.
This article highlights NDE methodologies as being used and developed for in-service inspection of critical components of Indian BWRs and PHWRs and their role in ageing management of these components.
IGSCC in BWRs
One of the significant failures reported in 1990s in BWRs all over the world, was that of reactor core shroud cracking. The cracks were confined to weld heat-affected zones (HAZ) and the mechanism of cracking was identified as intergranular stress corrosion cracking (IGSCC) and irradiation-assisted stress corrosion cracking (IASCC). The core shroud of the BWR at Tarapur is a 25mm thick AISI 304 austenitic stainless steel cylinder. It partitions the feedwater in the reactor vessel downcomer annulus region from the coolant water flowing upwards through the reactor core. It provides structural support to the core and maintains its geometry. The core shroud has nine circumferential welds. The top four welds, H1, H3, H4A and H4B, are accessible for inspection from inside. These four welds are inspected by visual and ultrasonic examination. While the visual examination is carried out using an underwater radiation-resistant camera, special probe holders and manipulators are used for ultrasonic examination.
The core shroud inspection has been carried out during every re-fuelling outage since 1995. No crack-like indication has been observed in any of the welds examined . The effectiveness of visual examination for detection of fine cracks was qualified on SRCS (sensitivity, resolution, contrast standard) panels. These panels comprise stainless steel plate to which four wires of varying diameters (15 to 70 microns) are attached. These panels are kept underwater and the wires are viewed by a radiation-resistant underwater camera kept at a distance applicable during actual in-service inspection. These panels were fabricated in the laboratory and used for on-site qualification.
For ultrasonic examination, the technique for IGSCC detection is based on angle beam examination of the heat affected zone. For circumferential welds H4A and H4B, the probe holder comprises three ultrasonic transducers: two angle beam and one normal beam. One of the angle beam transducers is used for the top HAZ and the other for the bottom HAZ. The normal beam transducer is used to pick up the vessel ID signal which is used as a reference for ensuring the radial beam direction during angle beam examination. The probe holder is housed in a specially designed mechanism called a CART (carriage for advancing and retracting transducer). The CART assembly (Figure 1) was designed and fabricated in collaboration with Nuclear Power Corporation of India Limited. The objective of CART is to achieve an extended coverage (instead of spot checking) during ultrasonic examination of the H4A circumferential weld. The CART is connected to a grapple-operated manipulator that can take it to the desired H4A weld azimuth location. The development of CART system helped to examine the larger length of weld joint in a very short time.
Since the bottom five welds are not accessible for any examination, extensive stress analysis for safety assessment of a shroud that is assumed to be cracked at various weld locations has been done. The analysis was carried out at normal operating loads and loads due to potentially dangerous incidents such as re-circulation line break (RLB), main steam line break (MSLB) and seismic events . The results indicate that the safety functions such as control rod movement, core spray and poison injection are not impaired under these abnormal conditions.
IGSCC is also a generic problem in primary pipelines of BWRs. These are examined periodically as per the guidelines of ASME B&PV Code Sec. XI. The code calls for 100% ultrasonic examination of the welds and the heat affected zones by using the angle beam shear wave technique, with a 10% wall thickness-deep machined notch as the reference defect standard. Over the years, many IGSCC failures have been observed in these pipelines and corrective action has been taken. The old pipelines have also been replaced with new pipelines of IGSCC-resistant material. The new pipelines are inspected after installation and also periodically during refuelling outages for monitoring IGSCC.
The limitation of conventional ultrasonic techniques based on amplitude comparison for evaluation of IGSCC depth is well known. The difficulty arises due to the poor reflectivity of IGSCC to ultrasonic waves. In order to overcome these limitations, two approaches were standardized. The first approach uses a known-depth IGSCC as a reference defect standard, instead of a machined notch. This approach deals with the difference in reflectivity of IGSCC in the component and the reference standard. The second approach employs tip-diffraction techniques for sizing IGSCC. In these techniques, sizing is based on monitoring the time of travel of the reflected/diffracted signals from the crack extremities. In order to standardize the above two approaches, IGSCC was generated in 25mm thick stainless steel plates. The plate containing a deep, circular groove was first sensitized at 677°C for one hour. Subsequently, weld was deposited in the groove under restraint. This ensured residual stresses of significant magnitude in the heat-affected zone. The plate was then exposed to polyphonic acid, known to generate IGSCC in sensitized austenitic steel at room temperature. After overnight exposure, IGSCC was generated in the plate (Figure 2). The intergranular nature of the crack was confirmed by in-situ metallography.
The plate containing IGSCC was then subjected to ultrasonic examination. Three different techniques are used for sizing. They are: using 10% wall thickness deep notch as reference defect standard (as per ASME B&PV Code Sec. XI guidelines); using a known depth IGSCC as reference defect standard; and the tip-diffraction technique. The results of this study are shown in Table 1. The results of the investigation indicate that the depth sizing accuracy for IGSCC is extremely poor using a notch as a reference defect standard. The sizing accuracy is improved by using known-depth IGSCC as reference standard, but the best accuracy is obtained by the tip-diffraction technique. [3,4].
Inspecting PHWR channels
Pressurized heavy water reactors (PHWRs) are the mainstay of the Indian nuclear power programme. The Indian PHWR  consists of a few hundred coolant channels. A coolant channel assembly comprises a horizontal pressure tube (PT) carrying fuel and hot coolant, enclosed by a concentric calandria tube (CT). These tubes are separated by four garter spring spacers, which prevent hot PT coming in contact with cold CT (Figure 3). There are 306 coolant channels in a typical 220MWe Indian PHWR, and 380 channels in 540MWe Indian PHWR.
PHWR operating conditions lead to the degradation of the pressure tube with respect to (i) dimensional change (creep and growth), (ii) deterioration in mechanical properties (hardening and embrittlement) thereby reducing its flaw tolerance, (iii) the growth of existing flaws, which were too small or ‘insignificant’ at the time of installation, and (iv) initiation and growth of new flaws. The pressure tube, which is made of zirconium alloy (zircaloy–2 or Zr - 2.5% Nb), undergoes corrosion in an aqueous environment during service. This reaction releases hydrogen, part of which is absorbed in the pressure tube material. The absorbed hydrogen is responsible for the two most commonly observed degradation mechanisms that limit the life of a pressure tube. Delayed hydride cracking (DHC) occurs in conditions of relatively high tensile stress and the presence of hydrogen. Hydride blisters form when pressure tubes sag and touch calandria tubes, forming a cold spot. hydrogen accumulates in the cold spot and forms a brittle zirconium hydride blister.
The integrity of pressure tubes is central to the safety of PHWRs. In order to ensure this at all times during its service, the pressure tubes are periodically examined by NDE techniques. Such inspection provides valuable input on the presence or absence of flaws in pressure tubes and their characteristics to designers, plant operators and regulatory authorities for fitness-for-service assessments.
In India, in-service inspection of PHWR coolant channels is carried out using a semi-automated channel inspection tool known as BARCIS (Bhabha Atomic Research Centre channel inspection system). BARCIS consists of inspection head (which carries ultrasonic and eddy-current sensors) and a special sealing plug and drive mechanism. The NDE capabilities in BARCIS include ultrasonic wall thickness measurement and ultrasonic flaw detection in the pressure tube, eddy current detection of garter spring location and tilt, eddy current estimation of annular gap between pressure tube and calandria tube, and eddy current flaw detection on the inside surface of the pressure tube. During in-service inspection of the coolant channel, the inspection head is moved from inside of the pressure tube in a systematic manner from one end to the other. Indications, if any, are recorded and evaluated for acceptance/rejection .
The authors’ laboratory, Bhabha Atomic Research Centre, represented India during the International Atomic Energy Agency coordinated research programme (CRP) on intercomparison of techniques for pressure tube inspection and diagnostics. A total of seven laboratories from six countries participated in this CRP. The primary objective of this CRP was to arrive at the most effective NDE techniques for pressure tube inspection.
The CRP involved blind tests on pressure tube samples containing artificial flaws that resemble real flaws of concern such as delayed hydride cracking, fretting damage on pressure tube ID due to bearing pad or debris, lap-type defects and laminar flaws. The inter-comparison of NDE techniques based on the results of pressure tube sample investigation highlights the most reliable and accurate NDE method (ultrasonic, eddy-current or a combination of both) and also a specific technique for that NDE method (time-of-flight monitoring, amplitude monitoring, C-scan image, etc.) for detection and characterisation of various kinds of flaws encountered in pressure tubes [10, 11].
As part of this CRP, seven pressure tube samples were examined by ultrasonic testing. The flaws in these samples were characterized by B-scan and C-scan imaging. The images provided useful information on the nature of flaws such as DHC, debris fret, bearing pad fret and lap-type flaws (see Figure 4).
Detection of uncracked hydride blisters in pressure tubes poses several challenges since they cannot be sensed by conventional ultrasonic techniques (there is not much difference in the acoustic impedance of zircaloy and zirconium hydride). As a result, the interface between the blister and the parent metal does not reflect any significant amount of incident ultrasonic energy. Ultrasonic examination techniques based on ultrasonic velocity ratio measurement, and ultrasonic B-scan and C-scan imaging, have therefore been standardized for this purpose . These techniques are based on difference in sound velocity in zircaloy and zirconium hydride. For in-situ applications, the actual thickness of the pressure tube is not known. As a result the actual sound velocity cannot be measured. Velocity ratio is a useful parameter for such cases (Table 2).
Ultrasonic imaging was used as a tool for detection of hydride blisters in pressure tubes. A pressure tube sample containing laboratory-grown blisters was scanned in an immersion condition using normal beam longitudinal wave and angle beam shear wave techniques. B-scan images were collected during the travel of the probe. These images are shown in Figure 5a & 5b. Figure 5a shows the B-scan image using longitudinal waves. Since the velocity of longitudinal wave is higher in zirconium-hydride compared to zircaloy, the time of travel from the back wall echo is less at the blister location. This is clearly seen in the image. While using the shear wave technique (Figure 5b), the reverse happens. This is because of the lower shear wave velocity in zirconium-hydride as compared to ziracaloy-2.
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